IR 05000295/1996013

Insp Repts 50-295/96-13 & 50-304/96-13 on 961001-03,28-30 & 1107.Violations Noted.Major Areas Inspected:Maint. Observed & Reviewed ISI Procedures,Personnel Certifications

ML20133A335
Person / Time
Site:	Zion  File:ZionSolutions icon.png
Issue date:	12/23/1996
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20133A321	List: IR 05000295/1996013 ML20133A317 ML20133A329
References
50-295-96-13, 50-304-96-13, NUDOCS 9612310053
Download: ML20133A335 (13)

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U.S. NUCLEAR REGULATORY COMMISSION REGION ll1 Docket Nos: 50-295;50-304 i

License Nos: DPR-39; DPR-48  :

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Report No: 50-295/96013(DRS): 50-304/96013(DRS) ;

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Licensee: Commonwealth Edison Company J

Facility: Zion Generating Station, Units 1 and 2 Location: 105 Shiloh Boulevard

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Zion, IL 60099

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l Dates: October 1-3,28-30, and November 7,1996 Inspectors: M. S. Holmberr,, Reactor Inspector .

K. Green-Bates I

Approved by: Wayne Kropp, Chief Engineering Branch 1

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9612310053 961223 PDR ADOCK 05000295 G PM

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Report Details

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j 11. Maintenance M1 Conduct of Maintenance  !

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M 1.1 Observation of Unit 2 Inservice Insoection (ISI) Work Activities l Insoection Scoce (73753. 73755. 73052)

t inspectors observed work, reviewed ISl procedures, personnel certifications, and e reviewed data associated with the following activities:

l * AtiE C9mbustion Engineering (CE) contracted personnel performing Eddy curreni examination (ET) and analysis of the steam generator (SG) tubes.

. * CE contracted personnel performing ET data acquisition and analysis for CE

sleeve installation.

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j * CE contracted personnel performing SG tube sleeve installation and upper sleeve welds.

• CE contracted personnel performing SG tube lower sleeve weld visual inspection.

i * CONAM personnel performing ultrasonic examination (UT) of line 2FWOO3 20" lateral te * Lambert MacGill Thomas, Inc. (LMT), personnel performing dye penetrant

(PT) examination of an 8" RHR hot leg safety injection line tee welds 14,15 and 16 identified on drawing COM-1-4402.

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Observations and Findinas

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l CE contracted personnel performed the ET of the Unit 2 SGs using multi-frequency ET equipment. Probes and equipment configurations used were equivalent or were demonstrated equivalent to EPRI NP 6201 Appendix H qualified equipment. The ET l inspection scope for each SG included:

, * Motor rotated pancake coil (MRPC) inspection of 100 percent of the hot leg

and cold leg side of the tube sheet region.

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e inspection of 100 percent of inservice row 1 and 2 U-Bends and inservice sleeves using a motor rotated + Point * piob t

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e A 100 percent full length inspection of inservice non-sleeved SG tubes using a bobbin coi ,

e MRPC inspection of all hot-leg side dented tube support plate intersections of greater than 5 volt i The inspector concluded that the examinations performed exceeded technical specification requirements and generic letter (GL) 95-03, "Circumferential Cracking ,

of Steam Generator Tubes," commitment Zion station "SG ET Data Analysis Guidelines," revision 2, required all free-span SG <

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tube degradation to be reported by primary and secondary analysts as non-quantifiable indications (NQl), requiring reinspection and evaluation with a motor ,

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rotating pancake coil (MRPC) type probe. This included indications, which would have formerly been dispositioned as SG tube manufacturing burnish marks with no further inspection / evaluation. The inspector considered this a conservative change in data analysis and was indicative of a focus on safet As a result of ET of the Unit 2 SGs, greater than 1 % of the SG tubes were ;

characterized as defective and the following SG tube repairs were planned:

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i F Re-rolled tubes Sleeved tubes Plugged tubes )

2174 226 438

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The predominant forms of tube degradation detected were axial oriented inside !

diameter connected ET indications located at the SG tube roll transition regio These indications were primarily in the hot leg side of the tubesheet and were indicative of primary water stress corrosion crackin ET also identified potential loose parts in each steam generator. Secondary side visual examinations confirmed the loose parts. Because of unsuccessful attempts at removal, the following parts will remain in th^ ateam generator * SG C less than 0.063" diamett ' ire j

• SG D 0.125" diameter wire (presumed to be weld wire)

0.125" diameter (presumed to be weld wire)

0.125" diameter wire (presumed to be weld wire)  !

