ML20113C936
| ML20113C936 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 02/15/1985 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20113C935 | List: |
| References | |
| NUDOCS 8504120145 | |
| Download: ML20113C936 (9) | |
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COMMONWEALTH EDISON BYRON NUCLEAR POWER STATION UNIT -1 PRESERVICE INSPECTION
SUMMARY
REPORT FEBRUARY 15, 1985 Station Address:
Byron Nuclear Power Station 4450 N. German Church Road Byron, Illinois 61010 Owner's Address:
Cornmonwealth Edison Company 72 West Adams Street Chicago, Illinois 60606 8504120145 050223 PDR ADOCK 05000454' G
9 BYRON UNIT 1 PSI
SUMMARY
REPORT The Preservice Inspection (PSI) of the Conunonwealth Edison Co. (CECO), Byron Nuclear Power Station Unit I was performed in compliance with the rules and regulations of Section XI Division 1, " Rules for Inservice Inspection of Nuclear Power Plant Components", of the American Society of Mechanical
-Engineers (ASME) Boiler and Pressure Vessel Code, 1977 Edition and Addenda through the Summer 1978 Addenda, as per the requirements of the Title 10, Part 50.55a of the Code of Federal Regulations (10CFR50.55a).
The Nondestructive Examination (NDE) PSI Program Plan for Class 1, 2 and 3 components was developed in accordance with the requirements and intent of Subsections IWA, IWB, IWC and IWD,Section XI, Division 1, of the ASME Code.
In an effort to increase the level of confidence in system integrity, CECO elected to expand the scope of preservice examinations for Class 2 systers, as delineated in Article IWC-2000, to include 100% of all non-exempt Class 2 piping systems and volumetric examination of pressure retaining welds in the four (4) Unit 1 Steam Generators.
In addition to the ASME Section XI requirements of examination, certain augmented inspections were required by the Nuclear Regulatory Commission. The Byron Unit 1 Augmented PSI requirements included: examination of High Energy lines in conformance with NUREG 0800, " Standard Review Plan", Section 6.6;
" Eddy Current Examination of Nonferromagnetic Steam Generator Heat Exchanger Tubing" in conformance with Section XI Appendix IV, and Regulatory Guide 1.83
" Revision 1, " Inservice Inspection of Pressurized Water Reactor Steam Generator Tubes"; and implementation of Regulatory Guide 1.150 " Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations".
Where the Code and/or Augmented requirement (s) was/were deemed to be impractical, specific request for relief were developed and submitted for the Commission's review.
Identification of Examination Requirements The PSI Program contains the NDE program tables. These tables are presented in a tabular format consistent with Tables IWB, IWC, and IWD-2500 of the ASME Code. The NDE program tables include the corresponding code category, item number, and component / weld selection in conformance with Class 1, 2 and 3 examination requirements and intent of Subsection IWA, IWB, IWC and IND of Section XI.
Program notec and relief requests are identified in the remarks column.
Exempted Components _
Class 1, 2, and 3 components (or parts of components) which were not included in the NDE program tables and are exempt from examination, as specified in Section XI Paragraph IWB and IWC-1220 " Components Exempt from Examination" and Table IWD-2500-1 " Examination Requirements", are identified in the NDE Program Plan together with the technical justification (s) for exempting the component / system. (0661M) i i
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Irplementation of PSI Program The implementation of the Byron Nuclear Power Station Unit 1 NDE PSI Program Plan, Revision 3 dated June 30, 1984, was performed by Ebasco Services Inc.
and Rockwell International in cooperation with Commonwealth Edison Co., or their designee. Specifically, the Ebasco PSI scope included Class 1 and 2 non-exempt systems and components requiring volumetric, surface, VT-1 visual examinations or a combination thereof, including Reactor Pressure Vessel manual ultrasonics (UT) and eddy current (ET) examinations of Steam Generator tubing.
. Visual Examinations (VT-2, VT-3 and VT-4) of Class 1, 2 and 3 components and their supports, identified under the PSI Program Plan and/or requiring reexamination (s) due to repairs or replacements under Section XI were performed by Commonwealth Edison Company, or their designee.
