ML20235H278

Forwards Rev 0 to Braidwood Unit 2 Preservice Insp Program & Presents Differences That Exist Between Unit 1 & 2 Preservice Insp Programs

ML20235H278
Person / Time
Site:	Braidwood
Issue date:	09/23/1987
From:	Hunsader S COMMONWEALTH EDISON CO.
To:	Murley T Office of Nuclear Reactor Regulation
Shared Package
ML20235H281	List: ML20235H278 ML20235H282
References
3606K, NUDOCS 8710010002
Download: ML20235H278 (6)

Text

__.

., 7 e'N Commonwealth Edison

) One First Ndional Plaza. Chicago, Illinois

( [

i

) Address R ,p!y to: Post Offic] Box 767

(, - Chicago, Illinois 60690 - 0767 September 23, 1987 Mr. Thomas E. Murley, Director Office of Nuclear Reactor Regulation U.S. Nuclear Regulatory Commission Washington, DC 20555 ATTN: Document Control Desk

Subject:

Braidwood Station Unit 2 Preservice Inspection (PSI) Program NRC Docket No. 50-457

Dear Mr. Murley:

This letter provides a description of the Braidwood Unit 2 PSI Program and presents the differences that exist between the Braidwood Unit 1 PSI Program and the Braidwood Unit 2 PSI Program.

The Braidwood Unit 2 PSI Program consists of a tabular listing of ASME Class 1, 2, and 3 components requiring inspection in a manner consistent with the ASME Section XI examination tables. Included in this listing are a series of Notes and Relief Requests. The Braidwood Unit 2 preservice inspection requirements are the same as those used for Braidwood Unit 1.

'3even (7) Notes and fourteen (14) Relief Requests are included in the program. Attachment A provides a cross reference list for comparison of the Braidwood Unit 1 and Unit 2 PSI program Notes. Attachment B provides a similar listing for Relief Requests. (Relief Request 2NR-8 is included for Braidwood Unit 2 and is not applicable to Braidwood Unit 1).

Braidwood Unit 2, Note 7 is similar to Braidwood Unit 1 Relief Requests INR-14 and 1No-15. To meet Table IWC-2500-1 requirements for surface and/or volumetric inspection, twenty-five percent (25%) of the welds in all other systems not specifically listed in the Table have been examined. Because other Class 2 welds were substituted for the inaccessible welds listed in Note 7, without compromising plant safety (therefore, no relief is being requested), it was felt that a Note and not a Relief Request was appropriate to document these welds.

Unit 2 components exempted from PSI examination per Section XI paragraphs IWB-1220 and IWC-1220, and not subject to examination per Table IWD-2500-1, correspond to the Unit 1 exempted components. There are some differences between units in the components exempted, due to design changes and minor installation differences.

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Y T.E. Murley September 23, 1987 Inspections have been performed to the requirements of the ASME,Section XI Code, 1977 Edition with addenda up to and including the Summer, 1978 Addenda. In addition to the Code required inspections, augmented inspections have been performed on high energy lines and shop examination reports were used in lieu of preservice field examinations for the reactor coolant pump flywheels. j The sample of Unit 2 components and parts of the pressure retaining a boundaries that have received PSI examination is essentially the same as the )

sample examined during the Braidwood Unit 1 PSI program. Volumetric, surface, and visual examinations were performed on Braidwood Unit 2  ;

components per Code or augmented requirements by Commonwealth Edison or  !

contractor (Pittsburgh Testing Laboratory and Phillips, Getschow Company) l personnel. The automated ultrasonic examination of the Braidwood Unit 2 )

reactor vessel was performed by Combustion Engineering to the requirements of Regulatory Guide 1.150. The eddy current examination of the Braidwood steam generator tubing was performed by Conam Inspection.

In June 1984, Commonwealth Edison received approval to utilize ASME Code Case N-401 which allowed the use of the ZETEC Model MIZ-18 instrument (records digitized data on magnetic tape). Similar to the Unit 1 inspection, one hundred percent (100%) of the tubes in the four (4) steam generators were examined using the MIZ-18 instrument.

Commonwealth Edison and Lambert, MacGill, and Thomas (LMT) ,

personnel have recently completed ultrasonic examination of the cast  !

stainless steel elbow and component welds in the Braidwood Unit 2 Reactor Coolant (RC) system. Automated ultrasonic equipment was utilized to examine  ;

four (4) welds in the Unit 7 RC system, two in Loop 2 (2RC-02-18, 2RC-02-31) and two in Loop 3 (2RC-03-04, 2RC-03-17). All other cast stainless /

component welds were examined manually. Both methods utilized 1.0 MHZ refracted longitudinal wave, focused transducers.

This is being submitted for NRC review. Please address any questions concerning this matter to this office.  !