In SG A, a 0.125" diameter wire was detected and removed in addition, a less than 0.063" diameter wire was confirmed in SG B and the engineering staff presumed the wire was removed during sludge lancing operations. The inspectors reviewed an engineering assessment to evaluate the effects on the SG tubes from fretting and impact sliding type wear for the loose parts remaining in SG C and SG D and for the wire presumed removed from SG B. The calculations performed estimated the time required to reach minimum wall based on data and methodology in Westinghouse WCAP-12510," Zion Units 1 and 2 Steam Generator Loose Object

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, Evaluation Licensing Report," issued in August 1990. The inspectort. concluded that the evaluations demonstrated a conservative margin with respect to length of i

the plant operating cycle and the time required to reach minimum tube wall !

thickness. The estimated minimum times required to reach minimum allowed tube wall thickness ranged from three to fourteen years. The engineering program manager stated that the affected tubes would be reinspected and wear assessed during the next SG inspection.

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Five SG tubes (hot leg side portions) had been selected and were scheduled to be removed for metallurgical evaluations of the tube degradation mechanism Additionally, thirty-three SG tubes were selected for in-situ pressure testing. These tubes contained representative examples of active SG tube degradation modes based on ET data. The in-situ pressure testing was performed to gather data on potential SG tube leakage and/or burst pressures to allow assessment of the as found integrity of SG tube The licenses staff identified fcur SG tubes conteining axial ET indications with measured lengths greater than that allowed by an e.visting calculation for tube structuralintegrity. The assessment to demonstrate that these four SG tubes were bounded by the licensing basis during the last operating cycle was considered an unresolved item pending review of this assessment by the NRC (50-295/ 1 96013-01(DRS); 50-304/96013-01(DRS)). The affecmd tubes were i

SG C i I

Tube location ET Indication indication Maximum Length Length Allowed Row Column Type Location Inches By Calculation *

1 13 mal U-bend 0.65 0.53 1 90 MAI U-bend 0.55 0.53 1 72 6At U-bend 0.59 0.53 19 48 sal TTS 0.46 0.43

+ 0.97 "  ;

MAI Multiple Axial Indications SAI Single Axial Indications TTS Top of SG tubesheet

Values listed were reported by the engineering staff based on a calculation, which was not considered final pending completion of quality assurance review During a secondary side SG pressure test, the licensee identified sixteen SG tubes with dampness / wetness at the primary side tubesheet face. Ten of these tubes had rsjecteble ET indications and six tubes had been sleeved with CE Tungsten inert Gas (TIG) welded sleeves, which had been installed in previous outages. The licensee identified flaws (e.g., porosity, blow holes or lack of fusion) in each of the six (damp / wet) sleeved tubes at the lower sleeve weld. As a result, the licensee

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reexamined the lower sleeve welds for all Unit 2 inservice sleeves. Upon completion of this inspection a total of eleven (eight in SG A and three in SG B) SG tubes with CE TIG welded sleeves, were identified as having rejectable flaws in the lower sleeve weld. These sleeves had been installed during outages in 1992 and 1995. The licensee investigation into the causes for returning to service CE TIG welded sleeves with rejectable indications in the lower sleeve weld and the causes of the potential leakage in non-sleeved SG tubes was still underway at the conclusion of this inspection. This was considered an unresolved item pending completion of the licensee's investigation and review of additional information (50-295/ 96013-02(DRS); 50-304/96013-02(DRS)). _Cpnclusions For the 1996 Zion Unit 2 refueling outage, the predominant forms of SG tube degradation detected were axial oriented eddy current indications at the SG tube roll transitions indicative of primary water stress corrosion cracking. Based on the following, the inspectors concluded that the licensee's safety focus in managing steam generator tube degradation has improved when compared to the 1995 refueling outage:

e the extensive scope of the SG eddy current testing (ET) performed, e the conservative requirement in the ET data analysis guidelines which required initial disposition of SG tube manufacturing burnish marks as repairable indications pending further ET, and e the extensive in-situ SG tube leakage / burst tests, and the five SG tubes removed for metallurgical analysi However, the inspectors had concerns with four tubes in SG C with ET indication lengths greater than existing calculations allowed for tube structural integrity and the causes of the potential SG tube leakage identified for ten tubes during the secondary side SG pressure testin M4 Maintenance Staff Knowledge and Performance M4.1 Potentially Defective SG Tubes Returned to Service Insoection Scone (73753,73755)

For the Units 1 and 21995 refueling outages, inspectors reviewed the SG ET cata, nonconformance reports and corrective actions taken for ET weld zone indicaticns (WZl) recorded for Combustion Engineering (CE) TIG welded sleeves. An unresolved item (50-295/96007-04(DRS); 50-304/96007-04(DRS)) concerning disposition of the WZI was identified during a previous inspection (see paragraph M 8.1 ).

