IR 05000302/1987032
| ML20237E941 | |
| Person / Time | |
|---|---|
| Site: | Crystal River  |
| Issue date: | 12/10/1987 |
| From: | Blake J, Economos N, Glasman M NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20237E923 | List: |
| References | |
| 50-302-87-32, IEB-83-007, IEB-83-7, NUDOCS 8712290272 | |
| Download: ML20237E941 (17) | |
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lo NUCLEAR REGULATORY COMMISSIOM
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REGION ll p
j 101 MARIETTA STREET, N.W.
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'r ATLANTA, GEORGI A 30323
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Report No.:
50-302/87-32 Licensee:
Florida Power Corporation 3201 34th Street, South St. Petersburg, FL 33733 Docket No.:
50-302 License No.:
DPR-72 Facility Name:
Crystal River.3 Inspection Cond e
ctober 19-23 and ovember 2-6, 1987-Inspectors:
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? IMa'n I) ate Signed N
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/2.IP, 7-Appro(edby:
J. J.
lake, Section Chief Dat'e Signed g
ering Branch-iv ion of Reactor Safety SUMMARY Scope:
This routine, unannounced inspection was in the area of Inservice Inspection (ISI): -Review of ISI Procedures and evaluation of. Inservice Inspection results; licensee action on previous inspection findings; letdown cooler 3C installation; licensee action on Temporary Instruction (TI) 2515(84);
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Results: One apparent violation was identified - Failure-to meet ASME Code Section IX welder performance qualification requirements, paragraph 8.
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8712290272 87 DR ADOCK O O2,
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REPORT DETAILS 1.
Persons Contacted Licensee Employees
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- P. F. Mckee, Director Plant Operations
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- W. Rossfeld, Manager, Nuclear Compliance
- K. R. Wilson, Manager Nuclear Licensing i
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- C. Brown, Manager Outages
- W. Neuman, Supervisor, Inservice Inspection (ISI)
- D.
Gulling, Nuclear ISI Specialist
- J. J. Warren, Principal Nuclear Mechanical Engineer
- J. C. Hicks, Material Technology Level III Corporate J. N. Hessinger, Supervisor Materials Quality Control F. V. Fusick, Supervisor Site Nuclear Engineering Services R. L. Dunaway, Level II Radiographer A. Petrowsky, Nuclear Structural Engineer Other licensee employees contacted included construction craftsmen, engineers, technicians, mechanics, and office personnel.
Other Organization Fluor Mechanical Services (Fluor)
R. Mahaley, Project Engineer NRC Resident Inspectors
- T. F. Stetka J. E. Tedrow
- Attended exit interview 2.
Exit Interview The inspection scope and findings were summarized on October 23, and November 6,1987, with those persons indicated in paragraph 1 above. The inspectors described the areas inspected and discussed in detail the inspection findings.
One violation was identified in the area of welder qualification (Violation - (302/87-32-01), Failure to Meet ASME Code Section IX Welder performance qualification requirements paragraph 8).
No dissenting comments were received from the licensee. -The licensee identified as proprietary one of the documents reviewed during this inspection.
However, this information is not included in this report.
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3.
Licensee Action on Previous Enforcement Matters (Closed) Unresolved (UNR) Item 85-23-02, WP5 Validity.
By memoranda entitled, "WPS Hi story," dated May 5, 1987, the licensee's Material Technology has described FPC's roll in the qualification of welding procedures from 1971 to the present.
The memo states that procedure qualifications dated circa 1971 were developed under FPC direction by mechanical contractor, Levsey/Jonas at the Anclote Plant construction'
site.
Also, the document states that, subsequent weld procedure specifications WPSs and supporting qualification records PQRs issued between the 1976 and 1980 timeframe were written and qualified by personnel within the FPC Organization.
Sometimes in 1980, the licensee issued the FPC Welding Manual and more recently the Nuclear V!elding.
Manual.
Within this timeframe, WPSs and PQRs ~ have been handled in accordance. with requirements of these two ' documents.
In a similar doc. ament issued on May 15, 1987, the licensee's cognizant welding engineer states that the Catalytic, Inc. (CI) welding program including (WPSs),
there of were reviewed and approved by FPC's Quality Programs Department who found them acceptable for use at CR-3.
Finally, the memo concludes that FPC was the owner and repair organization for FPC work done during the CI contract period.
