IR 05000282/2014008
| ML14162A593 | |
| Person / Time | |
|---|---|
| Site: | Prairie Island  |
| Issue date: | 06/11/2014 |
| From: | Robert Daley Engineering Branch 3 |
| To: | Davison K Northern States Power Co |
| Alan Dahbur | |
| References | |
| IR-14-008 | |
| Download: ML14162A593 (19) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION une 11, 2014
SUBJECT:
PRAIRIE ISLAND NUCLEAR GENERATING PLANT, UNITS 1 AND 2 EVALUATIONS OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000282/2014008; 05000306/2014008
Dear Mr. Davison:
On June 3, 2014, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications inspection at your Prairie Island Nuclear Generating Plant, Units 1 and 2. The enclosed inspection report documents the inspection results which were discussed on May 1, 2014, with Mr. S. Sharp, Site Operation Director, and on June 3, 2014 with Mr. J. Mathew, Design Engineering Manager, and other members of your staff.
The inspection 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.
The NRC inspectors documented one finding of very low safety significance (Green) in this report. This finding was determined to involve violations of NRC requirements. However, because of its very low safety significance and because the issue was entered into your Corrective Action Program, The NRC is treating the issue as Non-Cited Violation (NCV) in accordance with Section 2.3.2 of the NRC Enforcement Policy. Additionally, a licensee identified violation is listed in Section 4OA7 of this report.
If you contest the subject or severity of the 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 NRC Resident Inspector office at the Prairie Island Nuclear Generating Plant. In addition, if you disagree with the cross-cutting aspect assigned to the finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region III, and the NRC Resident Inspector at the Prairie Island Nuclear Generating Plant.
In accordance with Title 10, Code of Federal Regulations (CFR), Section 2.390 of the NRC's
"Rules of Practice," a copy of this letter, its enclosure, and your response (if any), will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records System (PARS) component of NRC's Agencywide Documents Access and Management System (ADAMS), accessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Docket Nos. 50-282; 50-306 License Nos. DPR-42; DPR-60
Enclosure:
Inspection Report 05000282/2014008; 05000306/2014008 w/Attachment: Supplemental Information
REGION III==
Docket Nos.: 50-282; 50-306 License Nos.: DPR-42; DPR-60 Report Nos.: 05000282/2014008; 05000306/2014008 Licensee: Northern States Power Company, Minnesota Facility: Prairie Island Nuclear Generating Plant, Units 1 and 2 Location: Welch, MN Dates: April 14 through June 3, 2014 Inspectors: A. Dahbur, Senior Reactor Inspector, Lead J. Gilliam, Reactor Inspector S. Sheldon, Senior Reactor Inspector D. Szwarc, Senior Reactor Inspector Approved by: Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Enclosure
SUMMARY OF FINDINGS
IR 05000282/2014008, 05000306/2014008; 04/14/2014 - 06/03/2014; Prairie Island Nuclear
Generating Plant, Units 1 and 2; Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications.
This report covers a two-week announced baseline inspection on evaluations of changes, tests, and experiments and permanent plant modifications. The inspection was conducted by Region III inspectors. One finding was identified by the inspectors. The finding was considered Non-Cited Violation (NCV) of NRC regulations. The significance of most findings is indicated by their color (i.e., greater than Green, or Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Cross-cutting aspects were determined using IMC 0310, Aspects Within the Cross Cutting Areas. Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. All violations of NRC requirements are dispositioned in accordance with the NRCs Enforcement Policy dated July 9, 2013. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 5, dated February 2014.
NRC-Identified
and Self-Revealed Findings
Cornerstone: Mitigating Systems
- Green.
The inspectors identified a finding of very low safety significance and associated NCV of the Prairie Island Nuclear Generating Plant Facility Operating License Condition 2.C.(4) for the licensees failure to implement the requirements as specified in the Fire Protection Program (FPP) for impaired safe shutdown equipment. Specifically, the licensee failed to establish appropriate compensatory measures when they identified lack of coordination between DC panel fuses and upstream panels supply fuse under fault conditions for several safe shutdown power supplies. The licensee replaced all miss-coordinated fuses and entered the issue into their Corrective Action Program.
