IR 05000282/2006006
| 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
SUBJECT:
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 NRC EVALUATION OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT 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 combined baseline 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 the completion of the inspection on March 24, 2006.
The inspectors examined activities conducted under your license as they relate to safety and compliance with the Commissions 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 safety significance was identified which involved a violation of NRC requirements. However, because this violation was of very low safety significance, not willful, and because it was entered into your corrective action program, the NRC is treating the issue as a Non-Cited Violation in accordance with Section VI.A.1 of the NRCs Enforcement Policy.
If you contest the subject or severity of a Non-Cited Violation, you should provide a response within 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 of Enforcement, 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 letter and its enclosure will be available electronically for public inspection in the NRC Public Document 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, Chief Engineering Branch 1 Division of Reactor Safety Docket Nos. 50-282; 50-306 License 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 III==
Docket No: 50-282; 50-306 License No:
Report No: 05000282/2006006 (DRS); 05000306/2006006 (DRS)
Licensee:
Facility: Prairie Island Nuclear Generating Plant Location:
Dates: March 6 through March 24, 2006 Inspectors: J. Neurauter, Senior Reactor Inspector, Team Leader Alan Dahbur, Reactor Inspector Approved by: D. Hills, Chief Engineering Branch 1 Division of Reactor Safety (DRS)
SUMMARY OF FINDINGS
IR 05000282/2006006 (DRS); 05000306/2006006 (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.
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 commercial nuclear 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, Design Control, having very low safety significance was identified by the inspectors.
Specifically, the licensee had not evaluated and updated the associated plant cable ampacity calculation to determine the potential consequences of adverse effects to cabling due to higher temperatures in the auxiliary feedwater (AFW) pump rooms and other auxiliary building areas. After identification by the inspectors, the licensee was able to demonstrate that even though the higher temperatures decreased the ampacity margins 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 cornerstone objective to ensure the availability, reliability, and capability of systems that mitigate transients 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 licensees preliminary evaluation determined that the higher temperatures in the AFW pump rooms and other auxiliary building areas would not prevent equipment important to safety from functioning. (Section 1R17.1.b.1)
Cornerstone: Barrier Integrity
No findings of significance were identified.
Licensee-Identified Violations
None.
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R02 Evaluations of Changes, Tests, or Experiments
.1 Review of 10 CFR 50.59 Evaluations and Screenings
a. Inspection Scope
From March 6 through March 24, 2006, the inspectors reviewed eight evaluations performed pursuant to 10 CFR 50.59. The inspectors confirmed that the evaluations were thorough and that prior NRC approval was obtained as appropriate. The inspectors 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 50.59 evaluation. 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 in Regulatory 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
.1 Review of Permanent Plant Modifications
a. Inspection Scope
From March 6 through March 24, 2006, the inspectors reviewed twelve permanent plant modifications 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 the modifications. 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 affects the barrier integrity cornerstone, was not selected as part of this inspection. This modification will be inspected at a later date in accordance with inspection procedure 71007, Reactor Vessel Head Replacement Inspection.
b. Findings
b.1 Failure to Consider Adverse Ampacity Effects of High Ambient Temperature Conditions in the Auxiliary Feedwater Pump Rooms
Introduction:
On March 15, 2006, the inspectors identified a Non-Cited Violation of 10 CFR Part 50, Appendix B Criterion III, Design Control, of very low safety significance (Green). Specifically, the licensee had not evaluated and updated the associated plant cable ampacity calculation to determine the potential consequences of adverse effects to cabling due to higher temperatures in the AFW pump rooms and other auxiliary building areas Discussion: Revised licensee calculation ENG-ME-021, Auxiliary Feedwater Pump Room Heat-up, indicated that the potential maximum ambient temperature in the AFW pump rooms could reach up to 127EF. The potential high ambient temperature could occur during post accident mitigation (an extended loss of off-site power) and when the initial ambient temperature in the rooms was at 104EF.
The licensee evaluated the effects of the high ambient temperature on the safety-related equipment (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 for an ambient room temperature of 127EF. This conclusion was also documented in the licensees 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 licensees 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 could also reach up to 122EF in several areas in the auxiliary building.
Prairie Island Engineering Manual for Electrical Cables Design, Fabrication and Installation Summary was based on an ambient temperature of 104EF. Other plant specific 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 104EF. The licensee failed to evaluate and update the cable ampacity calculation to evaluate the effects of potential high ambient temperatures on the 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 by cables 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 that safety 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 limit where 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 deficiency warranting 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 important to safety.
The finding was greater than minor in accordance with IMC 0612, Power Reactor Inspection Reports, Appendix B, Issue Screening, because it affected the mitigating system cornerstone objective to ensure the availability, reliability, and capability of systems that mitigate transients 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.
The inspectors determined the finding was of very low significance (Green) using IMC 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 licensees preliminary evaluation determined that the higher temperatures in the AFW pump rooms and other auxiliary building areas would not prevent equipment important 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 could experience. The Prairie Island calculation and engineering manual that did address cable ampacity were significantly less conservative, since temperatures of 104EF were assumed where temperatures in these areas could exceed 122EF.
Because this issue was of very low safety significance, not willful, and because it was entered in the licensees 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)
4OA2 Identification and Resolution of Problems
.1 Routine 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 evaluations and 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 corrective action 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
4OA6 Meetings
.1 Exit Meeting
The inspectors presented the inspection results to Mr. T. Palmisano and others of the licensees 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 Engineer
Nuclear Regulatory Commission
- J. Adams, Senior Resident Inspector
ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
None.
Opened and Closed
- 05000282/2006006-01; NCV Failure to Consider Adverse Ampacity Effects of High
- 05000306/2006006-01 Temperature Conditions in the Auxiliary feedwater Pump Rooms
Discussed
None.