IR 05000302/1986003
| ML20214D378 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 02/12/1986 |
| From: | Blake J, Economos N NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20214D356 | List: |
| References | |
| 50-302-86-03, 50-302-86-3, NUDOCS 8603050145 | |
| Download: ML20214D378 (8) | |
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.s eMic UNITED STATES
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NUCLEAR REGULATORY COMMISSION
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ATLANTA, GEORGI A 30323
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Report No.: 50-302/86-03
Licensee: Florida' Power Ccrporation 3201 34th Street South
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St. Petersburg, FL 33733 Docket No.: 50-302 License No.: DPR-72 l
Facility Name:
Crystal River 3 i
i Inspection Cenducted: Januar-10, and 14-16, 1986
Inspector 1 mm -irr d
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<7 Fate Kgned Approved by:
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'ake, Section Chief Date Signed i
ering Branch
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ivision of Reactor Safety
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SUMMARY l
Scope: This rcutine, announced inspection entailed 67 inspector-hours on site in the areas of inspector action on previous enforcement matters; inspector identified follow-up items; review and evaluation of ISI data; licensee actions
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relative to Reactor Coolant Pump "1A" shaft failure. January 7, 1986 was dedicated
l to respirator training and issuing of appropriate picture badge.
l Results: No violations or deviations were identified.
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REPORT DETAILS 1.
Persons Contacted Licensee Employees
- P. F. McKee, Plant Manager K. R. Wilson, Supervisor, Site Nuclear Licensing C. G. Brown, Nuclear Outage and Modification Assistant Manager M. D. Clary, Nuclear Mechanical Engineer Greg Halnon, RCP-A-Task Force Leader
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J. J. Warren, Nuclear Welding Engineer Eric Johnston, Site Nuclear Design Engineering F. V. Fusick, Supervisor, Nuclear Engineering Don Gulling, ISI Specialist Other Organization Babcock & Wilcox (B&W)
G. R. Stromer, Level III Examiner C. E. Thompson, Level II UT Examiner NRC Resident Inspector T. Stetka
- Attended exit interview 2.
Exit Interview The inspection scope and findings were discussed on January 16, 1985, with the licensee's representative indicated in paragraph 1 above. The inspector followup item listed below was discussed with appropriate personnel.
(0 pen) Inspector Followup Item (IFI) 302/86-03-01:
RC Pump "1A"
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Replacement Shaft QA Data Package, paragraph 5.
The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspector during this inspection.
3.
Licensee Action on Previous Enforcement Matters (92702)
(0 pen) Violation 50-302/85-17-01: Control of Field Welding Activities The licensee's letters of response dated July 3, 1985 and August 30, 1985, were reviewed by the staff.
The inspector held discussions with the Site Nuclear Welding Engineer and Nuclear Operations Engineering representative.
From these discussions, the inspector ascertained that a concerted effort is
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underway to-issue a new welding manual identified as the Nuclear Operat' ions
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Welding Manual (N0WM).
The new procedures will supersede existing welding control procedures used at CR-3.
Most of the procedures in the manual have been issued for final review / comment' while some others are in final draft form.
This effort is designed to streamline the welding programs at CR-3.
The procedures will be applicable to all nuclear welding activities performed by -FPC site organizations and by contractors on structures, components and piping systems.
A number of the procedures reviewed for content were'as follows.
Qualification of Welding Procedures
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Administrative Control of Welding
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Control of Welding Consumables
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Control of Welding Equipment
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It is anticipated that the new welding manual will be issued for j
implementation in March of this year. Therefore, this item will remain open pending a review of the approved document.
4.
Unresolved Items (92701)
Unresolved items were not identified during this inspection.
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Independent Inspection Effort (92706)
Reactor Coolant Pump "1A" Shaft Failure I
This work effort was a *ollowup to the Region's initial response to the licensee's notification on January 1,1986, that the plant tripped because of reactor coolant pump, "1A" failure.
Results of that inspection were documented in Report 50-301/86-04.
Upon arrival at the site on January 6, 1986, the inspector ascertained that the pump motor was still in place,
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although efforts were underway to remove it.