This item is considered closed.
(Closed) UNR Item (86-40-01), Letacwn Coolers Evaluation of Replacement Suitability t.nd Failure Analysis.
This item was identified when the inspector determined that the licensee could not provide an evaluation of the suitability of the replacement coolers including a failure analysis
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report as required by paragraph IWA-7220, " Verification of Acceptability",
ASME Code Section XI. In this instance, the time required to develop the f ailure analysis information and apply appropriate corrective. actions to the replacement component during an existing does not appear to be realistic. 'During this inspection, the inspectors had discussions with cognizant engineering personnel and reviewed the failure analysis report on letdown cooler "A".
According to this report, high cycle mechanical fatigue and flow induced tube vibrations contributed significantly to the letdown cooler failure Corrective actions include design changes and hardware modifications on the secondary side of the cooler which alters fluid flow and direction, minimize points of stress concentration, and improved supports to overcome flow-induced loads.
4.
Unresolved Items Unresolved items were not identified riuring this' inspection.
5.
- Temporary Instructions (TI) 2515/84, Primary Coolant System Pressure Isolation (Event V) Valves TI 2515/84, Verification of Compliance with Order for Modification of
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Licensee Primary Coolant System Pressure Isolation (Event) Valves the Reactor Safety Study (RSS), WASH-1400, identified in a PWR an intersystem loss of coolant accident (LOCA) that a significant contributor to risk of
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core melt accidents (Event V). The design examined in the RSS contained in-series check valves isolating the high pressure primary coolant system (pCS) from the low pressure injection system (LPIS) pip.ing. The 1 scenario
- which leads to the Event V accident is initiated by the failure of these check. valves to function as a pressure isolation barrier against reactor coolant system (RCS) pressure.
This ' causes an overpressurization and rupture of the LPIS low pressure piping which results in a LOCA that bypasses containment.
.To better define. the Event V concern, al1 light water reactor licensees.
were requested. by letter dated February 23, 1980, to. provide specific information on such valve configurations in s accordance with 10 - CFR 50. 54-( f).
In_ addition, licensees were asked to perform individual check valve leak testing before plant startup after the next scheduled outage.
Based en licensee responses and the ongoing unsatisfactory operational-experience at several plants, -the NRC staff concluded that a valve configuration of concern existed; meaning that, when pressure isolation was provided by two check. valves in-series, and when failure of one valve in the pair can go undetected for a substantial length of time, verification of valve integrity was required. The staff concluded that, since these valves are safety-related, they needed to be tested periodi-cally to ensure low probability of gross failure, as a result, the staff determined. that periodic examination of check. valves was required to be undertaken by the licensees to verify that each valve was seated properly and functioning as a pressure isolation device. Such testing was intended to reduce the overall risk of an intersystem LOCA.
On April 20, 1987, the Commission issued an Order for Modification of License Concerning Primary Coolant System Pressure Isolation Valves, to specified PWR and BWR plants, requiring that the above described testing o
be implemented. This Order included a Safety Evaluation Report (SER) and Technical Specification insert pages to require leak rate testing of Event V pressure isolation valves.
TI 2515/56 was issued on June 1, 1981, for followup inspection of implementation of the Event V orders.
It expired December 1983.
The present TI 2515/84, is intended to verify sati sfactory completion of licensee actions to implement the periodic testing of Event V valves as required by the aforementioned Order.
Within these areas, the below listed valves were selected at random for a review of test records and compliance with actions required by the-subject Order and TI 2515/84.
The valves are listed in Table 3.4-2 of Crystal-River 3 TS.
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Ractor Coolant System Pressure Isolation Valves System Valve Decay Heat / Low Pressure Injection CFV-1 DHV-2 CFV-3 DHV-1 This inspection effort included review of:
the plant's Technical Specification (TS) Table 3.4 Reactor Coolant System Pressure Isolation Valves a copy of the licensee's Event V Order and the original SER, procedure changes and records of hardware modifications as applicable.
The procedure for this leak test is SP-603, Rev. 4, " Decay Heat / Core Flood Check Valve Leak Testing Procedure."
This procedure was reviewed to verify that it is consistent with TS requirements including:
a.
An acceptable test method is used.
This would include a direct volumetric leakage rate measurement or other equivalent means capable of demonstrating that leakage rate limits given in the TS are not exceeded.
b.
