ML20211M629
| ML20211M629 | |
| Person / Time | |
|---|---|
| Site: | Byron  |
| Issue date: | 02/02/1987 |
| From: | COMMONWEALTH EDISON CO. |
| To: | |
| Shared Package | |
| ML20211M626 | List: |
| References | |
| NUDOCS 8702270330 | |
| Download: ML20211M629 (9) | |
Text
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COMMONWEALTH EDISON BYRON NUCLEAR POWER STATION UNIT -2 PRESERVICE INSPECTION
SUMMARY
REPORT FEBRUARY 2, 1987 Station Address:
Byron Nuclear Power Station 4450 N. German Church Road Byron, Illinois 61010 Owner's Address:
Commonwealth Edison Company 72 West Adams Street Chicago, Illinois 60606 0702270330 870202 PDR ADOCK 05000454 G PDR
BYRON UNIT 2 PSK
SUMMARY
REPORT The Preservice Inspection (PSI) of the Commonwealth Edison Co. (CECO), Byron Nuclear Power Station Unit 2, 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 through the Summer 1973 Addenda, as per the requirements of the Title 10, Part 50.55a of the Code of Federal Regulatione (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, Divisi 3n 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 systems, 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 2 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 2 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 requests 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 requiremente and intent of Subsection IWA, IWB, IWC and IWD of Section XI. Program notes and relief requerts are identified in the remarks column.
Exempted Components Class 1, 2, and 3 cocponents (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 IND-2500-1 " Examination Requirements", are identified in the NDE Program Plan together with the technical justification (s) for exempting the component / system.
(1220M/0144M)
Implementation of PSI Progrrm The implementation of the Byron Nuclear Power Station Unit 2 NDE PSI Program Plan, Revision 1 dated February 18, 1.086, 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 / automated 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 wake developed and qualified in conformance with ASME Sections V and XI, 1977 Edition through Summer '78 Addenda.
Visual examinations (VT-2, VT-3 and VT-4) of Class 1, 2, and 3 components and their 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 required procedures, oc ASME Section III required procedures which were evaluated by the Authorized Inspection Agency and found to be in excess of Section XI requirements 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 completed on November 3, 1986. The primary Authorized Nuclear Inspectors were Mr. J. llendricks and Mr. S. Felton, f rom Itartford Steam Boiler Inspection and Insurance Company of llartford, 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 pST Final Report's document, Volume 40, RPV-AUT Engineering Report 445ERC00002.
(1220M/0144M)
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Defici r.ciei ll .
l 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 L
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 l
In instances where specific requirements of Section XI were deemed to be
- ' impractical, requests for relief have been docketed, as described in l 10CFR50.55a (g) (5) (iv), and approved. The following is a brief description of the approved relief requests
, NR-1: Limited ultrasonic examination of Cast Stainless Steel to Cast
[ Stainless Steel welds on Reactor Coolant System due to poor acoustic properties.
NR-2: Limited ultrasonic examination of Cast Stainless Steel to Stainless Steel (Safe-End) welds on Reactor Coolant System due to poor acoustic properties.
NR-3: Limited ultrasonic examination of Cast Stainless Steel to Cast
- Carbon Steel welds on Reactor Coolant System due to poor acoustic properties. -
NR-4: Inaccessible welds due to saddle plates on Main Steam, Safety Injection, and Residual Heat Removal.
NR-5: Inaccessible welds due to wall and floor penetrations on Residual Heat Removal.
NR-6: Inaccessible welds due to support geometry and a floor penetration on Containment Spray.
i NR-7: Disassembly of Class 1 valves for internal visual examinations not commensurate with the increased level of safety achieved.
NR-8: 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-9: Limited ultrasonic examination of Pressuriser and Steam Generator Primary Nossle welds due to geometric constraints and clad inner surfaces.
NR-10: Limited ultrasonic examination of Reactor Pressure Vessel welds 4
due to geometry and permanent restraints.
e (1220N/0144M)
NR-ll: Lisitcd ultreccnic examination cf R :cter Procguro V Icol No: slo walda 'du3 to intornal geometry.
NR-12: Limited ultrasonic examination of Cast Stainless Steel to Stainless Steel welds on Reactor Coolant System due to poor acoustic properties.
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NR-13: Limited ultrasonic examination of Steam Generator Secondary L Nozzles and Residual Heat Exchanger Nozzles due to geometric constraints.
NR-14: Limited surface examinations on Containment Spray, Chemical and Volume Control, and Residual Heat Removal Pump Supports due to support geometry.
NR-15: Limited ultrasonic examination of a Excess Letdown Heat Exchanger shell circumferential weld due to four branch connections.
NR-16: Limited ultrasonic examination of Regenerative Heat Exchanger head circumferential welds due to geometric constraints.
NR-17: Limited ultrasonic examination on a pressurizer line weld due to geometric constraints.
NR-Note 5: Limited ultrasonic examination of Safety Injection welds due to geometric constraints.
Examination Summary:
The following is a summary of the examinations conducted during the Byron Unit 2 PSI. Please refer to the Ebasco PSI Final Report for specific information.
Piping Reactor Coolant:
A total of 843 pressure retaining welds in the Reactor Coolant System were l examined. Of these, 243 welds were examined by volumetric and surface l
methods and 609 welds were examined by a surface method only. Volumetric l examinations did not reveal any reportable indications. Reactor Coolant piping surface examinations revealed 6 welds with reportable indications.
l All 6 welds were subsequently repaired and reinspected with acceptable j results.
i A total of 704 visual examinations of pressure retaining bolting (VT-1)
! were performed with no reportable indications.