Discussions with onsite licensee management disclosed the formation of a task force, headed by the assistant outage manager, to oversee and direct all activities relative to the pump's failure and repair including design, engineering, maintenance inspection and testing.
Discussions with cognizant engineering personnel disclosed that a spare pump shaft assembly was on site and would be used as a replacement if an ultrasonic examination determined that the shaft in the RC pump had failed.
The inspector observed the replacement shaft assembly identified by serial number S/N-0531, in storage and requested that the licensee retrieve the QA package for review.
On January 9, 1986, the replacement shaft was sent to the vendor, Byron Jackson in Los Angeles, California, for inspection, testing, balancing, etc.
Babcock & Wilcox was contracted to ultrasonically examine the shaft of the failed pump, to determine whether the shaft had failed and if possible to identify the location of fracture / failure on the shaft. On January 14, 1986 the pump motor was lifted from RC pump 1A.
This was followed by removal of the spool piece which made the pump shaft end accessible. At this point, the shaft was ultrasonically examined (UT) by B&W using a procedure written for
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thickness measurements of material ranging from 0.10" to 100 inches in thickness, in this case 68 3/4 inches in length. - The procedure was identified as ISI-185, Rev. O " Ultrasonic Examination for Thickness Measurements of Pump Shaft From End Face of Shaft."
The examination was conducted with a 2.25 Mhz., O transducer, one inch in diameter.
A major reflector was detected 50 inches from the top of the shaft and another of a i
lesser magnitude at 56 inches.
The major reflector was observed for 360
while the lesser one was present at one location only.
This information along with the fact that the shaft could be jogged with a minimum force back and forth and sideways increased the' likelihood of a break at the predicted length.
See Attached sketch of shaft for location of break.
It should be noted that the fracture occurred at the machined circumferential groove at the 50 inch length indicated by the UT examination.
After removal of the coupling, seal packages and external interferences, e.g. snubbers, the top l
portion of the broken shaft was removed from the RC pump.
At this point, l
the inspector observed that failure occurred in the machined groove at or near the sharp re-entrant angles of the bottom edges.
Judging from the
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i appearance of the fracture surface, it seems that the failure was the result of a fatigue mechanism which may have initiated at the sharp
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J re-entrant angle due to a mechanical and/or a metallurgical flaw in
the area.
The broken shaf t segment was sent to the B&W laboratory
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in Lynchburg, Va. for a metallurgical examination to determine the cause of the failure.
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In view of these circumstances, the licensee decided to UT the shafts of the other three RC pumps. To perform this task, the licensee contracted B&W who will develop a more sensitive UT procedure which would have the capability
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of detecting small cracks that may be present along the length of the shaft.
l The inspector discussed with the licensee and B&W certain technical aspects of the new procedure including calibration block material, size and configuration; transducer frequency and size, acceptance criteria,
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l flaw / crack evaluation and inspection technique (s).
The licensee agreed to notify the inspector when B&W was ready to demonstrate the procedure's adequacy and examine the aforementioned shafts.
The inspector discussed, with cognizant personnel, the design and fabrication of certain fixtures and the lifting device used for the RC pump motor lift. Documents reviewed included design calculations, weld sketches,
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nondestructive test results and weld records and load test results.
The inspector requested, and the licensee agreed, to provide for review the QA data package for the replacement shaft as this becomes available. Review of this package was identified as inspector followup item (IFI), for tracking purposes, IFI 302/86-03-01 RC pump "1A" Replacement Shaft QA Data Package".
6.
Inservice Inspection - Review of Program (73051)
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The inspector held discussions with cognizant licensee personnel to review the extents of ISI examinations with respect to ASME Section XI, (74S75),
requirements for major components including nozzles and piping.
Specific i
areas of interest included review of IS0's of safety related piping systems
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which depict weld population and selection for ISI purposes to satisfy code requirements. The inspector also ascertained whether measures were in place
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to ensure that line modifications were reflected on ISI pipe sketches and that changes to weld populations-were considered in determining weld
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inspection requirements.
In order to verify implementation of these measures, the inspector reviewed
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the following isometrics:
Sketch No.