Very truly yours, l S. C. Hunsader l Nuclear Licensing Administrator I cc: S. Sends l

NRC RIII I Braidwood Resident Inspector l

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I ATTACHMENT A Braidwood Unit 2/Braidwood Unit 1 i Preservice Inspection Program Note Cross Reference List Braidwood Unit 2 Braidwood Unit 1 Note No. Note Description Note No.

l 1 Credit was taken for construction 1 y surface examinations in lieu of performing  !

preservice surface examinations.

Commonwealth Edison performed the preservice examinations for Braidwood Units 1 and 2. Contractor surface examination reports were reviewed by the Braidwood ANII to determine if the requirements of Subarticle IWB/IWC-2200 of ASME Section XI were met. If so, the examinations were not redone. No generic substitution was made.

2 volumetric examination of class 2 welds 2 was limited due to component geometry or interference.

It is important to note that the welds listed in Note 2 received a voluraetric examination consisting of axial and circumferential scans; this meets the ASME Code requirements. However it was attempted to complete an axial full "V" path scan, and this was not always attainable due to the listed obstructions / limitations.

l 3 Credit was taken for construction field 3 examinations in lieu of performing Section XI visual (VT-3 and VT-4) examinations.

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Pcga 2 of 2 Braidwood Unit 2/Braidwood Unit 1 Preservice Inspection Program Note Cross Reference List Braidwood Unit 2 Braidwood Unit 1 Note No. Note Description Note No.

1 i

4 Code Case N-401, allowing the use 4 of digitally encoded eddy current data, will be utilized as previously approved by the NRC.

In June, 1984, Commonwealth Edison received approval to utilize ASME Code case N-401. This allowed the use of the ZETEC Model MIZ-18 eddy current instrument, which records digitized data on magnetic  ;

tape, for the preservice examination of the Braidwood steam generator tubes. Appendix IV to Section XI, Article IV-5000 requires that data from steam generator tube examinations be recorded on both magnetic tape and strip chart.

5 Credit was taken for shop examinations 5 in lieu of field examinations on the '

reactor coolant pump flywheels.

J 6 System pressure tests were not executed 6 after Section III hydrostatic tests and )

prior to initial plant startup.

ASME Section XI. 1983 edition does not require a preservice leakage test. VT-2 examinations will be performed, however, in  !

those cases where a leak test is being {

substituted for a surface and/or volumetric  ;

t examination.

7 Several ASME Section XI Class 2 Welds were N/A  ;

inaccessible for examination.

However, to meet the Table IWC-2500-1 requirements of surface and/or volumetric inspection, 25% of the welds in all other systems not specifically listed in the Table are to be examined. Therefore other Class 2 welds were substituted for the inaccessible welds listed in this note without compromising plant safety. i 3606K

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ATTACKMENT B Braidwood Unit 1/ Byron Unit'1 Preservice Inspection Program Relief Request Cross Reference List Braidwood Unit 2 Braidwood Unit 1 Relief Request No. Relief Request Description Relief Request No. j 2NR-1 Surface examinations on Class 2 welds INR-1  !

were not performed due to locations under saddle plates.

2WR-2 Limited ultrasonic examinations were 1NH-2 performed on cast stainless steel elbow-to-cast pump or valve welds. j 2NR-3 Credit was taken for manufacturer's 1NR-3 test data in lieu of performing visual (VT-1) examinations on Class 1 valve internals.

-2NR-4 Ultrasonic examinations were not INR-4 performed on pressurizer and steam generator primary nozzle inner radii due to geometric constraints.

2NR-5 Limited ultrasonic examinations were 1NR-5 performed on cast stainless steel elbow-to-steam generator nozzle welds.

2NR-6 Limited ultrasonic examinations were INR-6 performed on cast stainless steel elbow, pump or valve-to-pipe welds.

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2NR-7 Limited ultrasonic examinations were INR-7 performed on cast stainless steel ]

elbow-to-reactor vessel nozzle I safe-end welds.

2NR-8 Ultrasonic examinations were not N/A performed on stainless steel weldolets due to weld geometry and material metallurgical properties.

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Braidwood Unit 2/Braidwood Unit 1

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Preservice Inspection Progrum Relief Request Cross Reference List Braidwood Unit 2" Braidwood Unit 1 Relief Request No. Relief Request Description Relief Request No.

2NR-9 Limited ultrasonic examinations INR-9

.were performed on reactor

!- . pressure vessel circumferential.

. welds'due.to inaccessibility.

2NR ' Limited ultrasonic' examinations lNR-10 were performed on horizontal letdown. heat exchanger shell circumferential weld due to inaccessibility.

2NR-ll' Limited ultrasonic examinations INR-ll were performed on excess letdown heat exchanger shell weld due to inaccessibility.

2NR-12 Limited ultrasonic examinations INR-12 were performed on steam generator and residual. heat exchanger secondary nozzles due to geometric constraints.

2NR-13 Volumetric examinations on Class 1 INR-13 Reactor Coolant welds were not performed due to location under a permanent restraint or geometric constraints.

2NR-14 Best effort volumetric examinations INR-16 were performed on Class 1 welds due to component-to-component orientation or interfering conditions.

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