IR 05000282/1999014

Insp Repts 50-282/99-14 & 50-306/99-14 on 990920-22. Violations Noted.Major Areas Inspected:Heat Sink Performance

ML20217M136
Person / Time
Site:	Prairie Island
Issue date:	10/19/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20217M116	List: IR 05000282/1999014 ML20217M111
References
50-282-99-14, 50-306-99-14, NUDOCS 9910270141
Download: ML20217M136 (8)

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U.S. NUCLEAR REGULATORY COMMISSION REGION lil Docket Nos: 50-282; 50-306

' License Nos: DPR-42; DPR-60 Report No: 50-282/99014(DRS); 50-306/99014(DRS)

Licensee: Northern States Power Company Facility: Prairie Island Nuclear Generating Plant Units 1 & 2 '

Location: 1717 Wakonade Dr. East Welch, MN 55089 l I

I Dates: September 20 - 22,1999 )

Inspector: James A. Gavula, Reactor Engineer Approved by: John M. Jacobson, Chief, Mechanical Engineering Branch

! Division of Reactor Safety l

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9910270141 991019 PDR ADOCK 05000282 G PDR

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l SUMMARY OF FINDINGS Prairie Island Nuclear Generating Plant, Units 1 & 2 NRC Inspection Report 50-282/99014(DRS); 50-306/99014(DRS)

This report covers the pilot baseline inspection for the biennial review of heat sink performanc The heat sink performance inspection covers an inspectable area under the Initiating Events and Mitigating Systems cornerstones for which there is no performance indicator. Adequate or superior performance is not reported. Inspection findings were evaluated according to their potential significance for safety, using the NRC's Significance Determination Process, and assigned colors of GREEN, WHITE, YELLOW, or RED. GREEN findings are indicative of issues that, while they may not be desirable, represent little effect on safety. WHITE findings ;

indicate issues with some increased importance to safety, which may require additional NRC j inspections. YELLOW findings are more serious issues with an even higher potential to affect j safe performance and would require the NRC to take additional actions. RED findings l represent an unacceptable loss of margin to safety and would result in the NRC taking j significant actions that could include ordering the plant to shut down. Those findings that cannot be evaluated for a direct effect on safety with the Significance Determination Process, f such as those findings that affect the NRC's ability to oversee licensees, are not assigned a j colo j i

Cornerstone: Mitigating Systems

Green: The inspector identified that a degraded heat exchanger tube on one of the emergency diesel generators had not been documented within the licensee's corrective action program. Although the licensee plugged the tube and corrected the specific deficiency, the lack of documentation within the corrective action program limited the !

licensee's ability to identify and trend a condition that had previously affected diesel l generator design functio {

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A non-cited violation of 10 CFR 50.59 was identified during closeout of an unresolved l item from the 1997 System Operational Performance Inspection. It was determined that prior NRC approval should havt aeen sought for the modification requiring manual actions to connect a nitrogen bottle on loss of instrument air. This issue had minimal impact on safety because corrective actions were taken when the issue was originally identified.

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Rooort Details

' REACTOR SAFETY Comerstones: Initiating Events, Mitigating Systems, Barrier Integrity ( 1R07 Nigrtjink Performance Inspection Scope The inspector reviewed the surveillance procedure for the most recent thermal i

performance tests on the component cooling heat exchangers Nos.11 and 12, the

!- results of the periodic inspection and cleaning for the emergency diesel generator No. D2 heat exchangers, and the residual heat removal heat exchanger l Observations and Findinas j l

The inspector observed that the component cooling thermal performance calculation, i performed u:1 der Work Order 9815409 for heat exchanger No.12, used incorrect and nonconservative information. The calculation used 1491 active tubes instead of 1490, because the assumed number of plugged tubes was incorrect. This error had a very small impact on the calculated heat removal capability considering the existing fouling factors and did not challenge the approximate 25 percent tube plugging margi However, the licensee considered the calculation to be part of the surveillance procedure and not a separate calculation, and therefore it was not checked and verified similar to other calculations. The licensee wrote General Action item No.19992760 to address this erro The inspector observed that, for the residual heat removal heat exchangers, the

~ licensee did not perform any testing, inspection, or maintenance to ensure proper heat transfer. The resolution of an industry-identified problem for baffle plate bypass leakage for these heat exchangsrs stated that intemal inspection should not be done unless heat ,

transfer degradation had been noted. The inspector questioned how heat transfer )

degradation could be noted if no specific activities were being performed in that regar i OTHER ACTMTIES 40A1 Identification and Resolution of Problems

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! Insoection Scooe i The inspector reviewed several condition reports in the iicensee's corrective action program that related to heat exchangers within the scope of the heat sink performance inspection to verify that the licensee adequately identified and resolved problems, in addition, the inspector reviewed Work Order 9713239, which contained the results of the last eddy current examination for the emergency diesel generator No. D2 heat l

exchangers.