The performance deficiency was determined to be more than minor because the finding was associated with the Mitigating Systems cornerstone attribute of Protection Against External Factors (Fire) and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to fire events prevent undesirable consequences (i.e., core damage). Specifically, the failure to establish compensatory measures for lack of fuse coordination degraded the defense and depth element of the Fire Protection Program. The finding represented a low degradation and therefore the inspectors determined that the finding screened as having very low safety significance (Green) in Task 1.3.1 of IMC 0609, Appendix F. The inspectors determined that the finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence for the licensees failure to follow instructions as specified in Procedure FP-E-CAL-01 Calculations. [H.8] (Section 1R17.2(b)(1))
Licensee-Identified Violations
Violations of very-low-safety significance or Severity Level IV that were identified by the licensee have been reviewed by the inspectors. Corrective actions taken or planned by the licensee have been entered into the licensees Corrective Action Program. These violations and corrective action tracking numbers are listed in Section 4OA7 of this report.
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events and Mitigating Systems
1R17 Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications
.1 Evaluations of Changes, Tests, and Experiments
a. Inspection Scope
The inspectors reviewed eight safety evaluations performed pursuant to Title 10, Code of Federal Regulations (CFR) 50.59 to determine whether the evaluations were adequate and that prior NRC-approval was obtained as appropriate. The inspectors also reviewed fourteen screenings where licensee personnel had determined that a 10 CFR 50.59 evaluation was not necessary. The inspectors reviewed these documents to determine if:
- the changes, tests, and experiments performed were evaluated in accordance with 10 CFR 50.59 and that sufficient documentation existed to confirm that a license amendment was not required;
- the safety issue requiring the change, tests or experiment was resolved;
- the licensee conclusions for evaluations of changes, tests, and experiments were correct and consistent with 10 CFR 50.59; and
- the design and licensing basis documentation was updated to reflect the change.
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.
This inspection constituted eight samples of evaluations and fourteen samples of screenings and/or applicability determinations as defined in IP 71111.17 04.
b. Findings
No findings of significance were identified.
.2 Permanent Plant Modifications
a. Inspection Scope
The inspectors reviewed six permanent plant modifications that had been installed in the plant during the last three years. This review included in-plant walkdowns for portions of the modified emergency Diesel Generators D1 and D2 starting air systems; verified actions to open batteries rooms doors as specified in procedures; Control Room Envelope and the connected Aux Building drains and the HELB Doors in Unit 2 Turbine Building Relay Room. The modifications were selected based upon risk significance, safety significance, and complexity. The inspectors reviewed the modifications selected to determine if:
- the supporting design and licensing basis documentation was updated;
- the changes were in accordance with the specified design requirements;
- the procedures and training plans affected by the modification have been adequately updated;
- the test documentation as required by the applicable test programs has been updated; and
- post-modification testing adequately verified system operability and/or functionality.
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.
This inspection constituted six permanent plant modification samples as defined in IP 71111.17 04.
b. Findings
- (1) No Compensatory Measure were established for Lack of Fuses Coordination associated with Safe Shutdown Power Supplies
Introduction:
The inspectors identified a finding of very low safety significance (Green)and associated NCV of the Prairie Island Nuclear Generating Plant Facility Operating License Condition 2.C.(4) for the licensees failure to implement the requirements as specified in the Fire Protection Program for impaired safe shutdown equipment.
Specifically, the licensee failed to establish appropriate compensatory measures when they identified lack of electrical coordination between DC panel fuses and upstream panels supply fuses under fault conditions for several safe shutdown power supplies.
The licensee replaced all miss-coordinated fuses and entered the issue into their Corrective Action Program.