The test procedure ensures that leakage rates obtained are for individual valves rather than for combined components.
c.
The test procedure requires that leakage rates received at test pressures less than the maximum potential pressure differential across the valve be adjusted by assuming leakage to be directly proportional to the pressure differential to the one-half power as noted in the SER which accompanies the Order.
d.
Acceptance criteria stated in the test procedure are in accordance j
with the TS.
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e.
Verify that the procedure identifies required corrective actions in event leakage rate results are unacceptable.
For the valves listed above, the inspectors reviewed records of test completed during previous outages as listed below.
Date Tested Results 8/31/87 Acceptable 6/14/86 Acceptable 9/4/85 Acceptable
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4/19/84 Acceptable 10/21/83 Acceptable 8/4/83 Acceptable i
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12/8/81 Acceptable 10/5/81 Acceptable Discussions with the cognizant engineer disclosed that rone of the subject
' isolation valves failed leak rate acceptance criteria stipulated in the TS and, therefore, no repair work has been performed during the timeframe inspected.
In addition, test data were reviewed and verify that:
a.
Test records contain major test data including upstream and downstream pressures, leak volume per unit time (or equivalent),
leakage rate adjustment calculations when required, and leakage rate acceptance criteria based on trending from previous tests where applicable.
b.
Recorded test frequency is in accordance with TS.
c As found leakage (i.e.,
prior to valve stroking, modification, adjustments, etc.) is recorded, d.
Leakage rates trending has been documented and adequately evaluated by the licensee in accordance with the TS requirement.
e.
No test data anomalies exist which indicate improper or inaccurate testing.
f.
Adequate corrective actions were taken for valves not meeting the acceptance criteria.
Within the areas inspected, no violations or deviations were identified.
6.
(Closed) IE Bulletin No. 83-07:
Apparently Fraudulent Products Sold by Ray Miller, Inc.
By memorandum from G. R. Westafer to J. P. O'Reilly dated March 22, 1984, the licensee responded to each of the action items as required by the Bulletin in the following manner:
Action Item 1 - FPC reviewed all certified material test reports in the document files for material supplied CR-3.
Material identified by this record search currently (March 22, 1984) installed in safety-related systems is tested below.
No Ray Miller material was identified in stock.
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Equipment Description of Materials Make-up Tank; Materials 1 - 4" Pipe schedule 40 welded x 0'7" (SS)
used for miscellaneous tank nozzles 1 - 4"x9" Nipple schedule 40 T0E (SS)
2 - 1 1/2" Coupling S/W 3000# (SS)
1 - 3/4" Pipe schedule 40 welded x 7'7" (SS)
1 - 3/4" Coupling S/W 3000# (SS)
2 - 1" Coupling S/W 3000# (SS)
1 - 3/4" Cap Scrd. 150# (SS)
1 - 1" Schedule 40 Pipe Welded x O'7" 1 - 1" Welded neck flange ASA 150# RF Sch. 40 Date of CMTR:
July 23, 1970 Buyer:
Ecthlehem Steel, Bethlehem, PA Seal Return Coolers; For each seal return cooler Materials used for 4 - 10" Caps Schedule 10 Inlet / Outlet Head and 1 - 4" Pipe Schedule 40 Seamless x 3'0" Return Head 1 - 10" Pipe Schedule 10 Welded x 2'0" cut Date of CMTR:
Janua ry 10, 1971 Buyer:
Basco Division of American Precision 2777 Walden Ave.
Buffalo, NY Miscellaneous Tubing; 78'3/8"x 0.065 wall tubing (seamless, SS)
Unknown Plant Location Date of CMTR:
April 21, 1969
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Buyer:
Capital Pipe and Steel Products 301 City Line Ave.
Bala-Cynyd, PA Electrical Penetrations 16 2" Pipe schedule 40 welded plain end x (208-211 and 401-404)
6'6 1/4" (SS)
51 2" Pipe schedule 40 welded plain end x 7'10 1/8" (SS)
16 2" Pipe schedule 40 welded plain end x 6' 81/2" Date of CMTR:
June 1, 1970 and July 8, 1970 Boyer:
Irwin Steel Fabricators P.O. Drawer #1388 Station C Canton, Ohio Action Item 2 -To determine the safety significance of the presence of the Ray Miller materials, Florida Power Corporation Performed a Failure Analysis for each affected safety-related system. The analyses for the Seal Return Coolers, Make-up Tank nozzles, Electrical Penetrations and tubing (3/8 inch) are presented as an Attachment to this report.
