IR 05000254/1997018

Insp Repts 50-254/97-18 & 50-265/97-18 on 970922-25. Violations Noted.Major Areas Inspected:Maintenance & Engineering

ML20198T322
Person / Time
Site:	Quad Cities
Issue date:	11/10/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20198T288	List: IR 05000254/1997018 ML20198T282 ML20198T311
References
50-254-97-18, 50-265-97-18, NUDOCS 9711140258
Download: ML20198T322 (11)

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U.S. NUCLEAR REGULATORY COMMISSIO t ..

REGION lli -

Docket Nos: 150-254;50-265 License Nos: DPR 29; DPR-30

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. Report Nos: 50-254/97018(DRS); 50 265/97018(DRS)

' Licensee: Commonwealth Edison Company Facility: Quad Cities Nuclear Power Statio Units 1 and 2 Location- 22710 206th Avenue North Cordova,IL 61242

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Dates:. September 22-25,1997

inspector: M. S. Holmberg l Approved by: J. A. Gavula, Chief, Engineering Branch 1 Division of Reactor Safety

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EXECUTIVE SUMMARY Quad Cities Nuclear Power Plant, Units 1 and 2 NRC Inspection Report 50-254/97018; 50-265/97018 This inspection included a review of the implementation of the inservice inspection (ISI)

program for the past Unit 1 and 2 refueling outages. Additionally, resolution of existing NRC open items associated with the ISI program was reviewed.-

Maintenance:

• A reactor vessel nozzle weld examination may not have met ASME Code requirements for volumetric examination, based on scan limitations. Further, this examination had not been documented as limited (Section M3). '

• In general, the licensee's ISI program implementation met ASME Section XI requirements and the licensee's planned initiative to revise and improve the ISI program was positive (Section M7).

• Adequate bases were not available to justify exclusion of two systems from the scope of the ISI program (Section M7).

• The number and variety of ISI issues documented in the problem identification forms, indicated that additional oversight is warranted, to assure that program weaknesses are

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corrected (Soction M7). .

• A violation was cited for failure to take corrective actions for the ineffective induction heat stress improvement (lHSI) process for the Unit 1 recirculation system welds (Section M8).

Enoineerino:

• The licensee's evaluation concluded that the ASME Code allowed fatigue usage factor for the reactor vessel head studs will be exceeded in 1998; however, the licensee has not taken any action to resolve this issue (Section E2).

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Reoort Details .-

II. Maintenance M3 Maintenance Pracedures and Documentation M3.1 Inservice insoection flSI) Examination Documentatiqa Insoection Scoce (73753. 73755)

The inspector reviewed a sampl0 of 15 individual inservice examination reports and IS!

personnel certifications for examinations recorded in the following documents:

o " Quad Cities Nuclear Power Station Units 1 and 2 Inservice Inspection (ISI) Post Outage (90 Day) Summary Report Refueling Outage Number O2R14 [Ouad Cities Unit 2 Refueling Outage 14]"

e " Quad Cities Nuclear Power Station Unit 1 ISI Post-Outage (90 Day) Summary Report Refueling Outage Number 01R14" e " Quad Cities Nuclear Power Station Unit Two ISI Post-Outage (90 Day)

Summary Report Refueling Outage Number Q2R12" e " Quad Cities Nuclear Power Station Unit One ISI Post-Outage (90 Day)

Summary Report Refueling Outage Number Q1R12" The inspector reviewed the licensee documented inservice examination records for the reactor vessel shell welds performed during 02R14 and the following relief request associated with this examination:

• " Quad Cities Nuclear Power Station Unit 2 Results of Augmented Examination of the Reactor Pressure Vessel (RPV) Shell Welds and Relief Request Pursuant to 10 CFR 50.55a(g)(6)(ii)(A) NRC Docket Number 50-265" Observations and Findinos Personnel certification records met ASME Code requirements for the selected examinations completed in refueling outages Q1R14 and O2R14. The examination records were complete with the appropriate calibration records and required examination data. For the selected examinations reviewed from the 90-day summary reports, the inspector identified no deviations from ASME Code requirements, except as discussed below:

