IR 05000266/1998007

Insp Rept 50-266/98-07 on 980310-13,16,17,0401-02,22,24 & 0504-05.No Violations Noted.Major Areas Inspected: Maint & Engineeing

ML20248K412
Person / Time
Site:	Point Beach
Issue date:	06/03/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20248K396	List: IR 05000266/1998007 ML20248K393
References
50-266-98-07, 50-266-98-7, NUDOCS 9806100109
Download: ML20248K412 (11)

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l U.S. NUCLEAR REGULATORY COMMISSION REGION lli Docket No: 50-266 License No: DPR-24 l

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Report No: 50-266/98007(DRS)

Licensee: Wisconsin Electric Power Company Facility: Point Beach Nuclear Plant, Unit 1 Location: 6610 Nuclear Road Two Rivers, WI 54241 Dates: March 10-13,16-17, April 1-2,22-24, and May 4-5,1998 Inspector J. Schapker, Reactor Engineer Approved by: J. Gavula, Chief Engineering Branch 1 Division of Reactor Safety 9906100109 980603 PDR ADOCK 05000266 9 PDR

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EXECUTIVE SUMMARY l

Point Beach Nuclear Plant, Unit 1 '

NRC Inspection Report 50-266/98007

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This routine inspection included a review of the Inservice Inspection Program and the Unit 1 part length control rod drive housing modification performed during the Unit 1 refueling outag The following specific observations were made:

Maintenance

• State-of-the art equipment and procedures were utilized to perform the inservice inspections. Licensee and contractor audits and surveillance were adequate to assure compliance to procedures and the steam generator examination guideline (Section M1)

. Inservice inspection procedures and data reviewed by the inspector complied with ASME Code,Section V and Section XI requirements. (Section M3)

Engineering

. The licensee's decision to remove the Unit 1 part length control rod drive housings in response to the leaking part length control rod drive housing at Prairie Island demonstrated a conservative decision brsed on safety. Observation and review of the work procedures confirmed that the modification was performed in accordance with applicable ASME Code and regulatory requirements. (Section E1)

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Report Details 11. Maintenance M1 Conduct of Maintenance M1.1 Observation of Inservice insoection (ISI) Activities Insoection Scone (73753. 73052. 50002. 73755)

The inspector observed the acquisition, analysis and resolution of the eddy current examination (ET) data for the Urtit 1 steam generators (SGs), and performed an independent review of portions of the ET data for indications resolved with the plus point coil. The inspector reviewed the ET data analysis guidelines, personnel certifications including site-specific analyst per'lormance demonstration examinations, acquisition procedums, quality assurance audits, certifications / calibration of ET equipment, and the visual emrnination of the secondary side of the SGs after sludge lancin The inspector observed work, reviewed the ISI plan, procedures, personnel and equipment certifications and reviewed data associated with the following activities:

. Ultrasonic examination (UT) of the reactor vessel welds

. ET of the steam generator tubing Observation and Findings Reactor Vessel (RV) Examination The inspector reviewed the performance of the mechanized ultrasonic and remote visual examinations of the intemals and interior surfaces of the RV. The UT procedures, equipment and personnel were qualified by Southwest Research Institute (SwRI) with oversight of the Electric Power Research Institute (EPRI) through the Performance Demonstration Initiative (PDI) for the circumferential and long seam welds using weld mockups containing fabricated flaws. Further, SwRI examiners were qualified to the applicable American Society of Mechanical Engineers (ASME) Code Section XI and procedure requirement The inspector observed the mechanized UT scanning of the RV circumferential and nozzle-to-shell welds, data acquisition, and data evaluation in progress. Indications were identified in the nozzle-to-shell welds. These indications were classified as fabrication related indications which had been identified in previous examinations performed by SwRI. Flaw sizing confirmed that the fabrication indications did not exceed the ASME Code allowable, and that no flaw growth had occurred. In addition, the inspector reviewed the fabrication radiographs for the RV nozzles which contained l

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fabrication flaws identified by the current UT examination. These flaws (slag / porosity inclusions) met the applicable ASME Code acceptance criteria at the time of the fabrication (1966-67) and the current ASME Code Section XI requirement The PDI UT procedures did not meet some requirements stated in the licensee's current commitment to the ASME Code. The licensee had submitted a relief request to the NRC identifying the deviations for the ASME Code with a technicaljustification of the proposed alternative. In the licensee's technicaljustification, comparison of the ultrasonic testing technique used in the PDI with that of the current ASME Code and Regulatory Guide 1.150, demonstrated that the PDI technique provided an equivalent or a better examination of the RV than examinations required by the goveming standard The NRC staff approved this relief reques Steam Generator Examination The SG tubing ET inspection scope included: 100% bobbin coil examination of all open tubes, a 20% sample of tubes using a motorized rotating pancake coils with a plus point coil at the hot leg side of the top of the tube sheet, a plus point coil examination of all row one and two U-bend tube areas and unresolved indications ider.tified by the bobbin coil. The inspector considered the equipment and procedures used for this inspection to be state-of-the art. A secondary side visual examination using a remote camera after sludge lancing was also performed. The licensee and contractor performed audits and surveillance of the steam generator inspection activities which were adequate to assure compliance with procedures and EPRI SG examination guideline The ET examination did not identify defective tubes that required repair. Anti-Vibration Bar (AVB) wear was the only reportable degradation identified. The maximum degradation identified for AVB wear was 28% up from a previous maximum of 25%