All NDE examinations, including evaluation of results, were performed by personnel qualified to the appropriate levels in the various NDE methods, per the requirements of the American Society for Nondestructive Testing SNT-TC-1A, 1975. Personnel performing VT-1 were qualified in accordance with the comparable levels of competency as defined in ANSI N45.2.6, 1973. The implementing NDE procedures were developed and qualified in conformance with ASME Sections V and XI, 1977 Edition through Summer '78 Addenda.
Visual examinations (VT-3 and VT-4) of Class 1, 2, and 3 component supports including evaluation of results, were performed by examiners qualified to ANSI N45.2.6-1973 and/or ASME Section XI, 1980 Edition through Winter 1981 Addenda. Visual examinations were performed in accordance with ASME Section XI Procedures, or ASME Section III procedures which were evaluated to be in excess of Section XI requirements by the Authorized Inspection Agency.
NDE Examinations All PSI nondestructive examinations, including evaluation of flaw indications, were performed in accordance with the requirements stipulated under Section XI Subarticle IWA-2200, " Examination Methods".
The Preservice Inspections were performed from May 8, 1981 to July 6, 1984.
The Authorized Nuclear Inspector was Mr. J. Becker from Hartford Steam Boiler Inspection and Insurance Company of Hartford, Connecticut, whose address is 120 South Riverside Plaza, Suite 416, Chicago, Illinois, 60606.
Regulatory Guide 1.150 Interpretation and Implementation Commonwealth Edison, Rockwell International, and Ebasco jointly reviewed Regulatory Guide 1.150 " Ultrasonic Testing of Reactor Vessel Welds During Preservice and Inservice Examinations", in order to formulate a plan to provide compliance with the regulatory guide within the constraints of technology, equipment, and schedule'. The degree to which the regulatory guide was implemented and the results obtained are reflected in the PSI Final Report's document 445ER000001 "Preservice AUT of Byron Unit 1 Reactor Vessel Welds, Engineering Report".
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j Deficiencies A deficiency report was initiated if necessitated by the following conditions:
A.
If reportable indications were revealed during nondestructive examinations.
B.
If inaccessible components were identified which required possible redesign and/or relief from NDE requirements.
C.
If any possible testing deviation occurred (i.e., components that do not lend themselves to surface and/or volumetric examination due to their material properties and/or geometrical configuration).
Summary of Relief Requests In instances where specific requirements of Section XI were deemed to be impractical requests for relief have been docketed, as described in 10CFR50.55a (g) (5) (iv), and approved. The following is a brief description of the approved relief requests:
NR-1:
Inaccessible circumferential weld on Chemical & Volume Control letdown line from Reactor Coolant Loop 3.
NR-2:
Inaccessible welds due to saddle plates on Main Steam, Safety Injection, and Residual Heat Removal.
NR-3:
Limited ultrasonic examination of Cast Stainless Steel to Cast Stainless Steel welds on Reactor Coolant System due to poor acoustic properties.
NR-4:
Limited ultrasonic examination of Cast Stainless Steel to Cast Carbon Steel welds on Reactor Coolant System due to poor acoustic properties.
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NR-5:
Limited ultrasonic examination of Safety Injection welds due to component geometry.
NR-6:
Limited ultrasonic examination of Cast Stainless Steel to Stainless Steel welds on Reactor Coolant System due to poor acoustic properties.
NR-7:
Disassembly of Class 1 valves for internal visual examinations not commensurate with the increased level of safety achieved.
NR-8:
Limited ultrasonic examination of Cast Stainless Steel to Stainless Steel (Safe-End) welds on Reactor Coolant System due to poor acoustic properties.
NR-9:
Limited ultrasonic examination of Reactor Pressure Vessel welds due to geometry and permanent restraints.
NR-10:
Limited ultrasonic examination of Reactor Pressure Vessel Nozzle welds due to internal geometry. (0661M)
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NR-ll:
Limited ultrasonic examination of Letdown Heat Exchanger shell circumferential weld due to flange bolting.
NR-12:
Limited ultrasonic examination of Excess Letdown Heat Exchanger shell circumferential veld due to four branch connections.