Within these areas, no deviations or violations were identified.
7.
Inservice Inspection Data Review and Evaluation (73755)
a.
Eddy-Current (EC) Examination of Once Through Steam Generator (OTSG)
"A" ISI activities during this refueling outage (6th) included the EC examination of OTSG "A".
B&W was responsible for data acquisition and analysis.
The controlling document for this work effort was approved procedure SP-305, Rev. 11 OTSG Tubes Eddy-Current Inspection, which references RG 1.83, Rev. I and ASME code case N-401, (I51-83-29).
Tube selectior was made by the licensee in
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accordance with CR-3 TS 4.4.5.2 req:.irements which translates to 6%
of the tubes in SG
"A" or 942 tubes selected at random.
Defects identified in two tubes (70-28 and 72-5) expanded the inspection to include an additional 12%, bringing the number of inspected tubes to a total of 3370 or 21.7%.
By review of interoffice correspondence and discussions with cognizant personnel, the inspectors ascertained that the defects have been characterized as " crack-like" indications with a 52-55% through-wall penetration. The defects were located at the top of the 12th tube support plate. Further eddy-current testing ECT us1ng a 8x1 probe confirmed the defects to be small, localized or crack-like indications, but could not identify the orientation of the cracks (i.e., axial or circumferential). ECT testing of the tubes in 1983 found those tubes to be free of indications.
A report describing results of eddy-current testing on OTSG "A" will be submitted to the Commission as required by CR-3 Technical Specifications.
In addition to the above, the inspectors reviewed Calibration Standard No. 1183506B-0 quality records as well as personnel qualifications for Level I, II and III examiners involved in the examination.
The below-listed procedures and corresponding qualifications were reviewed for technical content.
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ISI-418, Rev. 3 Technical Procedure for the Multi-frequency Eddy Through Steam l
Current Examination of Once
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Generator (OTSG) tubing in 177 Steam Generators using the MIZ-18.
l ISI-425, Rev. 10 Eddy Current Examination of Tubing by the Absolute Multifrequency (8x1) Technique.
Within the areas inspected, no deviations or violations were identified.
b.
Volumetric Examination (Radiography)
The inspectors reviewed the ISI NDE records (radiographs) described below.
The applicable code for this ISI is the ASME Boiler and Pressure Vessel Code,Section XI,1974 Edition with Addenda through S1975.
Records for the examination area listed below were reviewed to ascertain whether the records contained or provided reference to:
examination results and data sheets; equipment data; calibration data l
sheets; evaluation data; records on extent of examination; records l
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relative to deviation from program; disposition of findings;
re-examination af ter repair; and identification of NDE materials as l
applicable.
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Examination Area Radiographs:
Item Weld Size System C2.1.26.4 FW-12 6.6x430 Emergency Feedwater Pipe to Flange C2.1.16 FW-12 "A" SG 14"x.700 diam Main Feedwater Header
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C2.1.24 FW-130D Main Feedwater Header
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C2.1.25 MK-124-123 Tee to Feedwater Header
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C2.1.15 MK-124-123 Tee to Header
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C2.1.14 120-127 Feedwater Header to Cap C2.1.10 MS-38C 24"x1.00 Main Steam Within these areas, the inspectors noted that the penetrometer on one of the radiographs, C2.1.14 station 3-4, reviewed had been placed l
on the weld joint reinforcement.
The inspectors discussed this matter with the cognizant Level II interpreter who had the weld
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section in question re-shot, which corrected the problem.
Because this problem was not observed in any of the other ra,diographs reviewed during this inspection, the inspectors concluded that it was j
an isolated case and, therefore; no enforcement action was warranted.
Within the areas inspected, no violations or deviations were identified.
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8.
Letdown Cooler "3C" Installation (37700)
In an effort to lessen the vulnerability of the plant to lengthy forced
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outages caused by letdown cooler failures, the licensee is installing a third cooler in parallel with the other two, (MUC+HE-1C), as a spare. The
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work is being performed under modification approval record, MAR
- 87-08-03-01.
