ML20217A267

ISI Plan for Braidwood Nuclear Generating Station,Units 1 & 2,1st Insp Interval

ML20217A267
Person / Time
Site:	Braidwood
Issue date:	03/05/1998
From:	COMMONWEALTH EDISON CO.
To:
Shared Package
ML20217A240	List: ML20217A233 ML20217A267
References
NUDOCS 9803240349
Download: ML20217A267 (90)

Text

. ENCLOSURE 1 Revision 5 of the Inservice Inspection Program 9903240349 990305 PDR ADOCK 05000456 -

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Rev sion 5 INSERVICE INSPECTION PLAN FOR BRAIDWOOD NUCLEAR GENERATING STATION UNITS 1 AND 2 IST INSPECTION INTERVAL

, Commercial Service Date Unit 1: July 29,1988 Commercial Service Date Unit 2: October 17,1988 1

Braidwood Nuclear Station R.R. #1 Box 84 Braceville, Illinois 60407 .

Comed 1

P.O. Box 767 '

l Chicago, Illinois 60690 l

I

Revsion 5 ERAIDWOOD NUCLEAR STATION, UNITS 1 AND 2, INSERVICE INSPECTION PLAN, IST INTERVAL REVISION

SUMMARY

SHEET Effective Page(s) Rev.

i to 11 5 iii to iv 4 1-1 3 2-1 to 2-0 3 3-1 to 3-2 3 3-3 5 3-4 to 3-5 2 3-6 2 3-7 to 3-8 2 3-9 3 3-10 to 3-14 2 3-15 to 3-21 3 3-22 2 3-23 2 3-24 to 3-25 4 3-26 3 3-27 2 3-28 to 3-30 2 3-31 2 3-32 3 3-33 to 3-35 3 3-36 1 3-37 to 3-56 0 3-57 to 3-97 1 3-98 to 3-128 0 4-1 3 4-2 4 4-3 to 4-18 2 4-19 1 4-20 to 4-21 2 4-22 to 4-27 0 5-1 3 5-2 4 5-3 to 5-4 2 5-5 4 5-6 3 5-7 to 5-10 2 5-11 to 5-12 3 5-13 to 5-15 3 5 16 to 5-17 4 5-18 2 5-19 3 5-20 to 5-21 2 5-22 3 5-23 2 5-24 4 5-25 3 5-26 4 5-27 3 5-28 to 5-29 4 5-30 to 5-31 3 6-1 to 6-3 2 6-4 to 6-5 1 i

Revslen 5 BRAIDWOOD NUCLEAR STATION, UNITS 1 AND 2, INSERVICE INSPECTION PLAN, IST INTERVAL REVISION

SUMMARY

SHEET (Continued)

Effective Page(s) Rev.

7-1 to 7-3 1 7-4 to 7-7 2 7-8 to 7-9 1 8-1 3 9-1 to 9-7 3 1 10-1 3 j I

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E'V181Cn 4 TABLE OF CONTENTE Section Description Pace 1.0 Introduction and Plan Description 1-1 2.0 Piping P&ID's and Isometric Drawings 2-1 3.0 Relief Requests 3-1 4.0 Technical Approach and Positions 4-1 5.0 ISI Plan Summary 5-1 6.0 Program Plan for Component Support 6-1 7.0 Program Plan for Snubbers 7-1 8.0 Augmented Inservice Inspection Requirements 8-1 9.0 Exempt Component Summary 9-1 10.0 References 10-1 O

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Revision 4 LIST OF TABLES Table Description Page 2.0-1 List of piping and Instrumentation Diagrams 2 ,':

(sorted by System name) 2.0-2 List of Piping Isometrics 2-3 2.0-3 Inservice Inspection Line List Class 2-6 1 and 2 3.0-1 Relief Requests Summaries 3-2 3 4.0-1 Technical Approach and Position Summaries 4-2 5.0-1 ISI Plan Summary 5-2 5.0-2 Exam Method Abbreviations 5-31 9.0-1 Exempt Component Listing 9-2 i

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i iv

Revision 5 Table'3.0-1 Relief Request susumaries (sheet 2 of 2)

-NR-21 Alternate Hydrostatic Pressure Test Requirements for ASME Class 1, 2 and 3 repaired or replaced components.

NR-22 Alternate rules for 10 year Hydrostatic Pressure Testing for Class 3 systems.

NR-23 Alternate Examination of Nozzle to Vessel Welds - Residual Heat Removal Heat Exchangers.

NR-24 Alternative rules for the Inservice Inspection of the Pressurizer Surge Nozzle to Shell Weld and Nozzle Inner Radius Section.

NR-25 Alternative rules for the Inservice Inspection of the Pressurizer Seismic Lug Welds.

NR-26 Alternative rules for the Inservice Inspection of the Inaccessible Welds on Welded Attachments NR-27 Volumetric Examination of Reactor Vessel Nozzle to Vessel Welds NR-28 Alternative rules for the Inservice Inspection of Reactor Vessel Nozzle Inner Radius Sections (IRS)

HR-29 Alternative rules for the Inservice Inspection of Reactor Vessel Circumferential Shell Welds, Lower Head Circumferential Weld *ad Shell to Flange Weld.

NR-30 Alternative rules to Table IWC-2500-1, Category C-H: Pressure Testing of Containment Penetration Piping with attached nonclassed piping NR-31 Volumetric Examination of Reactor Pressure Vessel (RPV) Shell Weld NR-32 ASME Section XI Repair and Replacement Procedures for IWE and IWL Components {

J NR-33 Alternate Requirements for the Repair of Concrete Containment Reinforcing Steel NR-34 Volumetric Examination of Reactor Pressure Vessel (RPV) Circumferential Shell Welds NR-35 Volumetric Examination of Reactor Pressure vessel (RPV) Circumferential

-Shell Welds N3-36 Volumetric Examination of Reactor Pressure Vessel (RPV) Circumferential Shell Welds NR-37 Volumetric Examination of Reactor Vessel Nozzle to Vesael Welds NR-38 Volumetric Examination of Reactor Pressure Vessel (RPV) Lower Head Circumferential Welds

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e 3-3 .

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Revision I RELIEF REQUEST NR-25 COMPONENT IDENTIFICATION Code Classes: 1

Reference:

IWB-2500-1 Examination Categories: B-H Item Numbers B8.20

Description:

Alternate rules for the Inservice Inspection of the Pressurizer Seismic Lug Welds.

Component Numbers: IPZR-01-PSL-01, IPZR-01-PSL-02 1PZR-01-PSL-03, IPZR-01-PSL-04 2PZR-01-PSL-01, 2PZR-01-PSL-02 2PZR-01-PSL-03, 2PZR-01-PSL-04

_ CODE REQUIREMENT Subsection IWB, Table IWB-2500-1, Examination Category B-H, Item B8.00 requires surface or volumetric examination of Integrally Welded Attachments to the Pressurizer (Reference Figure IWB-2500-15).

RASIS FOR RELIEF Comed's Braidwood Nuclear Power Station, Units 1 and 2, conducts ISI activities in accordance with the 1963 Section XI Edition, 1983 Summer Addenda as required by Title 10, Code of Federal Regulations, Part 50, Section 55a, Paragraph g, Subparagraph 4 [10 CFR 50. 55a (g) (4 )] . Pursuant to 10 CFR 50.55a (g) (5) (iii), relief is requested on the basis that compliance with the specified Code requirement has been determined to be impractical.

j Braidwood Units 1 and 2 Pressurizer seismic lugs are welded to the Pressurizer shell (reference Attachment 1). There are 4 seismic lugs per unit, located 90 degrees l apart (reference Attachment 5). In order to perform examinations on the seismic lug welds, the outside surface of the lower vessel shell to lug area must be accessible.

The exam surf ace is not accessible since it is covered by the seisnde lug restraint ,

and lower Pressurizer shell insulation (reference Attachment 3 and 4). Also, the .

' configuration of the Pressurizer coffin limits access to the seismic lugs. The l impact of removing the seismic lug restraint, altering the Pressurizer coffin and  ;

removing the lower shell insulation is presented below.

The seismic restraint (Reference Attachment 1 and 2), which surrounds the lug, prohibits access needed to perform a meaningful surface exam. There are 4 restraints located abcut the 428' elevation, one.for each lug, which were not designed for removal. The top of the concrete floor at this location is at 420' 3" elevation. This floor, which is 2'6" thick, interferes with access to 2 of the 4 lugs (Reference Attachment 2, 3 and 5). Also, the Pressurizer coffin itself severely limits access to remaining 2 seismic restraints (Reference Attachment 5).  !

All of the restraints, which are embedded in the concrete, would require major modification to the existing Pressurizer coffin to allow for removal and access.

l This modification would require the redesign of the seismic restraint and l ' Pressurizer coffin to allow for periodic removal and access to the seismic restraints.

Implementation of this redesign would require significant engineering resources, construction resources and significant dose to plant personnel.

l lI -

L 3-57

Revisien 1 RELIEF REQUEST NR-25 (cont . )

Only the upper panels were designed with clips to provide for removal. Insulation on the lower shell of the Pressurizer prohibits access needed to perform a meaningful surface examination of the seismic lug weld areas. The removal of the insulation covering the lower Pressurizer shell to seismic lug area will result in high radiation exposure to plant personnel. The insulation on the pressurizer consists of panels which are fastened together. The lower panels are fastened together with screws. To provide access from below weald require scaffolding from the 401' elevation grating to the 428' elevation of the seismic restraint.s Also, to remove the Pre'ssurizer shell insulation would require removal of the screw fastenere. Access to these screws is limited by the floor and Pressurizer coffin (reference Attachment 3 and 5). As stated above, the insulation could be removed from the upper portions of the lugs. This can only be accomplished for 2 of the 4 seismic lugs, because access is prohibited by the Pressurizer coffin configuration (Reference Attachment 5). The current configuration of the seismic restraint also only allows lindted access for visual examination. To provide suitable access for all 4 seismic lug restraints would require major modifications and significant resources.