NR-13:
Limited ultrasonic examination of Pressurizer and Steam Generator Primary Nozzle welds due to geometric constraints and clad inner surfaces.
NR-14:
Limited ultrasonic examination of Steam Generator Secondary Nozzles and Residual Heat Exchanger Nozzles due to geometric constraints.
NR-15:
Limited surface examinations on Containment Spray, Chemical and Volume Control, and Residual Heat Removal Pump Supports due to support geometry.
NR-Note 5:
Visual examination of Reactor Coolant Pump internal surfaces not performed. Credit was taken for the manufacturer's surface examinations. The Reactor Coolant Pumps are integrally cast with no pump casing welds.
NR-Note 11: Limited ultrasonic examination of Excess Letdown Heat Exchanger shell weld and Safety Injection piping weld due to geometric constraints.
Examination Summary:
The following is a summary of the examinations conducted during the Byron Unit 1 PSI. Please refer to the Ebasco PSI Final Report for specific information.
Piping Reactor Coolant:
A total of 930 pressure retaining welds in the Reactor Coolant System were examined. Of these, 240 welds were examined by volumetric and surface methods and 690 welds were examined by a surface method only. Volumetric examinations did not reveal any reportable indications. Reactor Coolant piping surface examinations revealed 15 welds with reportable indications. All 15 welds were subsequently repaired and reinspected with acceptable results.
A total of 775 visual examinations of pressure retaining bolting (VT-1) were performed with no reportable indications.
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Safety Iniection:
A total of 973 pressure retaining welds in the Safety Injection System were examined. Of these, 665 welds were examined by volumetric and surface methods and 308 welds were examined by a surface method only.
Volumetric examinations revealed 2 welds with reportable indications. The indications were determined to be acceptable, as is, based upon evaluation. Safety Injection piping surface examinations revealed 16 welds with reportable indications. All 16 welds were subsequently repaired and reinspected with acceptable results.
A total of 9 integrally welded attachments were examined by a surface method with no reportable indications.
A total of 464 visual examinations of pressure retaining bolting (VT-1) were performed with no reportable indications.
Pressurizer Surge:
A to".31 of 156 pressure retaining welds in the Pressurizer Surge System were examined. Of these, 92 welds were examined by volumetric and surface methods and 64 welds were examined by a surface method only. Volumetric examinations did not reveal any reportable indications. Pressurizer Surge piping surface examinations revealed 2 welds with reportable indications.
The two welds were subsequently repaired and reinspected with acceptable results.
A total of 152 Visual examinations of pressure retaining bolting (VT-1) were performed with no reportable indications.
Residual Heat Removal:
A total of 480 pressure retaining welds in the Residual Heat Removal System were examined. Of these, 66 welds were examined by volumetric and surface methods and 414 were examined by a surface method only. Neither volumetric nor surface examinations revealed any reportable indications.
A total of 9 integrally welded attachments were examined by a surface method with no reportable indications.
A total of 72 visual examinations of pressure retaining bolting (VT-1) were performed with no reportable indications.
l Chemical and Volume Control:
A total of 278 pressure retaining welds in the Chemical and Volume Conbrol System were examined. All 278 welds were examined by a surface method with no reportable indications.
I One (1) integrally welded attachment was examined by a surface method with no reportable indications.
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A total of 98 visual examinations of pressure retaining bolting (VT-1) l were performed with no reportable indications.
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i Containment Spray:
A total of 342 pressure retaining welds in the Containment Spray System were examined. All 342 welds were examined by a surface method with no reportable indications.
Main Steam:
A total of 311 pressure retaining welds in the Main Steam System were examined. Of these, 275 were examined by volumetric and surface methods and 36 welds were examined by a surface method only. Volumetric examinations revealed 5 welds with reportable indications. Of these, 4 welds were further evaluated and found to be acceptable, as is.
The 1 weld was repaired and subsequently reinspected with acceptable results.
Main Steam piping surface examinations. revealed 1 weld with reportable indications. The 1 weld was subsequently repaired and reinspected with acceptable results.