The inspector identified that the inservice ultrasonic examinations of the Unit 2 reactor pressure vessel nozzle weld (NBB) completed April 5,1993, had scan limitations recordad on the examination data sheet. This examination was submitted to the NRC in

" Quad Cities Nuclear Power Station Unit Two ISI Post-Outage (90 Day) Sum' mary Report Refueling Outage Number Q2R12," and no liispection coverage limc tion had been recorded. The inspection clocumentation was insufficient to permit the inspector or l

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the licensco staff to quantify the extent of examination coverage actually obtained. The .,

inspector considered that ASME Code volumetric inspection requirements may not have /

been met and confirmed that an ASME Code relief request had not been approved by the NRC for this examination. The licensee subsequently issued problem identification form (Plf) Q1997-03604 to address this issue. This was considered on unresolved item pending demonstration by the licentee that ASME Code requirements for exar.iination coverage of this wold hart been met (URI 50 254(265)/97018-01).

The licer'see had completed an augmented inspection of the Unit 2 reactor vessel shell welds to meet 10 CFR 50.55a(g)(6)(ii)(A) rer!uirements in April 1997. Due to access restrictior.s the licensee was unable to achieve the ASME Cede Section XI required coverage for each shell weld and submitted c Code relief request on September 18, 1997. However the inspector identified the following issues with the licensco's September 18,1997 Code relief submittal:

• Vertical shel! welds and portions of circumferential welds had been secnned from only orie side of the weid, vice both sides. The submittal appeared to take examination credit for the examinations performed from one side of the wel The effectivo:ess of wold examinations performed from one side of a weld, had been previously questioned by the NRC. In a letter from the NRC dated April 23, 1997, to Carl R. Osman (Chairman of the industry Performance Demonstration initiative (PDl) Program) an Open item 95-0109 documented NRC concems for weak performance criterion for far side PD examination * The vessel was examined from the inside diameter surface and in this submittal the licensee stated *lt was determined that supplemental manual examinations from the outside surface were not practical due to the bioshield wall, insulation, and dose considerations." However, the licensee documerited in PlF 971821, that exterior vnsselinsulatiors had been removed during the examination to locate the vertical vessel shell welds. The licensee had not taken advantage of this opportunity to perform supplementary external shell weld examinations and this opportunity had not been identified in the licensee's submitta * An additional reactor vessel circumferential wold was identified in this submittal, which had not boon included in the orig!nal licensee plan for the augmented vessel inspection submitted on October 10,199 The inspectors discussed these issues with the cognizant Office of Nuclear Reactor Regulation (NRR) technical staff involved in review of this submittal. The inspector also discussed these issues wah the licensee staff who stated that they intended to resolve these issues with NRR. The licensee subsequently issued PlF Q1W7-03614 to track and resolve these issues. The inspector considered that these issues demonstrated a less than iigorous licensee staff effort for the augmented vessel examination relief submitta The inspector reviewed data and flaw inu . on evaluations recorded for a vertical and circumferential shell weld in the Unit 2 augmented reactor vessel examination and

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determined that all recorded indications were within ASME Code limit __

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. C2Drjusions The inspectors determined that ASME Code requirements had been met for select eraminations completed in the previous refucting outages for each unit (01R14 and Q2R14). However, 'he inspector identified a O2R12 inspection of a reactor vessel nozzle weld with scan limitations for which ASME Code Section XI volumetric inspection requirements appeared not to have been met. This examination had not been '

documented as limite M7 Quality Assurance in Maintenance Activities M7.1 ISI Program implementatiOD Insoection Sepoe (73051)

The licensee's ISI progrc ; plan the the third ISI interval nad been previously reviewed and accepted by the NRC and found to meet ASME Code Section XI Code requirements, except where relief had been granted by the NRC. Therefore, the inspector reviewed the following to assess the effectiveness of the licensee's ISI program implementatio * Licensee audit reports completed on ISI program elements in the past three years; QAA 04 95-10, OAA 04 96-04, OAA 04 97 03, QAA 04 97-0 *

