ML061810027
| ML061810027 | |
| Person / Time | |
|---|---|
| Site: | Braidwood  |
| Issue date: | 06/29/2006 |
| From: | Polson K Exelon Generation Co, Exelon Nuclear |
| To: | Document Control Desk, Office of Nuclear Reactor Regulation |
| References | |
| BW060061, EA-03-009 | |
| Download: ML061810027 (29) | |
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Exelo Gere aton Company, LLC B~idv cdStatio June 29, 2006 BW060061 U. S. Nuclear Regulatory Commission ATTN: Document Control Desk Washington, DC 20555-0001 Braidwood Station, Unit 1 Facility Operating License No. NPF-72 NRC Docket No. SIN 50-456
Subject:
Braidwood Station, Unit 1, 60-Day Response to First Revised NRC Order EA-03-009, Issuance of First Revised NRC Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors
Reference:
Letter from NRC, Issuance of First Revised NRC Order (EA-03-009) Establishing Interim Inspection Requirements for Reactor Pressure Vessel Heads at Pressurized Water Reactors dated February 20, 2004 The purpose of this letter is to provide the results of examinations performed at F3raidwood Station, Unit 1, in accordance with the requirements of the referenced NRC Order.
During the Spring 2006 (Al Ri2) refueling outage, Braidwood Station, Unit 1, completed both the nonvisual volumetric nondestructive examination in accordance with Order Section IV. C. (5) (b) and the visual inspection to identify potential boric acid leaks from pressure-retaining components above the reactor pressure vessel (RPV) in accordance with Section IV, paragraph D of the Order.
Section IV. E. of the Order requires that the results of this examination be submitted to the NRC within 60 days after returning the plant to operation. The Braidwood Station, Unit 1, Spring 2006 refueling outage ended on May 3, 2006 and therefore the inspection results must be submitted by July 2, 2006.
U.S. Nuclear Regulatory Commission Page 2 June 29, 2006 Detailed reports describing the examination results are provided in two attachments to this letter. In summary, there were no indications of cracking in any of the penetrations, and there was no evidence of a leakage path along the reactor vessel head penetration shrink-fit regions. Note that ten penetrations (i.e., numbers 42, 49, 54, 63, 65, 66, 71, 72, 77, and 78), which had limited inspection coverage due to the physical length of the penetration tube, will require a relaxation request.
The examination of one penetration (number 74) was limited circumferentially (approximately 60 degrees) due to the surface condition of the inside diameter surface of the penetration tube causing probe lift-off. EGC will also submit a relaxation request to address the lack of coverage for this penetration.
The visual inspection performed in accordance with Section IV, paragraph D of the Order did not identify any boric acid leaks from pressure-retaining components above the RPV head or any boron deposits on the mirror insulation above the RPV head.
Should you have any questions or desire additional information regarding this letter, please contact Mr. Dale Ambler, Regulatory Assurance Manager, at (815) 417-2800.
Respectfully, Keith J. Kolson Site Vice President Braidwood Nuclear Generating Station Attachments:
- 1) Al Ri 2 Reactor Pressure Vessel Head Visual Inspection Results 2)
Braidwood Unit 1 Al R12 Reactor Vessel Head Penetration Examination/Westinghouse Report WDI-PJF-1303195-FSR-001
ATTACHMENT 1 Al Ri 2 Reactor Vessel Head Visual Inspection Results
ATTACHMENT 1 Al Ri 2 Reactor Vessel Head Visual Inspection Results The visual examinations performed on the Braidwood Unit 1 reactor pressure vessel (RPV) head during the Spring 2006 refueling outage are contained in the first revised NRC Order EA-03-009 (Order),Section IV. D, which states:
During each refueling outage, visual inspections shall be performed to identify potential boric acid leaks from pressure-retaining components above the RPV head. For anyplant with boron deposits on the surface ofthe RPV head or related insulation, discoveredeither during the inspections required by this Order or otherwise and regardless of the source of the deposit, before returning the plant to operation the Licensee shall perform inspections of the affected RPV head surface and penetrations appropriate to the conditions found to verify the integrity of the affected area and penetrations.
VISUAL EXAMINATION RESULTS During the Braidwood Station Unit 1 Spring 2006 refueling outage (Al R12), walk downs were performed with the unit in Mode 3, shortly after reactor shutdown. The walk downs were performed in accordance with the requirements of the Braidwood Station Boric Acid Corrosion Control program and paragraph IV. D of the Order. No evidence of reactor coolant leakage or degradation associated with boric acid leakage was noted during these walk downs.
ATTACHMENT 2 Westinghouse Report WDI-PJF-l 303195-FSR-00i Braidwood Unit 1 Al Ri 2 Reactor Vessel Head Penetration Examination April 2006
ATTACHMENT 2 Braidwood Unit 1 A1R12 Reactor Vessel Head Penetration Examination The nonvisual NDE examinations (ultrasonic and eddy current) performed on the Braidwood Station Unit 1 reactor pressure vessel (RPV) head during the Spring 2006 refueling outage are contained in the first revised NRC Order EA-03-009 (Order),Section IV, paragraphs C.(3) and C.(5)(b).