By document review, the inspectors ascertained that the modification will result in changes to Technical Specifications Table
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3.6-1 to add valve MUV-505A and to Sections 5.3, 9.1 and 9.5 of the FSAR.
l The licensee's safety evaluation performed pursuant to 10 CFR 50.59 requirements dated September 18, 1987, showed no unreviewed safety questions had resulted from this modification.
The spare letdown cooler features minor design changes that will eliminate previously identified
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failure sites and mitigate flow induced vibration problems on the secondary side of the cooler; this matter is discussed further under paragraph 3 of this report.
The piping up to and including the new l
I isolation valve inside containment is identified as Class 1.
Applicable codes and standards are as follows:
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Primary Side - Weld fabrication, inspection and testing, USAS B31.7-69 with Radiographic Code Case 72 l
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Secondary Side - Weld fabrication, inspection and testing USAS B31.1-67
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Materials, e.g.,
piping and fittings - ASME Section III through latest Addenda.
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Hydrostatic lesting - ASME Section XI 80W81
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Preservice Inspection of Welds - ASME Section XI 74S75
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Repairs and Replacement - ASME Section XI 77578 The new cooler was manuf actured by Graham Manufacturing Company, Inc.,
under Purchase Order F9047060-D.
FLUOR, under cor. tract with FPC is responsible for welding activities which are being performed in accordance with FPC's Nuclear Operations Welding (NOW) Manual. Welders and weld procedures, qualified to the latest issue in effect of ASME Code Section IX.
Welds on the primary side were fabricated with the GTA process using a consumable insert welding procedure, CR3-8/8-TS, Rev. 4.
Repairs, when required, were performed with a GTA welding procedure except with an open butt joint design.
Radiography was performed by the licensee, a.
Inspection of Welds Completed welds and others in process were inspected for compliance with code and regulatory requirements.
These were as follows:
Weld Size Configuration Status MU-85-232 3" x.438 Flange to Tee Complete
- MU-85-237 3" x.438 Pipe to Tee Complete
- MU-85-238 3" x.438 Flange to Tee Complete MU-85-239 3" x.438 Pipe to Tee Complete MU-85-276 3" x.438 Valve to Pipe Complete MU-85-277 3" x.438 Valve to E11 Complete MU-75-279 3" x.438 Pipe to Ell Complete MU-85-280 3" x.438 Pipe to Flange Complete MU-85-252 2 1/2 x.438 Pipe to Tee In Process
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MU-85-253 2 1/2 x.438 Pipe to Tee In Process MU-85-254 2 1/2 x.438 E11 to Tee Complete MU-85-256 3" x.438 Pipe to Reducer Complete MU-85-224 2 1/2 x.438 Pipe to Tee In Process MU-85-227 2 1/2 x.438 Pipe to Valve In Process In addition, the inspectors checked these welds for the following attributes and/or conditions as applicable to assure that work was conducted in accordance with a " traveler"; welding procedures and/or drawings; WPS assigned in accordance with applicable code; technique and sequence were specified; materials as specified; geometry as specified; fitup and alignment as specified; gas shielding and purging as specified; technique was as specified.
Welding consumables were as specified and consistent with the code; gas flow was controlled as specified; welding equipment was as specified; interpass temperature was controlled and consistent with the applicable codes; interpass cleaning was performed as specified; process control system has provision for repairs consistent with applicable codes; weld repairs were conducted in accordance with specified procedures; base metal repairs were properly documented; and welder identification was properly documented.
Completed welds were checked for weld reinforcement, arc strikes and weld spatter, cracks, laps, porosity, slag, oxide film, and undercut did not exceed prescribed limits.
b.
Welding Consumables The welding materials storage and :--ur station was inspected.
Procurement records, including receipt inspectien reports were reviewed for conformance with applicable specification and code requirements. Materials checked were as follows:
Type Size Heat Control No.
E-316 3/32"/
C5329T316 QCI-114653 E-316 1/8"/
05329T316 QCI-115782 E-308 3/32"/
C4264T308 QCI-111785 E-308 1/8"/
98795 QCI-115772 E-308 1/16"/
A4402R308 QCI-68938 c.
Materials (Piping, Fitting Valves)
Pertinent quality records of parts / components selected at random were reviewed to ascertain whether the records were in conformance with estabitshed procedures and whether the records reflected material /
component characteristics consistent with applicable code require-ments in material test reports, NDE records, certifications, quality releases, nonconformance/ deviation reports, receipt inspection reports, and code data reports as applicable.