Even if the non removable insulation is removed (Reference Attachment 3, 4& 5),

full surface examination of the seismic lugs would not be achieved. The Pressurizer coffin, concrete floor and seismic restraint geometry would greatly limit access to all sides. The resulting coverage would only be a small percentage of the weld volume. The limited data obtained from these examinations do not provide a -

compensatory increase in quality and safety to justify the hazards of personnel radiation exposure to obtain the data. When the removable insulation panels are removed, it is estimated that 5.74" of surface per accessible lug will be achievable. This accessible portion of surface can be visually inspected. It is expected that only a best effort Liquid Penetrant (PT) exam can be performed on the accessible exposed surfaces. Access and clearance interferences will lindt how well the surf ace of the examination volume can be prepped for the PT examination.

Because the examination is being performed on slightly rusted carbon steel components, which will receive a best effort surface prep, that a white to pinkish back ground will be expected after developing. Even with a pinkish background, detection of relevant indications will still be possible. Also, bleed out from the lower edge of the non removable insulation will interfere with some of the-accessible exam volume. This volume of interference will depend upon the amount of bleed out and will mask any relevant indication.

PROPOSED ALTERNATE PROVISIONS A VT-1 of the upper surfaces of the 2 accessible lugs when the removable insulation ,

panels are removed. It is estimated that 5.74" per accessible lug (4" of the top and .87" of each side) will be achievable when the removable insulation panels are removed. This is approximately 28.74 of the total exam volume for one lug. Also, a best effort surface inspection (Liquid lenetrant) will be performed on those portion of the lug that are inspectable when the removable insulation panels are removed.

In conjunction with the above proposed alternative technique, the periodic VT-2 examinetions in accordance with the requirements of ASME Section XI, Table IWB-2500-1, Examination Category B-P and applicable Reactor Coolant system monitoring requirements specified in the Technical Specifications will provide reasonable assurance of continued structural integrity of the Pressurizer shell.

APPLICABLE TIME PERIOD This relief request will be required for the first ten-year Inspection Interval.

APPROVAL STATUS Relief granted per SER dated 12/11/97.

3-58

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Revision 1 NR-25 Attachment 5 (Drawing not to scale)

PRESSURIZER COFFIN

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Plan at Elevation 428'-7 1/2" UNIT 1 AS SHOWN UNIT 2 OPP. HAND 3-63

Fevision 1 RELIEF REQUEST NR-26 CCatPONENT IDENTIFICATION Code Class (es)* 2

Reference:

IWC-2500-1 Examination Categories: C-C Item Numbers: C3.20

Description:

Alternative rules for the Inservice Inspection of Inaccessible Welds on Welded Attachments Component Number (s): Unit 1 Welds:

• 1FW-06-33, *1FW-10-29, *1FW-11-29, *1FW-12-29

• 1RH-03-37B, *1RH-04-73A, *1RH-05-21B, IRH-07-25A

• 1RH-08-02A, 1RH-09-45A, *1SI-04-02A, *1SI-12-23A ISI-26-09A Unit 2 Welds:

• 2FW-06-01, *2FW-10-24, *2FW-11-25, *2FW-12-25

• 2RH-04-03, *2RH-05-35, *2RH-08-03, *2RH-09-28 *

• 2SI-04-03, *2SI-12-19A, *2SI-26-03

(*) denotes welds selected for inspection during the interval.

CODE REQUIREMENT Subsection IWC, Table IWC-2500-1, Examination Category C-C, Item C3.20 requires surface examination of the Integrally Welded Attachments to Piping (Reference Figure IWC-2500-5).

BASIS FOR RELIEF Comed's Braidwood Nuclear Power Station, Units 1 and 2, conducts ISI activities in accordance with the 1983 Section XI Edition, 1983 Summer Addenda as required by Title 10, Code of Federal Regulations, Part 50, Section 55a, Paragraph g, Subparagraph 4 [10 CFR 50.55a (g) (4)] . Pursuant to 10 CFR 50.55a tg) (5) (iii), relief is being requested on the basis that compliance with the specified Code requirement a has been determined to be impractical.

Some penetrations at Braidwood were originally designed where one of the integral attachment welds is inside the flued head penetration assenaly, thus making the welds inaccessible for inservice inspection. Access from oitside of the closed end

.of the flued head penetration assembly for examiners is prosibited by the integral attachment. Access from the open end of the penetration is severely restrained due to geometry and clearance. See Attachments 2, 2, 3, 4 and 5 for penetration details. The integral attachment weld is set back some distance inside the flued head penetration assembly and the clearance between the pipe and penetration sleeve is small. See Table 1 on Attachment 6.

To satisfy the Code requirement to perform a surface examination of this weld, modification to the flued head penetration assembly and/or piping to allow access would be required. Braidwood would incur significant engineering and installation costs to perform such a modification without a compensating increase in the level of quality and safety to justify such modifications.

3-64 ..

1

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Revision 1 RELIEF REQUEST NR-26 (cont.)

PROPOSED ALTERNATE PROVISIONS When a weld is scheduled for inspection, a surface examination of the accessible weld on the exposed outside surface of the penetration will be performed. In conjunction with the above proposed alternative technique, the periodic VT-2 examinations in accordance with the requirements of ASME Section XI, Table IWC-2500-1, Examination Category C-H will provide reasonable assurance of continued structural integrity of the piping systems.

APPLICABLE TIME PERIOD This relief request will be required for the First Ten Year Inspection Interval.

APPROVAL STATUS

• Belief granted per SER dated 12/11/97.

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Revisien 1 NR-26 ATTACHMENT 2 (Drawing not to scale) l PENETRATION ATTACHMENT l

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Revision 1 NR-26 ATTACHMENT 5 (Drawing not to scale)

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3-70 i

Revision 1 NR-26 ATTACHMENT 6 TABLE 1 Neld ID Pipe Pipe Attachment Thickness Penet. Penet.

Size Thickness Thickness concrete size Clearance To Ta Ten Pe.

TYPE 2 PENETRATIONS (ATTACHMENT 1) 1RH-04-73A 12" 0.375" 2.0" 3' 6" 24" 4.9" 1RH-08-02A 12" 0.375" 2.0" 3' 6" 24" 4.9" ISI-04-02A 8" 0.906" 2.0" 3' 6" 24" 7.0" ISI-12-09A 12" 1.125" 2.0" 3' 6" 24" 4.9" 2RH-04-03 12" 0.375" 2.0" 3' 6" 24" 4.9" 2 RH- 0 8-03 12" 0.375" 2.0" 3' 6" 24" 4.9" 2SI-04-03 8" 0.906" 2.0" 3' 6" 24" 7.0" 2SI-12-19A 12" 1.125" 2.0" 3' 6" 24" 4.9" TYPE 3 PENETRATIONS (ATTACHMENT 1)

IFW-06-33 6" 0.432" 2" 4'6" 16" 3.8" 1FW-10-29 6" 0.432" 2" 4 ' 6" 16" 3.8" l 1FW-11-29 6" 0.432" 2" 4 ' 6" 16" 3.8" 1FW-12-29 6" 0.432" 2" 4 ' 6" 16" 3.8" 1SI-26-09A 8" 0.906" 2" 3' 6" 24" 7.0" 2FW-06-01 6" 0.432" 2" 4 ' 6" 16" 3.8" 2 FW 2 9 6" 0.432" 2" 4 ' 6" 16" 3.8" 2FW-11-29 6" 0.432" 2" 4 ' 6" 16" 3.8" 2fw 12-29 6" 0.432" 2" 4'6" 16" 3.8" 2SI-26-03 8" 0.906" 2" 3' 6" 24" 7.0" DETAIL 4 PENETRATIONS (ATTACHMENT 4) 1RH-03-37B 8" 0.375" 1" 3' 0" 18" 4.3" 1RH-07-25A 8" 0.375" 1" 3' 0" 18" 4.3" .

DETAIL 27 PENETRATIONS (ATTACHMENT 3) .

,1RH-05-21B 8" 0.375" 1" 3' 0" 20" 5.2" 1RH-09-45A 8" 0.375" 1" 3' 0" 20" 5.2"

~.'RH-05-35 8" 0.375" 1" 3' 0" 20" 5.2"

~iRH-09-28 8" 0.375" 1" 3' 0" 20" 5.2" l

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3-71

Revision 1 RELIEF REQUEST NRJ27 COMPONENT IDENTIFICATION -

Code Class (es): 1

Reference:

IWB-2500-1 Examination Categories: B-D j Item Numbers: B3.90

Description:

Volumetric Examination of Reactor Pressure Vessel Nozzle to Vessel Welds Component Number (s): Unit 1 Welds:

1RV-01-006, 1RV-01-009, 1RV-01-010, 1RV-01-013 CODE REQUIREMENT Subsection IWB, Table IWB-2500-1, Examination Category B-D, Item B3.90 requires essentially 100% volumetric examination of the region described in Figure IWB-2500-7 for Reactor Pressure Vessel (RPV) nozzle-to-vessel welds.

RASIS FOR RELIEF Comed's Braidwood Nuclear Power Station Units I conducts ISI activities in accordance with the 1983 Section XI Edition, 1983 Summer Addenda as required by Title 10, code of Federal Regulations, Part 50, Section 55a, Paragraph (g),

Subparagraph (4) (10 CFR 50. 55a (g) (4 )) . Pursuant to 10 CFR 50. 55a (g) (6) (1) , relief is requested on the basis that the code requirement to examine essentially 100% of the welds' volume is impractical due to geometric interference.