Paragraph lV.C.(3) of the Order states:
...The requirements ofparagraph IV.C.(5)(b) must be completedat least once prior to February 11, 2008, and thereafter, at least every 4 refueling outages or every 7years, whichever occurs first.
Paragraph IV.C.(5)(b) of the Order states:
Foreach penetration, perform a nonvisual NDE in accordance with either(I), (ii) or (iii):
(i)
Ultrasonic testing of the RPV head penetration nozzle volume (i.e.,
nozzle base material) from 2 inches above the highest point of the root ofthe J-groove weld (on a horizontal planeperpendicular to the nozzle axis) to 2 inches below the lowest point at the toe of the J-groove weld on a horizontal plane perpendicular to the nozzle axis (or the bottom of the nozzle if less than 2 inches [see Figure IV-1J);
OR from 2inches above the highest point of the root ofthe J-groove weld(on a horizontal plane perpendicular to the nozzle axis) to 1.0-inch below the lowestpoint at the toe of the J-groove weld (on a horizontal plane perpendicularto the nozzle axis) and including all RPVhead penetration nozzle surfaces below the J-groove weld that have an operating stress level (including all residual andnormal operation stresses) of20 ksi tension and greater (see Figure IV-2).
In addition, an assessmentshall be made to determine ifleakage has occurred into the annulus between the RPV head penetration nozzle and the RPV head low-alloy steel.
(ii)
Eddy current testing or dye penetrant testing ofthe entire wetted surface of the J-groove weld and the wetted surface ofthe RPVhead penetration nozzle base material from atleast two inches above the highestpoint of the root ofthe J-groove weld (on a horizontal plane perpendicular to the nozzle axis) to 2 inches below the lowest point at the toe ofthe J-groove weld on a horizontal plane perpendicular to the nozzle axis (or the bottom of the nozzle ifless than 2inches
[see Figure lV-3J; OR from 2 inches above the highest point of the root of the J-groove weld (on a horizontal planeperpendicular to the nozzle axis) to 10-inch below the lowestpoint at the toe ofthe J-groove weld (on a horizontal plane perpendicular to the nozzle axis) and including all RPV head penetration nozzle surfaces below the J-groove weld that have an operating stress level(including all residual and normal operation stresses) of 20 ksi tension and greater (see Figure lV-4).
ATTACHMENT 2 Braidwood Unit 1 Al Ri 2 Reactor Vessel Head Penetration Examination (iii)
A combination of (i) and (ii) to cover equivalent volumes, surfaces and leakpaths ofthe RPV head penetration nozzle base material and J-groove weld as describedin (i) and (ii). Substitution of aportion of a volumetric exam on a nozzle with a surface examination may be performed with the following requirements:
1.
On nozzle material below the J-groove weld, both the outside diameter and inside diameter surfaces of the nozzle must be examined.
2.
On nozzle material above the J-groove weld, surface examination of the outside diameter surface ofthe nozzle is permitted provided a surface examination ofthe J-groove weld is also performed.
The ultrasonic and eddy current examinations for Braidwood Unit 1 required by the Order were performed by Westinghouse. There were no indications of cracking in any of the penetrations and there was no evidence of a leakage path along the reactor vessel head penetration shrink-fit regions. Detailed discussion of the examinations performed and results are contained in Westinghouse report WDI-PJF-1 303195-FSR-001 that is attached. The referenced Examination Procedures, Technical Justifications, Calibration and Examination Data, and Volumes 2 and 3 are not included in the attached report.
The following terms are referenced in the Westinghouse report:
Acronym Term BBP B and B Prime BWP Backwall Perturbation CBH Cleared By History IPA Indication Profile Analysis LOL Lack of Lateral Wave MAI Multiple Axial Indications MCI Multiple Circumferential Indications NDD No Detectable Defect P11 Penetration Tube Indication PVI Permeability Variation Indication SAl Single Axial Indication SCI Single Circumferential Indication SGI Surface Geometry Indication SSS Shallow Surface Scratch WVI Weld Volume Indication
Braidwood Unit 1 Page 1 of 22
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Reactor Vessel Head Penetration Examination Braidwood Unit 1 Al Ri 2 Reactor Vessel Head Penetration Examination April 2006 Final NDE Report WDI-PJF-1 3031 95-FSR-001 Westinghouse Electric Company Nuclear Services Waltz Mill Service Center P.O. Box 158 Madison, Pennsylvania 15663 USA
Braidwood Unit 1
ØWestinghouse Reactor Vessel Head Penetration Examination Page 2 of 22 Table of Contents Volume 1 Examination Summary 1.0 Introduction 2.0 Scope of Work 2.1 CRDM Penetration Tube Ultrasonic and Supplementary Eddy Current Examinations from the Tube ID 2.1.1 CRDM Penetration Tube 7010 Open Housing Scanner Examinations 2.1.2 CRDM Penetration Tube Gapscanner Trinity Probe Examinations 2.2 Eddy Current Wetted Surface Examinations 2.2.1 Vent Line Tube ID and J-WeId Eddy Current Examinations 3.0 Examination Results 3.1 CRDM Penetration Tube Ultrasonic and Supplementary Eddy Current Examinations from the Tube ID 3.2 Eddy Current Wetted Surface Examinations 3.2.1 Vent Line Tube and J-Weld Eddy Current Examinations 4.0 Examination Coverage 5.0 Discussion of Results 6.0 References Appendix A:
Braidwood Unit 1 RVHP Examination Coverage Summary Examination Procedures Technical Justifications Calibrations Data Electronic Copies Examination Data 2 Disks
Braidwood Unit 1
- WB5TIn~hOUSe Reactor Vessel Head Penetration Examination Page 3 of 22 Volume 2 Personnel Certifications Equipment Certifications Volume 3 Examination Results Ultrasonic and Eddy Current Examinations SørinQ 2006 1.
