GNRO-2012/00045, Response to Request for Additional Information Regarding Request ISI-17

Response to Request for Additional Information Regarding Request ISI-17

ML12131A408
Person / Time
Site:	Grand Gulf
Issue date:	05/09/2012
From:	Perino C Entergy Operations
To:	Office of Nuclear Reactor Regulation, Document Control Desk
References
GNRO-2012/00045, ISI-17
Download: ML12131A408 (13)

Text

• Entergy

.-=.'" Entergy Operations, Inc.

P.O. Box 756 Port Gibson, Mississippi 39150 Tel: 601-437-2800 Christina Perino Licensing Manager GNRO-2012/00045 May 9,2012 u.s. Nuclear Regulatory Commission (NRC)

Attn: Document Control Desk Washington, D.C. 20555

Subject:

Grand Gulf Nuclear Station Response to Request for Additional Information Regarding Relief Request ISI-17 Grand Gulf Nuclear Station (GGNS), Unit 1 Docket No. 50-416 License No. NPF-29

References:

1. Grand Gulf Nuclear Station Relief Request ISI-17 Repair Plan for lSI Weld N06B-KB, dated May 2, 2012, (GNRQ-2012/00040; ADAMS Accession No. ML12124A245)

2. NRC Electronic Transmission of Questions, Request for Additional Information regarding Relief Request ISI-17, dated May 8,2012 (GNRI-2012/00045; TAC No. ME8525)

Dear Sir or Madam:

Grand Gulf Nuclear Station (GGNS) requested approval of Relief Request ISI-17 to repair degraded weld N06B-KB at the "C" nozzle in the Residual Heat Removal (RHR) 1 Low Pressure Core Injection (LPCI) system in correspondence dated May 2, 2012 (see Reference 1). A draft Request for Additional Information (RAI) was received on May 7,2012 from the NRC and was followed up with a conference call on May 8, 2012 to clarify. The NRC Project Manager followed up with the final RAI (Reference 2). Attached is GGNS's response.

This letter contains no new commitments.

If you have questions or require additional information concerning this report, please contact Mr. Stephen Scott at (601) 368-5456.

Sincerely,

~;t~

CLP\rrj

Enclosure:

Response to RAI Attachment(s): (see next page)

GNRO-2012/00045 Page 2 of2 Attachments(s): 1. Sketch of Weld Overlay on Weld N06B-KB

2. Structural Integrity Letter: Report No. 1200536.401.RO cc:

NRC Senior Resident Inspector Grand Gulf Nuclear Station Port Gibson, MS 39150 U.S. Nuclear RegUlatory Commission ATTN: Mr. Elmo E. Collins, Jr.

Region Administrator, Region IV 1600 East Lamar Boulevard Arlington, TX 76011-4511 U. S. Nuclear Regulatory Commission ATIN: Mr. Alan Wang, NRR/DORL Mail Stop OWFN/8 B1 Washington, DC 20555-0001

Enclosure to GNRO-2012100045 Response to RAI

Relief Request 181-17 Request for Additional Information (RAI) Questions and Responses

1. On page 5 of the relief request, the licensee stated that it will perform fracture mechanics analyses to predict crack growth of the detected flaw based on intergranular stress corrosion cracking (IGSCC) and fatigue. (1) Discuss the crack growth rate of the IGSCC that will be used for the crack growth calculation and provide the reference. (2) Discuss briefly the fracture mechanics analyses (Le., methodology, technical basis, input, initial flaw size). (3)

Discuss the acceptance criteria for the results of the fracture mechanics analyses.

Entergy Response: See Structural Integrity Letter: Report No. 1200536.401.R0 (Attachment 2 to GNRO-2012100045).

2. On Page 7 of the relief request, item (d) states that PT (penetrant test] examination will be performed in accordance with the ASME Code,Section III. Confirm that the acceptance criteria for the PT results (indications detected) will be based on the ASME Code,Section III, NB-5000.

Entergy Response: The PT examination acceptance criteria of the weld overlay will comply with NB-5000 of the 1992 Edition of ASME Section III. However, the PT acceptance criteria of the base material adjacent to the weld overlay will comply with NB-2500 of ASME Section III. This PT acceptance criteria is specified in Appendix Q, paragraph Q-4100(b) of the 2004 Editioni2005 Addenda of ASME Section XI for acceptance of the weld overlay. It should be noted that this weld overlay PT acceptance criteria is consistent with Code Case N-74D-2. To comply with preservice examination requirements of N-504-4, paragraph (i), the PT examination will also comply with the acceptance standards of IWB-3514-2.

3. Provide (1) the nominal diameter of the pipe and pipe wall thickness, (2) Figure 2 of the relief request shows some dimensions of the overlay and nozzle configuration. Provide a more detailed design drawing of the weld overlay, nozzle, safe end and pipe.

