NG-98-1365, Forwards Response to Violations Noted in Insp Rept 50-331/98-08.Corrective Actions:Matl Removed from Drywell & Addl Closeout Insp Was Completed Satisfactorily on 980514

Forwards Response to Violations Noted in Insp Rept 50-331/98-08.Corrective Actions:Matl Removed from Drywell & Addl Closeout Insp Was Completed Satisfactorily on 980514

ML20238F687
Person / Time
Site:	Duane Arnold
Issue date:	08/27/1998
From:	Franz J IES UTILITIES INC., (FORMERLY IOWA ELECTRIC LIGHT
To:	NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM)
References
50-331-98-08, 50-331-98-8, NG-98-1365, NUDOCS 9809040198
Download: ML20238F687 (7)

Text

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ALLI ANT UTILITIES IES Unhues Inc.

IES Utilities 87 E)$EEIIY""" "'"

a I'alo,IA 52324 97M5 Ofhce: 319.851/611 August 27, J 998 rax: ainstan6 NG-98-1365 ""'"""87'""'

U. S. Nuclear Regulatory Commission Attn: Document Control Desk Mail Station 0-Pl-17 Washington, D.C. 20555-0001

Subject:

Duane Arnold Energy Center Docket No: 50-331 Op. License No: DPR-49 Reply to a Notice of Violation Transmitted with Inspection Report 98008 File: A-105, A-102

Dear Sir:

This letter and attachment are provided in response to the Notice of Violation transmitted with NRC Inspection Report 98008.

There are no new commitments made in this letter.

If you have any questions regarding this matter, please contact my office.

Sincerely, J . Franz Vice President, Nuclear

Attachment:

Reply to a Notice of Violation Transmitted with Inspection Report 98008 cc: J. Karrick E. Protsch 1 D. Wilson k I R. Laufer (NRC-NRR)

J. Caldwell(Region III)

NRC Resident Office DOCU

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9809040198 980827 PDR ADOCK 05000331 0 PDR; i

, Attachment to NG-98-1365 Page 1 of 6

. . Alliant-IES Utilities Reply to a Notice of Violation Transmitted with Inspection Report 98008 VIOLATION 1:

10 CFR 50, Appendix B, Criterion V, " Instructions, Procedures, and Drawings," states, in part, that " Activities affecting quality shall be prescribed by documented instructions, s procedures, or drawings, of a type appropriate to the circumstances and shall be accomplished in accordance with these instructions, procedures, or drawings."

Integrated Plant Operating Instruction (IPOI) 7, Revision 51, "Special Operations,"

Attachment 2, " Primary Containment Closeout," Step 4(e), requires the licensee to verify that the drywell general area is free of trash, tools, and loose articles.

Contrary to the above, on May 13,1998, the licensee conducted a primary containment closecut inspection, an activity affecting quality, and failed to verify that the drywell general area was free of trash and tools. On May 13, after the licensee conducted its primary containment closeout inspection in accordance with IPol 7, the inspectors found trash and a tool in the drywell general area consisting of tape rolled into balls, exposed fibrous insulation on a section of drywell cooling piping, tie-wrap pieces, nails, glass, sheet metal screws, wiring used to fasten insulation covers, and a carpenter's ruler.

This is a Severity Level IV Violation.

RESPONSE TO VIOLATION 1:

1. REASON FOR Tile VIOL,ATION Attachment 2 to IPOI 7, step (4) states to " Verify the following free of trash, tools, loose articles, and obstructions," then lists under vessel platform, drywell to torus downcomers, drywell to torus jet deflectors, downcomer ring header, and drywell general area as the specific areas to inspect and signoff. Altogether,37 signoffitems are included in the attachment to perform the primary containment closeout inspection. Item (4)(e), the drywell general area, is one of the 37 items required to be completed and signed off.

On May 13,1998, a team of senior plant personnel performed the drywell closecut inspection in accordance with IPOl 7. Subsequent to this inspection, which had concluded the drywell general area to be " free" of trash and loose debris, an NRC inspection identified additional loose debris amounting to approximately 1/4 of a trash bag. The small amount of material l

identified posed no threat to equipment performance.

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~ Attachment to 1 NG-98-1365

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Page 2 of 6 The reason for the violation was the overly prescripdve guidance contained in IPOI 7 regarding removal of debris. The intent of the procedural step is to ensure general area cleanliness meets requirements to prevent a threat to safety system operation. The drywell general area is a large area composed of structural beams, conduit, and many components which create difTiculty in meeting an acceptance criterion of" free." All identified items found during the inspection were removed, but some items remained unidentified. The quantity ofitems missed by the inspection did not threaten operation of equipment.