All RPV welds are examined using remotely operated underwater volumetric inspection techniques. Underwater volumetric inspection techniques are utilized to meet ALARA concerns due to the high radiation levels in these areas. The outlet (Hot Leg) nozzles are constructed with an integral extension on the I.D. surface which mates with the internal core Sarrel. The extension provides a flow path for reactor coolant from the core into the hot leg nozzles. The integral extensions partially obstructs the circumferential scan for reflectors transverse to the weld (Reference Attachment 1). The integral extension, that confines the movement of the transducer '

package, along with the curvature of the RPV shell combine to limit full Code volume coverage when scanning in the direction parallel to the weld (Reference Attachment 2). This configurat.4on limits the examination aggregate volume coverage obtained for each weld and adpacent base metal to approximately 84% instead of the Code required essentially 100+ examination coverage.

Compliance with the a)plicable Code requirements may be accomplished by redesigning and modifying the ID of the Hot Leg nozzles and/or the building structure surrounding the RPV at the nozzles' elevation. Braidwood Unit 1 RPV was designed with a RPV shield wall (Reference. Attachment 3 and 4). This wall impedes access to the CD of the RPV shell for insulation removal, surface preparation and ultrasonic inspection. Modifying the nozzle ID surface would incur extensive radiation exposure to s%ation personnel and could be detrimental to the component. When designing,. fabricating and installing these welds, strict ASME Section III quality controls and procedures were used that minimized the introduction of fabrication defects. Additionally, the periodic VT-2 examinations in accordance with the 3-72

i Revision 1 RELIEF REQUEST NR-27 (cont.)

)

requirements of ASME Section XI, 7491e IWB-2500-1, Examination Category B-P and applicable Reactor Coolant system monitoring requirements specified in the Technical Specifications will provide reasonable assurance of continued structural integrity of the Reactor Vessel, Comed has recently performed these volumetric examinations to the fullest extent practical, i.e. 84i, during the A1R06 refuel outage and no recordable indications (NRI) were detected. The NRI results of the examination provide further assurance that unacceptable inservice flaws have not developed in the' subject welds. Thus, the modification of the nozzles and/or the building structure to increase examination volume coverage from 84% to essentially 100t would incur unnecessary radiological exposure and significant engineering costs without a compensating increase in the level of quality and safety. '

PROPOSED ALTERNATE PROVISIONS The Reactor Vessel outlet (Hot Leg) nozzle welds will be examined to the fullest extent practical using the available underwater yolumetric inspection techniques.

APPLICARIK TIME PERIOD This relief request will be required for the First Ten Year Inspection Interval.

APPROVAL STATUS Relief granted per SER dated 12/11/97.

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3-73

Revision 1 NR-27

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Section A-A Interference Detail 3-77

Pevision 1 RELIEF REQUEST NR-28 COMPONENT IDENTIFICATION Code Class (es): 1

Reference:

IWB-2500-1 Examination Categories: E-D Item Numbers: B3.100

Description:

Alternative rules for the Inservice Inspection of Reactor Pressure Vessel Hozzle Inner Radius Sections (IRS)

Component Number (s ) : Unit 1 Welds:

1RV-01-015, 1RV-01-016, 1RV-01-019, 1RV-01-020 CODE REQUIREMENT Subsection IWB, Table IWB-2500-1, Examination Category B-D, Item B3.100 requires essentially 100% volumetric examination of the inner radius region described in Figure IWB-2500-7 for Reactor Pressure Vessel (RPV) nozzle inner radius sections.

BASIS FOR RELIEF Comed's Braidwood Nuclear Power Station Unit 1 conducts ISI activities in accordance with the 1983 Section XI Edition, 1983 Summer Addenda as required by Title 10, Code of Federal Regulations, Part 50, Section 55a, Paragraph g, Subparagraph 4 [10 CFR

50. 55a (g) (4 ) ) . Pursuant to 10 CFR 50.55a (g) (6) (1), relief is requested on the basis i that the code requirement to examine essentially 100s of the each inner radius section is impractical due to the geometry.

All of the inner radius sections of the RPV nozzles are examined using remotely operated underwater volumetric inspection techniques. Underwater volumetric )

inspection techniques are utilized to meet ALARA concerns due to the high radiation l levels in these areas. Examination of the inner radius sections of the inlet (Cold i Leg) nozzles is lindted by the nozzle geometry (Reference Attachment 1 and 2). The design of the underwater volumetric inspection equipment was unable to scan the radius area where it transitions from the shell into the nozzle bore. This geometry

• l limits the volumetric examination aggregate volume coverage obtained for the nozzle inner radius section to about 824 instead of the essentially 100% code required exam volume.

Compliance with the applicable Code requirements may be accomplished by redesigning and modifying the nozzles' inner radius section and/or the building structure surrounding the RPV at the nozzles' elevation. Braidwood Unit I was designed with a RPV shield wall (Reference Attachment 3 and 4). This wall impedes access to the OD of the RPV shell for insulation removal, surface preparation and ultrasonic inspection. Modifying the nozzle ID surfaces surface would incur extensive radiation exposure to station personnel.

Based on industry experience, no operating issues to date have been identified in PWR nozzle inner radius sections. If an inservice flaw were to develop in this region, the flaw would be expected to initiate at the ID surface. A visual inspeztion (VT-1) of the entire ID IRS surface provides reasonable assurance 3-78 -

Revisien 1 RELIEF REQUEST NR-28 (cont.)

that unallowable inservice flaws have not developed in the subject area.

Additionally, the periodic VT-2 examinations in accordance with the requirements of ASME Section XI, Table IWB-2500-1, Examination Category B-P and applicable Reactor Coolant system monitoring requirements specified in the Technical Specifications will provide reasonable assurance of continued structural integrity of the Reactor Vessel. Comed has recently performed these volumetric examinations to the fullest extent practical, i.e. 82%, during the AIR 06 refuel outage and no recordable indications (HRI) were detected. The NRI results of the examination provide further assurance that unacceptable inservice flaws have not developed in the subject areas.

Thus, the modification of the nozzles and/or the building structure to increase examination coverage from 82% to essentially 100t would incur unnecessary radiological exposure and significant engineering costs without a compensating increase in the level of quality and safety.

PROPOSED ALTERNATE PROVISIONS The Reactor Vessel inlet (Cold Leg) nozzle inner radius sections were examined to the fullest extent practical using the available underwater volumetric inspection technique. In conjunction with the partial volumetric examination, a supplemental VT-1 of the nozzle inner radius area was conducted from the interior of the Reactor Vessel using underwater camera equipment.

APPLICABLE TIME PERIOD This relief request will be required for the First Ten Year Inspection Interval.

APPROVAL STATUS Relief granted per SER dated 12/11/97.

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3-83

• Pevision 1 RELIEF REQUEST-NR-29 COMPONENT IDENTIFICATION Code Class (es): 1

Reference:

IWB-2500-1 Examination Categories: B-A Item 'Jumbers: Bl.11, Bl.21, Bl.30

Description:

Alternative rules for the Inservice Inspection of Reactor Vessel Circumferential Shell Welds, Lower Head Circumferential Weld and Shell to Flange Weld.

Component Number (s): Unit 1 Welds:

1RV-01-003, 1RV-cl-004, 1RV-01-005, 1RV-02-001, 1RV-02-002 Unit 2 Welds:

2RV-01-003, 2RV-01-004, 2RV-01-005, 2RV-02-001, 2RV-02-002 CODE REQUIREMENT Subsection IWB, Table IWB-2500-1, Examination Category B-A, Items Bl.11, Bl.21 and Bl.30 requires volumetric examination be performed on the above Reactor Vessel Welds. These volumetric examinations are to be performed in accordance with IWA-2232. IWA-2232 states that the inspections "shall be conducted in accordance with Article 4 of Section V", and amended by Section XI. Braidwood is currently committed to the ASME Section XI 1983 Edition with Summer 1983 Addenda of Section V and XI.

i BASIS FOR RELIEF Relief is requested pursuant to the provision of 10 CFR 50.55a (a) (3) (1), the proposed alternative would provide an acceptable level of quality and safety.

Braidwood is requesting relief from the Section XI, 1983 Edition with Summer 1983 Addenda, Paragraph IWA-2232 requirements which requires these examinations to be conducted in accordance with Article 4 of Section V, 1983 Edition with Summer 1983 '

Addenda and as amended by Section XI. The attached Table A identifies the specific applicable Section V and XI requirements and the proposed corresponding alternative Performance Demonstration Initiative (PDI) technique.

The Electric Utility industry has developed a program to qualify ultrasonic inspection techniques. This program, Performance Demonstration Initiative (PJI), is designed to meet the intent of Appendix VIII of the ASME Code,Section XI, 1952 I l

Edition with 1993 Addenda. This program, PDI, used a variety of test blocks to evaluate transducer designs, scanning requirements and flaw rizing techniques.

Braidwood has contracted with Framatome Technologies (FTI) t> use the URSULA manipulator to perform the 10 Year Ultrasonic (UT) Reactor Vessel inspections. It is  :

the Braidwood'.s intent to use the FTI technique qualified in December 1995 to the PDI Program at the EPRI NDE Center that meets the intent of Appendix VIII of the 1992 Edition with 1993 Addenda. This FTI qualified PDI technique consists of scanning the examination volume, weld and base metal, as follows:

3-84 -i

Revision 1 RELIEF REQUEST NR-29'(cont.)

(1) For flaw detection, the examination volume will be scanned in two directions, one perpendicular and one parallel to the weld axis. The examination volume is scanned from one direction such that all the examination angles pass through the entire examination volume of interest-for each transducer. If full coverage is limited, scanning from both directions will be performed when coverage can be maximized.

(2) For flaw characterization (or flaw sizing), the examination will be conducted from two_ opposing directions, in the direction perpendicular to the plane of the flaw, when feasible.