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b. Observations and Findinas The inspectors identified two examples of failure to take prompt corrective actions to evaluate the severity and cause of the ET indications using a qualified nondestructive examination (NDE) technique prior to returning potentially defective SG tubes to service. The two examples pertained to eddy current testing (ET) using a rotating + Point probe which detected the following sleeved SG tubes with weld zone indications (WZl):

o Twenty-nine sleeved SG tubes during the February 1995 Zion Unit 2 refueling outage (RFO).

e Fifty-two sleeved tubes during the October 1995 Zion Unit 1 RF During both RFOs, the affected tubes were visually examined after ET and documented as acceptable on nonconformance reports by CE, the sleeving vendo No direct correlation was established between the visual examinations and ET indications. Based on meeting ultrasonic and visual weld examination acceptance criteria, the licensee:

e Returned to service 26 sleeved of the 29 Unit 2 SG tubes that had ET WZI without performing an assessment of the impact on SG tube integrit Corrective actions were not implemented until March 25,1996, when the licensee completed an operability assessment (96-01283, revision 0) for the sleeved SG tubes which contained ET WZl. The March 25,1996, operability assessment documented that the ET WZI were potentially the result of weld defects (lack of fusion, oxide film inclusions and sleeve weld outside diameter suck-back) caused by inadequate cleaning of the parent tube during sleeve installation. This assessment also concluded that regulatory requirements for structural and leak tight integrity were met for the affected SG tubes, e Returned to service 43 of the 52 Unit 1 sleeved SG tubes that had ET WZl without performing an assessment of the impact on SG tube integrit Corrective actions were not implemented until February 1996 (Z1M05 outage), when the affected 43 SG tubes with WZI were removed from service (plugged).

When initially detected, the natures of the SG ET WZls were unknown and ISI !

personnel assumed ET WZI were related to weld geometry. NRC staff reviews have concluded that the NDE techniques used were not qualified to size the indications revealed by ET, since the nature of the defects was unknown. Consequently, the decision to return these tubes to service was inappropriate, since the impact on SG tube integrity was unknown. Appendix B to 10 CFR Part 50, Criterion XVI, requires in cases of significant conditions adverse to quality, that the measures shall ensure that the cause of the condition is determined and corrective action is taken to preclude repetition. In this case, measures to identify conditions adverse to quality (e.g., eddy current testing, visual, and ultrasonic techniques) were implemente .

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However, the measures were not capable of determining the severity of the condition and the cause of the condition. Failure to take prompt corrective actions to evaluate the severity and cause of the ET indications using a qualified NDE technique prior to returning the affected SG sleeves to service is considered a violation to 10 CFR Part 50 Appendix B Criterion XVI, " Corrective Actions" (50-295/96013-03(DRS); 50-304/96013-03(DRS)).

Inspectors reviewed the ET data for 72 SG tubes with ET weld zone indications identified for the 1996 Unit 2 inspection. ET data analysis guidelines used for the '

1996 ET data analysis were explicit and contained examples of ET data responses j to known flaws / defects within the sleeve welds. However, no ET data analysis i guidelines existed for the 1995 2T data analysis. Inspectors concluded that the i increased number of ET WZls identified in the 1996 inspection when compared to l the February 1995 inspection was largely due to specific explicit data analysis l guidelines and improved analyst training used for the 1996 inspectio )

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M4.2 ASME Code Reauirements Not Met For Pioe Suonort Insoections ) Insoection Scooe (73755,7305.2)

l On September 5,1996, a resident inspector identified excessive gaps on pipe I supports in the auxiliary feedwater and safety injection systems. Inspectors reviewed documentation associated with ASME Code required inspections '

completed during past outages on these pipe supports to determine if Code and l NRC requirements had been me !

! Observations and Findinas 1

Piping supports identified by the resident inspector, with excessive clearance l between the building structure and support baseplate (herein referred to as gap) for safety injection and auxiliary feedwater systems, included the following supports:

e For Unit 1 - SlH-1058, SIRS-1229, SlH-1059, SlH-1057, SlH-1057A, SlH-1057B, SlH-1055, SIRS-1230, SlH-105 e For Unit 2 - 2MSRS-1133B, ASH 2123, FWRS-2250, FWRS-2251, FWRS-2253, FW RS-225 For the supports identified above, the gaps were measured and found to be outside tolerances specified in procedure NWSP-S-05, revision 3, " Concrete Expansion Anchors." This condition was documented in Problem identification Form 96-242 These supports had been previously examined (1992 for Unit 1,1990 and 1992 for Unit 2) as part of the ISI program, in accordance with procedure VT-3/4-1, revision 2, "VT-3/4 Visual Inspection Performed for Section XI." 10 CFR 50.55a(g)(4)(ii) requires the inservice examination of components to comply with the ASME Code incorporated by reference. The ASME Code,1980 Edition Winter 1981 Addenda,Section XI, Table IWF-2500-1, requires a VT-3 inspection on