recorded in 1995. All other AVB wear was less than 20% with similar growth rates. The licensee reduced the reporting critella for dents from ten volts to three volt Supplemental plus point examination of 48 dents in SG 11 and 63 in SG 12 did not identify any degradation). Seven indication codes were called in the hotleg of SG 11 and five in SG 12; subsequent plus point examinations did not confirm any degradatio c. Concbsiqns State-of-the art equipment and procedures were utilized to perform the inservice inspections. L,1censee and contractor audits and surveillance were adequate to assure compliance to procedures and the EPRI SG examination guideline !

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M3 Maintenance Procedures and Documentation M3.1 Review of Nondestructive Examination Data (73755) Insoection Scooe The inspector reviewed portions of the licensee's ISI program procedure and NDE data recorded in accordance with ASME Section XI requirements, Observations and Findinas All applicable ISI procedures were approved by the Authorized Nuclear Inservice Inspector, and in accordance with ASME Code Section V and XI requirement Conclusions Inservice Inspection procedures and data reviewed by the inspector complied with ASME Code Section V, and Section XI requirement Ill. Enaineerina E1 Conduct of Engineering Insoection Scooe (37550)

Modifications to the reactor vessel head wera performed by removal of the part length control rod drive housings (PLCH) and installation of head adapter plugs (HAP). The inspector reviewed the following engineering documents in support of the modifications and observed work activities:

. Modification Design Package MR-98-014

. Safety evaluation: SE-98-060

. Installation Work Plan: IWP-98-014

- Weld procedure specification: WSI A08165 revision D

. Material tests report / laboratory analysis of two HAPS

. Observed the removal of one PLCH

. Review of visual recordings of welding the HAP seal welds

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l Observations and Findings i The licensee conservatively elected to remove the Unit 1 PLCH's and cap the

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penetrations in response to a PLCH leak, caused by a fabrbation flaw, at Prairie IsleM The part length control rods were not used and therefore performed no safety related function. The pector reviewed the design change and safety evaluation documents, and confirma inat the proposed modifications complied with applicable ASME Code and regulatory requirements. Modification travelers contained adequate work and inspection instructions. The contractor's weld procedures complied with ASME Code Section IX,1986 Edition requirements. The modification safety analysis and design documents were appropriately reviewed by the licensee for the replacement HAP's. The effectiveness of licensee reviews was evidenced by the types of issues identified by the licensee such as the identification of a material specification discrepancy, that was resolved by the testing of both heats of material to verify the composition and material i propertie A relief request had been approved by the NRC to allow visualinspection of the welds instead of liquid penetrant examinations, due to the limited accessibility and high radiation area. The inspector observed the removal of the PLCHs and reviewed the taped recordings of the HAP seat welding and visualinspection documentation following installation. Visual examinations were performed using a high resolution camera and evaluated by a Level 11 visual inspection examiner. All welds were acceptable. Prior to start up, a visual inspection during the system pressure test was scheduled to be performe Conclusion The licensee's decision to remove the Unit 1 PLCH's in response to the leaking PLCH at Prairie Island demonstrated a conservative decision based on safety. Observation and I review of the work procedures confirmed that the modifications were performed in accordance with applicable ASME Code and regulatory requirement E8 Mir..:ellaneous Engineering issues E (Closed) Insoection Follow uo item 50-301/96004-04(DRP): The licensee had failed to mark the replacement SG weld centerlines as required by ASME Section XI, Appendix lll,1986 Edition. Although the vendor did not mark the weld centerlines, markings j

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adjacent to the welds were used to identify the weld centerlines based on WEC procedure 3288, Revision 22," Guidelines for RT/UT Layout of Weld." The licensee has I

included the WEC procedure in the ISI outage report for referencing the location of weld centerlines for UT examination E8.2 (Closed) insoection Follow uo item 50-301/96014-03 (DRS): The inspector had identified a concern with potential sensitization of existing weld material associated with the reactor coolant loop welds. Further review by the inspector confirmed that the