Performance-based UT techniques provide a higher degree of reliability for detection and characterization of flaws when compared to the conventional amplitude-based UT techniques that are currently required by ASME Section XI, 1989 Edition and earlier code editions and addendas. The performance-based demonstration requires the inspection equipment, procedures, and examiners to be tested on flawed specimen representing materials and configurations similar to those found in actual plant conditions. The NRC Staff has acknowledged the improvement achieved by performance-based UT techniques in the recently issued proposed Generic Letter 96-XX:

Effectiveness of Ultrasonic Testing Systems In Inservice Inspection Prograns (Federal Register Notice of December 31, 1996; 61 FR 69120). Additionally, the NRC Staff has assessed the PDI program activities and found that PDI has established and executed a well-planned and effective program to test UT equipment, procedures, and examiners on selected portions of Appendix VIII, which include reactor vessel inspection technique. This assessment is documented in a letter from J. Strosnider (NRC) to B. Sheffel (PDI) dated March 6, 1996 and referenced in the above mentioned '

proposed Generic Letter.

I PROPOSED ALTERNATE PROVISIONS Braidwood proposes to use FTI's underwater volumetric inspection techniques to inspect the reactor vessel Circumferential shell welds, lower head Circumferential weld and shell to flange weld. FTI's inspection techniques have been demonstrated and qualified to the PDI Program which meets the intent of the rules of Appendix VIII of the ASME Code,Section XI, 1992 Edition with 1993 Addenda. These techniques will be used in place of the currently required Section XI, 1983 Edition with Summer 1983 Addenda, techniques.

APPLICABLE TIME PERIOD This relief request will be required for the First Ten Year Inspection Interval. .

APPROVAL STATUS Felief granted per SER dated 5/13/97.

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i Revision 1 RELIEF REQUEST: NR-30 CCMPONEart IDENTIFICATION _

Code Class: 2

Reference:

Table IWC-2500-1 Examination Category: C-H Item Numbets: C7.30, C7.40, C7.70, and C7.80

Description:

Alternate Rules to Table IWC-2500-1, Category C-H:

Pressure Testing of Containment Penetration Piping with Attached Nonclassed Piping CODE REQUIREMENT ASME Section XI, Table IWC-2500-1, Examination Category C-H, requires the performance of a visual VT-2 examination during a system pressure test on Code Class 2 pressure retaining components. Note 7 of this table states, "The pressure boundary includes only those portions of the system required to operate or support the safety system function up to and including the first normally closed valve (including a safety or relief valve) or valve capable of automatic closure when the safety function is required."

RAsIs FOR RELIEF s

Comed's Braidwood Nuclear Power Station, Units 1 and 2, conducts ISI activities in l l

accordance with the 1983 Section XI Edition, 1983 Summer Addenda as required by l Title 10, Code of Federal Regulations, Part 50, Section 55a, Paragraph g, Subparagraph 4 (10 CFR 50.55a (g) (4)] . Pursuant to 10 CFR 50.55a (a) (3) (i), relief is requested on the basis that the proposed alternative would provide an acceptable level of quality and safety.

Specifically, Braidwood Station requests relief to perform 10 CFR 50 Appendix J 1eakage testing in lieu of the pressure test required by ASME Section XI, Table IWC-2500-1, Examination Category C-H on the code Class 2 Containment Penetration piping with attached nonclassed piping.

The applicable components are piping lines and valves which are portions of non-safety related systenu that penetrate the primary reactor containment. At each a containment penetration, the process pipe is classified Code Class 2 and provided with isolation valves that are either locked shut during nornal operation, capable of automatic closure, or capable of remote closure to support the containment safety function. The balance of piping outsid e the isolation valves is non-code and therefore outside the scope of the ASME Boiler & Pressure Vessel Code,Section XI.

These components perform no other safs i.f function. The only reason that the penetration piping is clas.A 'ted as C; u 2 is because of its function as part of the containment pressure bnuncary. Th+ remaining portion of the system is non nuclear related and the int *gr1 % ;f H system in relation to its primary function is not within the scope of Section XI. Since containment integrity is the only safety related. function, it is :ogical to test the Class 2 penetration portion of the system to the Appendix J criteria.

Theprimaryre$ctorcontainment integrity, including all containment penetrations, is periodically verified by performing leakage tests in accordance with 10 CFR 50, Appendix J. The Appendix J test frequency provides assurances that the containment pressure boundary is being maintained at an acceptable level while monitoring for deterioration of-seals, valves and piping. If a pipe existed with a through-wall flaw, the isolation valves located on both sides of the containment wall would 347 -

Revision 1 RELIEF REQUEST NR-30 (cent.j prevent any release outside containment. Multiple through-wall flaws or leakage paths occurring simultaneously inside and outside of containment between the isolation valves in a pipe segment is unlikely. Each of the Code Class 2 lines and their associated isolation valves are tested during an Appendix J 1eakage test at a pressure not less than 44.4 psig (Peak calculated containment pressure). The Appendix J leakage tests are performed at intervals in accordance with the requirements of the Braidwood Technical Specifications.

Performance of' these Appendix J 1eak tests will verify the integrity of the subject Code Class 2 lines and valves at the containment-penetrations. The performance of ASME Section XI, Examination Category C-H pressure tests on these same lines will provide little, if any, additional verification of primary reactor containment integrity and impose a burden of duplicate testing. Duplicate testing results in a significant increase in total amount of work force and radiological exposure without a compensating increase in the level of quality or safety.

Per the preceding information, Braidwood Station requests relief to use the Appendix J test as an optional. alternative to ASME Section XI requirements for pressure testing the Code Class 2 containment penetration components on the basis that the Proposed Alternate Provisions provide an acceptable level of quality and safety.

The proposed alternative is consistent with the requirements of Code Case N-522.

PROPOSED ALTERNATE PROVISIONS Braidwood Station will perform 10 CFR 50, Appendix J leakage tests as an optional alternative to the Section XI required pressure test on the subject primary reactor containment penetration piping and associated valves. When implementing the Appendix J leakage test and invoking this relief request, peak design pressure and procedures for the detection and location of through-wall flaws will be used.

PERIOD FOR WHICH RELIEF IS REQUESTED Relief is requested for the first inspection interval.

APPROVAL STATUS Relief granted per SER dated 12/11/97.

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1. 181 ^ I RELIEF REQUEST NR 831 COMPONENT IDENTIFICATION -

Code Class (es): 1

Reference:

IWB-2500-1, Table IWB-2500-1 Examination Categories: B-A Item Number (s): Bl.11

Description:

Volumetric Examination of Reactor Pressure Vessel (RPV) Circumferential Shell Weld.

Component Number (s) : Unit 1 Weld: 1RV-02-002 CODE REQUIREMENT Subsection IWB, Table IWB-2500-1, Examination Category B-A, Item Number B1.11 requires essentially 100+ volumetric examination of the RPV Circumferential Shell welds as detailed in Figure IWB-2500-1.

RASIS FOR RELIEF Comed's Braidwood Nuclear Power Station Units 1 conducts ISI activities in accordance with the 1983 Section XI Edition, 1983 Summer Addenda as required by Title 10, code of Federal Regulations, Part 50, Section 55a, Paragraph (g),

Subparagraph (4) [10 CFR 50. 55a (g) (4 ) } . Pursuant to 10 CFR 50.55a (g) (6) (i), relief is requested on the basis that the code requirement to examine essentially 100% of the weld volume is impractical due to geometric interference.

10CFR 50.55 a (g) (6) (ii) (A) (1) revokes all relief requests with respect to volumetric examination coverage for welds specified in Item Bl.10. Portions of a previously granted relief request, NR-9, addressed limited exam coverage on the Braidwood RPV shell welds.

Examination of the subject RPV shell weld was conducted on Braidwood Unit I during AIR 06 refuel outage (Spring 1997). During this exam at Braidwood Unit 1, physical obstructions and geometry prevented ultrasonic (UT) coverage in excess of 90% of the required volume. The examination of the Lower Shell

• Course-to-Dutchman weld, IRV-02-002, is restricted by six (6) core barrel locating lugs welded to the inner surface of the vessel approximately 2.5 inches above the welu centerline (See Attachment 1). These lugs obstruct the automated UT examination tool f rom examining the Code required volume of the weld and base material under and below each lug in both the circumferential and perpendicular scan directions (156* total for all 6 lugs, see Attachment 2, 3 and 4).

All weld metal and base material can be examined between the lugs (204* total length between all 6 lugs). The 6 lug interferences limit the examination aggregate volume coverage obtained for the weld and adjacent base metal to approximately 81t of the code required volume.

Compliance with the applicable Code requirements may be accomplished by redesigning and modifying the RPV and/or the building structure surrounding the vessel (s) .

Access for manual inspections from the OD of the RPV is limited because of the clv..-

proximity of the building structure to the RPV shell (See Attachment 1).

Examination of the Code requ' ired examination volume was completed to the maximum extent practical using alternate UT techniques qualified to the highest standard

~

3-89

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}

b'V18100 1 RELIEF REQUEST NR-31 (cont.)

available. RPV examinations were conducted from the I.D. of the vessel. Access to allow examination from the O.D. (shell side) of these welds is restricted due to the structural concrete surrounding the vessel. The examination techniques employed have been demonstrated and qualified to the Performance Demonstration Initiative (PDI) Program which meets the intent of the rules of Appendix VIII of the ASME Code,Section XI, 1992 Edition with 1993 Addenda. These techniques were used in place of the currently required Section XI, 1983 Edition with Summer 1983 Addenda, techniques (Ref erence Felief Request NR-29) . Although the techniques have been qualified at PDI for single direction scanning, examinations were performed from two directions, when required (i.e., performed from both sides of the weld on the same surface, where I

feasible).

l Strict ASME Section III quality controls were used when designing, fabricating, and installing these RPV welds.

Preservice practical were performed on these welds.

(PSI) examinations to the fullest extent PSI relief request INR-9 was submitted to the Staff and approved for these lug interferences. Comed has recently performed these ultrasonic examinations to the fullest extent practical, i.e. 814, during the AIR 06 refuel outage and no unacceptable indications to applicable Section XI standards were detected. The results of the examination provide further assurance that unallowable inservice flaws have not developed in the subject weld. In addition to UT, visual examinations (VT-1 and VT-2) of the weld also verifies its integrity. Thus, the modification of the RPV and/or the building structures to increase examination volume coverage from 81% to essentially 1001 would incur unnecessary radiological exposure and significant engineering expenses. Braidwood Station believes this course of action is a hardship without a compensating increase in the level of quality and safety.