Parts / components selected for this review were as follows:
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Item Control No.
Heat No.
Type 3" Anchor / Darling QCI-119317 Globe Value S/NE-6375-12-2 Tag #4MUS-V-424 3" Sched. 160 Tee QCI-117968 JXMZ SA-403/WP316 2 1/2" Sched. 160 Tee QCI-119334 JMXZ SA-403/WP304 3" Sched. 160 Pipe QCI-119865 469702 SA-312/304SS 3" Sched. 160 Pipe QCI-117070 K22391 A/SA312/316SS 3" Sched. 160, 90 Ell QCI-11931)
JXEL SA-403/316SS 3" Sched. 160, 90 Ell QCI-117313 F5163 SA-403/316SS 3" Blind Flange QCI-117968 WNRF SA-403/316SS Sched. 160
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3" x 2 1/2" Reducer QCI-117069 LAKP SA-403/304SS d.
Review of NDE Results - Radiographs Radiographs of completed field-fabricated welds were selected at random to ascertain whether the methods / techniques and extent of the examination complied with the applicable NDE procedure RT-003 Rev. 2; findings were properly recorded and evaluated by qualified personnel; whether if programmatic deviations were recorded as required; if personnel; instrument calibrations were designated, and qualifi-cations / certifications were on file.
The applicable code for this activity was discussed in the paragraph above. Records selected for this review were as follows:
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Radiographs MU-85-280 3" x.438 Pipe to Flange MU--85-279 3" x.438 Pipe to Elbow MU-85-256 3" x.438 Pipe to Reducer MU-85-254 2 1/2" x.438 Elbow to Tee The qualification records of six radiographer were reviewed to verify compliance with ASNT-TC-1A Recommended Practice requirements for this discipline.
e.
Welder Qualification Performance qualification records of welders involved in this work effort were reviewed for compliance with ASME Code Section IX requirements.
Original and current status records were reviewed for the following welders:
508, 511, 520, 521, 524, 525, and 530.
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Within these areas, the inspectors noted and subsequently discussed
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with cognizant personnel that all seven welders had been qualified to weld with the gas tungsten arc (GTA) welding process on a 0.223" thick test coupon.
Under the rules of ASME Section IX, QW-452, welders qualified on a 0.223 inch thick coupon are qualified to deposit weld metal to a maximum thickness of 0.446 inches with the process used in the qualification. Contrary to this requirement, on October 22, 1987, the inspectors determined that three of the seven aforementioned welders had fabricated field welds whose thickness l
ranged between 0.500 to 0.600 inches, and therefore, had exceeded the limits of their qualifications.
The licensee's cognizant engineer acknowledged this code violation and stated that the welders in question would be qualified by radiography under the provisions of
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ASME Code Section IX QW-304.1.
This failure to qualify welders in accordance with ASME Code Section IX requirements is in violation of paragraph (a)(1) of 10 CFR 50.55.a. This violation was identified as 302/87-32-01, Failure to meet ASME Code Section IX welder performance qualification requirements.
Attachment:
Failure Analysis
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ATTACHMENT FAILURE ANALYSIS Seal Return Coolers 3A and 3B (MUHE-2A, 2B)
The seal return coolers are two full-capacity (205 GPM) heat exchangers which transfer the heat from the seal return wa+er (controlled bleedoff from the reactor coolant pump seals) to the nuclear services closed cycle cooling water system.
The materials used to fabricate the tube side inlet / outlet nozzles, tubeside inlet / outlet head, and tubeside return head have been identified as supplied by Ray Miller, Inc. The materials in question serve as integral parts of the seal return cooler and should any of these materials fail, the equipment could not function properly.
However, during normal plant operation, only one of the straight tube design coolers is required to be online at any given time. Since Crystal River Unit 3 has provisions to isolate the seal return coolers (via manual valves MUV-77, 78, 79 and 80), either heat exchanger can be lined up.
Therefore, should a failure of the subject material occur, the redundant cooler can be aligned and normal plant operation can continue without any effect on public health or safety.
All materials listed in Enclosure 2 for the seal return coolers shall remain in service.
Should any of these materials fail, the standby seal return cooler will be placed in service and the failed material will then be replaced.