Problem identification reports (PIFs) generated by the licensee from January of

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1996 through September 1997, documenting problems with ISI program related

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activitie Qbe.rvations and Findings in OAA 04-9510, issued August 31,1995, the licensee identified several deficiencies in the ISI program and described ten recommendations for program improvement. This report stated,"The current ISI program does not satisfy ASME Code requirement While the required examinations are completed, they are not properly evalua~ed, '

or reported." The inspector completed a reviewed of selected (15) examinations documented in the inservice examination 90-day summary reports (discussed in section M3.1) to independently ast ess this conclusion. The inspectors review of subsequent ISI audit reports indicated that the ISI program had improved and was acceptabl However, the inspectors review of PlFs as discussed below indicated that problems continued to persist in the implementation of the licensee's ISI progra The inspector reviewed PlFs issued since January 1996, associated with the ISI program. The inspector considered the following PIFs to address potentially significant /

issues and reviewed in detail the root causes and licensee corrective actions:

PIF 96-1072: " Eleven welds had NDE surface inspections performed during the ISI scope in U-1 drywell that were not required by the ISI group."

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e PIF 96-1463: ' Contractor performed scan No. 7 (thickness across weld and

- base metal) bn U.T. data sheet NDT DS-CS-2A without performing calibration per procedure."

e PlF 96 2585: "During final review of the Reactor Vessel Level Instrumentation Piping packages, it was determined that a NIS 2 form was not completed."

e PlFs 961838,961938, and 96-2727: Control rod drive bolting and local power range tuonitor flange bolting work had been performed without considering .

ASME Cooe requirements and documentatio !

e PlFs96-847,and 961817: Discrepancies were documented associated with .

qualification and certification documentation for personnel performing ISI >

inspections, o PlF Q1997-03555: " Assumptions used in crack growth rate calculations not being monitored."

e PlF Q1997 03576: "lSI Group unable to verify accuracy of Q1R14 NDE examination records."

The inspector determined that these issues appeared to have been adequately resolved or dispositioned. However, the number and variety of issues indicated that opportunities existed for further improvement in the implementation of the ISI program. The licensee had previously issued nuclear tracking system (NTS) item 254 2519710101 to track an

) action item to revise the ISI program. The ISI lead program engineer stated that the ISI b program would undergo a significant revision, which would be completed and submitted to the NRC in early 1998. The inspector considered this a good inillativ The inspector identified that licensee :esolution of PIF 96-2039, "The ISI Program may not include all required systems," appeared to be inadequate. In response to PlF 96-2039, the licensee determined the containment atmospheric monitoring syctem (CAM)

and the atmospheric containment atmosphere dilution system (ACAD) were not required to be in the ISI program. This determination was based, inpart, on NUREG-800,

" Standard Review Plan for the Review of Safety Analysis Reports for Nuclear Power Plants" page 3.2.2-4 statement: "In accordance with Regulatory Guide 1.26, if upon postulated failure of the system, off site dose does not exceed 0.5 rem then this system is not ISI classified." The licensee was unable to demonstrate that the failure of the ,

ACAD or CAM systems would not result in an off site dose greater than 0.5 rem. The licensee subsequently initiated PlF Q1997-03607 to address this issue. This is an unresolved item pending completion of a licensee analysis demonstrating that the failure of the ACAD or CAM systems would not result in exceeding the 0.5 rem off site dose (URl 50 254(265)/97018-02), Conclusions in general, the licensee's ISI program implementation met ASME Code Section XI requirements; the inspector considered the licensee's p:anned initiative to re' vise and

- improve the ISI program to be good, However, the inspector was concemed that the

ACAD and CAM syctems had been excluded from the ISI program scope without

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adequate technical justification. Additionally, the number and variety of ISI issues identified in PlFs, iridicated that additional oversight is warranted to assure that program weaknesses are correcte M8 Miscellaneous Maintenance issues (92902)