System No.
Line and Size
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SK1C Rev. 1 131 12"/ Decay Heat
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SK3C Rev. 3 151, 153 14"/ Core Flood
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SK5C Rev. 2 173 2i"/ High Pressure Injection to Pump A-2
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Discharge SK6C Rev. 2 175 21"/ High Pressure Injection to Pump B-1
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Discharge
SK115C Rev. 3 280 6"/ Connects MUV-73 &
MUV-62 to Makeup Pump 1-A
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& 1-B SK9C Rev. 2 191 21/ Connects Pump B-2 Suction to Letdown Cooler j
Inlet Within the areas inspected, no violations or deviations were identified.
7.
Inservice Inspection - Data Review and Evaluation
A review of selected ISI data obtained over the last outage (Refueling Outage #5) was documented in Report 50-302/85-17.
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Data selected for review and evaluation during this inspection included evaluation reports on indications found by volumetric examination of certain
i welds in the pressurizer support to shell lugs, steam generator A shell to
nozzle belt and in the reactor pressure vessel.
The specific welds, I
findings and evaluations are as follows:
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Pressurizer Support to Shell Four Reportable Accept. per fracture i
i Lug X-Y Axis - B2.8.7 indications, 400 analysis document, j
401, 402 and 403
- 32-1158360 R0 1-Evaluation Rept.
l Pressurizer Support to Shell Eight Reportable Same as above Lug Y Axis - 62.8.9 indications; 201, 202, 400 to 405; Evaluation Report No.85-024
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Pressurizer Support to Shell Ten Reportable Same as above Lug Y-Z Axis - B.2.8.11 indications, 200 and 400 to 408 -
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Evaluation Report
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No.85-025
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Steam Gen. "A" Shell to Noz.
Eight Reportable Belt MK-2-3 C1.1.2 indications; 200-205 and 400-401, Evaluation Report No.85-033 Indications #204, #205 are same as #400 and #401, Accept, criteria IWB
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3511/1 ASME Section XI,1977 Edition per relief request to the Commission dated March 24, 1983 and fracture analysis report #32-1158427-00, remaining indications are acceptable.
I REACTOR PRESSURE VESSEL Long Seam, A2 Shell (X-Y)
Reportable Indications, All laminar per i
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Bl.1.5 One hundred thirty one, IWA-3360. Those Evaluation Report requiring evalua-No.85-015 tion were found acceptable per IWB-3510.2 Long Seam, A-2 Shell (Z-W)
Reportable Indications Same as above B1.1.6 One hundred thirty seven Evaluation Report No.85-016
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Long Seam, A-1 Shell (Y-Z)
Two Reportable Laminar flows
Bl.1.3 Indications Evaluation per IWA-3360.
Report 85-017 Acceptance per
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IWB-3510.2.
i Long Seam, Al Shell (W-X)
Reportable Indications Same as above B1.1.2 One hundred fifteen Evaluation Report 85-018
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Circle Seam Upper Core Belt Seven Reportable Five flows are
i MK-A7 to B1 (WR-1) B1.1.1 Indications, Evaluation Laminar per Report 85-019 IWA-3360.2 accep-
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tance per
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IWB-3510.2 Other two flaws evaluated as
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acceptable.
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Circle Seam Lower Core Belt Three hundred three All except four
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MK-Al to MK-A2 (WR-1) 81.4 Reportable Indications were evaluated Evaluation Report 85-020 as laminar per IWB-3510.2 Other
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four classified
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acceptable.
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Inlet Nozzle to Vessel from One Reportable Subsurface flow Vessel ID (X-Y) Bl.4.4 Indication Evaluation Acceptable per Report 85-021 IWB-3512
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Outlet Nozzle to Vessel from Four Reportable Subsurface flaws
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Vessel 10 (X Axis) 81.4.7 Indications Evaluation Acceptable per Report 85-022 IWB-3512
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Documentation and evaluation of these indications appear to be consistent
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I with applicable Code and Regulatory requirements.
Within the areas inspected,'no violations or deviations were identified.
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Attachment:
j Reactor Coolant Pump Shaft i
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