IR 05000413/1993022

Insp Repts 50-413/93-22 & 50-414/93-22 on 930719-23.No Violations Noted.Major Areas Inspected:Rhr Sys Pipe Failure, Emergency Diesel Generator Cooling Water Heat Exchanger Failure & Engineering Backlog

ML20056E732
Person / Time
Site:	Catawba
Issue date:	08/10/1993
From:	Blake J, Kleinsorge W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20056E718	List: IR 05000413/1993022 ML20056E717
References
50-413-93-22, 50-414-93-22, NUDOCS 9308250117
Download: ML20056E732 (11)

Text

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UNITED STATES

/pa arcq% NUCLEAR REGULATORY COMMISSION i y" ?$ REGloN 11

y'( ' 9 101 MARIETTA STREET. N.W., SUITE 2900 i 7,,

: E ATLANTA, GEORGIA 30323 0199 t !

%...../

Report Nos.: 50-413/93-22 and 50-414/93-22  !

l Licensee: Duke Power Company  ;

422 South Church Street l Charlotte, NC 28242 l Docket Nos.: 50-413 and 50-414 License Nos.: NPF-35 and NPF-52 Facility Name: Catawba 1 and 2 Inspection Conduct,ed;/-;Jul'y'I9:23,1993 Inspector: , ,

Wilijam P,/ Klein'sorge P.Ef s

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NOG /h

~Date' Signed '

Re3ctor' spector / -

l Approved by: _ /zs -

8 /v '/3 I Jeropfe'l Blak Chief Date Signed Maurials an rocess Section jEineering ranch

! Vivision of Reactor Projects i l'

! SUMMARY  !

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Scope:

This routine, announced inspection was conducted in the areas of Residual Heat Removal (RHR) system pipe failure, Emergency Diesel Generator Cooling Water Heat Exchanger failure, Condenser expansion bellows failure, and engineering l backlog.

l Results:

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The licensee has taken reasonable and necessary steps to prevent a recurrence of the RHR pipe failure. The actions taken by the licensee relative to the l Emergency Diesel Generator Cooling Water Heat Exchanger failure, and the l Condenser expansion bellows failure, were appropriate to the circumstance The engineering backlog is well controlle In the areas inspected, violations or deviations were not identified

9308250117 930811 PDR ADDCK 05000413 g PDR

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REPORT DETAILS 1. Persons Contacted Licensee Employees S. Bradshaw, Shift Operations Manager P. Dill, Radiation Protection Manager

• J. Forbes, Engineering Manager R. Jones, Instrumentation and Electrical Superintendent J. Lowery, Compliance Specialist ,

• W. McCollum, Station Manager J. Roach, Security Manager Other licensee employees contacted during this inspection included engineers, operators, security force members, technicians, and adminis-

, trative personnel.

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! NRC Resident Inspector

• R. Freudenberger, Senior Resident Inspector

• Attended exit interview 2. Residual Heat Removal (RHR) System Line Break On January 31, 1993, with Unit 2 in Mode four, Reactor Coolant (RC)

system at 340 'F and 375 psig, and preparations in progress to establish RHR cooling using the 2A RHR train, the RHR system sustained a water hammer severing RHR system vent valve ND-5 from the system. The break occurred at the toe of socket weld 2ND-32-14 in the pipe downstream of elbow H, as depicted on DPC drawing CN-2ND-32, Revision The elbow and the upstream fracture surface, as well as the downstream fracture surface and pipe segment, were sent to Babcock and Wilcox (B&W) for metallurgical evaluation. The licensee determined that the water hammer was the result of a procedural deficiency. The licensee's initial actions and the initial NRC inspection efforts are discussed in NRC inspection report 50-413,414/93-0 B&W report " Catawba 0.75-Inch Weld Socket Failure Analysis", LTC:940046-01:01 dated April 1993, concluded that the cause of the failure was high cycle fatigue. Two other shallow fatigue cracks were discovered in two other welds of the failed component. Suspected secondary cracks in the pipe nipple near the main fractur.e were probably secondary fatigue cracks that initiated at grinding cracks. The additional cracks in the compo-nent were identified with the solvent removable fluorescent, liquid penetrant examination techniqu The licensee examined numerous welds, using the solvent removable visible liquid penetrant technique, in vent assemblies of similar configuration to the failed assembly, on both Units I and 2. These examinations found no cracks. The welds in the assembly submitted to B&W, were not liquid

penetrant examined, prior to submittal to B&W, in order to protect the fracture surfaces from contamination with the penetrant dye, which could l compromise the metallurgical examination of those surfaces.