M8.1 [Clostad) Violation 50-25S265)/95004 01 The licensee failed to complete qualification review sheets for a contracted ISI worker as required by procedures. The licensee completed the required certification reviews for the NDE worker involved. The inspector reviewed the licensee's closure activities and found them to be adequate. This item is close M8.2 (CJQ$2dllRSRCG10LEoJIow-uo item 50 254(265)/96004-04: The inspector previously identified a concern over the pctentially ineffective induction heat stress improvement (IHSI) process which was utilized on recirculation system welds. This concern was prompted by the detection of substantial cracking in eight recirculation line pipe welds during the 1996 Unit i refueling outage, which called into question the effectiveness of the IHSI process used to mitigate weld crack initiation and growth. The licensee's evaluation (NTS #254100 96-00404) of the IHSI piocess effectiveness stated: " Based on past Comed and industry experience, it is apparent that the resultant lHSI residual ,

stress field was not consistently achieved." The licensee had submitted this evaluation to the NRC in *Ouad Cities Nuclear Power Station Unit 1 ISI Post-Outage (90 Day)

Summary Report Refueling Outage Number Q1R14," dated November 22,1996. The inspector considered this action inadequate for the following reasons; p * The submittal of the licensee investigation of the IHSI process in the 90-day ISI di summary report had not effectively communicated the licensco's findings to the cognizant NRC technical staf * The NRC's letter, dated May 10,1996, * Flaw evaluation of Recirculation System Piping Welds, Quad Cities Nuclear Power Station, Unit 1," stated: "However, to ensure safe plant operation in the long-term, the staff requests that eacn Category C weld treated with the IHSI process at Quad Cities Unit 1, should be inspected in accordance with IGSCC [intergranular stress corrosion cracking)

Category D schedule (100 percent every two fuel cycles) until such time that the root causes of the observed cracking are identified and the IHSI process applied to each Category C weld is determined to be effective in mitigating IGSCC.* The licensee had not t&en any actions to implement this recommendation or to document the technical basis for not implementing this recommendatio * Inspection report 50 254(265)/96004, dated June 7,1996, requested that the

. licensee submit their evaluation and planned corrective actions for the potentially ineffective IHSI process. The licensee had not taken corrective actions for the potentially ineffective IHSI process on Unit i nor had the licensee submitted planned corrective actions to the NRC as requeste * The same IHSI process, which the licensee had evaluated and determined to be less than fully effective in Unit 1 had bemi used on the Unit 2 recirculation system welds. The licensee had not completed a documented evaluation of the IHSI process effectiveness for the Unit 2 Category C recirculation pipe weld _ _ . _ . _ _ _ __..__ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ ._

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However, the extensive cracking identifie<l in Unit 1 had prompted the license 9 to ,

perform exarninations of seven additional recirculation system welds (above the !

outage examinetion scope requirements)in Q2R14. The licensee had not found evidence of additional cracking in the Unit 2 recirculation welds selected for Q2R1 The licensee's failure to take corrective actions for the ineffective IHSl process for the Unit i recirculation pipe welds is considered a violation of 10 CFR 50, Appendix B,  ;

Criterion XVI, * Corrective Actions,"(VIO 50-254(265)/97018-03(DRS)).

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At the conclusion of this inspection the licensee staff stated that they would address the Issues discussed above in a submittal to the NRC by November 21,1997. Additionally, the licensee initiated PIF Q1997-03606 to document and track resolution of this issu ,

M8.3 (Closed) Insoector Follow-uo item 50-254(265)/96004-011; The combined effect of the core spray system and core shroud cracking could divert some flow from the core in a i loss of coolant accident and could impact the peak clad temperature (PCT) recorded in the Updated Final Safety Analysis Report (UFSAR). The changes to the PCT for this cracking were appropriately evaluated and documented in a licensee letter dated April 24,1996, " Quad Cities Nuclear Power Station Units 1 and 2 Plant Specific ECCS (Emergency Core Cooling System] Evaluation Changes 10CFR50.46 Report DPR 29 and DPR-30." This item is close lll. Enghteering

E2 Engineering Support of Facilities and Equipment E UFSAR Review a, jnsoection Scoce(37700)