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component supports, which includes the support up to the building structur i Section XI of the ASME Code, paragraph IWA-2213(b), required the following:

"The VT-3 visual examination may require, as applicable to determine

, structural integrity, the measurement of clearances, detection of physical i

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displacement, structural adequacy of supporting elements...."

Contrary to these requirements, procedure VT-3/4-1, revision 2, used for performing VT-3 Code inspections, lacked acceptance criteria for this gap and did not contain explicit requirements to measure / verify this gap or clearance. Failure of procedure l VT-3/4-1 to contain the necessary acceptance criteria to verify clearances (gap between piping support plates and building structure)in accordance with VT-3 )

inspection requirements of IWA-2213(b) of Section XI,1980 Edition Winter 1981 j

, Addenda of the ASME Code is considered a violation of 10 CFR 50, Appendix B, i Criterion V (50-295/96013-04(DRS); 50-304/96013-04(DRS)). l Cons quently, the visual examinations performed were inadequate in that l clearances between the building structure and the support baseplate were not l verified to be within established criteria specified in procedure NWSP-S-05. ISI staff l stated that inspection of base plates during the course of visual inspection would be I supplemented by a Zion Component Examination Detail Instruction Form that will add the additional criteria not provided by procedure VT-3/4-1.

. M4.3 Reactor Vessel Insoection ISI Data Review Insoection Scone (73755)

The inspectors reviewed ultrasonic testing (UT) data recorded and coverage estimates for examinations of the Unit 1 reactor vessel performed in November of 1993 and the Unit 2 reactor vessel performed in December of 199 Observations and Findinas On October 1,1996, the licensee identified limitations that prevented performing a

. 100 percent weld volume inspection of reactor vessel welds performed in 1993 for Units 1 and 2. These limitations were caused by internal vessel appurtenances that physically limited UT scanning ability. This condition was documented on Problem Identification Form 96-2993, 10 CFR 50.55a(g)(6)(ii)A," Augmented Examination of Reactor Vessel," requires the vessel to be exami Ted in accordance with table IWB-2500-1 of the 1989 Edition,Section XI of the ASME Code. This table specified the inspection of essentially 100 percent (e.g., greater than 90 percent) of each reactor vessel shell weld (Category B-A). Four reactor vessel shell welds in each Unit were estimated by the licensee to have less than the 90 percent minimum volumetric coverage specified by table IWB-2500-1, without an NRR approved alternative. However,10 CFR 50.55a(g)(6)(ii)A(3) allowed deferring until the end of the first period of the next ISI interval the augmented vessel inspection requirements for licensees with fewer than

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l 40 months remaining on the ISIinterval in effect on September 8,1992. The ISI l staff engineer reported that the first period of the current ISIinterval would end December 31,1997, for Unit 1 and September 19,1998, for Unit 2. These dates incorporated the ASME Code Section XI, IWB-2412(b) option for extending the period by one year. At the conclusion of this inspection, the licensee was unsure of how compliance with the 10 CFR 50.55a(g)(6)(ii) requirements would be accomplished. The ISI program manager acknowledged opportunity for improvement in staff knowledge and implementation of the augmented vessel examination requirements. The licensee's actions to comply with the augmented examination of the reactor vessel is considered an inspection followup item pending further review by the NRC (50-295/96013-05(DRS); 50-304/96013-05(DRS)).

M4.4 Conclusions on Maintenance Staff Knowledge and Performance i

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Following the 1995 inspections of Unit 1 and Unit 2 SGs, the decision to return SG tubes with weld zone indications to service was considered by the inspector to be nonconservative and did not demonstrate a good staff safety focus. However, the new ET data analysis guidelines for the 1996 ET data analysis should improve '

performance in this area. The inspectors also concluded that the lack of a defined success path and schedule for complying with NRC requirements governing the augmented examination of the reactor vessel and the violation of ASME Code VT-3 visualinspection requirements for pipe supports,is indicative of staff weaknesses in i knowled Ce and implementation of these requirement j i

M8 Miscellaneous Maintenance issues j M 8.1 (Closed) Unresolved item (50-295/96007-04(DRS): 50-304/96007-04(DRS)):

