IR 05000282/2006006

IR 05000282-06-006 (Drs); 05000306-06-006 (Drs); 03/06/2006 - 03/24/2006; Prairie Island Nuclear Generating Plant, Units 1 and 2; Evaluation of Changes, Tests, or Experiments (10 CFR 50.59) and Permanent Plant Modifications

ML061220751
Person / Time
Site:	Prairie Island
Issue date:	04/28/2006
From:	Dave Hills NRC/RGN-III/DRS/EB1
To:	Thomas J. Palmisano Nuclear Management Co
References
IR-06-006
Download: ML061220751 (19)

Text

April 28, 2006SUBJECT:PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 NRC EVALUATION OF CHANGES, TESTS, OR EXPERIMENTS ANDPERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000282/2006006 (DRS); 05000306/2006006 (DRS)

Dear Mr. Palmisano:

On March 24, 2006, the U.S. Nuclear Regulatory Commission (NRC) completed a combinedbaseline inspection of the Evaluation of Changes, Tests, or Experiments and Permanent Plant Modifications at the Prairie Island Nuclear Generating Plant. The enclosed report documents the results of the inspection, which were discussed and others of your staff at thecompletion of the inspection on March 24, 2006.The inspectors examined activities conducted under your license as they relate to safety andcompliance with the Commission's rules and regulations and with the conditions of your license.

The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.Based on the results of the inspection, one NRC-identified finding of very low safetysignificance was identified which involved a violation of NRC requirements. However, becausethis violation was of very low safety significance, not willful, and because it was entered intoyour corrective action program, the NRC is treating the issue as a Non-Cited Violation in accordance with Section VI.A.1 of the NRC's Enforcement Policy.If you contest the subject or severity of a Non-Cited Violation, you should provide a responsewithin 30 days of the date of this inspection report, with the basis for your denial, to the U.S.

Nuclear Regulatory Commission, ATTN: Document Control Desk, Washington, DC 20555-0001, with a copy to the Regional Administrator, U.S. Nuclear Regulatory Commission -

Region III, 2443 Warrenville Road, Suite 210, Lisle, IL 60532-4352; the Director, Office ofEnforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector Office at the Prairie Island Nuclear Generating Plant facility.

In accordance with 10 CFR 2.390 of the NRC's "Rules of Practice," a copy of this letterand its enclosure will be available electronically for public inspection in the NRC PublicDocument Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).

Sincerely,

/RA/David E. Hills, ChiefEngineering Branch 1 Division of Reactor SafetyDocket Nos. 50-282; 50-306License Nos.

Enclosure:

Inspection Report 05000282/2006006 (DRS); 05000306/2006006 (DRS)

Document Room or from the Publicly Available Records (PARS) component of NRC's document system (ADAMS). ADAMS is a

REGION IIIDocket No:50-282; 50-306License No:Report No:05000282/2006006 (DRS); 05000306/2006006 (DRS)Licensee:Facility:Prairie Island Nuclear Generating PlantLocation:Dates:March 6 through March 24, 2006 Inspectors:J. Neurauter, Senior Reactor Inspector, Team LeaderAlan Dahbur, Reactor InspectorApproved by:D. Hills, ChiefEngineering Branch 1 Division of Reactor Safety (DRS)

SUMMARY OF FINDINGS

IR 05000282/2006006 (DRS); 05000306/2006006 (DRS); 03/06/2006 - 03/24/2006; PrairieIsland Nuclear Generating Plant, Units 1 and 2; Evaluation of Changes, Tests, or Experiments (10 CFR 50.59) and Permanent Plant Modifications.The inspection covered a two-week announced baseline inspection on evaluations of changes,tests, or experiments and permanent plant modifications. The inspection was conducted by two regional based engineering inspectors. One Green Non-Cited Violation (NCV) was identified.

The significance of most findings is indicated by their color (Green, White, Yellow, Red), using Inspection Manual Chapter 0609, "Significance Determination Process (SDP.)" Findings for which the SDP does not apply, may be Green, or be assigned a severity level after NRC management review. The NRC's program for overseeing the safe operation of commercialnuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Revision 3,dated July 2000.A.Inspector-Identified and Self-Revealed Findings

Cornerstone: Mitigating Systems

Green.

A Non-Cited violation of 10 CFR Part 50, Appendix B, Criterion III, "DesignControl," having very low safety significance was identified by the inspectors.

Specifically, the licensee had not evaluated and updated the associated plant cableampacity calculation to determine the potential consequences of adverse effects to cabling due to higher temperatures in the auxiliary feedwater (AFW) pump rooms andother auxiliary building areas. After identification by the inspectors, the licensee wasable to demonstrate that even though the higher temperatures decreased the ampacitymargins for the affected cabling, it did not decrease the margins to the limit where the cabling would fail if called upon to provide power to equipment important to safety.The finding was more than minor because it affected the mitigating system cornerstoneobjective to ensure the availability, reliability, and capability of systems that mitigatetransients and accidents, and if left uncorrected, the finding could become a more significant safety concern. Specifically, if left uncorrected, the licensee may not account for high temperature conditions in plant areas that could adversely affect the ampacity of cabling that supply power to equipment important to safety. This finding was of very low safety significance because, the licensee's preliminary evaluation determined that thehigher temperatures in the AFW pump rooms and other auxiliary building areas wouldnot prevent equipment important to safety from functioning. (Section 1R17.1.b.1)

Cornerstone: Barrier Integrity

No findings of significance were identified.