Entergy Response: See sketch on Attachment 1 which provides dimensions of the nozzle, safe end, weld overlay and nozzle configuration. It should be noted that Figures 1 and 2 in the original relief request were based on NDE data, and provided a preliminary overlay design thickness that was selected to optimize the future examination coverage. The sketch provided in this RAI response reflects nominal design values and variable dimensions that are dependent on the final contour of the overlay deposit. The outside diameter of the nozzle to safe end transition varies, therefore the wall thickness and the overlay thickness will vary depending on the axial location. To avoid confusion, it should also be recognized that Figures 1 and 2 in the original relief request show the nozzle on the right hand side with the safe end on the left hand side of the sketch, while the sketch provided in this RAI response shows the nozzle on the left hand side with the safe end on the right hand side of the sketch.

Relief Request 151-17 Request for Additional Information (RAI) Questions and Responses

4. (a) Discuss any surface preparation for the nozzle, safe end, and pipe prior to weld overlay installation. (b) If the surface configuration of the weld/nozzle is changed by the surface preparation process, discuss whether the weld be reexamined folloWing the surface preparation.

Entergy Response: The weld crown was removed to achieve the required flatness prior to performing the Appendix VIII, Supplement 10 volumetric examination that identified the flaw in the dissimilar metal weld. The only additional surface preparation performed was light buffing to achieve a clean surface for performing pre-overlay surface examinations, and to achieve a bright metal surface for welding. The design configuration of the nozzle, weld, and safe end assembly was not changed by the surface preparation.

Attachment 1 to GNRO-2012100045 Sketch of Weld Overlay on Weld N06B-KB

Attachment 1 to GNRO-2012/00045 Relief Request 181-17 Request for Additional Information (RAI) Questions and Responses 2.79" (Approx.)

2.41" (Approx.) 2.54" 3.54" 11.625" Dia. 0.75" (Approx.)

SKETCH OF WELD OVERLAY ON WELD N06B-KB

Attachment 2 to GNRO-2012/00045 Structural Integrity Letter: Report No. 1200536.401.RO

e Structural Integrity Associates, Inc.*

5215 Hellyer Ave.

Suite 210 san Jose, CA 95138-1025 Phone: 408-978-8200 Fax: 408-978-8964 www.slructinl.com mlaylor@slructint.com May 8, 2012 Report No. 1200536.401.RO Quality Program: IX1 Nuclear D Commercial Mr. Doug Jones Manager, Engineering Programs Grand Gulf Nuclear Station - Entergy 7003 Bald Hill Road Port Gibson, MS 39150

Subject:

Grand Gulf Nuclear Station - Request for Additional Information (RAI) for RR 17

- SI Response to RAI Question No.1

Reference:

1. E-mail from Entergy (J. Gobell) to Structural Integrity Associates (M.

Taylor) dated May 07, 2012;

Subject:

GG Request for Additional Information (RAI) for RR17

2. Grand Gulf Nuclear Station (GGNS) Relief Request ISI-17 Repair Plan for lSI Weld N06B-KB

Dear Mr. Jones:

In the Reference 1 e-mail, Entergy requested that Structural Integrity Associates (SI) prepare a draft response to question number 1 of the subject RAI from the NRC regarding the Reference 2 relief request. Question number 1 is stated as follows:

1. On page 5 of the relief request, the licensee stated that it will perform fracture mechanics analyses to predict crack growth of the detected flaw based on intergranular stress corrosion cracking (IGSCC) and fatigue. (1) Discuss the crack growth rate of the IGSCC that will be used for the crack growth calculation and provide the reference. (2) Discuss briefly the fracture mechanics analyses (i.e.,

methodology, technical basis, input, initial flaw size). (3) Discuss the acceptance criteria for the results of the fracture mechanics analyses.

SI's draft response is provided in Attachment 1 of this report.

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Mr. Doug Jones May 8, 2012 Report No. 1200536.401.RO Page 2 of2 Prepared by: Verified by:

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,.r ", ",,/,",,-_r~~_<...../t~-_J May 8, 2012  :,.... May 8,2012 Moses Taylor, Jr., P.E. Date Anthony J. Giannuzzi Date Senior Associate Senior Associate Approved by:

/l1J1t.Jt~ ) May 8, 2012 Moses Taylor, Jr., P.E. Date Senior Associate cc: Joseph Weicks Jamie GobeIl Christina Perino Stephen Scott William Sims Michael Briley Keith Brinson Jeffery Seiter Dennis Wiles Hal Gustin Dan Sommerville Angah Miessi

Attachment I SI Response to Question No. I ofRAI for Grand Gulf Nuclear Station Relief Request RR 17 (1) SI proposes to use the IGSCC crack growth rate for normal water chemistry provided in Figure A-21 ofEPRI BWRVIP-59-A [I]. The crack growth rates from Figure A-21 are as follows:

da/dt (in/hr) = 1.6 x 10- 8 (K)25 (K< 25 ksi-in 1/ 2) da/dt =5 x 10- 5 in/hr (K 2: 25 ksi_in I/2)

Although the Grand Gulf Nuclear Station employs both hydrogen water chemistry (HWC) and on-line noble chemistry (OLNC) as a global IGSCC mitigation technique against the initiation and propagation ofIGSCC and these BWR water chemistry practices have been shown to be effective in preventing the initiation and propagation of IGSCC of most reactor pressure vessel (RPV) nozzles, for this particular LPCI nozzle location, it appears that no significant benefit is achieved [2]. Accordingly, the more conservative normal water chemistry crack growth rates will be used.