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The licensee replaced relief valve 2ND3 during the outag The licensee l indicated that examination of the seating surface of that relief valve, l which was in t.ervice during the pipe failure, had suffered a great deal i of damage. This damage is indicative of a chattering and vibrating l relief valve. Subsequent to the relief valve replacement, the licensee

! has collected vibration data on the piping assembly which replaced the

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assembly that had failed. The data was collected during mid-loop operations and while the RHR system was brought into service. The data l did not indicate levels of vibration great enough to cause a fatigue

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failure. With the replacement of the relief valve which appeared to have been chattering and vibrating, the environment has changed such that the vibration levels are not significan It is a logical conclusion, that the unsupported valve assembly was subjected to high cycle vibration emanating from the chattering nearby relief valv Fatigue cracks started from notches at the six and 12 o' clock positions on the horizontal pipe nipple. The final failure was the result of the water hamme .

The licensee has changed their operating procedures to the minimize the water hammer effects.

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Thc inspector interviewed licensee personnel and reviewed the B&W failure analysis report and vibration analysis data. With the reduction in vibration levels in the area of the failed assembly, the verification of l the absence of cracks in other similar configurations in both Units 1 and

2, and the modification of operating procedures, it appears that the I licensee has taken reasonable and necessary steps to prevent a recurrence of the failur Within the areas examined, no violations or deviations were identifie . Diesel Generator Cooling Water Heat Exchanger (DGCWHX) Failures ,

As reported in NRC Inspection Report 50-413,414/92-29 and discussed in l NRC Inspection report 50-413,414/93-05, the site had two tube failures in the Unit 2 DGCWHX During the spring 1993 Unit 2 refueling outage, the licensee contracted Cramer and Lindell to Eddy Current (EC) examine 100% of the unplugged tubes in both Unit 2 DGCWHXs, using a Cecco 1 probe. The Cecco 1 probe is sensitive to circumferential flaws. Heretofore the licensee EC examined the DGCWHXs in both Units using a bobbin coil which is not sensitive to circumferential indications. As a result of this latest examination, the licensee plugged 22 tubes in the 2A DGCWHX, and 48 tubes in the 2B DGCWH . . - _ . . - - - . - . _ - .- -_ - _- . . _ _ .

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REPORT DETAILS 1. Persons Contacted  !

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Licensee Employees  !

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S. Bradshaw, Shift Operations Manager  !

P. Dill, Radiation Protection Manager i

! *J. Forbes, Engineering Manager  !

R. Jones, Instrumentation and Electrical Superintendent J. Lowery, Compliance Specialist t

• McCollum, Station Manager i J. Roach, Security Manager l

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Other licensee employees contacted during this inspection included  !

engineers, operators, security force members, technicians, and adminis- ,

trative personne NRC Resident Inspector

• R. Freudenberger, Senior Resident Inspector l

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• Attended exit interview 2. Residual Heat Removal (RHR) System Line Break  ;

On January 31, 1993, with Unit 2 in Mode four, Reactor Coolant (RC)

system at 340 *F and 375 psig, and preparations in progress to establish  !

RHR cooling using the 2A RHR train, the RHR system sustained a water  ;

hammer severing RHR system vent valve ND-5 from the system. The break  ;

occurred at the toe of socket weld 2ND-32-14 in the pipe downstream of l elbow H, as depicted on DPC drawing CN-2ND-32, Revision 8. The elbow and j the upstream fracture surface, as well as the downstream fracture surface and pipe segment, were sent to Babcock and Wilcox (B&W) for metallurgical j evaluation. The licensee determined that the water hammer was the result i of a procedural deficiency. The licensee's initial actions and the '

initial NRC inspection efforts are discussed in NRC inspection report 50- ,

413,414/93-0 B&W report " Catawba 0.75-Inch Weld Socket Failure Analysis", LTC:940046-  !