2. CORRECTIVE STEPS TAKEN AND THE RESULTS ACHIEVED The material was removed from the drywell and an additional closeout inspection was completed satisfactorily on May 14,1998.

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IPOl 7, step (4) has been revised to more clearly specify management expectations and l

remove the ambiguity.

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3. CORRECTIVE STEPS THAT WILL BE TAKEN TO AVOID FURTIIER VIOLATIONS Corrective actions to avoid further violations have been completed.

4. DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance was achieved on May 14,1998, when the items were removed from the drywell and the closeout inspection was completed satisfactorily.

VIOLATION 2:

Technical Specification 6.8.1.5 requires that written procedures involving nuclear safety, including applicable check-offlists and instructions, covering preventive and corrective maintenance operations which could have an affect on the nuclear safety of the facility, be implemented and maintained.

Maintenance Directive 24, Revision 12," Post Maintenance Testing Program," Step 5.0(1), I requires that post maintenance testing be performed, if appropriate, following corrective maintenance on components regardless of Quality Level to verify the ability of the system or component to perform its intended function. l Contrary to the above, on December 16,1997, adequate post maintenance testing was not performed following corrective maintenance on the primary containment isolation system (PCIS) Group 1 relay (A71B-K057) for the recirculation sample line outboard isolation valve. Specifically, on April 3,1998, the relay failed to perform its intended function when l

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, Attachment to NG-98-1365 Page 3 of 6 an inadvertent PCIS Group 1 isolation signal was received. The failure was the result of mechanical binding caused by improper retainer replacement during the corrective maintenance and relay reassembly on December 16,1997.

This is a Severity Level IV Violation.

RESPONSE TO VIOLATION 2:

1. REASON FOR THE VIOLATION r

As stated in Licensee Event Report (LER) 331/98-002, the cause of recirculation sample line outboard isolation valve CV4640 failing to close in response to the Group 1 isolation signal was the mechanical binding of Outboard Main Steam Drain &

Reactor Sample Isolation relay A71B K057 which prevented sending the isolation signal to CV4640 and MO4424. The relay failed tc, return to its deenergized state when the coil was deenergized due to mechanical binding of the movable contact 1 carrier. As a result, CV4640 and MO4424 did not receive an isolation signal. The failure mechanism of A71B-K057 was misalignment of the contact retaining clip which caused restriction of the movable contact carrier. The misalignment of the contact retaining clip is believed to have occurred during reassembly of the relay

- following coil replacement on December 16,1997. The relay is a General Electric I (GE) model CR120A.

The coil replacement was perfonned in response to GE Service Information Letter

- (SIL) 229, Supplement 1, which is associated with relay coil aging. A review of Corrective Maintenance Action Request (CMAR) A38252, which provided instructions for the replacement of the coil in relay A71B-K057, shows that post maintenance testing required verification that the relay picked up when energized, but did not require verification that the relay dropped out when deenergized. The isolation fimetion of the relay requires it to drop out to initiate closure of the associated isolation valves and should have been verified in accordance with post maintenance testing practices. Therefore, the reason for the violation was less than adequate attention to detail in that the maintenance planner and operations department reviewer (Maintenance TOSS) of the work failed to specify testing of the drop out i function. The maintenance procedure used to perform the specified functional test was a contributor to the violation in that it also did not include testing of the drop out function.

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, Attachment to NG-98-1365 Page 4 of 6

2. , CORRECTIVE STEPS TAKEN AND TIIE RESULTS ACIllEVED Relay A71Il-K057 was replaced with a new relay. Post maintenance testing was performed to verify that the new relay would return to its deenergized state after being energized.

The maintenance procedure for CR120A relays was revised to ensure adequacy of the functional testing requirements.

Additional relay maintenance procedures (other than CR120A relays) were reviewed for adequacy of functional testing requirements. As a result of this review, a procedure change was initiated to also add a step to the procedure for replacement of

' Agastat relays to ensure adequate functional testing of Agastat control relays. No other procedural enhancements were identified.

Other safety related CR120A relays, in which the coils had been replaced in response to GE SIL 229, Supplement 1, were tested to assure appropriate drop out when de-energized. This activity was completed and no additional failures occurred.

3. CORRECTIVE STEPS TIIAT WILL HE TAKEN TO AVOID FURTIIER VIOLATIONS Corrective actions to avoid further violations have been completed. ,

4. DATE WilEN FULL COMPLIANCE WILL BE ACIIIEVED I

Full compliance was achieved on April 18,1998, after completion of the corrective I maintenance to the relay (CMAR A47786) and appropriate post maintenance testing was I completed.

VIOLATION 3:

10 CFR 50.73(a)(2)(ii)(B) requires, in part, that the licensee report any event or condition that resulted in the nuclear power plant being in a condition that was outside the design basis of the plant within 30 days after discovery of the event or condition.

