ML021050239

Transmittal of Meeting Handout Materials for 04/10/2002 Meeting to Discuss Proposed Repairs and Modifications to the Reactor Pressure Vessel Head

ML021050239
Person / Time
Site:	Davis Besse
Issue date:	04/10/2002
From:	Sands S NRC/NRR/DLPM/LPD3
To:	NRC/NRR/DLPM/LPD3
References
Download: ML021050239 (32)

Text

NRC FORM 658 U.S. NUCLEAR REGULATORY COMMISSION (9-1999)

TRANSMITTAL OF MEETING HANDOUT MATERIALS FOR IMMEDIATE PLACEMENT IN THE PUBLIC DOMAIN This form is to be filled out (typed or hand-printed) by the person who announced the meeting (i.e., the person who issued the meeting notice). The completed form, and the attached copy of meeting handout materials, will be sent to the Document Control Desk on the same day of the meeting; under no circumstances will this be done later than the working day after the meeting.

Do not include proprietary materials.

DATE OF MEETING The attached document(s), which was/were handed out in this meeting, is/are to be placed 04/10/2002 in the public domain as soon as possible. The minutes of the meeting will be issued in the near future. Following are administrative details regarding this meeting:

Docket Number(s) 50-346 Plant/Facility Name Davis-Besse Nuclear Power Station TAC Number(s) (if available)

Reference Meeting Notice Accession # ML020880332 Purpose of Meeting (copy from meeting notice)

To discuss proposed repairs and modifications to the reactor pressure vessel head at the Davis-Besse Nudear Power Stalio:

NAME OF PERSON WHO ISSUED MEETING NOTICE ITLE Stephen Sands Project Manager OFFICE NRR DIVISION DLPM BRANCH PD 111-2 Distribution of this form and attachments:

Docket File/Central File PUBLIC NRC FORM 658 (9-1999)

PRINTED ON RECYCLED PAPER This form was designed using InForms

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Agenda Introduction - John Wood Inspection Results Mark McLaughlin Repair Concept Jim Powers Final Reactor Core Configuration Robb Borland FENOC

Meeting hOijective Present results of the Davis-Besse Nuclear Power Station reactor pressure vessel-head inspections and the.repair concept FENOC 3

N0 M I, 4:,.

Ins ectioii Results Davis-Besse shutdown for Refueling Outage February 16, 2002 Reactor Pressure Vessel Head (RPV) Inspections performed in response to NRC Bulletin 2001-01

• Performed ultrasonic (UT) examinations on all Control Rod Drive Mechanism Nozzles

• UT results independently verified by EPRI

• o0o

• Performed visual inspections of RPV head FENOC

,IF ý FuleliuS Framatome ANP Inc.

completed UT examination on all 69 CRDM nozzles using the under-head circumferential probe and subsequent confirmatory testing using the top-down UT on suspect nozzles 6

CIO Area of Degradation Nozzle with Axial Indication -0 Nozzle with Axial and Circumferential Indication - 0

FENOC Inspectiei esuMs Reactor Vessel Head and Service Structure Source: EPRIDEI Control Rod Ddve Spare Nozzle M~ 4Insulation f I Vessel Head Side view 7~ cotL S I

Inspection ReSultS Bolts Typical B&W Control Rod Drive Nozzle Low-Alloy Steel Reactor Vessel Head FENOC

Inspection Results

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Extent of Condition Investigation

• Remove Nozzles 2 and 11"

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• Liquid Penetrant Examination (PT) on bores

• Remove Nozzle 3 wastage area around and PT bore Area of Degradation FENOC 9

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In-iU-K FARTHEST LIMENSI 1 3/16 9 1/16 3/12/02 eanuls 0 -

11 2002.3. 16'S\\37 Reactor Head Degradation - Nozzle 3 11 FENOC cGO

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fRW Unnapt ZNoOZZLE 2 NOZZLE 6

/NOZZLE 11 Overview Repair will consist of two phases:

"* Installation of welded plugs in Nozzles 2 and 11

"* Restoration of removed wastage area around Nozzle 3 with a forged disk Three affected control rod drives to be relocated to spare nozzles NOZZLE 1 \\

FENOC S-ý 13 C04"

Repair Concuif Design Criteria

° Repair will meet design requirements of American Society of Mechanical Engineers (ASME) Boiler &

Pressure Vessel Code (BPVC)Section III

° Includes all normal." ff-normal and accident transient cycles and is design, rd for remaining licensed plant life Repairs will.-be performed b am consisting of personnel frornD BS raiatome ANP Inc., and Welding Services, Inc.