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licensee had complied with ASME Code Section ill and IX requirements for heat inputs, interpass temperature controls, and preheat temperature requirements which mitigated the possibility of sensitization of the weld V. Management Meetinas X1 Exit Meeting Summary The inspector presented the inspection results to members of licensee management at the conclusion of the inspection on May 5,1998. The licensee acknowledged the findings presented and did not identify any of the potential report input discussed as proprietar m

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I PARTIAL LIST OF PERSONS CONTACTED l

Licensee

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K. Crawley, Plant Engineering

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A. Flentje, Sr. Regulatory Compliance Engineer l

T. Hanna, ISI Engineer l J. Oswald, Plant Engineering C, Prothro, ISI Engineer M. Reddemann, Plant Manager l J. Schweitzer, Manager Site Engineering P. Wild, Plant Engineer Southwest Research Institute R. Riddles, Lill UT Framatone R. Merriman, Lill ET Zftles N. Farrenbaugh, Lill ET C. Mathison, Lill ET NflG F. Brown, Senior Resident inspector P. Louden, Resident Inspector INSPECTION PROCEDURES USED IP 73753: Observation of ISI examinations IP 73052: Review of ISI procedures IP 73051: Review of ISI program IP 73755: Review of ISI data IP 50002: Steam Generators IP 37550: Review of Engineering / modification

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LIST OF ITEMS OPENED, CLOSED AND DISCUSSED Open None Closed IFl 50-301/96014-03(DRS) Sensitization of weld material in reactor coolant loop piping weld IFl 50-301/96004-04(DRS) Centerline marking of steam generator fabrication welds Discussed None LIST OF DOCUMENTS REVIEWED PTB-AUT-14/1/1/1 Automated Ultrasonic inside Surface Examination of Pressure Piping Welds PTB-AUT15/15/2/1 Automated Ultrasonic Inside Surface Examination of Ferritic Reactor Pressure Vessels Greater Than 4.0 inches in Thickness PTB-VT7/0/0 Visual Examination of Nuclear Power Components SwRI-AUT2/10/1 Automated Ultrasonic Inside Surface Examination Indication Resolution and Sizing SwRl-AUT5/4/0 Southwest Research Institute par Device and Attachments Operation SwRI-AUT8/3/0 Southwest Research Institute par Device Calibration SwRI-AUT34/3/0 Southwest Research Institute par Device Checkout SwRI-AUT36/1/1 Checkout and Operation of the 8-Channel Enhanced Data Acquisition System swr!-AUT38/1/0 Automated Ultrasonic System Performance Verification l SwRI-EDAS2/3/1 Enhanced Data Acquisition System-Il Performance Verification Procedure (Test Plan)

SwRI-NDE4/1/0 Onsite NDE Records Control SwRI-NDE6/0/0 Use of Customer notification forms l

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l SwRI-PDI-AUT1/2/0 Automated Inside Surface Ultrasonic Examination of Ferritic Vessel Wall Greater Than 4.0 inches in Thickness SwRI-PDI-AUT2/2/0 Automated Inside Surface Ultrasonic Flaw Evaluation and Sizing NPL 98-0185 Licensee response to GL 97-05: SG Tube inspection Techniques Point Beach Nuclear Power Plant, Units 1 & Relief Request RR-1-02, RR-1-17: Ultrasonic examination of welds from the internal diameter of pipe in leu of surface examination for reactor coolant outlet nozzles and safety injection safe end to nozzle welds. RR-1-18: Current Code uses of Performance Demonstration Initiative procedures MR-98-014 Design Change Package for the Unit 1 PLCH removal and plug installatio WR-9803215 Work request for the PLCH removal and plug installatio lWP 98-014 Installation Work Plan for the PLCH removal and plug installatio WPS-A08165 Weld procedure specification for the HAP seal weld MLH-98-012 Design and ASME Code Section ll1 Evaluation of CRDM Adapter Plug Fillet Weld at Point Beach Unit 1 (Structural Integrity Associates; inc.)

WEC 3288 Guidelines for RT/UT Layout of Weld Revision 22 SE 98-060 10CFR50.59 /72.48 screening and safety evaluatio S1-400-07 Multi-Frequency Eddy Current Examination of Tubing PB-1 Revision 0 Site Specific ET Data Analysis Guidelines PB-Unit 1

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LIST OF ACRONYMS USED ASM American Society of Mechanical Engineers AVB Anti-Vibration Bar EPRI Electric Power Research Institute ET Eddy current examination HAP Head adapter plugs ISI Inservice Inspection NDE Nondestructive examination NRC Nuclear Regulatory Commission PLCH Part length control rod drive housing PDI ~ Performance Demonstration initiative I UT Ultrasonic Examination RV Reactor Vessel

.SG Steam Generator SwRI Southwest Research Institute

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