During a previous inspection (50-295/96007(DRS); 50-304/96007(DRS)), {

inspectors identified an unresolved item concerning the technical acceptability of the ;

licensee's decision to return sleeved SG tubes with eddy current testing (ET) weld zone indications (WZl) to service, based on visual and ultrasonic examination During this inspection, the inspectors identified a violation (see paragraph M4.1) for failure to take prompt corrective actions (e.g., evaluate the severity and cause of the ET indications using a qualified NDE technique) prior to returning potentially defective SG tubes to service. This item is close ,

I lil. Enaineerina E2 Engineering Support of Facilities and Equipment l

E Insoector Uodated Safety Analvsis Reoort (UFSAR) Review While performing the inspections discussed in this report, the inspectors reviewed UFSAR sections:  ; Classification of Structures Components and Systems 9 l l

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3.6.2. Main Steamline and Feedwater Piping

5. Compliance with Codes and Code Cases l 5. Steam Generators l

The inspectors did not identify any UFSAR discrepancies as a result of this special l revie ;

l l V. ManaaementIAeetings  !

X1 Exit Meeting Summary i At the conclusion of the inspection on October 30,1996, and final phone exit on

, November 7,1996, the inspector met with licensee representatives identified herein

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and summarized the scope and findings of the inspection activities. The inspector questioned licensee personnel as to the potential for proprietary information in the likely inspection report material discussed at the exit. No proprietary information was identified.

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PERSONNEL CONTACTED Licensee

Commonwealth Edison Comoany (Comed)

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• G. Schwartz, Station Manager

, #*R. Skowzgird, Programs Lead Engineer

'J. Blomgren, SG and RPV Group

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"M. Sears, SG and RPV Group

1. • S. Wilson, SG and RPV Group

• H. Smith, SG and RPV Group

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• B. Griffin, Engineer

• K. Moser, Engineer

} *T. Cook, SG Program Engineer Maintenance Engineering i *S. Davis, ISI Program Engineer Maintenance Engineering

1. • D. Saccommando, Nuclear Licensing Division

• D. Beutel, Regulatory Assurance

U.S. Nuclear Reaulatorv Commission (NRC)

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1 *R. Westberg, Senior Resident inspector E. Cobey, Resident inspector

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lilinois Deoartment of Nuclear Safety (IDNS)

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"J. Yesinowski, inspector s

Hartford Steam Boiler Insoection and Insurance Comoany (HSB)

l D. Oakley, ANil The NRC inspector also contacted and interviewed other licensee and contractor employees.

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' Denotes those present at the exit interview on October 30,1996.

# Denotes those contacted for the final exit interview per teleconference on November 7, 1996.

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INSPECTION PROCEDURES USED e

IP 73753: Inservice inspection IP 73755: Inservice inspection, Data Review and Evaluation IP 73052: Inservice inspection, Review of Procedures ITEMS OPENED, CLOSED, AND DISCUSSED

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50-295/96013-01(DRS); URI Four SG tubes with ET indications lengths greater than 50-304/96013-01(DRS) current calculations for structural integrity.

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50-295/96013-02(DRS); URI Causes for potential SG tube leakage identified during

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50-304/96013-02(DRS) secondary side SG pressure tes /96013-03(DRS); VIO Failure to take prompt corrective actions prior to

50-304/96013-03(DRS) returning potentially defective SG tubes to service, j

50-295/96013-04(DRS); VIO Failure to meet ASME Code VT-3 visualinspection 50-304/96013-04(DRS) requirements for piping suppor /96013-05(DRS); IFl implementation of 10 CFR 50.55a(g)(6)(ii)A 50-304/96013-05(DRS) requirements for augmented reactor vessel examination Closed 50-295/96007-04(DRS); URI Technical acceptability of returning SG tubes with ET l 50-304/96007-04(DRS) WZI to service.

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. LIST OF ACRONYMS USED

i i ANil ' Authorized Nuclear Inservice Inspector ASME American Society of Mechanical Engineers

CE Combustion Engineering l EPRI Electric Power Research Institute ET Eddy Current Testing FAC Flow Accelerated Corrosion

! GL Generic Letter 1 'IFl Inspection Follow-up Item l lP Inspection Procedure 1

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IR_ inspection Report ISI Inservice Inspection MAI Multiple Axial Indications NRC Nuclear Regulatory Commission PlF Problem Identification Form PT Dye Penetrant Testing RFO Refueling Outage sal Single Axial Indications SG Steam Generator TS Technical Specification l TTS Top of SG tubesheet I URI Unresolved item UFSAR Updated Safety Analysis Report 1 UT Ultrasonic Testing '

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VIO Violation WZI Weld Zone Indication (s) ,

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