(2) The crack models shown in Figure I are taken from the library of the fracture mechanics program pc-CRACK for the fracture mechanics and crack growth analysis. A crack model of a full circumferential crack in a cylinder with tJR = 0.2 is used with the axial stresses, and a model of a semi-elliptical longitudinal crack in a cylinder (with 0.1 < tJR < 1.0) is used with the hoop stresses.

The fatigue crack growth law for Alloy 600 in high purity BWR water containing 300 ppb dissolved oxygen, is obtained from NUREG/CR-6721 [3]:

CGRenv = CGRair + 4.4 X 10-7 (CGRair)033 (I) where CGRair (m/cycle) = da/dN =C A600 (l - 0.82 Rr 2.2 (L\K)4.1 (2) l12 where L\K is in MPa'm , and constant CA600 is given by a third-order polynomial of temperature T (OC) expressed as Using the applicable operating temperature in OF, assuming an R ratio of 0.9, and converting the unit of L\K from MPa--Vm to Ksi--Vin, the fatigue crack growth law in Equation (2) is converted to an equation of the form dA/dN = Constant x L\K41 The applicable fatigue cycles for the thermal transient events are distributed evenly over 40 years. For conservatism, the thermal and pressure ranges are assumed the full fluctuation from Attachment I to 1200536.40IRO I I)BInIcfInI IIIIt(Jt", Assocllles, Inc.-

Attachment 1 SI Response to Question No.1 ofRAI for Grand Gulf Nuclear Station Relief Request RR 17 zero to maximum stress. The full piping moment is conservatively added to the maximum K value for all cycles. An initial crack depth as reported in the examination results is used in the fatigue crack growth calculation. Fatigue crack growth threshold is assumed to be zero.

In addition to fatigue crack growth, IGSCC must be considered when the steady-state normal operating stress intensity factors are shown to be tensile for some flaw depths. Sustained steady-state normal operating stresses are the only stresses that need to be considered for the SCC growth analysis. The sustained stress intensity factors due to residual stress (at normal operating temperature and pressure) and full piping loads, conservatively including deadweight and seismic loading, are used in the SCC growth analysis.

(3) A lO-year inspection interval is assumed, based on EPRI BWRVIP-59-A [1]; however, a 40-year crack growth period is conservatively evaluated for both fatigue crack growth and IGSCC growth (for tensile stress regions in the DMW). No IGSCC crack growth is assumed for compressive stress intensity factor regions in the DMW. The initial crack shall not grow to exceed the design basis of the weld overlay during the inspection period. The acceptance criteria are in accordance with ASME Code,Section XI, IWB-3600 rules.

References:

1. BWRVIP-59-A: BWR Vessel and Internals Project, Evaluation of Crack Growth in BWR Nickel Base Austenitic Alloys in RPV Internals, EPRI, Palo Alto, CA: 2007. 1014874.

2. BWRVIP-219: BWR Vessel and Internals Project, Technical Basis for On-Line Noble Chem' Mitigation and Effectiveness Criteria for Inspection Relief, EPRI, Palo Alto, CA: 2009. lO19071.

3. NUREG/CR-6721, 'Effects of Alloy Chemistry, Cold Work, and Water Chemistry on Corrosion Fatigue and Stress Corrosion Cracking ofNickel Alloys and Welds,' Argonne National Laboratory, Argonne, IL, April 2001. to 1200536.401RO 2

Attachment 1 SI Response to Question No.1 ofRAI for Grand Gulf Nuclear Station Relief Request RR 17

~_.l..---I t--_...1..~---1_--I_X t

co: inner surfuce stress (x==O) a R

REQUIRED INPUTS:

t: wan thickness a: maxiIl1.lm. crack depth (a-x 5 O.8t) a) Full circumferential crack in a cylinder StressILoad IIlput Stress Coefficients ./

Coeffs. from Stress Tat,le ./

Stress T~ble x

(

/~ih-; ....,,* ..1

_I:

21-' ,

~

II Stress Intensity FaClors (1 D)

Stress Intensity FaClors (2D) cr(X) = C~'" C1x ... C2 X*'" C3 x3 Crack ITimensions: a c

./

./

~

Component Dimension>: t Rj I

Range: 0.05 <a,[<= 0.85 0.1 <= 3.!C <=1.0 0.1<=t'R<<=1.0 Source: [9]

b) Axial crack in a cylinder Figure 1. pc-Crack Flaw Models Used in Crack Growth Evaluation Attachment I to 1200536.401RO 3