The Updated Final Safety Analysis Report, Section 3.6.1.2.4.5, in describing the compartment pressure analysis model stated,"The analyses conducted yielded pressure versus time relationships that were used to determine if the structures surrounding the affected compartments or vital equipment within the compartment were capable of withstanding the pressurization. In all cases the analyses indicated that sufficient vent area was available to prevent over pressurization."

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e Attachment to NG-98-1365 Page 5 of 6 Contrary to the above, on May 5,1998, the licensee discovered that the high pressure coolant injection system compartment wall, a structure surrounding the affected compartment, was  !

not capable of withstanding the pressurization due to a high energy line break in the high pressure coolant injection room. This condition was outside the design basis as described in the Updated Final Safety Analysis Report. As ofJune 5, a licensee event report had not been submitted. l This is a Severity Level IV Violation.

RESPONSE TO VIOLATION 3:

1. REASON FOR TIIE VIOLATION Within 2 days after the identification of the need to add support to the liigh Pressure Coolant injection (HPCI) corridor wall, and after review of the Updated Final Safety Analysis Report (UFSAR) and other applicable design basis documents, DAEC established the position that if the additional engineering analyses being performed showed any impact on the ability to achieve and maintain safe shutdown, was found to impact operability of safety related equipment, or that the Reactor Building (RB) 757' I level of the reactor building could have become Environmental Qualification (EQ) harsh, this condition would have been reported as operation in a condition that was outside the design basis of the plant pursuant to 10 CFR 50.73 (a)(2)(ii). As stated in LER 331/98-007, (submitted on July 8,1998), none of these thresholds were attained as a result of the '

analyses.

'Ihe potential yielding of the corridor wall under new analytical models was considered by the DAEC to be outside an individual component's design basis.110 wever, the reporting requirement is specific to outside the design basis "of the plant," which, for this condition, was interpreted by DAEC using the above thresholds. The definition of

" design bases" per 10 CFR 50.2 uses the terms " functions" or " functional goals." When I reviewing this event from the perspective of the design basis of the plant (10 CFR 50.73 (a)(2)(ii)(B) states "In a condition that was outside the design basis of the plant") and considering UFSAR section 3.6.1.2," Description," the ability of the plant to be shutdown and maintained in a safe shutdown condition was considered the applicable

" functional goal" of the plant.

'UFSAR Section 3.6.1.2.4 and its statements regarding analysis modeling are not described under the heading of" design bases." The DAEC UFSAR, section 3.6.1.1,

" Design Bases," states, "Not required for FSAR," which is consistent with Regulatory Guide 1.70," Standard Format and Content of Safety Analysis Reports for Nuclear Power Plants," Revision 3, November 1978. Statements from the widely publicized Nine Mile Point blowout panel reporting issue (contested violation and related docketed

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, Attachment to NG-98-1365

. . . Page 6 of 6 correspondence) indicate the location ofinformation in the UFSAR was critical to dec'isions on reporting of events under the outside design basis criterion. The fact that the cited UFSAR statement is not located under the heading of design bases was considered applicable to the position taken.

Another consideration in determining the reporting of this issue was the context of the companion reporting requirements of 10 CFR 50.73(a)(2)(ii). Specifically," seriously degraded principal safety barriers" and "unanalyzed condition that significantly compromised plant safety" and the 1-hour requirement of the associated 10 CFR 50.72 event notification requirement (not made in this case due to shutdown plant conditions) convey a relatively high threshold for reporting. The potential and momentary yielding of the IIPCI wall in the event of a high energy line break that was analyzed to show no affect on any safety function, was not considered to meet the threshold implied by these terms.

The reason for the violation is the industry-wide differences in interpretation of this particular reporting requirement.

The DAEC does not contest this violation.

2. CORRECTIVE STEPS TAKEN AND THE RESULTS ACHIEVED LER 331/98-007, mailed on July 8,1998, reported the HPCI corridor wall issue.

The DAEC participated in the initial Boiling Water Reactor Owner's Group meeting with the NRC to pursue 10 CFR 50.72 and 10 CFR 50.73 rulemaking. .

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3. CORRECTIVE STEPS THAT WILL BE TAKEN TO AVOID FURTHER  !

VIOLATIONS I The DAEC has also volunteered to support the newly formed Nuclear Energy Institute (NEI) task force to provide industry input and support to the rulemaking in the area of event reporting. Insights and benchmarking gained from participation in this effort will be applied to implementation of the current regulations at DAEC if new reporting issues arise.

4. DATE WHEN FULL COMPLIANCE WILL BE ACHIEVED Full compliance was achieved on July 8,1998, when LER 331/98-007 was mailed.

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