01:01 dated April 1993, concluded that the cause of the failure was high cycle fatigue. Two other shallow fatigue cracks were discovered in two other welds of the failed component. Suspected secondary cracks in the pipe nipple near the main fracture were probably secondary fatigue cracks .

that initiated at grinding cracks. ' The additional cracks in the compo-

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nent were identified with the solvent removable fluorescent, liquid penetrant examination technique.

I The licensee examined numerous welds, using the solvent removable visible liquid penetrant technique, in vent assemblies of similar configuration to the failed assembly, on both Units 1 and 2. These examinations found no cracks. The welds in the assembly submitted to B&W, were not liquid penetrant examined, prior to submittal to B&W, in order to protect th . - . . . - . - - - . . . . . . . - -. - . . - - . - . - . -

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3 Although the Unit 1 DGCWHXs have not experienced tube failure, the licensee intends to EC examine 100% of tubes in both of the Unit 1 DGCWHXs using the Cecco 1 probe at the next Unit I refueling outag Failure analysis discovered that the Unit 2 DGCWHXs were flushed on a weekly basis while the Unit 1 DGCWHXs were not. The licensee, consider-ing that there could be a possible link between.the Unit 2 flushing and the circumferential tube failures, discontinued the flushing operation Attempts to instrument the floating tube sheet were unsuccessfu Engineering is continuing to investigate the source of the axial stresses that caused the circumferential failures. In addition the licensee is evaluating a tube material substitutio The inspector reviewed documents including Problem Investigation Process (PIP) reports, EC examination reports, and metallurgical analysis reports, and interviewed licensee personne The actions taken by the licensee were appropriate to the circumstance Within the areas examined, no violations or deviations were identifie . Unit 1 "E" Extraction Line Steam Leak and Condenser Tube Leak On May 27, the licensee detected a small loss of efficiency and a corresponding drop in shell pressure in the "E" feedwater heater. This indicated that the existing leak from the "E" extraction steam line expansion joint located in the "C" condenser had increased causing a decrease in the efficiency of the condenser. Conditions stabilized l following a reduction in efficiency equivalent to 50 Megawatts.

l On June 5, chemical analysis revealed a main condenser tube lea Power I

was reduced to 30 % to facilitate inspection in the "C" condenser water l box. One tube near the top of the tube sheet was found to be leakin The top row of tubes were plugged and the licensee returned to powe On June 9, an additional tube was identified as leaking. The licensee reduced power and plugged the two rows of tubes under the top row of tubes in the "C" condenser, and returned to power.

On June 12, Unit I was brought off line to repair the suspect ruptured HE expansion joint within the "C" condenser. Upon entry into the condenser i

the, extensive damage was noted to the 18-inch dia.1HE5 expansion joint and 10-inch dia. 1HD6 expansion joint. The upstream bellows Pc. 3, liner l Pc. 4 and shroud Pc. 10A were blown away on the 18-inch IHE5 expansion ,

I joint. Considerable stainless steel sheet metal and lagging was stripped j off feedwater heater IG3 and the 18-inch and 10-inch extraction lines I connected to the damaged expansion joint l Sheared pieces of lagging materials, including sheet metal, were scat-tered across the top of the exposed tubing of both condenser bundles ICI and 102. Obvious impingement damage was noted along the tubes in the top ;

row 205 of bundle 101. Tubes L205-10 through 14 and 16 were severed I between support plates 4 and 5 of bundle 1C1. The IHE5 expansion joint

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rupture coincided with the same support plate area where the tubes were severed in bundle 1C1. Numerous pieces of sheared lagging, indicated above, were found below the condenser bundles on the false bottom and along the sides of the bundles wedged between the tubes and the condenser wal EC examination of the perimeter tubes in bundles 101 and 102 identified approximately 93 blocked tubes in bundle ICI and approximately 76 blocked

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tubes in bundle 1C2. The two expansion joint assemblies were replaced in