Inspectors reviewed UFSAR Section 3.9.1.1.1," Reactor Pressure Vessel Fatigue Evaluation" and the supporting documentation identified below:

• SASR 89 02 " Vessel Fatigue Evaluation Considering Revised Thermal Cycles

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For Quad Cities Nuclear Station Units 1 and 2," Revision ,

• Letter from R. Stols (Commonwealth Edison Company (CECO)) to T.E. Murley (NRC), dated October 2,1990, Submitting GE Report SASR 89 02 " Vessel Fatigue Evaluation Considering Revised Thermal Cycles For Quad Cities Nuclear Station Units 1 and 2," Revision * Letter from L. Olshan (NRC) to T. Kovach (CECO), dated February 13,1991,

• lssuance of Amendment (TAC Nos. 75374 and 75375)." Observations and Findinas in the October 2,1990 letter, the licensee stated: "The report concludes that'the cumulative fatigue usage for the 40 year design life, taking into account the revised duty j cycles, is less than the allowable value of 1.0 for all components except the closure

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studs. The closure studs are predicted to reach a fatigue usage of 1.0 in 1998 which will  !

then require a more' rigorous visual examination or replacement." At the conclusion of i the inspection period, the licensee could not demonstrate that an NRC approved "more i rigorous visual examination," had been performed in lieu of replacement of the reactor vcssel head closure studs. Additionally, the licensee had not scheduled a Unit 1 or 2 shutdown to replace the head studs prior to 1998, when the head studs would reportedly

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exceed the ASME Code allowable fatigue usage factor of 1.0. This is considered an unresolved item pending replacement of reactor vessel head studs or NRC approval of a licensee submittal to operate with reactor vessel head studs beyond the Code allowed fstigue usage factor of 1.0 (URI 50-254(265)/97018-04(DRS)).

. Conclusions The licensee had not taken any action to resolve the issue that the Code allowed fatigue usage factor would be exceeded in 1998 for the reactor vessel head steds,

V. Management Meeti.ngs X1 Exit Meeting Summary

The inspectors presented the inspection results to members of licensee management at i the conclusion of the inspection on September 25,1997, The licensee acknowledged +

the findings presented and did not identify any of the potential report input discussed as proprietar :

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PARTIAL LIST OF PERSONS CONTACTED Commonwealth Ejjsgn ,

L. W. Pierce, Station Vice President D. Cook, Opera 9ons Manager

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- G. Perkins, ISI Group Lead ,

T. Wojcik, ISI S. Darin, Deputy Engineering Manager L

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C. Peterson, Megulatory Affairs -

11S NRC:

K. Walton, Resident inspector L. Collinc,- Resident inspector .

Illinois Deoartment of Nuclear Safety; j B. Gant.er -  !

INSPECTION PROCEDURES USED IP 73753 INSERVICE INSPECTION IP 72051 INSERVICE INSPECTION REVIEW OF PROGRAM IP 73755 INSERVICE INSPECTION REVIEW OF DATA IP ?.7700 ' DESIGN CHANGES AND MODIFICATIONS ,

IP 92902 FOLLOWUP-MAINTENANCE

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ITEMS OPENED, CLOSED or DISCUSSED QRitnftd -

URI 50-254(265)/97018 01 ASME Code requirements for nozzle N8B t.yamination may not have been met (Section M3).

URI 50-254(265)/97018-02 ACAD and CAM systems excluded from the ISI program, without supporting analysis (section M7).

VIO 50-254(265)/97018-03 Corrective actions were not taken for the ineffective IHSI treatment on Unit 1 recirculation system welds (Section M8).

URI 50-254(265)/97018-04 Operation of the reactor closure studs with a cumulative usage factor above 1.0 (Section E2).

C10Hd VIO 50-254(265)/95004-04 Failure to complete qualification review sheets for contracted ISI worker (Section M8).

IFl 50 254(265)/96004-04 Ineffective IHSI process on Unit 1 Recirc welds (Section M8).

IFl 50-254(265)/96004-05 Effect of the core spray and core shroud cracking on peak clad temperature in a LOCA (Section M8).

Dhcuned None

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