B.Licensee-Identified Violations

None.

REPORT DETAILS

1.REACTOR SAFETYCornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity1R02Evaluations of Changes, Tests, or Experiments (71111.02).1Review of 10 CFR 50.59 Evaluations and Screenings

a. Inspection Scope

From March 6 through March 24, 2006, the inspectors reviewed eight evaluationsperformed pursuant to 10 CFR 50.59. The inspectors confirmed that the evaluationswere thorough and that prior NRC approval was obtained as appropriate. Theinspectors also reviewed seventeen screenings where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. In regard to the changes reviewed where no 10 CFR 50.59 evaluation was performed, the inspectors verified that the changes did not meet the threshold to require a 10 CFR 50.59evaluation. The evaluations and screenings were chosen based on risk significance, safety significance, and complexity. The list of documents reviewed by the inspectors is included as an attachment to this report.The inspectors used, in part, Nuclear Energy Institute (NEI) 96-07, "Guidelines for 10 CFR 50.59 Implementation," Revision 1, to determine acceptability of the completed evaluations and screenings. The NEI document was endorsed by the NRC inRegulatory Guide 1.187, "Guidance for Implementation of 10 CFR 50.59, Changes,Tests, and Experiments," dated November 2000. The inspectors also consulted Part 9900 of the NRC Inspection Manual, "10 CFR Guidance for 10 CFR 50.59,Changes, Tests, and Experiments."

b. Findings

No findings of significance were identified.

1R17 Permanent Plant Modifications

.1Review of Permanent Plant Modifications

a. Inspection Scope

From March 6 through March 24, 2006, the inspectors reviewed twelve permanent plantmodifications that had been installed in the plant during the last two years. The modifications were chosen based upon risk significance, safety significance, and complexity. The inspectors reviewed the modifications to verify that the completed design changes were in accordance with the specified design requirements and the licensing bases and to confirm that the changes did not adversely affect any systems'safety function. Design and post-modification testing aspects were verified to ensure the functionality of the modification, its associated system, and any support systems. The inspectors also verified that the modifications performed did not place the plant in an increased risk configuration.

The inspectors also used applicable industry standards to evaluate acceptability of themodifications. The list of modifications and other documents reviewed by the inspectors is included as an attachment to this report.The Prairie Island Unit 1 reactor vessel head replacement modification, which affectsthe barrier integrity cornerstone, was not selected as part of this inspection. This modification will be inspected at a later date in accordance with inspection procedure71007, "Reactor Vessel Head Replacement Inspection."

b. Findings

b.1Failure to Consider Adverse Ampacity Effects of High Ambient Temperature Conditionsin the Auxiliary Feedwater Pump RoomsIntroduction: On March 15, 2006, the inspectors identified a Non-Cited Violation of10 CFR Part 50, Appendix B Criterion III, "Design Control," of very low safetysignificance (Green). Specifically, the licensee had not evaluated and updated the associated plant cable ampacity calculation to determine the potential consequences ofadverse effects to cabling due to higher temperatures in the AFW pump rooms and other auxiliary building areasDiscussion: Revised licensee calculation ENG-ME-021, "Auxiliary Feedwater PumpRoom Heat-up," indicated that the potential maximum ambient temperature in the AFWpump rooms could reach up to 127F. The potential high ambient temperature couldoccur during post accident mitigation (an extended loss of off-site power) and when the initial ambient temperature in the rooms was at 104F. The licensee evaluated the effects of the high ambient temperature on the safety-relatedequipment (i.e., motor driven AFW pump motors, motor operated valves, motor control centers, transformers and hot shutdown panel) located in the AFW pump rooms. The evaluation was documented in calculation ENG-ME-021, Revision 2 and concluded that the operability of the safety-related equipment located in the rooms was acceptable foran ambient room temperature of 127F. This conclusion was also documented in thelicensee's 10 CFR 50.59 screening number 2469, Revision 0. However, the licensee failed to address the effects of these heightened temperatures on the ampacity of electrical cables in the rooms. The inspectors also reviewed the licensee's 10 CFR 50.59 Safety Evaluation Number 1037 "Affect of Revised Unit 1 Main Steam Line Break on Auxiliary Building Environment," which identified that the ambient temperature couldalso reach up to 122F in several areas in the auxiliary building. Prairie Island Engineering Manual for Electrical Cables Design, Fabrication andInstallation Summary was based on an ambient temperature of 104F. Other plantspecific evaluations (i.e. Calculation ENG-EE-019 and Safety Evaluation 369) which have previously evaluated potential cable ampacity issues were also based on an ambient temperature of 104F. The licensee failed to evaluate and update the cableampacity calculation to evaluate the effects of potential high ambient temperatures onthe ampacity of electrical cables located in these rooms. Since higher temperatures adversely affect the ampacity of electrical cables, the higher temperatures in the AFW pump rooms and other plant areas had the potential to adversely affect the functionality and/or operability of equipment important to safety fed by cabling in these rooms. The inspectors were concerned that the possibility existed that some of the equipment fed bycables located in these areas may not function due to possible faulting of the supply cables. As a result of the inspectors' concerns, the licensee issued Action Request CAP 01018612. After performing a preliminary evaluation that assessed cabling in AFW pump rooms and the auxiliary building areas, the licensee determined that there was no evidence thatsafety related structures, systems, and components would not function as required. While the higher temperatures decreased the ampacity margins for the affected cabling, the licensee preliminarily determined that the margins it did not decrease to the limitwhere the cabling would fail if called upon to provide power to equipment important to safety.Analysis: The inspectors determined that this issue was a performance deficiencywarranting a significance evaluation, since the licensee failed to account for high temperature conditions in the AFW pump rooms and other several rooms located in the auxiliary building that adversely affected cables supplying power to equipment importantto safety. The finding was greater than minor in accordance with IMC 0612, "Power ReactorInspection Reports," Appendix B, "Issue Screening," because it affected the mitigating system cornerstone objective to ensure the availability, reliability, and capability ofsystems that mitigate transients and accidents, and if left uncorrected, the finding couldbecome a more significant safety concern. Specifically, if left uncorrected, the licensee may not account for high temperature conditions in plant areas that could adversely affect the ampacity of cabling that supply power to equipment important to safety. The inspectors determined the finding was of very low significance (Green) usingIMC 0609, Appendix A, "Significance Determination of Reactor Inspection Findings for the At-Power Situations," because the inspectors answered "no" to all five questions under the Mitigating Systems Cornerstone column of the Phase 1 worksheet. In particular, the licensee's preliminary evaluation determined that the higher temperaturesin the AFW pump rooms and other auxiliary building areas would not prevent equipmentimportant to safety from functioning.