Makeup Tank (MUT-1)
The makeup tank serves as a surge tank for the makeup system and provides NPSH for the makeup pumps. The tank also serves as a receiver for letdown flow for the reactor coolant systems, reactor coolant pump seal return, chemical addition, and introduces hydrogen and nitrogen into the makeup system. Due to the fact that temporary changes in the primary system coolant volume are seen in the tank, the tank inventory varies. However, the normal water inventory is 2250 gallons.
The materials used to fabricate miscellaneous nozzles on this tank have been identified as being supplied by Ray Miller, Inc.
The nozzles which are affected by the Ray Miller material study are as follows:
System Piping Connections:
a)
MU System inlet and outlet conns.
4" BW, Pipe Neck i
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N and H Gas Vent Inlet 3/4", 3000# SW Coupling
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2
c)
Tank Vent 1", 3000#, SW Coupling
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Tank Relief 1",150#, RF Flange 1", BW, Pipe Neck Alarms and Interlock:
a)
High and Low Level Indication 1 1/2", 3000#, S'W Coupling b)
High and Low Pressure Indication 1", 3000#, SW Coup.ing Considering the effects of failure in any of the above referenced nozzles, it has been determined that the worst case failure will occur at the makeup system outlet connection.
A failure occurring at this nozzle will have detrimental effects on the entire makeup system.
A catastrophic failure (worst case) of the MUT-1 outlet nozzle will immediately result in destruction of the operating makeup pump.
Due to the loss of the operating pump, the entire makeup system will be temporarily out of service.
During normal plant operation the main functions of the makeup system are to provide for purification and recirculation of the reactor coolant and also provide seal injection water for the reactor coolant pumps (RCPs). Should the makeup system be lost, the primary concern will be loss of RCP seal injection water.
However, the preliminary evaluation indicates that even if the seal (s)
fail (i.e., leakage through seals), this will have minor effects on immediate availability of the reactor coolant pumps.
The makeup system will remain out of service until the operator realizes what has occurred via the makeup tank low-low level alarm and takes action by aligning the backup makeup pump to an alternate makeup water supply (i.e.,
borated water storage tank).
Once the backup makeup pump is started and has come up to its rated capacity, the makeup system functions will be restored.
All materials listed in Enclosure 2 for the makeup tank shall remain in service. A review of the material specifications indicate that fraud on this type of material is unlikely. Specifically, the materials were not required to be low carbon content or sold as " seamless", thus the types of fraud already attributed to Ray Miller are not applicable.
Furthermore, this is a low pressure, low temperature system and thus, catastrophic failure is very unlikely.
Electrical Penetrations (P-208, 209, 210, 211, 401, 402, 403, 404)
The materials used to fabricate the penetration feedthroughs have been identified as supplied by Ray Miller, Inc.
The subject penetrations contain the power supply wiring for the RCP motors as well as provide a leak tight containment atmosphere boundary, Upon evaluation of the penetration layout, it has been determined that the feedthrough needs to fail in two different exact locations to cause loss of containment function. No failure (s) of penetration electrical function are likely to result from mechanical nonconformance...
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All materials listed in Enclosure 2 for the electrical penetrations shall remain in service. In order to breach containment, the feedthroughs must have two failure points which occur at the penetration end plates. Since a double failure at specific locations is very unlikely to occur, continued service of these materials is justifiable by engineering judgement and required testing per 10 CFR 50 Appendix J.
3/8" Tubing The 78 feet of 3/8" tubing identified as supplied by Ray Miller, Inc., can not
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be specifically located in Crystal River Unit 3.
However, there is no (
documented evidence of any 3/8" tubing failures since plant startup.
A large percentage of the tubing utilized during construction of Crystal River (
Unit 3 serves as instrument air line connections for pneumatically operated valves. All valves which receive emergency safeguard signals have provisions built into them for system protection. These safety provisions are in the form of accumulator or failsafe valve design. Therefore, should an instrument air
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line break occur, the valve will still perform its safety function.
If the tubing was utilized in the primary fluid system, Crystal River Unit 3 has the capability to control the reactor cooling system inventory (via the makeup system) for up to a 3/8" line break. Therefore, should the tubing fail, the primary fluid system can still perform its safety function.
All materials listed in Enclosure 2 for miscellaneous tubing shall remain in service. Since there is no documented evidence of any similar tubing failures, the continued use of the subject tubing is justified by engineering judgement.
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