PROPOSED ALTERNATE PROVISIONS The ultrasonic examination of the Braidwood Unit 1 RPV shell weld, 1RV-02-002, was performed to the maximum extent practical. In conjunction with the partial ultrasonic examination, a supplemental VT-1 of the RPV shell weld was conducted from the interior of the RPV using underwater camera equipment.

APPLICABLE TIME PERIOD I This relief request will be required for the First Ten Year Inspection Interval.

APPROVAL STATUS I

Relief granted per SER dated 12/11/97. .

m 3-90

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Revision 1 Braidwood Stadon ist interval insenice inspection Plan e

RELIEF REQUEST: NR-32 (Page 1 of 3)

COMPONENT IDENTIFICATION Code Class: MC, CC .

References:

IWA-4000 and IWA-7000 IWE-4000 and IWE-7000

• IWL-4000 and IWL-7000 Examination Category: E-A, E-B, E-C, E-D, E-F, E-G, and E-P L-A, L-B Item Number: All Class MC and CC components listed in Table IWE-2500-1 and IWL-2500-1

Description:

ASME Section XI Repair and Replacement Procedures for IWE and IWL Components Component Numbers: All Class MC and CC components subject to Repair and Replacement rules of IWA-4000, IWA-7000, IWE-4000, IWE-7000, IWL-4000 and IWL-7000.

CODE REQUIRDIENT 10 CFR 50.55a (g) (4) (v) requires class MC and CC pressure retaining components and their integral attachments to meet the applicable repair and replacement requirements of the ASME Boiler and Pressure Vessel Code Section XI, 1992 Edition with the 1992 Addenda, Articles IWA-4000, IWA-7000, IWE-4000, IWE-7000, IWL-4000 and IWL-7000.

BASIS FDR RELIEF Relief is requested from immediate compliance with the repair and replacement requirements of Subsections IWE and IWL. Pursuant to 10 CFR 50.55a (a) (3) (ii),

relief is requested on the basis that immediate compliance with the aforementioned requirements would result in unusual difficulty without a compensating increase in the level of quality and safety.

A revision to 10 CFR 50.55a was published on August 8, 1996, which endorses Subsections IWE and IWL of the ASME Boiler and Pressure Vessel Code Section XI, 1992 Edition with the 1992 Addenda. This revision requires the completion of an expedited examination by September 9, 2001. However, in a letter to the Nuclear Energy Institute (NEI) dated November 6, 1996, the NRC staff clarifies that all ,

repair and replacement activities within the scope of subsections IWE and IWL which are conducted after September 9, 1996 must be conducted in accordance with the applicable rules of Subsections IWE and IWL of ASME Section XI, 1992 Edition with the 1992 Addenda.

Inmediate compliance with the repair and replacement rules of ASME Section XI, 1992 Edition with the 1992 Addenda for IWE components is impractical because substantial time and resources must be expended for the following major efforts:

1. i CONTAINMENT STRUCTURE COMPONENT CLASSIFICATION: The containment structures at Braidwood Station were constructed to the proposed rules of the 1973 ASME Section III, Division 2 Code.

In order to comply with 10 CFR 50.55a (g) (4) (v),

it will be decessary to identify and reclassify all containment components to Class MC and CC classification criteria. This effort will include the retrieval and review of all applicable fabrication and installation documentation, the development of a basis document to identify the correct classification boundaries and the eventual development of an inservice 3 - 95 .

Revision 1 Braidwood Station 1st Intenal inservice Inspection Plan RELIEF REQUEST: NR-32 (Page 2 of 3) inspection program to govern all IWE/IWL-related activities at Braidwood Station.

2. PROCEDURE REVISIONS: .The requirements of Subsections IWE and IWL must be incorporated into applicable ctation procedures. The current Inservice Inspection program, (which includes the repair and replacement program), for Braidwood Station is currently based on the rules of ASME Section XI, 1983.

Edition with the Summer 1983 Addenda, and only addresses the inservice inspection requirements for Class 1, 2, and 3 pressure retaining components and component supports. Therefore, multiple procedures that control Code repair and replacement activities must be revised to incorporate the unique requirements of Subsections IWE and IWL.

3. EXAMINER TRAINING AND CERTIFICATION: The unique examiner qualification required by subsections IWE and IWL must be incorporated into the existing Commonwealth Edison (Comed) certification and training program. The existing Comed certification and training program only addresses the certification.

requirementa for Class 1, 2 and 3 pressure retaining components and component supports. The Comed certification and training program must be revised to incorporate the unique requirements of Subsections IWE and IWL.

Since the containment structures at Braidwood Station were constructed to the rules of Section III, all repair and replacement activities conducted on these components have been subje ted to the Comed Quality Assurance (QA) Manual (Commonwealth Edison Company, Topical Report CE-1-A, Section 2, 3.1) which implements the requirements of 10 CFR 50, Appendix.B. The Comed QA program requires repair and replacement activities to be conducted in accordance with the original design specifications using approved procedures. This approach assures applicable design bases are maintained. As stated in Comed Quality Assurance Program Topical Report CE-1-A, current revision, the Comed Quality Assurance Program complies with the quality requirements of 10 CFR 50, Appendix B, ASME Section III NCA-4000, and ANSI /ASME HQA-1. In addition, the containment structure integrity is verified by periodic pressure tests in accordance with Appendix J and the surveillance of the post-tensioning system. Post-tensioning system testing and examinations are performed in accordance with Technical Specification required programs, Braidwood Procedure BwVP 200-15, " Containment Vessel Tendon Inspection Requirements" and implementing procedure, BwvS 6.1.1.1-

-1, " Containment Vessel Tendon Test". .

These Approved procedures incorporate the requirements of NRC proposed Revision 3 of Regulatory Guide 1.35. I For the above reason 1 the immediate application of the requirerents of Articles IWA-4 000, IWA-7000, IWE-4000, IWE-7000, IWL-4000 and IWL-7000 imposes added administrative burden (such as requirement for a repair / replacement plan and NIS-2 form) without prov2 ding a compensating increase'in the level of quality or safety.

PROPOSED ALTE3GULTE PROVISIcets Through September 8, 1997, all repair and replacement activities conducted on Class MC and CS containment structure components and their integral attachments at Braidwood Station will be performed in accordance with the. existing Comed QA Program requirements. Compliance with ASME Section XI, 1992 Edition with the 1992 Addenda, Articles IWA-4000, IWA-7000, IWE-4000, IWE-7000, IWL-4000 and IWL-17000 will begin on September 9, 1997.

3 - 96 * -

Revision I Braidwood Station 1st Interval Inservice inspection Plan RELIEF REQUEST: NR-32 (Page 3 of 3)

Should the need arise to complete Repair / Replacement activities on any Class MC and CC pressure retaining components or their integral attachments prior to September 9, 1997, these activities will continue to be controlled using approved Nuclear Work Requests in accordance with Braidwood Administrative Procedure BwAP 1600-1, " Action / Work Request Processing Procedure". These Nuclear Work Requests will be classified as " Nuclear Safety Related" and thus, their preparation, review, approval, implementation and associated post Repair / Replacement testing is governed by the Comed QA Manual (Commonwealth Edison company, Topical Report CE-1-A, Section 2, 3.1). The Comed Quality Assurance Program complies with the quality requirements of 10 CFR 50 Appendix B, ASME Section III NCA-4000, and ANSI /AEME NQA-1. Post-tensioning system testing and examinations will be performed in accordance with Technical Specification required programs, Braidwood procedures BwVP 200-15,

" Containment Vessel Tendon Inspection Requirements", and BwVS 6.1.6.1-1,

" Containment Vessel Tendon Test". These approved procedures incorporate the requirements of proposed Revision 3 of Regulatory Guide 1.35. In addition, the applicability of Section XI repair / replacement requirements for Class CC concrete containment and post-tensioning system components, Class MC pressure retaining components and their integral attachments and metallic shell and penetration liners of Class CC components will be controlled in accordance with Braidwood administrative Procedure BwAP 1600-5, "ASME Section XI Repair / Replacement Requirements".

APPLICABLE TIME PERIOD Relief is requested for the first ten-year interval of the Inservice Inspection Program for Braidwood Units 1 and 2 until September 9, 1997. I APPROVAL STATUS Pending NRC Review.

O e

3 - 97 * -

Revision 0 RELIEF REQUEST: NR-33 (Page 1 of 3)

COMPONENTIDENTIFICATION Unit: Braidwood Unit 1 Code Class: CC

References:

IWL-4000 10 CFR 50.55a 1992 Edition with 1992 Addenda

Description:

Altemative Requirements for the Repair of Concrete Containment Reinforcing Steel.

CODE REOUIREMENT 10CFR50.55a(g) has invoked ASME Section XI(1992 Edition with 1992 Addenda) for the repair activities of containment structures. ASME Section XI, Subsection IWL-4220 requires that the

" damaged reinforcing steel shall be repaired by any method permitted in the original Construction Code or in ASME Section III, Division 2, with or without the removal of the damaged steel".

The original Construction Code for Braidwood was the 1973 ASME Section III, Division 2 (issued for trial use and comment) Subsection CC-4334.7, and only allowed welding of reinforcing steel with carbon equivalents equal to or less than 0.55%.