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.- Qbiervations and Findinas The eddy current examination for emergency diesel generator No. D2 identified air cooler heat exchanger tube No.16-2 as having greater than 90 percent wall loss, and

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recommended immediate plugging. The accompanying report noted that this tube had been identified for the same indication in 1989, with the same recommendation; however, the tube had not been plugged at that time. The inspector observed that the licensee had not documented this situation within their corrective action program, and that a past leak in a similar diesel generator heat exchanger, the lube oil cooler, had resulted in inis diesel generator being unable to perform its design function. Although the licensee plugged the tube and corrected the specific deficiency, the lack of documentation within the corrective action program limited the licensee's ability to identify and trend a condition that had previously affected the diesel generator design function. Because the current heat exchanger tube degradation did not affect the diesel generator design function, this finding was determined to be of very low risk significance and was categorized as " Green." The licensee subsequently entered this issue into their corrective action program as Non-Conformance Report No.1999290 A4 Other

.1 LQiosed) Vioiation 50-282/95014-01(DRS): 50-306/95014-01(DRS): Violation of Design Control involving Waterhammer Analysis For Containment Fan Coolers. This issue was subsequently covered by Generic Letter 96-06, " Assurance of Equipment Operability and Containment inMgrity During Design-Basis Accident Conditions," which is being resolved by the Ofnce of Nuclear Reactor Regulation through TAC Nos. M96854 and M96855. This item is close .2 (Closed) Unresolved item 50-282/96008-10 (DRS): Operability Questions Regarding the Containment Fan Coolers. This issue was subsequently covered by Generic Letter 96-06, " Assurance of Equipment Operability and Containment integrity During Design-Basis Accident Conditions," which is being resolved by the Office of Nuclear Reactor Reguiation through TAC Nos. M96854 and M96855. . This item is closed.

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.3 (QigyJ) Violation 50-282/97290-01013(DRS): 50-306/97290-01013 (DRS): Violation of Test Control involving Auxiliary Feedwater Acceptance Criteria. The licensee reviewed the inservice testing procedures for the safety-related pumps and verified that the acceptance criteria considered both the ASME Section XI criteria and the Updated Final Safety Analysis Report design requirements and specified the more limiting value. The procedures for the auxiliary feedwater pumps were revised appropriately. This item is close .4 (Open) Violation 50-282/97290-01023(DRS): 50-306/97290-01023 (DRS): Violation of 50.11(e) involving Failure to Update the Updated Final Safety Analysis Report Auxiliary Feedwater Accident Flow Rates. In their resporse to the Notice of Violation, dated November 14,1997, the liusnsee stated that plans were to complete the remaining 14 items before the Updated Safety Analysis Report submittal of late 1998. This commitment was not met, and there are currently six items, where the activities necessary to allow the Updated Safety Analysis Report to be revised, have not been completed. Three of these items involve revisions to the high energy line break

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analysis, and the others require the submittal of a license amendment request for boron dilution. This is being tracked by licensee commitment tracking number 1998014 .5 (Closed) Violation 50-282/97290-01033(DRS): 50-306/97290-01033 (DRS): Violation of Corrective Action involving Failure to Review Auxiliary Feedwater Acceptance Criteri The licensee took a number of corrective actions to improve its correcta action program, including establishment of a new corrective action system. All corrective

- actions associated with this violation are completed. The licensee's corrective action program was evaluated in August - September 1999, under the pilot inspection process for Problem identification and Resolution. This inspection concluded that the licensee's corrective action process was acceptable. Therefore, this violation is close .6 (Closed) Insoector Followuo item 50-282/97008-01(DRS): 50-306/97008-01(DRS):

Review of Auxiliary Feedwater Flow Model. This licensee's flow model of the auxiliary feedwater system has been benchmarked and no adjustments to the results were mad This item is close .7 (Closed) Unresolved item 50-282/97008-09(DRS): 50-306/97008-09(DRS):

Determination of Acceptability of Using Manual Action to Connect Nitrogen Bottle on Loss of instrument Air. This item was forwarded to the Office of Nuclear Reactor Regulation where it was determined that the operator actions were sufficiently complex that errors of either commission or omission needed to be considered. Therefore, they determined that NRC approval should have been sought before this modification was implemented. The licensee took immediate corrective actions to remove the air bottl The failure to recognize that prior NRC approval was required resulted in a violation of

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- 10 CFR 50.59. This Severity Level IV violation is being treated as a Non-Cited Violation (NCV 50-282/306-99014-01) consistent with Appendix C of the NRC Enforcement Polic .8 (Closed) Unresolved item 50-282/97008-10(DRS): 50-306/97008-10(DRS):