Enforcement:

10 CFR Part 50, Appendix B, Criterion III, "Design Control" states, in part,that measures shall be established to assure that applicable design basis are correctly translated into specifications, drawings, procedures and instructions. Contrary to the above, the licensee did not have a design basis calculation for cable ampacity that supported the high temperatures that the AFW pump rooms and other plant areas couldexperience. The Prairie Island calculation and engineering manual that did address cable ampacity were significantly less conservative, since temperatures of 104F wereassumed where temperatures in these areas could exceed 122F.Because this issue was of very low safety significance, not willful, and because it wasentered in the licensee's corrective action program as CAP 01018612, this violation is being treated as an NCV, consistent with Section VI.A of the NRC Enforcement Policy.

(NCV 05000282/2006006-01; 05000306/2006006-01)

OTHER ACTIVITIES (OA)

4OA2Identification and Resolution of Problems.1Routine Review of Condition Reports

a. Inspection Scope

From March 6 through March 24, 2006, the inspectors Action Process documents that identified or were related to 10 CFR 50.59 evaluationsand permanent plant modifications. The inspectors reviewed these documents to evaluate the effectiveness of corrective actions related to permanent plant modifications and evaluations for changes, tests, or experiments issues. In addition, corrective action documents written on issues identified during the inspection were reviewed to verify adequate problem identification and incorporation of the problems into the correctiveaction system. The specific corrective action documents that were sampled and reviewed by the team are listed in the attachment to this report.

b. Findings

No findings of significance were identified.

OTHER ACTIVITIES

4OA6Meetings.1Exit MeetingThe inspectors presented the inspection results to Mr. T. Palmisano and others of thelicensee's staff, on March 24, 2006. Licensee personnel acknowledged the inspection results presented. Licensee personnel were asked to identify any documents, materials, or information provided during the inspection that were considered proprietary other than those returned. No additional proprietary information was identified.ATTACHMENT:

SUPPLEMENTAL INFORMATION

KEY POINTS OF CONTACT

Licensee

T. Palmisano, Site Vice President

C. Mundt, Design Engineering Manager

J. Kivi, Senior Regulatory Compliance Engineer

S. Thomas, Design Engineering Supervisor

L. Gunderson, Mechanical Design Engineer

C. Sansome, Mechanical Design EngineerNuclear Regulatory Commission

J. Adams, Senior Resident Inspector

Attachment

ITEMS OPENED, CLOSED, AND DISCUSSED

Opened

None.Opened and

Closed

05000282/2006006-01;05000306/2006006-01 NCVFailure to Consider Adverse Ampacity Effects of HighTemperature Conditions in the Auxiliary feedwaterPump RoomsDiscussedNone.

Attachment

LIST OF DOCUMENTS REVIEWED

The following is a list of licensee documents reviewed during the inspection, includingdocuments prepared by others for the licensee.

Inclusion on this list does not imply that NRCinspectors reviewed the documents in their entirety, but rather, that selected sections orportions of the documents were evaluated as part of the overall inspection effort.

Inclusion of a