BASIS FOR RELIEF '

Pursuant to 10 CFR 50.55a(a)(3)(ii), reliefis requested on the basis that compliance with the specified requirements would result in hardship or unusual difficulty without a compensating increase in the level of quality and safety, To facilitate replacement of the steam generators at Braidwood Unit 1, a construction opening will be created in the containment building during the Steam Generator Replacement outage (SGRO). Cor.tainment concrete removal activities involve the use of automated chipping machines to remove concrete. Once exposed, the reinforcing steel will be cut and removed. As part of the coatainment restoration effort, stub ends of the reinforcing steel that remain in-place at the edges of the opening will be Cad-welded with replacement reinforcing steel to reconstruct pre-outage reinforcing steel patterns. For containment restoration, all reinforcing steel must have the required cross section which was specified in the original design specification. Although every attempt will be made to protect the reinforcing steel (stub end areas) during the concrete removal process, physical damage (nicks) in tne reinforcing steel base metal may occur. As per the original construction specification, L-?.722, nicks in excess of 1/8" will be evaluated by engineering. Unacceptable nicks mun be removed or repaired prior to restoration of the containment wall structure. Welding is the only way to repair the base metal to restore its original cross sectional area.

3 - 98

itevision 0 RELIEF REQUEST: NR-33 (Page 2 of 3)

The structural design approach used for the Braidwood Unit 1 Containment Structure is that the capacity to resist design basis loads is provided by the unbonded post-tensioning system. The.

post-tensioning system provides the effect of an inward pressure which counteracts the effects created by a postulated accident. The reinforcing steel bar at the Braidwood Unit 1 is classified as temperature steel and is added to the containment for crack control and to minimize shrinkage and temperature effects and does not have the primary role ofcarrying applied loads.

Using material certifications, it has been determined that the reinforcing steel in the area of the .

construction opening at Braidwood has_a carbon equivalent in excess of 0.55% (approximately

- 0.65%). Therefore, reliefis requested from the 1973 ASME Section III, Division 2 code requirement CC-4334.7 which only allows for welding of reinforcing steel which has carbon equivalents equal to or less than 0.55%.

' The Steam Generator Replacement Project at the R. E. Ginna Nuclear Power Station used AWS

. DI.4-92 for base metal repair and structural welding of their reinforcing steel as part of their -

containment restoration activities. Ginna has a non-ASME containment structure and, as such, was using AWS standards consistent with their original construction design basis. The containment dome at Ginna is a conventional reinforced concrete structure with #18 reinforcing steel bars. In the Ginna case, the reinforcing steel plays a primary role of carrying applied loads.

Compliance with the AWS standards provided Ginna with an acceptable level of safety as per their design and licensing basis.

I Since the Braidwood containment reinforcing steel does not have the primary role of carrying applied loads, the Ginna application of AWS Dl.4-92 for the welding of containment reinforcing steel represents a more safety significant use than what is required for Braidwood. Use of AWS Dl.4-92 will provide a level of safety and quality equivalent to the current Braidwood design and licensing basis.

Welding procedures and welding personnel to be used in the containment restoration activities at Braidwood following Steam Generator Replacement will meet the requirements of AWS Dl.4-

92. Procedural controls that comply with the requirements of Table 5.1, Table 5.2, and Paragraph 5.7 of AWS Dl.4 92 for welding electrode selection, storage, pre-heat, and interpass temperatures ensure that the requirements of the AWS standard will be met for base metal weld I repair of reinforcing steel with carbon equivalent content in excess of 0.55%.

-By employing approved procedures and using welders qualified per the AWS DI.4-92 code requirements, weld repair for damaged reinforcing steel will be performed to a level of quality and safety which has previously been accepted for use in containment structure applications.

.3 - 99

' ~

U 2

RELIEF REQUEST: NRo33 (Page 3 of 3) i If strict compliance with 10CFR50.55a(g) were required, base metal repair by welding would not be allowed by 1973 ASME Section III, Division 2 for the Braidwood Unit I reinforcing steel due to its carbon equivalent content being in excess of 0.55%. As a result, if a stub end of the reinforcing steel (where Cad-weld splices will be attached during the restoration of the containment opening) were nicked in excess of 1/8" and evaluated as unac::eptable, additional chipping of concrete to expose an undamaged section of reinforcing steel would be required.

Additional concrete chipping unnecessarily increases the size of the construction opening and increases the possibility of additional damage to more reinforcing steel, thus requiring further expansion of the opening size. The expansion on any side of the containment opening would require extension of the elevated work platform being used for the containment opening restoration. This would require additional safety measures and structures to ensure worker safety.

Additional concrete removal to avoid reinforcing steel weld repair would not provide any compensating increase in quality or safety since, in either case, the containment is being restored to its original design basis condition. Weld repair of containment reinforcing steel in accordance with AWS Dl.4 92 would avoid unnecessary removal of additional containment concrete and potential undesirable work conditions. Accordingly, use of AWS DI.4-92 is requested based on the hardship provisions of 10CFR50.55a(a)(3)(ii). .

I PROPOSED ALTERNATE PROVISIONS The American Welding Society (AWS) Standard DI.4-92 is proposed as an alternative to perform i

l weld repair of containment reinforcing steel. AWS Dl.4-92 allows welding of reinforcing steels with any carbon equivalent value provided the welding procedures are approved to the requirements of the standard including the following: low hydrogen electrodes of the appropriate strength level are used, the electrode storage conditions are controlled to preserve their low hydrogen characteristics, and the appropriate minimum preheat and interpass temperatures are maintained.

APPLICABLE TIME PERIOD 4

This n' lief will be required for the Braidwood Unit 1 Steam Generator Replacement Outage which is rcheduled for the Spring Refueling Outage in 1998

- 6PPROVAL STATUS Pending NRC Review.

3 - 100

• "1810" O RELIEF REQUEST NRJ34 COMPONENT IDENTIFICATION ~

Code Class (es): 1

Reference:

10CFR50.55a (g) (ii) (A) (2)

IWB-2500-1, Table IWB-2500-1 Examination Categories: B-A Item Number (s): Bl.11

Description:

Volumetric Examination of Reactor Pressure Ve'ssel (RPV) Circumferential Shell Welds.

Component Number (s): Unit 1 Weld: 1RV-02-002 CODE REQUIREMENT 10CFR50. 55a (g) (ii) ( A) (2) requires that all licenses shall augment their Reactor Pressure Vessel (RPV) examination by implementing once, as part of the inservice inspection interval in effect on September 8, 1992, the examination requirements for the RPV shell welds specified in Item Bl.10 of Examination Category B-A, " Pressure Retaining Welds in Reactor Vessel", in Table IWB-2500-1 of Subsection IWB of the 1989 Edition rf Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code. For the purpose of this augmented examination, " essentially 100%" as used in Table IWB-2500-1 means more than 90% of the examination volume of each weld, where the reduction in coverage is due to interference by another component, or part geometry.

RASIS FOR RELIEF Pursuant to 10 CFR 50.55a (g) (ii) (A) (5), relief is requested from the requirement to examine more than 904 of the examination volume of the RPV circumferential shell weld, 1RV-02-002, on the basis that the alternative to the examination requirements would provide an acceptable level of quality and safety.

10CFR 50. 55 a (g) (6) (ii) (A) ( J ) revokes all relief requests with respect to i volumetric examination coverage for RPV shell welds specified in Item Bl.10.

Portions of a previously granted First Interval relief request, NR-9, ,

addressed limited exam coverage on the Braidwood RPV shell welds.

Augmented examination of the subject RPV shell weld was conducted on Braidwood {

Unit 1 during AIR 06 refuel outage (Spring 1997). During this exam at l Braidwood Unit 1, physical obstructions and geometry prevented ultrasonic (UT) coverage in excess cf 904 of the required volume. The examination of the Lower Shell course-to-Detchman weld, IRV-02-002, is restricted by six (6) core barrel locating lugs welded to the inner surface of the vessel approximately 2.5 inches above the weld centerline (See Attachment 1). These lugs obstruct l the automated UT examination tool from examining the Code required volume of '

weld and base material under and below each lug in both the circumferential and perpendicular scan directions (156' total for all 6 lugs, See Attachment 2, 3,. and 4) .

• Adl weld metal and base material can be examined between the lugs (204' total length between all 6 lugs). The 6 lug interferences limit the examination aggregate volume coverage for the weld and adjacent base metal to approximately 814 of the Code required volume.

1 3 - 101 -

i

Revision 0 ,

RELIEF REQUEST NR-34 (cont.)

Compliance with the-applicable Code requirements may be accomplished by redesigning and modifying the RPV and/or the building structure surrounding the vessel (s). Access for manual inspections from the OD of the RPV is limited because of the close proximity of the building structure to the RPV shell (See Attachment 1).

Examination of the Code required examination volume was completed to the maximum extent practical using alternate UT techniques qualified to the highest standard available. RPV examinations were conducted from the I.D. of the vessel. Access to allow examination from the O.D. (shell side) of these welds is restricted due to the structural concrete surrounding the vessel.

The examination techniques employed have been demonstrated and qualified to the Performance Demonstration Initiativs (PDI) Program which meets the intent

.of the rules of Appendix VIII of the ASME Code,Section XI, 1992 Edition with 1 1993 Addenda. These techniques were used in place of the currently required Section XI, 1983 Edition with Summer 1983 Addenda, techniques (Reference Relief Request NR-29). Although the techniques have been qualified at PDI for single direction scanning, examinations were performed from two directions, when required (i.e., performed from both sides of the weld on the same surface, where feasible).

Strict ASME Section III quality controls were used when designing, fabricating, and installing these RPV welds. Preservice examinations to the fullest extent practical were performed on these welds. Preservice Inspection (PSI) relief request INR-9 was submitted to the Staff and approved for these lug interferences. Comed has recently performed these ultrasonic examinations to the fullest extent practical, i.e. 81%, during the AIR 06 refuel outage and no unacceptable indications to applicable Section XI standards were detected.

The results of the examination provide further assurance that unallowable inservice flaws have not developed in the subject weld. In addition to UT, visual examinations (VT-1 and VT-2) of the weld alsc verifies its integrity.

Thus, the modification of the RPV and/or the building structures to increase examination volume coverage from Bli to more than 90% would incur unnecessary radiological exposure and significant engineering expenses. Braidwood Station believes this course of action is a hardship without a compensating increase in the level of quality and safety.