Third party design analysis by Structural Integrity Associates Mock-ups will be used to demonstrate effectiveness of cutting, welding and examination techniques FENOC 14

Repair ConcPit Applicable Codes Design code for the Reactor Vessel was ASME Section III, 1968 Edition, Summer 1968 Addenda Design code to be used for the repair is the ASME BPVC Section III, 1989 Edition ASME BPVC Secti XI, 1995 Edition, with 1996 Addendum is, governig: inservice inspection code for Non-destructive ex'aminations (NDE) of repair will be performed in accordance with Section III FENOC 15

Nozzles 2& 11 PLUG (MATERIAL GRADE ALLOY 690)

CROSS SECTION NOZZLES 2 RPV

& I HEAD 1

Repair Sequence

"* Machine and perform Liquid Penetrant (PT) examination of bore

"* Machine plug to match bore

• Insert and weld plug using remote machine Gas Tungsten Arc Welding Ambient Temperature Temper Bead and Alloy 52 Weld Filler Material

• Perform PT and Ultrasonic (UT) examination on completed weld 16 FENOC

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~y 690, §1B=564TLUN\\ SN06690 1

IFIS 40C Risk

1. 1. Azale3 IInfdIdcentlArea Repair Sequence

1. 11 PENETRATION PLUG

• Inspect walls using Liquid ERNiCrFe-7 (ALLOY 52)

SA 533CR R

B Penetrant Examination (PT)

//*°FORGING--NSERT T

Butter the bore surface using BUTTRGING INERTDT Ambient Temperature Temper Bead welding process Machine and after 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> hold inspect using PT and UT examination NiCrFe FORGING INSERTex m n to SB-564NUNSFN06690 (ALLOY 690)

Fit up and weld in forged disk

• Weld to be inspected using PT REPAIR CROSS SECTION and Radiographic (RT) examination FENOC 18

NozzleS u38dllsceutifres Repair Sequence

1. 11 PENETRATION PLUG

• Inspect walls using Liquid ERNirFeo7 (ALLOY 52)

SA-533 GR 8 Penetrant Examination (PT)

NFORGING INSERT TO e Buffer the bore surface using Ambient Temperature Temper Bead welding process

"" Machine and after 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br /> hold inspect using PT and UT examination NiCrFe FORGING INSERT

%SB-564 SN06690 (ALLOY 690) eN o

Fit up and weld in forged disk e Weld to be inspected using PT REPAIR CROSS SECTION and Radiographic (RT) examination FENOC 18

RepairfCniceut Confirmatory Action Letter - Repair Plan NRC Approvals per 10 CFR 50.55a Penetrations #2 & #11

- Approval to use Ambient Temperature Temper Bead Welding - Code Case N-63 8 Methodology (Consistent with those granted to other plants for CRDM Nozzle Repairs)

- Approvals include:

- Interpass Temperature Qualification (Section XI IWA-46 10 (b))

- Impact Testing toneet ASME BPVC Section III (Section XI IWA-4632 (b))

Penetration #3 - Weld Buttering

- Approval to use"AmbienT, mperamper Bead Welding - Code Case N-63 8 Methodology.

- Approvals include:

- 100 In2 Limitation (Section XI IWA-4631 (b))

- Interpass Temperature Qualification (Section XI IWA-46 10 (b))

- Preheat/Interpass Temperature Monitoring (Section XI IWA FENOC 19 S...............-,4610 (a))

RepairConlcept Post Repair and Inspection Testing

° Liquid Penetrant Examination Radiographic Examination Code Case N-416-1

- System leakage test at full temperature and pressure FENOC 20

"-LINN

Final BReactor Core Contiguraion Overview

"* Total number of control rod assemblies (CRAs) remains the same

"* Number of individual CRAs in each control rod group remains thew-same

• Original Cycle 14 fuel loading pattern maintained FENOC 22

Final Reactor Core Contiguration Proposed Changes

"* Three CRAs moved to new core positions using existing spare CRDM nozzles

"* Eight CRAs exchanged between two control rod groups to mainta.inappropriate core symmetry

"* Cycle 14 reload analysisredone by Framatome ANP for the new CRA pattern FENOC 23

<4 RP 0

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F E

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B A

2 Non-Rodded Locations 3

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Normal CRA Locations 7

E1 New CRA Locations 9

(Nozzles 15, 16,21) 1o tl Spare CRDM Nozzles 12 (HV = Head Vent) 13 14 APSR Locations SNCQ2 FENOC 2

CRA Relocation Group 1 Relocations Original Group 1 R

P 0

N M

L K

H G

F E

D C

B

-A New Group 1 R

P 0

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H G

F E

S2 3

4 5

6 7

8 9

"10 11 12 13 14

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3 D

C B

A 4

5 6

7 8

9 10 11 12 13 14 15 4

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CRA Rliocation Group 3 Relocations Original Group 3 R

P 0ON M L

K H

G F

E D

C B

A New Group 3 R

P0 N

M L

K H

G F

E D

C B

A 1

2 3

4 6

7 8

9 10 11 12 13 14 15 i i 4i--I 4

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1 2

3 4

5 6

7 8

9 10 11 12 13 14 15 FENOC 26 m 4-4-4-4--f i

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CIA Ielocafioo Group 6 Relocations Original Group 6 R

P ON M

L K

H G

F E

D C

B A

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0 1

2 3

4 5

6 8

9 10 11 12 13 14 15 New Group 6 R

PG0 N

M L

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G F

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A 1

2 3

4 5

6 7

8 9

10 11 12 13 14 15 FENOC 27 i

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CUA Relocation CRA Relocation Effect All CRA worths (total, group, stuck, ejected, dropped) well within those assumed in USAR safety analyses Rod insertion limits meet shutdown margin requirements FENOC 28

Final Reactor Core Contiguiration NRC approvals in accordance with Confirmatory Action Letter FENOC 29