Determination of Acceptability of Instrumentation Setpoint Uncertainties and of Administrative Control of Setpoints. This issue is under review by the Office of Nuclear Reactor Regulation as a generic concem, and when the review has been completed, the licensee will be informed of any necessary actions by separate correspondenc Therefore, this item is close .9 (Closed) Violation 50-282/97008-13(DRS): 50-306/97006-13(DRS): Design Control Violation involving Untimely Corrective Action on Cable Tray Separation Issue. The technical issues involved with the violation were reviewed and closed in inspection report 50-282/306-98005. The remaining corrective actions dealt with the corrective action program. As discussed above (Violation 282/306-97290-01033), the licensee took a number of steps to improve its corrective action program, including establishment of a new corrective action system. All corrective actions associated with this violation are completed. -The !;censee's corrective action program was evaluated in August -

September 1999 under the pilot inspection process for Problem identification and Resolution. This inspection concluded that the licensee's corrective action process was acceptable. - Therefore, this violation is close . .* -

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l 40A5 Manaaement Meetinos

-.1 Exit Meetina Summarv i

The inspector presented the inspection results to Mr. T. Amundson and other members of licensee management in an exit meeting on September 22,1999. The licensee acknowledged the information and findings presented. No proprietary information was identifie i

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! l PARTIAL LIST OF PERSONS CONTACTED Licensee ,

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T. Amundson, General Superintendent Engineering l T. Breene, Supe.Intendent, Nuclear Engineering i T. Downing, System Engineer

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P. Hajovy, System Engineer -

M. Heller, Supe,intendent, Mechanical Systems / Programs S. Hiedeman, Superintendent, Mechanical Systems l T. Silverberg, General Superintendent, Operations, Acting Plant Manager NRC M. Kunowski, Project Engineer  !

l INSPECTION PROCEDURE USED IP 71111.07 (draft): Heat Sink Performance ITEMS OPENED, CLOSED, AND DISCUSSED Opened 282/306/99014-01 NCV Violation of 10 CFR 50.59 for Manual Actions During Loss of Instrument Ai Closed 282/306/99014-01 NCV Violation of 10 CFR 50.59 for Manual Actions During Loss l of Instrument Ai /306/95014-01 VIO Design Control for Containment Fan Cooler Waterhammer Analysi /96008-10 URI Operability of Containment Fan Cooler /306/97290-01013 VIO Violation of Test Control for Auxiliary Feedwater Acceptance Criteri /306/97290-01033 VIO Violation of Corrective Action involving Failure to Review Auxiliary Feedwater Acceptance Criteri /306/97008-01 IFl Review of Auxiliary Feedwater Flow Mode /306/97008-09 URI Determination of Acceptability of Using Manual Action to Connect Nitrogen Bottle on Loss of instrument Ai /306/97008-10 URI Determination of Acceptability of Instrumentation Setpoint Uncertainties and of Administrative Control of Setpoint /306/97008-13 VIO Design Control Violation involving Untimely Corrective Action on Cable Tray Separation issu Discussed 282/306/97290-01023 VIO Violation of 10 CFR 50.71(e) Involving Failure to Update the Final Safety Analysis Report Auxiliary Feedwater Accident Flow Rate .

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LIST OF ACRONYMS USED CFR Code of Federal Regulations DRS Division of Reactor Safety IFl inspection Followup Item NCV Non-Cited Violation NRC Nuclear Regulatory Commission URI Unresolved item VIO Violation LIST OF DOCUMENTS REVIEWED Miscellaneous Work Request 9815409, SP1304 U1 CC Hx Performance Tes Work Request 9809590, SP 2304 U2 CC Hx Performance Tes Work Request 9713239 D2 DG 18 Month insp,4/27/98, PINGP 1066 For Work Request 9604984, D2 DG 18 Month insp,9/22/96, No PINGP 1066 Form Availabl Work Request 9407735, D2 DG 18 Month insp,314/95, PINGP 1066 For Operating Experience Assessments: Westinghouse IG94002,"CCW HX Fouling and Containment Response."

Generic Letter 89-13 Implementing Program, Section NSP Materials and Special Processes Report,5/1/98, Eddy Current Examination of D2 Air, Oil, and Jacket Cooler Heat Exchanger Condition Report Forms Condition Report 19970880, Potentially Undersized EDG Oil Cooler Condition Report 19980144, IR 97008 Action 1 Condition Report 19981123, Unit 2 AFWP Surveillance Procedure Condition Report 19983244,22CC Hx Divider Plate Bendin Procedures Surveillance Procedure 1304 U1 Component Cooling Heat Exchanger Performance Tes PM 3001-2-D2, Diesel Generator 18 Month inspection, Rev.1 !

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