Braidwood Unit 1 Al R12 Reactor Vessel Head Inspection, Calibrations/Data Sheets, OHS & Gapscanner Reports, Penetrations #1
- 78 Vent Line Tube and J-Qroove Weld Eddy Current Examinations SDrinp 2006
- 1. Al Ri 2 Braidwood Unit 1 Vent Line ID and Vent Line J-Weld Inspection Raw Data and Results
Swesflnghouse Braidwood Unit 1 Reactor Vessel Head Penetration Examination Page 4 of 22
1.0 INTRODUCTION
During the Braidwood Unit 1 Al Rl2 outage in the spring of 2006, Westinghouse performed nondestructive examinations (NDE) of the seventy-eight control rod drive mechanism (CRDM) penetration tubes and the vent line in the reactor vessel head.
The purpose of the examination program was to identify evidence of primary water stress corrosion cracking (PWSCC) that might be present on the outside diameter (OD) and inside diameter (ID) surfaces of the head penetration tubes and to assess whether leakage might have occurred into the annulus at the tube-to-head interface.
Examinations were performed using procedures and techniques demonstrated through the EPRI/MRP protocol [l} and/or Westinghouse internal demonstration programs, and applied consistent with the requirements of the February 20, 2004, first revision to USNRC Order EA-03-009, Establishing Interim Inspection Requirements for Reactor Vessel Heads at Pressurized Water Reactors [2}.
The Braidwood reactor vessel head is a Westinghouse design. The head was manufactured by Babcock & Wilcox (B&W) in Mt. Vernon, IN. The alloy 600 penetration tubes are shrunk fit into the reactor vessel head and attached with alloy 182/82 partial penetration J-groove welds. The vent line is an alloy 600 tube attached to the reactor vessel head with an alloy 182/82 partial penetration weld.
The penetration tubes in the Braidwood reactor vessel head are machined from heats of material supplied by B&W Tubular Products. The penetration tubes measure 4.0 on the OD and have an ID dimension of 2.75.
The vent line, 1 schedule 160, has a nominal wall thickness of 0.25 and a nominal OD of 1.315.
A summary of the heats of material in the head is provided in Table 1-1.
Table 1-1: Braidwood Unit 1 RV Head Penetration Material Heats Material Supplier Heat #
B&W 90877 B&W 90704 B&W 90878 B&W 80059 B&W 91112 B&W 90862 B&W 90745
~~WeshflghOUSe Braidwood Unit 1 Reactor Vessel Head Penetration Examination Page 5 of 22 There are a variety of configurations for the 78 penetration tubes, each configuration requiring special consideration for examination. The penetration tube configurations are as follows:
Fifty-three penetration tubes with thermal sleeves installed Two heat junction thermocouple locations with modified thermal sleeves installed Twenty-three penetration locations without thermal sleeves The Braidwood reactor vessel head is in the low susceptibility category as defined in the first Revision to USNRC Order EA-03-009.
Paragraph IV.C (5) of the first Revision to USNRC Order EA-03-009 specifies:
a) Bare metal visual examination of 100% of the RPVhead surface (including 360° around each RPVhead penetration nozzle), and b) For each penetration, perform a nonvisual NDE in accordance with either i, ii or iii:
i.
Ultrasonic testing of each RPVhead penetration nozzle volume (i.e., nozzle base material) from two (2) inches above the highestpoint of the root of the J-groove weld to 2 inches below the lowest point at the toe of the J-groove weld on a horizontal plane perpendicular to the nozzle axis; OR from 2 inches above the highestpoint of the root of the J-groove weld to 1 inch below the lowest point at the toe of the J-groove weld and including all RPV head penetration nozzle surfaces below the J-groove weld that have an operating stress level of 20 ksi tension and greater. In addition, an assessment shallbe made to determine if leakage has occurred into the annulus between the RPV head penetration nozzle and the RPV head low alloy steel.
ii.
Eddy currentor dye penetrant testing of the entire wetted surface of the J-groove weld and the wetted surface of the RPV headpenetration nozzle base material from at least 2 inches above the highest point of the root of the J-groove weld to 2 inches below the lowest point at the toe of the J-groove weld on a horizontal plane perpendicular to the nozzle axis; OR from 2inches above the highest point of the root of the J-groove weld to 1 inch below the lowest point at the toe of the J-groove weld and including all RPVhead penetration nozzle surfaces below the J-groove weld that have an operating stress level of 20 ksi tension and greater.
ill.