Description:
On November 02, 2010, during a revision to ENG-EE-012, 125 Volts DC System Coordination Study, electrical coordination issues were identified in which non-selective fuse coordination existed between a DC panel fuses and the upstream fuses to the panels. The non-selective coordination was shown to exist for high values of current which would only be expected under fault conditions. The circuits identified were as follows: PNL 131-11, PNL 131-13, PNL 131-18, PNL 131-19, PNL 171-1, PNL 171-2, PNL 181-1, PNL 181-2, PNL 231-13, PNL 231-18, PNL 231-19, PNL 271-1, PNL 271-2, PNL 281-1,and PNL 281-2. This issue was documented in the Corrective Actions Program as AR 01256681. Complete review and verification of ENG-EE-012 Revision 1 under EC 16914 was completed on February 01, 2011 and confirmed that mis-coordination identified in the parent AR remained present. An evaluation of fuse replacement options was performed under AR 01256681 assignment, which developed recommended replacement fuse sizes/types to achieve coordination. Fuse replacements and selections were performed under ENG-EE-012, Revision 2 per EC 18289. All work orders for fuse replacement were completed by November 16, 2013.
During the inspectors review of the engineering changes and corrective actions associated with this issue, the inspectors questioned if this issue was reviewed by the Fire Protection Program group. Specifically, the inspectors were concerned if the effect of lack of fuses coordination on the safe shutdown power supplies and the safe shutdown analysis (SSA) were evaluated during the licensees review and if any compensatory measures were established. In response to the inspectors concern, the licensee indicated that no Fire Protection Engineers review was performed for this issue and hence no compensatory measures were established during the non-compliance time frame. The licensee entered the inspectors concern into their Corrective Action Program as AR 01428822 and identified that at the time of AR 01256681 issuance, the SSA Calculation GEN-PI-026, Revision 5D was the analysis of record. At that time there was no documented use of ENG-EE-012 as an input to GEN-PI-026, however, Calculation GEN-PI-026 provided the basis for fuse coordination for several power supplies credited in the SSA. The failure to properly identify Calculation ENG-EE-012 as an input to Calculation GEN-PI-026 resulted in a potential impact on the SSA with no review by the FFP Engineer.
The licensee entered the inspectors concern into their Corrective Action Program as AR 01428822 and planned to update GEN-PI-026 to include ENG-EE-012 as an input; verify that the SSA contains all required inputs to validate the output of the analysis; determine why ENG-EE-012 was not used as an input to GEN-PI-026; and issue actions to resolve any identified process weakness
Analysis:
The inspectors determined that the failure to establish compensatory measures was contrary to the Prairie Island Nuclear Generating Plant License Condition associated with the Fire Protection Program and was a performance deficiency. Specifically, the licensee failed to establish compensatory measures as specified in Procedure F5 Appendix K for lack of fuse coordination associated with safe shutdown power supplies.
The performance deficiency was determined to be more than minor because it was associated with the Mitigating Systems cornerstone attribute of Protection Against External Events (Fire) and affected the cornerstone objective of ensuring the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Specifically, the failure to establish appropriate compensatory measures degraded the defense and depth element of the Fire Protection Program.
In accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Phase I - Initial Characterization of Findings, Table 3, the inspectors determined that the finding affected the implementation of the administrative controls of the FPP. Therefore, screening under IMC 0609, Appendix F, Fire Protection Significance Determination Process, was required. The inspectors determined that the finding represented a low degradation because the compensatory measures were only associated with lack of coordination between fuses for safe shutdown power supplies and not because of hot work or other fire protection impairment. Therefore, the inspectors determined that the finding screened as having very-low-safety significance (Green) in Task 1.3.1 of IMC 0609, Appendix F.
This finding had a cross-cutting aspect in the area of Human Performance, Procedure Adherence for the licensees failure to follow instructions as specified in Procedure FP-E-CAL-01 Calculations, which is the governing procedure for preparing, reviewing, and managing configuration control of calculations and analysis. Specifically, the licensee failed to identify and review the SSA when updated Calculation ENG-EE-012 because the SSA did not reference the calculation as an input. [H.8].
Enforcement:
Prairie Island Nuclear Generating Plant facility Operating License Condition 2.C.(4), for both Units 1 and 2, required the licensee to implement and maintain in effect all provisions of the approved Fire Protection Program as described in the Updated Safety Analysis Report (USAR).
Procedure F5 Appendix K was a written procedure which covered Fire Protection Program implementation in that the procedure provided functional requirements, compensatory action, surveillance requirements, and reporting requirements of Fire Protection systems.