PROPOSED ALTERNATE PROVISIONS The ultrasonic examination of the Braidwood Unit 1 RPV shell weld, 1RV-02-002, was performed to the maximum extent practical. In conjunction with the partial ultrasonic examination, a supplemental VT-1 of the RPV shell weld was ,

conducted from the interior of the RPV using underwater camera equipment.

APPLICABLE TIME PERIOD This relief request will be required for the First Ten Year Inspection Interval.

APPROVAL STATUS Pending NRC review.

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Revision 0 RELIEF REQUEST NR COMPONENT IDENTIFICATION Code classies): 1

Reference:

- IWB-2500-1, Table IWB-2500-1 Examination Categories: B-A Item Nunber (s) : Bl.ll

Description:

Volumetric Examination of Reactor Pressure Vessel (RPV)-Circumferential Shell Weld.

Component Number (s ) : Unit 2 Weld: 2RV-02-002 CODE REQUIREMENT Subsection IWB, Table IWB-2500-1, Examination Category S-A, Item Number B1.11 requires essentially 100% volumetric examination of the RPV Circumferential Shell welds as detailed in Figure IWB-2500-1.

RASIS FOR RELIEF Comed's Braidwood Nuclear Power Station Unit 2 conducts ISI activities in accordance with the 1983 Section XI Edition, 1983 Summer Addenda as required by Title 10, code of Federal Regulations, Part 50, Section 55a, Paragraph (g), Subparagraph (4) [10 CFR 50.55a(g)(4)]. Pursuant to 10 CFR 50.55a (g) {5) (iii), relief is requested on the basis that the code requirement to examine essentially 100% of the weld volume is impractical due to geometric interference.

10CFR 50.55 a (g) (6) (ii) (A) (1) revokes all relief requests with respect to

. volumetric examination coverage for welds specified in Item Bl.10. Portions of a previously granted relief request, NR-9, addressed limited exam coverage l on the Braidwood RPV shell welds.

Examination of the subject RPV shell weld was conducted on Braidwood Unit 2 during A2R06 refuel outage (Fall 1997). During this exam at Braidwood Unit 2, physical obstructions and geometry prevented ultrasonic (UT) coverage in excess of 90% of the required volume. The examination of the Lower Shell Course-to-Dutchman weld, 2RV-02-002,-is restricted by six (6) core barrel locating lugs welded to the inner surface of the vessel approximately 2.5 inches above the weld centerline (See Attachment 1). These lugs obstruct the automated UT examination tool from examining the Code required volume of the weld and base material under and below each lug in both the circumferential and perpendicular scan directions (156* total for all 6 lugs, see Attachment 2, 3 and 4). All weld metal.and base material can be examined between the

-lugs (204' total length between all 6 lugs). The 6 lug interferences limit the examination aggregate volume coverage obtained for the weld and adjacent base metal to approximately 81% of the Code required volume.

' Compliance with the applicable Code requirements may be accomplished by redesigning and modifying the RPV and/or the building structure surrounding the vessel (s) .

Access for manual inspections from the OD of the RPV is limited because of the close proximity of the building structure to the RPV shell (See Attachment 1).

Examination of the Code required examination volume was completed to the maximum

. extent practical using alternate UT techniques qualified to the highest standard 3-107 i

p ,

Revision 0

, RELIEF REQUEST NR-35 (cont.)

available. RPV examinations were conducted f rom the I.D. of the vessel. Access to allow examination from the O.D.~ (shell side) of these welds is restricted due to the structural concrete surrounding the vessel. The examination techniques employed have been demonstrated and qualified to the Performance Demonstration Initiative (PDI)' Program which meets the intent of the rules of Appendix VIII of the ASME Code,Section XI, 1992 Edition with 1993 Addenda. These techniques were used in place of the currently required Section XI, 1983 Edition with Summer 1983 Addenda, techniques (Reference Relief Request NR-29) . Although the techniques have been qualified at PDI for single direction scanning, examinations were performed from two directions,

.when required (i.e., performed from both sides of the weld on the same surface, where feasible).

Strict ASME Section III quality controls were used when designing, fabricating, and installing these RPV welds. Preservice (PSI) examinations to the fullest extent.

practical were performed on these welds. PSI relief request 2NR-9 was submitted to the Staff and approved for these lug interferences. Comed has recently performed these ultrasonic examinations to the fullest extent practical, i.e. 81%, during the A2R06 refuel outage and no unacceptable indications to applicable Section XI standards were detected. The results of the examination provide further assurance that unallowable inservice flaws have not developed in the subject weld. In

_ addition to UT, visual examinations (VT-1 and VT-2) of the weld also verifies its integrity. Thus, the modification of the RPV and/or the building structures to increase examination volume coverage from 81% to essentially 100% would incur unnecessary radiological exposure and significant engineering expenses. Braidwood Station believes this course of action is a hardship without a compensating increase-in the level of quality and safety.

PROPOSED ALTERNATE PROVISIONS  ;

i The ultrasonic examination of the Braidwead Unit 2 RPV shell weld, 2RV-02-002, was performed to the maximum extent pract3.al. In conjunction with the partial ultrasonic examination, a supplemental VT-1 of the RPV shell weld was conducted from the interior of the RPV using undetweter camera equipment.

APPLICABLE TIME PERIOD This relief request will be required for the First Ten Year Inspection Interval.

APPROVAL STATUS Pending NRC Review.

3-108

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Pevrsion t RELIEF REQUEST NR-36 Cl4PONENT IDENTIFICATION Code Classles): 1 References 10CFR50.55aIg>(ii)(A)(2)

IWB-2500-1, Table IWB-2500-1 Examination Categories: B-A Item Number (s): Bl.11

==

Description:==

volumetric Examination of Peactor Pressure ' Jesse; (RPV) Circumferential Shell Welds.

Component Number (s): Unit 2 Weld: 2RV-02-002 CODE REQUt NT 10CFR50. 55a t g) (ii) ( A) (2) requires that all licenses shall augment their Reactor Pressure Vessel (RPV) examination by implementing once, as part of the inservice inspection interval in effect on September 8, 1992, the examrnatien requirements for the RPV shell welds specified in Item Bl.10 of Examination Category B-A, " Pressure Retaining Welds in Reactor Vessel", in Table IWB-2500-1 of subsection IWB of the 1989 Edition of Section XI, Division 1, of the ASME Boiler and Pressure Vessel Code. For the purpose of this augmented examination, " essentially 100%" as used in Table IWB-2500-1 means more than 90t of the examination volume of each weld, where the reduction in coverage is due to interference by another component, or part geometry.

BASIS FOR RELIEF Pursuant to 10 CFR 50.55a (g) (ii) U4) (5), relief is requested from the requirement to examine more than 90% of the examination volume of the RPV circumferential shell weld, 2RV-02-002, on the basis that the alternative to the examination requirements would provide an acceptable level of quality and safety.

10CFR 50.55 a (g) (6) (ii) U4) (1) revokes all relief requests with respect to volumetric examination coverage for RPV shell welds specified in Item Bl.10.

Portions of a previously granted First Interval relief request, NF-9, addressed limited exam coverage on the Braidwood RPV shell welds.

Augmented examination of the subject RPV shell weld was conducted on Braidwood Unit 2 during A2R06 refuel outage (Fall 1997). During this exam at Eraidwood Unit 2, physical obstructions and geometry prevented ultrasonic (UT) coverage in excess of 904 of the required volume. The examination of the Lower Shell Course-to-Dutchman weld, 2RV-02-002, is restricted by six (6) core barrel locating lugs welded to the inner surface of the vessel approximately 2.5 inches above the weld centerline (see Attachment 1). These lugs cbstruct the automated UT examination tool from examining the Code required volume of weld and base material under and below each lug in both the circumferential and perpendicular scan directions (156* total for all 6 lugs, See Attachment 2, 3, and 4). All weld metal and base material can be examined between the lugs (204* total length between all 6 lugs). The 6 lug interferences limit the examination aggregate volume coverage for the weld and adjacent base metal to approximately 814 of the Code required volume.

3 - 113

i l

(

I Revision RELIEF REQUEST NR-36 (cont.)

Compliance with the applicable Code requirements may be a t:cmplis hea . ' .

redesigning and modifying the RPV and/or the building structure surz:u ning l the vessel (s). Access for manual inspections frem the CD cf the FJ' ::

limited because of the close prcximity of the building structure :: tha EPV shell (See Attachment li.

Examination of the C de' required examinatien volume was : mpleted :: n.-

maximum extent pract1:a1 using alternate UT techniques qualified :: : -

l l

highest standard avaliable. RPV examinations were :enducted f rcm :ne :... cf l the vessel. Access to allow examination f rom the O.D. ishell side) ; these l

welds is restricted due to the structural concrete surrounding the versel.

The examination techniques employed have been demonstrated and qualifiea to the Performance Dem:nstration Initiative (PDI! Tregram which meets :ne ;ntent of the rules of Appendix VIII of the ASME Code,Section XI, 1992 Ed :::n with 1993 Addenda. These te.:hniques were used in place of the current 1; r e gt.i re d Section XI, 1983 Edition with Summer 1983 Addenda, techniques (Eeferenre Relief Request UR-291. Although the techniques have been qualified at TLI fer single direction scanning, examinations were performed from two directions, when required (i.e., performed f cm both sides of the weld on the san +

surface, where feasibler.

Strict ASME Section III quality controls were used when designing, fabricating, and installing these RPV welds. Preservice examinations t: the fullest extent practical were performed on these welds. Preservice In2rection (PSI) relief request 2MP-9 was submitted to the Staff and approved f : these lug interferences. CemEd has recently performed these ultrasonic examinations to the fullest extent practical, i.e. 814, during the A2R06 refuel cutage and no unacceptable indications to applicable Section XI standards were detected. l The results of the examination provide further assurance that unallowable inservice flaws have not developed in the subject weld. In addition tc UT, visual examinations (VT-1 and VT-2) of the weld also verifies its integrity. 1 Thus, the modification of the RPV and/or the building structures to increase l examination volume coverage from 814 to more than 904 would incur unnecessary j radiological exposure and significant engineering expenses. Braidwood Station  !