A combination of (i) and (ii) to cover equivalent volumes, surfaces and leak paths of the RPV head penetration nozzle base material and J-groove welds described in (i) and (ii).
Braidwood Unit 1
~~weshnghousa Page 6 of 22 Reactor Vessel Head Penetration Examination For plants in the low susceptibility category, inspections specified in paragraph IV.C (5)
(a) are required every third refueling outage or every 5 years, whichever comes first. If an inspection meeting the requirements of paragraph IV.C (5) (a) was not performed during the last refueling outage prior to February ii, 2003, an inspection meeting those requirements is required within the first 2 refueling outages after February ii, 2003. An inspection meeting the requirements of paragraph IV.C (5) (b) is required at least once prior to February ii, 2008 and at least every 4 refueling outages or every 7 years, whichever comes first, thereafter.
The examination program selected for Braidwood during the Al Ri 2 outage included ultrasonic examinations of the 78 CRDM penetration nozzles with leakage assessment in accordance with paragraph IV.C (5) (b) (i) of the Revised NRC Order.
For the vent line, the wetted surface examination option using eddy current techniques was selected in accordance with Section IV.C (5) (b) (ii) of the Revised NRC Order.
Stress distribution curves were developed in advance of the examination which identified the hoop stress distributions below the attachment welds on the OD surfaces of penetration tubes. A fracture analysis was performed and the results were presented in the form of flaw tolerance charts for both surface and through wall flaws. If indications of PWSCC had been identified, the charts were available to determine the allowable safe operation service life [3].
A contingency plan was in place to address geometric conditions at penetration locations where access of the Trinity blade probes in the penetration/tube annulus might be limited. The contingency plan included equipment and procedures necessary to perform wetted surface examinations in accordance with Section IV.C (5) (b) (ii) of the Revised Order.
The following Westinghouse field service procedures and field change notices (FCN5) were approved for use at Braidwood Unit 1.
WDI-ET-002, Rev. 7 Eddy Current Inspection of J-Groove Welds in Vessel Head Penetrations WDI-ET-003, Rev. 9 IntraSpect Eddy Current Imaging Procedure for Inspection of Reactor Vessel Head Penetrations WDI-ET-004, Rev. 10 IntraSpect Eddy Current Analysis Guidelines WDI-ET-008, Rev. 7 lntraSpect Eddy Current Imaging Procedure for Inspection of Reactor Vessel Head Penetrations With Gap Scanner WDI-UT-0i0, Rev. 12 - IntraSpect Ultrasonic Procedure for Inspection of Reactor Vessel Head Penetrations, Time of Flight Ultrasonic & Longitudinal Wave & Shear Wave
~~We~hn~hOUSe Braidwood Unit 1 Reactor Vessel Head Penetration Examination Page 7 of 22 WDI-UT-0i3, Rev. 10 IntraSpect UT Analysis Guidelines WDI-STD-iol, Rev. 5 RVHI Vent Tube J-Weld Eddy Current Examination WDI-STD-i 14, Rev. 4 RVHI Vent Tube ID & CS Wastage Eddy Current Examination WDI-STD-l5i, Rev.i-Reactor Vessel Head Inspection for Byron Units i&2 CAE/CBE and Braidwood Units 1&2 CCE/CDE WCAL-002, Rev. 7 Pulser/Receiver Linearity Procedure 2.0 SCOPE OF WORK The reactor vessel head penetration examination scope at Braidwood included all seventy-eight CRDM penetration tubes and the vent line.
The examination methodology selected for each penetration was dependent upon the penetration tube configuration and penetration-specific conditions.
1.
Twenty-three penetration tubes without thermal sleeves were examined from the ID using the Westinghouse 7010 Open Housing Scanner (OHS).
2.
Fifty-five penetration tubes; fifty-three with thermal sleeves and two heat junction thermocouple locations, were examined from the ID using the Westinghouse Gapscanner and Trinity blade probes.
3.
The vent line tube eddy current examination was performed with an array of 16 pIus-Point probes and a low frequency bobbin coil. The vent line J-groove weld eddy current examination was performed with an array of 28 plus~Pointcoils.
The delivery system used for the CRDM examinations was the Westinghouse DERI 700 manipulator.
The DERI 700 is a multi-purpose robot that can access all head penetrations and provides a common platform for all CRDM examination end effectors. The manipulator consists of a central leg, mounted on a carriage, which in turn is mounted onto a guide rail. The manipulator arm, with elbow and removable wrist, is mounted onto the carriage, which travels vertically along the manipulator leg.
The DERI 700 was used to deliver 1) the Westinghouse 7010 Open Housing Scanner for ultrasonic and 2) supplementary eddy current examinations of open penetration locations and the Westinghouse Gapscanner end effector for Trinity probe examinations of penetration locations containing thermal sleeves.