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accordance with Minor Modification CNCE 60087. With repairs completed the licensee returned to power. The licensee suspects that the failure was caused by fatigue in the upstream bellows of the 18-inch expansion ,

join Samples of the failed bellows have been submitted for metallur- :

gical failure analysi ]

The inspector reviewed documents including EC reports, PIP reports and the Minor Modification, interviewed licensee personnel, examined the i condenser and the removed damaged materials. The actions taken by the l licensee were appropriate to the circumstance l l

This issue is further discussed in NRC inspection reports Nos 50- 1 r 413,414/93-17, and 50-413,414/93-1 l

! I Within the areas examined, no violations or deviations were identifie l 5. Engineering Backlog

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The engineering backlog at the Catawba station consists of the following: 1 (1) 180 active Modifications (NSM)s; (2) 27 inactive NSMs; (3) 10 active Design Studies (DS)s; (4) 8 inactive dss; (5) 513 open Minor Modifica-i tions (MM)s; and (6) 355 Station Problem Reports (SPR) To evaluate the engineering backlog the inspector reviewed procedures, i interviewed licensee personnel, and examined selected engineering i documents as described belo Documents Examined ID Revision Title

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O Catawba Nuclear Station Modification Manual Through interviews with the shift operations manager, the security manager, the radiation protection manager and the instrumentation and

' electrical superintendent the inspector determined that the customers of engineering were participants in the prioritizing and scheduling of the engineering backlog. They indicated that regulatory issues were promptly addressed but convenience items took longe l

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The licensee has 180 active NSMs and 27 inactive NSMs, The age of NSM i backlog items is shown in Figure 1. Figure 1 depicts the number of work items initiated each year that remain in the current backlog. To evaluate the age and appropriateness of the NSM backlog the inspector

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examined the following active and inactive NSM package The licensee's NSM data base indicated that 53 NSM have been cancele !

The inspector selected the below indicated canceled NSMs for independent l review, to assess the basis for cancellatio l

Nuclear Station Modification Packages Examined l ID Type Subject-

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CN50078 Active Add dampers and instruments & controls to enhance  !

7/24/84 the control room ventilating system  ;

CN10809 Active When filter units on VA system are shut down, 9/30/85 ventilation to health physics, chemistry, hot i machine shop and chemistry' labs is also shut  ;

down. Design to evaluate and correct the prob- '

le CN20207 Active Delete or replace the jacket water standpipe 5/7/86 level glass with a more reliable typ l CN1095710 Active Move fire hose cabinet IRF264 to make fire hose 3/9/86 more accessibl CN20402 Active Provide computer point for FWST sump HI level, 2/9/87 588' 7" l CN11190 Active Install power supply in radiation monitor 8/14/89 cabinet CN500ll Inactive Add composite sampler to the WT final discharge (no date) on a proportional flow basi CN50271 Inactive Provide Annunciator in ADIO fire detection panel l 10/23/85 in the control room and pull cable from IRFPS5930 l to provide annunciator input CN50308 Inactive Installation or upgrade of barriers in utility 1 l 3/17/86 ports l

CN50340 Inactive Delete doors MLP1, MSP1, and MSV1 from the 10/9/86 security system CN11055 Inactive Provide individual fuses for MFWIV actuators SV's 5/13/87 and provide reverse bias diodes for the SV coils CN11196 Inactive Install new flow glass for w m cca50 with a 9/18/89 three way valve l

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l Nuclear Station Modification Packages Examined

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ID Type Subject CN50236 Canceled Extend the demineralized water inlet on the YM 9/24/85 storage tank below the water level to minimize nitrogen entrainmen CN20458 Canceled Install shims in NV letdown heat exchanger to 7/22/87 reduce U-tube vibration CN50387 Canceled Install chemical addition system to deliver mol- l

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7/27/88 luscide to the RN pumps intake pits CN11233 Canceled Delete main steam leak detection system 8/17/90 CN50423 Canceled Install backdraft dampers in VC ductwork j 3/15/91

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CN11265 Canceled Replace NS heat exchanger IA and 1 /27/91 The licensee has 335 open SPRs. The age of SPR backlog items is shown in Figure 1. To evaluate the age and appropriateness of the SPR backlog the inspector examined the following SPR package Station Problem Reports Examined ID Type Subject i CNCE0191 Open Replace IRCPG/5080/5100/5120/5140 with 0-100 PSI l 7/31/85 gauges Marsh model E0148B i