I believes this course of action is a hardship without a compensating Ancrease in the level of quality and safety.

PROPOSED ALTERNATE PROVISIONS The ultrasonic examination of the Braidwood Unit 2 RPV shell weld, 2RV-02-002, was performed to the maximum extent practical. In conjunction with the  ;

partial ultrasonic examination, a supplemental VT-1 of the RPV shell weld was  ;

conducted from the interior of the RPV using underwater camera equipment.  ;

APPLICABLE TIME PERIOD This relief request will be required for the First Ten Year Inspe::::n Interval. )

APPROVAL STATUS Pending URC review.

l i

3 - 114 1

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! Revision 0 NR-36 I

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Pevisien 0 NR-36 Attachment 3 (Drawing not t scale)

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3 - 118 j

Fevision C RELIEF REQUEST NR-37 COMPONENT IDENTIFICATIOLI Code Class (es): 1 Peference: IWB-2500-1 Examination Categories: B-D Item Humbers: B3.90

==

Description:==

Volumetric Examination of Peactor Pressure Vessel No::le to Vessel Welds component Number (s): Unit 2 Welds:

2RV-01-006, 2RV-01-009, 2RV-01-010, 2RV-01-013 CODE REQUIREMENT Subsection IWB, Table IWB-2500-1, Examination Category B-D, Item B3.90 requires essentially 100+ volumetric examination of the region described in Figure IWB-2500-7 for Reactor Pressure Vecsel (RPV) nozzle-to-vessel welds.

BASIS FOR RELIEF Comed's Braidwood Nuclear Power Station Unit 2 conducts ISI activities in accordance with the 1983 Section XI Edition, 1983 Summer Addenda as required by Title 10, Code of Federal Regulations, Part 50, Section 55a, Paragraph (g), Subparagraph (4) [10 CFR 50.55a (g) (4 ) ) . Pursuant to 10 CFR 50.55a (g) (5) (iii), relief is requested on the basis that the code requirement to examine essentially 1004 of the welds' volume is impractical due to geometric interference.

All RPV welds are examined using remotely operated underwater volumetric inspection techniques. Underwater volumetric inspection techniques are utilized to meet ALARA concerns due to the high radiation levels in these areas. The outlet (Hot Leg) no::les are constructed with an integral extension on the I.D. surface which mates with the internal core barrel. The extension provides a flow path for reactor coolant from the core into the hot leg nozzles. The integral extensions partially obstructs the circumferential scan for reflectors transverse to the weld (Reference ). The integral extetsion, that confines the movement of the transducer package, along with the curvature . f the RPV shell conbine to limit full Code volume coverage when scanning in the dire tion parallel to the weld (Reference Attachment 2). This configuration limits the examination aggregate volume coverage cbtained for each weld and adjacent base metal to approximately 81; instead of the Code required essentially 1004 examination coverage.

Compliance with the applicable Code requirements may be accomplished by redesigning and modifying the ID of the Hot Leg no::les and/or the building structure surrounding the.RPV at the nozzles' elevation. Braidwood Unit 2 RPV was designed with a RPV shield wall (Reference Attachment 3 and 4). This wall impedes access to the OD of the RPV shell for insulation removal, surface preparation and ultrasonic inspection. Mod,1fying the nozzle Ib surface would incur extensive radiation exposure to station personnel and could be detrimental to the ecmponent.

When designing, fabricating and installing these welds, strict ASME Section III quality controls and procedures were used that minimized the introduction of fabrication l l

defects. Additionally, the periodic VT-2 examinations in accordance with the l

3-119

Fevisi n n RELIEF REQUEST NR-37 ' (cont. )

requirements of ASME Section XI, Table IWB-2500-1, Examination Cate;:r; S-F and applicable P.eactor CO:lant system monitoring requirements specified a the Te:hni:a1 Specifications will provide reasonable assurance of continued stru:tural integr ty of the Reactor Vessel, Comed has recently performed these volumetri: examinati ns to the fullest extent practical, i.e. 81 , during the A2R06 refuel eutage and nc recordable indications (URI) were detected. The URI results of tne examination provide further assurance that unacceptable inservi:e flaws have not developed in the subject welds. Thus, the modificatien of the n ::les and/or the ruilding structure to in:rease examination volume coverage fr:m B1- to essentially 100- w:uld incur unnecessary radiological exposure and significant engineering ::sts without a compensating increase in the level of quality and safety.

inOPOtED ALTERNATE PROVISIONS The Peactor Vessel outlet (H t Leg) noz:le welds will be examined t: :ne fullest extent practical using the.available underwater volumetric inspertic: techniques.

APPLICABLE TIME PERIOD This relief request will be required for the First Ten Year Inspe:ti:n Interval.

APPROVAL STATUS Pending NRC Review.

I I

J

)

3-120 J

Revision 0 NR-37 ATTACHMENT 1 (Drawing not to scale)

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3-121 j t

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Interference Detail 3-123

1 Revision C RELIEF REQUEST NR-38

~ COMPONENT IDENTIFICA' TION Code Class (es): 1

'Peterences' IWB-2500-1, Table IWB-2500-1 Examination Categories: B-A Item Humber(s): Bl.21 Descriptions Volumetric Examination of Reactor Pressure Vessel (RPV) Lower Head Circumferential Wela.

. Component !! umber (s ) : Unit 1 Weld: 1RV-02-001 Unit 2 Weld: 2RV-02-001

' CODE REQUIREMENT-subsection IWB. Table IWB-2500-1, Examination Category B-A, item Humber Bl.21 l requires essentially 100t volumetric examination of the accessible length of the Reactor Pressure vesse] (RPV) Lower Head Circumferential weld as detailed in Figure IWB-2500-3.

RAsIs FOR RELIEF 1 Comed's Braidwood Nuclear Power Station Units 2 conducts ISI activities in

'accordance with the 1983 Section XI Edition, 1983 Summer Addenda ~as required by Title 10, code of Federal Regulations, Part 50, Section 55a, Paragraph (g),

Subparagraph:(4) [10 CFR 50.55a tg) (4)] . Pursuant to 10 CFR 50.55a (g) (5) (iii),

relief isltequested'on the basis that the code requirement to examine essentially 100% of the accessible length of weld volume is impractical due to geometric interference.

Examination lof the' subject RPV Lower Head Circumferential weld was conducted

'on Braidwood Unit l'during AIR 06 refuel outage (Spring 1997) and on Braidwood Unit 2 during A2R06, refuel outage (Fall 1997). During these exams at Braidwood Units - l' and 2, physical obstructions and geometry prevented' ultrasonic (UT) coverage in excess of 90% of the accessible length of ' weld volume. .The examination of the Lower Head Circumferential welds,11RV-02-001 and-2RV-02-001,.is restricted by instrumentation nozzle penetratiens (See Attachment'2). These instrumentation nozzle penetrations obstruct the

' automated UT examination tool from examining the accessible length of Code weld and base material volume in both.the circumferential and perpendicular scan directions. ' All weld metal and base material can be examined between the instrumentation' nozzle penetrations. The instrumentation nozzle penetrations interferences limit the examination aggregate volume coverage obtained for the subject-welds and'adjacentLbase' metal to approximately 864 of the code

. required volume.- ,

compliance with.the applicable Code-requirements may be accomplished by redesigning and modifying the.RPV and/or the building structure surrounding the vessel (s).

Access for manual inspections.from the CD of the RPV is limited because of the close

. proximity lof the building structure to the RPV shell (See Attachment 1).

, l Examination'of;the Code required examination volume was completed to the maximum-extentfpractical1using alternate UT; techniques' qualified to the1 highest standard 3-125

e a Fevision O.

RELIEF REQUEST MR-38 (cont.)-

available. RPV examinations were conducted from the I.D..of the vessel. Access to allow examination from the 0.D. of these. welds is restricted due to the' structural ~

concrete surrounding the-vessel. The examination techniques employed have been demonstrated and qualified to the Performance-Demonstration Initiative (PDI) Program-which meets:the intent of the rulesgot Appendix VIII of the ASME Code,'Section XI, 1992 Edition with 1993 Addenda. These techniques were used in place of the 4 currently required Section XI, 1983 Edition with Summer 1983 %ddenda,. techniques

(Reference. Relief Request NR-29). Although the techniques have been qualified at PDI-for' single direction. scanning, examinations were performed-from.two directions,-

when required (i.e., performed from both sides of the weld on the same surface, where-feasible).

Strict ASME'Section III quality centrols were used when designing, fabricsting and installing these RPV welds. Preservice (PSI) examinations to the fullest extent practical were performed on these welds. PSI relief request 1 (2 ) NR-9. were ; submitted to.the Staff and' approved for these lug interferences. Comed has recently' performed these ultrasonic examinations to the fullest extent practical, i.e. 86t, during;the

.A1R06 and A2R06 refuel outages and no unacceptable indications to applicable Section~

' ~XI standards were detected. .The results of these examinations provide further assurance that unallowable inservice flaws have not deveicped in the subject welds.

Id addition to'UT, visual examinations (VT-2) of the-weld also verifies its integrity. Thus, the modification of the RPV and/or the building structures to

-increase examination volume coverage from 86't to essentially 100e would incur

unnecessary. radiological exposure and significant engineering expenses. 'Braidwood Station believes this course of action is a hardship without a compensating increase in the level of. quality and safety.

PROPOSED ALTEMatATE PROVISIosts

.The' ultrasonic examination of the Braidwood Units 1 and 2 Lower Head Circumferential welds, 1RV-02-001 and 2RV-02-001, will be performed to the maximum extent practical.

, ~ APPLICABLE TIME PERIOD This relief request will be required for the First Ten Year Inspection Interval.

APPROVAL STATUS Pending NRC Review.

6

'3-126-

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