S Reactor Vessel Head P:netration Examination Page 8 of 22 The Westinghouse 7010 Open Housing Scanner delivers an examination wand containing ultrasonic and eddy current probes to the ID surface of open reactor vessel head penetrations. The scanning motion is in a vertical direction moving from a specified height above the weld, in this case at least 2.0, to the bottom of each penetration. The probe is indexed in the circumferential direction. With the open housing scanner, multiple examination probes are delivered simultaneously. These include time-of-flight diffraction ultrasonic (TOFD-UT) probes oriented in the axial and circumferential directions, 0°ultrasonic probe to identify variations in the penetration tube-to-reactor vessel head shrink fit area that might indicate a leak path in the annulus between the tube and the head, and a supplementary eddy current probe for identification of circumferential and axial degradation on the ID surfaces of the penetration tubes The Gapscanner end effector delivers Trinity blade probes into the annulus between the ID surface of the head penetration tube and the OD surface of the thermal sleeve. The typical annulus size is 0.125. The Trinity blade probes include a TOFD UT transducer pair for detection of axial and circumferential degradation, a 0°ultrasonic transducer to identify variations in the penetration tube-to-reactor vessel head shrink fit area that might indicate a leak path in the annulus between the tube and the head, and a supplementary crosswound eddy current coil. The scanning motion is in a vertical direction moving from a specified height above the weld, in this case at least 2.0, to the bottom of each penetration. The probes are indexed in the circumferential direction.
2.1 CRDM Penetration Tube Ultrasonic and Supplementary Eddy Current Examinations from the Tube ID All seventy-eight penetration tubes were ultrasonically examined from the tube ID surface in accordance with Section lV.C (5) (b) (i) of the Revised NRC Order. Methods for leakage assessment were incorporated into these examinations.
2.1 ~.1CRDM Penetration Tube 7010 Open Housing Scanner Examinations 7010 Open Housing Scanner examinations were conducted on twenty-three reactor vessel head penetrations without thermal sleeves.
Examinations of these vessel head penetrations included:
1.
TOFD ultrasonic techniques in accordance with WDI-UT-01 0, Rev. 12
IntraSpect Ultrasonic Procedure for Inspection of Reactor Vessel Head Penetrations, Time of Flight Ultrasonic Longitudinal Wave & Shear Wave, 2.
straight beam ultrasonic techniques to identify possible leak paths in the shrink fit region between the head penetrations and the reactor vessel head, also in accordance with WDI-UT-0i0, Rev. 12, and
- westinghouse Braidwood Unit 1 Reactor Vessel Head Penetration Examination Page 9 of 22 3.
supplementary eddy current examinations on the penetration tube ID surfaces in accordance with and WDI-ET-003, Rev. 9 - IntraSpect Eddy Current Imaging Procedure for Inspection of Reactor Vessel Head Penetrations.
2.1.2 CROM Penetration Tube Gapscanner Trinity Probe Examinations Examinations were performed with the Gapscanner end effector and Trinity probes on fifty-five penetration tubes; fifty-three with thermal sleeves and two heat junction thermocouple locations, from the penetration ID surfaces.
Examinations of these vessel head penetrations included:
1.
TOFD ultrasonic techniques in accordance with WDI-UT-0l 0, Rev. 12
IntraSpect Ultrasonic Procedure for Inspection of Reactor Vessel Head Penetrations, Time of Flight Ultrasonic Longitudinal Wave & Shear Wave, 2.
straight beam ultrasonic techniques to identify possible leak paths in the shrink fit region between the head penetrations and the reactor vessel head, also in accordance with WDI-UT-010, Rev. 12, and 3.
supplementary eddy current examinations in accordance with and WDI-ET-008, Rev. 7
- IntraSpect Eddy Current Imaging Procedure for Inspection of Reactor Vessel Head Penetrations.
2.2 Eddy Current Wetted Surface Examinations Wetted surface examinations were conducted on the vent line and the vent line weld using eddy current techniques in accordance with Section IV.C (5) (b) (ii) of the Revised NRC Order.
2.2.1 Vent Line Tube ID and J-WeId Eddy Current Examinations The vent line tube eddy current examination was performed with an array of 16 plus-Point probes and a low frequency bobbin coil in accordance with WDI-STD-1 14, Rev.
4 - Head Vent ID Eddy Current Inspection. The vent line J-groove weld eddy current examination was performed with an array of 28 plus-Point coils in accordance with WDI-STD-l01, Rev. 5, RVHI Vent Tube J-Weld Eddy Current Examination.
c~wesr~nghouse Braidwood Unit 1 Reactor Vessel Head Penetration Examination Page 10 of 22 3.0 EXAMINATION RESULTS 3.1 CRDM Penetration Tube Ultrasonic and Supplementary Eddy Current Examinations from the Tube ID Table 3-1 provides a summary of results from the 7010 Open Housing Scanner reactor vessel head penetration nondestructive examinations.