Procedure F5 Appendix K, Section 7.20, stated, in part, if a fire protection-related system, structure, or component (SSC) not addressed in Section 7.0 is determined to be impaired, appropriate compensatory measures shall be determined by the Fire Protection Program Engineer and their basis documented in a CAP.
Contrary to the above, from February 1, 2011 through November 16, 2013, the licensee failed to implement the approved Fire Protection Program as specified in Procedure F5, Appendix K, Section 7.2. Specifically, the licensee failed to establish appropriate compensatory measures for lack of fuse coordination associated with safe shutdown power supplies. The licensee entered this issue into their Corrective Action Program as AR 01428822 and planned to revise affected calculations and resolve any identified process weaknesses. Because this violation was of very-low-safety significance and it was entered into the licensees Corrective Action Program, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy (NCV 05000282/2014008-01; 05000306/2014008-01, No Compensatory Measure were established for Lack of Fuses Coordination associated with Safe Shutdown Power Supplies).
OTHER ACTIVITIES
4OA2 Identification and Resolution of Problems
a. Inspection Scope
The inspectors reviewed several corrective 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 of changes, tests, and experiments. 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 inspectors are listed in the to this report.
b. Findings
No findings of significance were identified.
4OA6 Management Meetings
.1 Exit Meeting Summary
On May 1, 2014, the inspectors presented the inspection results to Mr. S. Sharp, and on June 3, 2014, to Mr. J. Mathew and other members of the licensee staff. The licensee personnel acknowledged the inspection results presented and did not identify any proprietary content. The inspectors confirmed that all proprietary material reviewed during the inspection was returned to the licensee staff.
4OA7 Licensee-Identified Violations
The following violation of very low significance (Green) or Severity Level IV was identified by the licensee and is a violation of NRC requirements which meets the criteria of the NRC Enforcement Policy for being dispositioned as an NCV.
- The licensee identified a Severity Level IV violation of 10 CFR 50.59, (Changes, Tests, and Experiments, for the failure to demonstrate in a written evaluation that prior NRC-approval was not required for changes made to an accident analysis. Specifically, the licensee incorrectly concluded in written Evaluation 1102, Waste Gas Tank Rupture Dose Analysis, Revision 0 that higher activity levels and dose rates at the Exclusion Area Boundary and Low Population Zone associated with extended plant life due to license extension did not result in a more than minimal increase in the consequences of an accident previously evaluated in the UFSAR.
The performance deficiency was determined to be more than minor because it was associated with the Radiation Safety cornerstone attribute of program and process and affected the cornerstone objective of ensuring adequate protection of public health and safety from exposure to radioactive materials released into the public domain as a result of routine civilian nuclear reactor operation. The inspectors determined the violation was of Severity Level IV because the associated finding was of very low safety significance (Green) as there was no actual radioactive material release. The licensee entered this issue into their Corrective Action Program as AR 1417573 and AR 1427150 and intended to submit a license amendment request for review by the NRC.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- J. Hallenbeck, Site Engineering Director
- S. Sharp, Site Operation Director
- N. Haskell, Corporate Engineering Director
- H. Butterworth, Nuclear Oversight Manager
- J. Kivi, Employee Concern Program Manager
- B. Rogers, Procurement. Manager
- W. Partridge, Mechanical Design Engineering
- S. Kerus, Procurement Supervisor
- I. Nordby, Regulatory Assurance
- J. Mathew, Design Engineering Manager
- R. Schaefer, Design Engineering Supervisor
- D. Vincent, Regulatory Assurance
- H. Storgen, Site Engineering
- B. Wegner, Mechanical Design Engineering
Nuclear Regulatory Commission
- G. Shear, Director, Division of Reactor Safety
LIST OF ITEMS
OPENED, CLOSED AND DISCUSSED
Opened/Closed
- 05000282/2014008-01 NCV No Compensatory Measure were established for Lack
- 05000306/2014008-01 of Fuses Coordination associated with Safe Shutdown Power Supplies (Section 1R17.2(b)(1))
Attachment