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CNCE0870 Open Correct drawing errors for Unit I containment 2/29/86 mechanical equipment building sump pump CNCE1494 Open Rewire limit switches for proper operation of 11/19/87 IBBSV1750 and IBBSV1780 CNCE1901 Open Rewire contacts for CNILTPS 5070 and 5071 to agree 8/9/88 with vendor drawings i CNCE2214 Open Revise documents to reflect existence of valves 3/6/89 ISVS814 and 815 CNCE2547 Open Determine correct setpoint temperature of switch 11/1/89 and correct documentation as necessary  !

l The licensee has 513 open mms. The age of MM backlog items is shown in Figure 1. To evaluate the age and appropriateness of the MM backlog the inspector examined the following MM packages.

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Minor Modifications Examined l

ID Type Subject l

CNPR02867 Open KRFS6290 and KRFS6320 can not be calibrated to 9/15/87 the desired set point j CNPR02892 Open Reactor coolant pump seal leakoff stand pipe  !

10/1/87 has design conditior, af 65 psia at 200 degrees ' During startup of A pump the stand pipe may I be isolated from the NCDT and be over l pressurize l CNPR03290 Open Valves 1(2) CF-90,1(2) CF-89,1(2) CF-88 and l 3/11/88 1(2) CF-87 on 1(2) MC2 are currently labeled S/G A, B, C, D, CF CONT. ISOL. BYPASS re- i spectively 1 CNPR03634 Open The rod control (IRE) system power cabinet 8/22/88 power supplies are subject to failure during power operations without providing any indica-tion of alarm of failure CNPR04587 Open Flow diagrams for systems providing flow to 12/13/89 radiation monitors show the flow switches in the QA portions of the sample line. Flow swit-ches for EMF system are not Q l CNPR04677 Open The error rate for the security access control 2/6/90 system is extremely hig Engineering Backlog ,

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Based on the sample of NSMs, mms and, SPRs examined the oldest items are operational and personnel convenience items. Based on the sample of canceled NSM the cancellations were appropriat Figure 1 indicates that the majority-of backlog items are relatively new (three years old or less).

Total Total Total Modifications' Station .

Minor NSM Problem Modifications Reports MM SPR Backlog Total 7/93 207 335 513 Closed in last 12 Months 120 283 336 As can be seen by the table above, the number of items closed in the last i 12 months represents a substantial portion of the current backlog total If this trend continues the backlog should continue to~ shrin The licensee does track the engineering backlog, however the inspector found no evidence that they trended the dat By interview and independent review of selected records, the inspector determined that the engineering backlog is well controlle ;

i Within the areas examined, no violations or deviations were identifie ;

l Exit Interview '

i The inspection scope and results were summarized on July 23, 1993, with !

those persons indicated in paragraph 1. The inspector described the l areas inspected. Although reviewed during this inspection, proprietary !

information is not contained in this report. Dissenting comments were ;

not received from the license l Acronyms and Initialisms B&W -

Babcock and Wilcox CNCE -

Modification DGCWHX -

Diesel Generator Cooling Water Heat Exchanger DS -

Design Study FWST -

Refueling Water Storage Tank IRE -

Rod Control MFWIV -

Main Feedwater Isolation Valve MM -

Minor Modification ND -

Residual Heat Removal NPF -

Nuclear Power Facility NRC -

Nuclear Regulatory Commission NS -

LPSW Intake Screen backwash NSM -

Modification NV -

Chemical and Volume Control l

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! 9 Acronyms and Initialisms Cont's l Professional Engineer PIP -

Problem Investigation Process psia -

Pounds per Square Inch Absolute RC -

Reactor Coolant RCDT -

Reactor Coolant Drain Tank RF -

Fire protection RHR -

Residual Heat Removal System RN -

Nuclear Service Water SPR -

Station Problem Report SV -

Main Steam Vent to Atmosphere VC -

Control Area Heating Ventilating and Air Conditioning WT -

Sanitation and Water Treatment l

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