Table 3-1: Open Housing Scanner Examination Results 11 NDD NDD 13 PTI/BBP/NDD PTI/BBP/NDD NDD NDD 18 NDD NDD NDD NDD 19 PTI/BBP/NDD PTI/BBP/NDD NDD NDD 20 NDD NDD NDD NDD 21 NDD NDD NDD NDD 22 NDD NDD NDD SSS/NDD 23 NDD NDD NDD NDD 24 NDD NDD NDD NDD 25 NDD NDD NDD NDD 26 NDD NDD NDD NDD 27 NDD NDD NDD NOD 28 NDD NDD NDD NDD 29 NDD NDD NDD NDD 62 NDD NDD NDD NDD 64 NDD NDD NDD NDD 74 LCG/NDD LCG/NDD NOD SSI/NDD 75 NDD NDD NDD NDD 76 NDD NDD NDD NOD 77 NDD NDD NDD NDD 78 NDD NOD NOD NOD Legend BBP: Band B Prime IPA: Indication ProfileAnalysis NDD: No Detectable Defect PT!: Penetration Tube Indication LCG: Loss of Contact due to Geometry SSI: Surface Scratch Indication SSS: Shallow Surface Scratch No detectable degradation characteristic of PWSCC was reported in any of the penetrations examined with the 7010 Open Housing Scanner. There was no evidence of leakage in the annulus between the penetration nozzles and the reactor vessel head.
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Braidwood Unit 1 Reactor Vessel Head Penetration Examination Page 11 of 22 Table 3-2 provides a summary of results from Gapscanner examinations performed with Trinity Probes.
Table 3-2: Trinity Probe Examination Results Penetration #
PCS24 TOFD 00 Leak Path Supplementary Tube ID ECT 1
NDD NDD NDD 2
NDD NDD NDD 3
NDD NOD SCI/NOD 4
PTI/BBP/NDD NDD NDD 5
NDD NDD NDD 6
NDD NOD SCI/NOD 7
PTI/BBP/NDD NDD NOD PTI/IPA/LCG/NDD LCG/NDD 34 NOD NDD NOD 35 PTI/IPA/NDD NDD NOD 36 NOD NOD NOD 37 NOD NOD NOD 38 PTI/IPA/NOD NDD NDD 39 NOD NOD NOD 40 NDO NOD NOD 41 NOD NOD NOD
- west~nghouse Braidwood Unit 1 Reactor Vessel Head Penetration Examination Page 12 of 22 Penetration #
PCSZ4 lOFt) 0° Leak Path Supplementary Tube 10 ECT 42 NOD NDO NOD 43 NOD NDD NOD 44 NOD NDD NOD 45 PTIIIPNNDD NOD NOD 46 PTIIBBP/NDD NOD NOD 47 PTI/1PAJNDD NOD NOD 48 NOD NDO SGI/NDD 49 PTI/BBP/NOD NOD NOD 50 NOD NDD NDO 51 PTI/BBP/NOD NOD NDD 52 PTI/BBP/IPA/NDD NOD NDD 53 NDD NDD NDD 54 PTI/BBP/NOD NOD NDD 55 PTI/BBP/NDD NDO NDD 56 NOD NDD NDD 57 PTI/1PNNDD NOD NDD 58 NOD NDD NDD 59 NOD NDD NDD 60 LCGINDD NOD NOD 61 NOD NDD NOD 62 63 NOD NDD NOD 64 65 NOD NDO CCG/NDD 66 PTI/IPA/NDD NDD NOD NOD 69 NOD NDO 70 NOD NOD NDD NDO P
71 LCG: Loss of Contactdue to Geometry CCG: Craze Cracking Geometry SO!: Surface Geometry Indication 67 68 NOD 72 NDD NOD_________
NOD F
NOD NOD NDD NDD I
P Legend BBP: B and B Prime IPA: Indication ProfileAnalysis NDD: No Detectable Defect PT!: Penetration Tube Indication
%5westinghouse Braidwood Unit 1 Reactor Vessel Head Penetration Examination Page 13 of 22 Indicates Penetrations Examined With the Open Housing Scanner No detectable degradation characteristic of PWSCC was reported in any of the penetration tubes examined with the Trinity Probes. There was no evidence of leakage in the annulus between the penetration nozzles and the reactor vessel head.
3.2 Eddy Current Wetted Surface Examinations 3.2.2 Vent Line Tube and J-WeId Eddy Current Examinations Results of the eddy current examinations of the vent line and vent line J-groove weld are summarized in Table 3-4.
Table 3-4 Vent Tube and J-Groove Weld Eddy Current Results Penetration #.
ArrayECT Results Vent Line Weld NOD Vent Line Tube NOD Legend NDD: No Detectable Degradation No detectable degradation characteristic of PWSCC was identified during the eddy current examinations of the vent line J-groove weld and the vent line tube ID surface.
4.0 EXAMINATION COVERAGE 4.1 Penetration Tube Configuration and Examination Summary The configuration of a sleeved Braidwood CRDM penetration tube is illustrated in Figure 4-1.
This figure represents the tube-to-head geometry at the penetration 00 azimuth, or downhill side of the tube. The bottom ends of all penetration tubes are threaded on the OD surface and have a chamfer on the ID surface. The threads extend from the bottom of the tube to an elevation of approximately 1.00 where a thread relief is machined. The top of the thread relief is 1.13 above the bottom of the tube. The distance from the top of the thread relief to the bottom of the fillet of the J-groove weld, identified as A, varies based on location of the penetration in the head. These distances are generally longer for penetrations at inboard locations and become progressively shorter for penetrations located further away from the center of the head. The ID surface chamfers are machined at a 20°angle from the bottom of the tube to an elevation of 0.76.