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(1220M/0144M)
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- Safe.ty Inicction! ,
A total of 1009 pressure retaining welds in the Safety Injection System -
were examined. Of these,.447 welds were examined by volumetric and surface methods.and 581 welds were examined by a. surface method'only.
Neither volumetric nor surface examinations revealed any reportable indicatings.
A total of 21 integrally welded attachments were examined by a surface j method with no reportable indications. I l
A total of 464 visual examinations of pressure retaining bolting (VT-1) were performed.with no. reportable indications.
Pressurizer Surce:
A total of 112' pressure retaining welds in the Pressurizer Surge System were examined. Of these, 85 welds were examined by volumetric and surface methods and 27 welds were examined by a surface method only. Volumetric examinations did not reveal any reportable indications. Pressuriser Surge i piping surface examinations revealed 1 weld with a reportable indication. I The weld was 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 592 pressure retaining welds in the Residual Heat Removal System were examined. Of these, 72 welds were examined by volumetric and surface methods and 520 were examined by a surface method only.
- Volumetric and surface examinations revealed 1 reportable indication each. Both welds were subsequently repaired and reinspected with acceptable results.
A total of 4 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.
Chemical and Volume Control:
A total of 272 pressure retaining welds in the Chemical and Volume Control System were examined. All 272 welds were examined by a surface method.
with no reportable indications.
l One (1) integrally welded attachment was examined by a surface method with no reportable indications.
A total of 98 visual examinations of pressure retaining bolting (VT-1)
! were performed with no reportable indications.
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i (1220M/0144M)
Containment Spery:
A total ~of 400 pressure retaining welds in the Containment' Spray System-were examined. All 400 welds were examined by a surface method with no j- reportable indications.
Main Steam: . . .
A total of!321 pressure retaining welds in the Main-Steam System were examined. Of these, 293 were examined by volumetric and surface methods and 28 welds were examined by a surface method only. Volumetric
- examinations revealed 4 welds with reportable indications. Of these, 1 weld was further evaluated and found to be acceptable, as is. The 3' remaining welds were 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.
l A total of 32 integrally welded attachments were examined by a surface
' method with no reportable indications.
Feedwater:
t A total of 338 pressure retaining welds in the Feedwater System were examined. Of these, 186 welds were examined by volumetric and surface
- methods and 152 welds were examined by a surface method only. Volumetric
' examinations did not reveal any reportable indications. Feedwater piping surface examinations revealed 2 welds with reportable indications. Both welds were subsequently repaired and reinspected with acceptable results.
}- A total of 4 integrally welded attachments were examined by a surface method with no reportable indications.
Components:
Reactor Pressure Vessel:
l A total of 3 pressure retaining welds in the Reactor Pressure Vessel were
- examined by manual ultrasonics with no reportable indications.
l All 54 sets of Reactor Pressure Vessel Closure Head Bolting (studs, nuts, .
j washers) and the ligaments between the stud holes were examined by the i required examination methods revealing 3 reportable indications in the reactor head studs. The 3 studs with defects were subsequently repaired l
j and reinspected with acceptable results.
All 45 pressure retaining welds in the Control Rod Drive Housing were examined by a surface method with no reportable indications.
1 l
A total of 23 sets of pressure retaining bolts for the Control Rod Drive Housings 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.
l examined by automated ultrasonics with no reportable indications.
(1220M/0144M)
~ ~ 'Preacurizer: .
A total of 21 pressure-retaining welds in the Pressurizer Vessel.were examined. .A11.21 welds were examined by a volumetric method. These examinations revealed 1 indication which is unacceptable to ASME Section XI. Further investigation showed that removal of the indication and subsequent. repair would involve significant risk to the integrity of the vessel. As'a result, Commonwealth Edison does not intend to remove the indication. To ensure that the indication does not grow to a critical size during plant operation, augmented inservice inspections of the indication will be performed.
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 which are unacceptable to ASME Section XI. Further investigation showed that removal and repair of the indications would involve significant risk to
.the integrity of the vessels. As a result, Commonwealth Edison does not intend to remove the indications. To ensure that the indications do not grow to critical sizes during plant operation, augmented inservice inspections of the indications will be performed.
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 17 of the PSI Final Report for listing) were examined.
Of these, 2 welds were examined by a volumetric and surface method, 2 were examined by surface method, and the remaining 20 welds were examined by a volumetric method only. Volumetric examinations revealed 1 reportable indication which was subsequently further evaluated and found to be acceptable as is.
A total of 12 integrally welded attachments were examined by a surface method with no recordable 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.
(1220M/0144M) t-
' ~' Eddy Curr nt Eddy current examination of the Unit 2 Steam Generators was conducted during the months of October and November, 1985. The examinations included 100% of the steam generator tubing full length from the inlet side. The following quantities of indications were identified:
SteakGenerator"A"-9tubeswiththroughwallindications.
Steam Generator "B" - 8 tubes with through wall indications.
Steam Generator "C" - 12 tubes with through wall indications.
Steam Generator "D" - 16 tubes with through wall indications.
Of the above indications, 42 are less than 20% through wall and the remaining 3 are 25%, 28%, and 37% through wall. All 45 indications originated on the outside diameter of the tubes.
A total of 23 tubes were identified as having expansion transition locations above the top of the tube sheets.
Other conditions, such as denting, ridging, and permeability variations were also noted, but found to be within acceptable limits.
No tubes were plugged as a result of the above examinations.
Augmented PSI:
Main Steam:
A total of 14 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 2 Preservice Inspection are located in Volume 49 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 50 of the same report.
(1220M/0144M)