Feedwater:
A total of 332 pressure retaining welds in the Feedwater System were examined. Of these, 156 welds were examined by volumetric and surface methods and 176 welds were examined by a surface method only. Volumetric examinations did not reveal any reportable indications. Feedwater piping surface examinations revealed 25 welds with reportable indications. All 25 welds were subsequently repaired and reinspected with acceptable results.
A total of 9 integrally welded attachments were examined by a surface method with no reportable indications.
Components:
Reactor Pressure Vessel:
A total of 3 pressure retaining welds in the Reactor Pressure Vessel were examined by manual ultrasonics with no reportable indications.
4 All 54 sets of Reactor Pressure Vessel Closure Head Bolting (studs, nuts, washers) and the ligaments between the stud holes were examined by the required examination methods with no reportable indications.
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All 45 pressure retaining welds in the Control Rod Drive Housing were examined by a surface method with no reportable indications.
Reactor Pressure Vessel (AUT):
A total of 13 pressure retaining welds in the Reactor Pressure Vessel were 4
examined by automated ultrasonics. There was one reportable indication detected during this inspection, and it was subsequently evaluated. No repairs or corrective actions were required as a result of this inspection.
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e, Pressurizer:
A total of 21 pressure retaining welds in the Pressurizer Vessel were examined. All 21 welds were examined by a volumetric method. Volumetric examinations revealed 3 reportable indications in circumferential weld PC-01.
Subsequent evaluations resulted in all 3 welds being acceptable as is.
However, circumferential weld PC-01 will be examined during the first period of the Byron Unit 1 ISI to monitor for further degradation as required by the Nuclear Regulatory Commission.
All 16 sets of manway bolting were visually (VT-1) examined with no reportable indications.
The Pressurizer Vessel support skirt weld was examined by a surface method with no reportable indications.
A total of 44 pressure retaining welds in the Steam Generators were examined. Of these, 12 welds were examined by volumetric and surface methods and the remaining 32 welds were examined by a volumetric method only. Volumetric examinations revealed 11 reportable indications in 8 circumferential welds. Of these, 5 indications were subsequently repaired and reinspected with acceptable results. The remaining 6 indications, after further evaluation and/or reinspections, were found to be acceptable as is.
All 128 sets of primary manway bolting were visually (VT-1) examined with no reportable indications.
Miscellaneous Pressure Vessels:
A total of 24 pressure retaining welds in Miscellaneous Pressure Vessels (refer to Volume 18 of the PSI Final Report for listing) were examined.
Of these,~ 4 welds were examined by a volumetric and surface method and the remaining 20 welds were examined by a volumetric method only. Neither volumetric examinations nor surface examinations revealed any reportable indications.
Component Supports All non-exempt Class 1, 2, and 3 component supports received a preservice examination.
Reportable indications identified during visual examinations were repaired in accordance with ASME Section III. All non-exempt Class 1, 2, and 3 component supports are acceptable as verified by the ANII acceptance of the NIS-1 Form.
Eddy Current:
Eddy current examination of the Unit 1 Steam Generators was conducted during the period from February to May of 1982. The following tubes exhibited obstructions and could not be fully examined:
Steam Generator #1, Row 12, Column 2 Steam Generator #2, Row 12, Column 113 (0661M)
T Both tubes were subsequently plugged by Westinghouse. Other conditions, such as denting, ridging, permeability variations and percent through wall penetration, were also noted, but found to be within acceptable limits.
Following this examination, a modification consisting of expanding the tubes in the feedwater preheater section of each steam generator was performed. Westinghouse then performed eddy current examination on these tubes in October of 1983 and found no reportable indications.
Augmented PSI:
Main Steam:
A total of 11 pressure retaining welds in the Main Steam System were examined volumetrically under the Augmented Program with no reportable indications.
Feedwater:
A total of 47 pressure retaining welds in the Feedwater System were examined volumetrically under the Augmented Program with no reportable indications.
Form NSI-l Owner's Data Report for Inservice Inspections:
The NIS-1 Forms for Byron Unit 1 Preservice Inspection are located in Volume 43 of the Ebasco Preservice Inspection Final Report. Also, the associated NIS-2 Forms for repairs made prior to the NIS-1 completion are located in Volume 44 of the same report.
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