5 Westinghouse Braidwood Unit Reactor Vessel Head Penetration Examination Page 14 of 22 Figure 4-1: Illustration of Axially Oriented TOFD Examination Coverage on Braidwood Unit 1 Penetration Geometry at 00 (Downhill Side) 4.2 Ultrasonic Testing Coverage in Accordance With Section IV.C (5) (b) (i) of the Revised NRC Order The ultrasonic method demonstrated through the EPRI/MRP Protocol for detection of circumferential and axial degradation on the OD and ID surfaces of CRDM penetration tubes is the time-of-flight diffraction (TOFD) technique. The TOFD technique is a pitch/catch ultrasonic method, where longitudinal waves are transmitted into the tube at an angle by a transmitter (T) and reflects off of the backside of the tube to a receiver (R),
as shown in path 1-2 in Figure 4-1. A lateral wave also travels on the tube ID surface between the transmitter and receiver as shown in path 3. The transmitting and receiving elements are mounted on a shoe with a probe center spacing of 0.925. ID TOFD coverage is provided by the lateral wave to the elevation of the chamfer the tube on the ID surface. With an axially oriented TOFD transducer pair in the Trinity Probe, OD coverage becomes completely effective at an elevation just above the top of the thread relief.
Carbon Steel Buttering Thermal Sleeve Stainless Steel A
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Braidwood Unit 1 Reactor Vessel Head Penetration Examination Page 15 of 22 The presence of the thread relief results in a slight masking of the ultrasound to the OD surface to an elevation conservatively estimated at 0.20 above the thread relief. In this area, however, OD initiated degradation would be detected once the depth of the degradation exceeded the depth of the masked area. With a circumferentially oriented TOFD transducer pair, included in the Open Housing Scanner, OD coverage is extended to the elevation of the top of the chamfer, approximately 0.76 above the bottom of the tube. In the threaded region, cracks extending deeper than the threads will be detected.
Examination coverage on the ID surfaces of the fifty-five penetration tubes examined with Trinity Probes and twenty-three penetration tubes examined with the Open Housing Scanner extended from the top of the chamfer in each tube to at least 2.0 above the uppermost elevation of the weld. The extent of coverage was verified for each penetration by 1) confirmation that tube entry signals at the elevation of the chamfer were evident in the eddy current and ultrasonic data, and 2) direct measurements from the TOFD UT C-scans which demonstrated scan coverage elevations were in excess of 2.0 above the uppermost elevation of each weld.
Examination coverage on the OD surfaces of the twenty-three penetration tubes examined with the Open Housing Scanner extended from the top of the chamfer in each tube to at least 2.0 above the uppermost elevation of the weld. Forthose tubes examined with Trinity Probes OD coverage extended from just above the elevation of the thread relief to at least 2.0 above the welds. The extent of coverage was verified for each examination of each penetration by 1) confirmation that TOFD responses were evident from the thread relief and 2) direct measurements from the TOFD UT C-scans which demonstrated scan coverage elevations were in excess of 2.0 above the uppermost elevation of each weld.
Examination coverage measured for each penetration location during the spring 2006 examination program is provided in Appendix A.
5.0 DISCUSSION OF RESULTS Penetration tube ultrasonic examination data were analyzed in accordance with WDI-UT-Ol 3, Rev. 10 IntraSpect UT Analysis Guidelines.
Eddy current data were analyzed in accordance with WDI-ET-004, Rev. 10 lntraSpect Eddy Current Analysis Guidelines Inspection of Reactor Vessel Head Penetrations. The screening and resolution process for ID indications is summarized in the logic chart in Figure 5-1 and the process for OD indications is summarized in the logic charts in Figures 5-2 and 5.3.
Data sheets and printouts of the results of each examination performed on each penetration are found in Volume 3.
Results from the TOFD ultrasonic and eddy current examinations of the seventy-eight CRDM penetrations and head vent line identified no indications characteristic of PWSCC.
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Braidwood Unit 1 Reactor Vessel Head Penetration Examination Page 16 of 22 Figure 5 Penetration Tube ID Indication Screening ET/UT:
ID INDICATION SCREENING SAl, SCI MAI, MCI
+/-~~LOLYES~SGIORPVI/NDD
<006W/OHS
<0.04 W/ SWORD TOFD TIP SIGNA YES
~NO REPORT DEPTH, 0.06 < d < 0.16 LENGTH REPORT FOR DISPOSITION
- RESCAN, OPTIONAL RESCAN HIGHER DENSITY W/ PCS 18 OR PCS 10 REPEAT REPEAT ANALYSIS STEPS ANALYSIS STEPS
Braidwood Unit 1
~Swestin~house Reactor Vessel Head Penetration Examination Page 17 of 2 Figure 5 Penetration Tube OD Indication Screening Within Weld Zone UT:
OD INDICATION SCREENING WITHIN WELD ZONE YES NO INDICATIONS VISIBLE IN PRIOR INSPECTION DATA GROWTH L> +0.2 D> + 0.06
Westinghouse Braidwood Unit 1 Reactor Vessel Head Penetration Examination Page 18 of 22 Figure 5 Penetration Tube OD Indication Screening Above or Below Weld Zone UT: OD INDICATION SCREENING ABOVE/BELOW WELD ZONE YES INDICATIONS VISIBLE IN PRIOR INSPECTION DATA SURFACE CONNECTIVITY TESTS RESULTS NO NO NDD
T
~ 5 Westinghouse Braidwood Unit 1 Reactor Vessel Head Penetration Examination Page 19 of 22
6.0 REFERENCES
[1] EPRI/MRP89 Technical Report, Materials Reliability Program: Demonstrations of Vendor Equipment and Procedures for the Inspection of Control Rod Drive Mechanism Head Penetrations (MRP-89), EPRI, Palo Alto, CA: July, 2003.
[2] USNRC Letter EA-03-009, Issuance of First Revised NRC Order (EA-03-009)
Establishing Interim Inspection Requirements for Reactor Vessel Heads at Pressurized Water Reactors, February 20, 2004.
[3] WCAP-16394-P, Rev. 0, Structural Integrity Evaluation of Reactor Vessel Head Penetrations to Support Continued Operation: Byron and Braidwood Units 1 and 2, Westinghouse Electric Company LLC, February 2005.
5Westinghouse Braidwood Unit 1 Reactor Vessel Head Penetration Examination Page 20 of 22 Appendix A: Braidwood Unit 1 RVHI Exam Coverage Summary Table A lists the coverage achieved above and below the J-Groove welds on the OD surfaces of each penetration tube location. This table incorporates the procedure, WDI-UT-01 3, Rev.10.
At all 78 penetration locations, coverage above the welds exceeded 2.0 inches above the highest point of the root of the J-Groove weld.
Coverage extends at least 1.0 below the lowest point at the toe of the J-Groove welds at all penetrations except # 42, 49, 54, 63, 65, 66, 71, 72, 77, and 78.
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Braidwood Unit 1 Reactor Vessel Head Penetration Examination Page 21 of 22 Pen #
Coverage Below the Weld
@0° 1.0 Below Weld @ 0° Coverage Above the Weld
@1800 2.0 Above Weld © 180° Measured in inches Y or N Measured in inches YorN 29 1.16 Y
4.24 Y
30 1.24 Y
4.16 Y
31 1.48 Y
4.08 Y
32 1.36 Y
4.04 Y
33 1.48 Y
3.96 Y
34 1.28 Y
4.24 Y
35 1.28 Y
3.80 Y
36 1.20 Y
4.04 Y
37 1.08 Y
4.72 Y
38 1.24 Y
4.00 Y
39 1.64 Y
3.72 Y
40 1.32 Y
4.56 Y
41 1.12 Y
4.72 Y
42 0.92 N
4.08 Y
43 1.20 Y
4.36 Y
44 1.24 Y
3.76 Y
45 1.60 Y
4.04 Y
46 1.16 Y
4.08 Y
47 1.36 Y
4.48 Y
48 1.04 Y
4.84 Y
49 0.76 N
4.32 Y
50 1.04 Y
4.32 Y
51 1.40 Y
3.92 Y
52 53 1.44 1.24 Y
Y 4.00 4.04 Y
Y 54 0.92 N
4.76 Y
55 1.20 Y
4.00 Y
56 1.44 Y
3.84 Y
57 1.16 Y
3.80 Y
58 1.24 Y
4.12 Y
59 1.04 Y
4.28 Y
60 1.08 Y
4.28 Y
61 1.08 Y
4.88 Y
62 1.24 Y
5.04 Y
63 0.92 N
3.80 Y
64 1.04 Y
4,12 Y
65 0.92 N
4.52 Y
66 0.92 J
4.36 Y
67 1.00 Y
4.04 Y
68 1.04 Y
4.04 Y
Westinghouse Braidwood Unit 1 Reactor Vessel Head Penetration Examination Page 22 of 22 Pen #
Coverage Belowthe Weld
@00 1.0 Below Weld © 0° Coverage Above the Weld
@180° 2.0 Above Weld © 180° Measured in inches Y or N Measured in inches YorN 69 1.20 Y
4.08 Y
70 1.04 Y
4.68 Y
71 0.88
~J 4.40 Y
72 0.92 N
4.68 Y
73 1.00 Y
4.96 Y
74* (see note below) 1.08 Y
4.20 Y
75 1.36 Y
6.12 Y
76 1.12 Y
4.24 Y
77 0.92 N
4.32 Y
78 0.84 N
4.56 Y
- During inspection of penetration 74, the open housing scanner shoe lifted off from approximately 0 to 60 degrees, just above the weld. An informational boroscope video recording was performed via manual jump to view the areas of concern. The pictures below show these areas; with the picture on the left showing small deposits and the picture on the right showing a larger deposit. A letter was transmitted to Exelon under CCE-06-4l which contains the NDE coverage evaluation and crack growth assessment for a sub surface flaw. Between ECT and UT, there was enough coverage to preclude the existence of any surface connected indications on the ID. Since ID coverage has been obtained, primary water stress corrosion cracking (PWSCC) is not a concern.
Available leak path results showed no suspect areas. Per CCE-06-41, the valid assumption for a crack in this region would be due to fatigue, which exhibits much slower crack growth rates relative to that due to PWSCC. Refer to CCE-06-4l for more information on this.