IR 05000346/2014007
| ML15050A150 | |
| Person / Time | |
|---|---|
| Site: | Davis Besse, Trojan  |
| Issue date: | 02/17/2015 |
| From: | Robert Daley Engineering Branch 3 |
| To: | Lieb R FirstEnergy Nuclear Operating Co |
| References | |
| IR 2014007 | |
| Download: ML15050A150 (20) | |
Text
UNITED STATES ary 17, 2015
SUBJECT:
DAVIS-BESSE NUCLEAR POWER STATION EVALUATIONS OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS BASELINE INSPECTION REPORT 05000346/2014007
Dear Mr. Lieb:
On January 9, 2015, the U.S. Nuclear Regulatory Commission (NRC) completed an Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications Inspection at your Davis-Besse Nuclear Power Station. The enclosed inspection report documents the inspection results which were discussed on January 9, 2015, with you 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.
One NRC-identified finding of very-low safety significance (Green) was identified during this inspection. This finding was determined to involve a violation of NRC requirements. However, because of the very-low safety significance and because the issue was entered into your corrective action program, the NRC is treating the issue as a Non-Cited Violation (NCV) in accordance with Section 2.3.2 of the NRC Enforcement Policy.
If you contest the violation or significance of the NCV, 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 copies to the Regional Administrator, U.S. Nuclear Regulatory Commission - Region III; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, DC 20555-0001; and the Resident Inspector at Davis-Besse Nuclear Power Station. In addition, if you disagree with the cross-cutting aspect assigned to any 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 Davis-Besse Nuclear Power Station. In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 2.390, Public Inspections, Exemptions, Requests for Withholding, 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 NRCs Public Document Room or from the Publicly Available Records (PARS)
component of the NRC's Agencywide Documents Access and Management 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/
Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Docket Nos. 50-346;72-017 License No. NPF-3
Enclosure:
Inspection Report 05000346/2014007 w/Attachment: Supplemental Information
REGION III==
Docket Nos: 50-346;72-017 License No: NPF-3 Report No: 05000346/2014007 Licensee: FirstEnergy Nuclear Operating Company (FENOC)
Facility: Davis-Besse Nuclear Power Station Location: Oak Harbor, OH Dates: November 3, 2014 - January 09, 2015 Inspectors: Jasmine Gilliam, Acting Senior Reactor Inspector (Lead)
Loyd Cain, Senior Resident Inspector (Vogtle)
Ijaz Hafeez, Reactor Inspector Lionel Rodriguez, Reactor Inspector (Observer)
Approved by: Robert C. Daley, Chief Engineering Branch 3 Division of Reactor Safety Enclosure
SUMMARY OF FINDINGS
IR 05000346/2014007; 11/03/2014 - 01/09/2015; Davis-Besse Nuclear Power Station;
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 based engineering inspectors. One finding of very-low safety significance was identified by the inspectors. The finding was considered a Non-Cited Violation (NCV) of U.S. Nuclear Regulatory Commission (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: Initiating Events
- Green.
The inspectors identified a finding of very-low safety significance (Green) and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III,
Design Control for the licensees failure to install and control the Reactor Coolant Pump (RCP) seal cavity vent flexible hoses per the design basis analysis. Specifically, the licensee failed to correctly translate the design basis installation configuration and installation fatigue analysis in calculation SP-274-I, Pipe Stress Analysis: Reactor Coolant Pump 1-1-1 Seal Cavity Vent, into specifications, drawings, procedures, and instructions. The licensee entered this finding into their Corrective Action Program (CAP) to review the lack of controls over the installation of the flexible hoses, but determined that the flexible hoses remained operable.
The performance deficiency was determined to be more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern.
Specifically, the failure to install and control the flexible hoses in accordance with the design basis analysis could lead to failure of the hoses due to operation beyond their analyzed limits. The finding screened as of very-low safety significance (Green)because the finding could not result in exceeding the Reactor Coolant System (RCS)leak rate for a small Loss of Coolant Accident (LOCA) after a reasonable assessment of degradation, and it could not have likely affected other systems used to mitigate a LOCA resulting in a total loss of their function after a reasonable assessment of degradation.
The inspectors determined this finding had an associated cross-cutting aspect, Design Margins, in the Human Performance cross-cutting area. Specifically, the licensee did not carefully guard and change the RCP seal cavity vent lines, which form part of the RCS fission product barrier, through a systematic and rigorous process. [H.6]
(Section 1R17.2.b.(1))
Licensee-Identified Violations
No violations were identified.
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R17 Evaluations of Changes, Tests, and Experiments and Permanent Plant Modifications
.1 Evaluation of Changes, Tests, and Experiments
a. Inspection Scope
The inspectors reviewed 6 safety evaluations performed pursuant to Title 10, Code of Federal Regulations (CFR) 50.59 to determine if the evaluations were adequate and that prior U.S. Nuclear Regulatory Commission (NRC) approval was obtained as appropriate.
The inspectors also reviewed 13 screenings and 2 applicability determinations 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 6 samples of evaluations and 15 samples of screenings and/or applicability determinations as defined in Inspection Procedure (IP) 71111.17-04.
b. Findings
No findings 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 control room; emergency diesel generators fuel oil storage tank level transmitter and indicator; fire door 519 A and 520A; and dampers CV 5325 B/C. 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) Failure to Install and Control Reactor Coolant Pump (RCP) Seal Cavity Vent Flexible Hoses Per Design Basis Analysis
Introduction:
The inspectors identified a finding of very low safety significance (Green)and associated Non-Cited Violation (NCV) of 10 CFR Part 50, Appendix B, Criterion III, Design Control for the licensees failure to install and control the RCP seal cavity vent flexible hoses per the design basis analysis. Specifically, the licensee failed to correctly translate the design basis installation configuration and installation fatigue analysis in calculation SP-274-I, Pipe Stress
Analysis:
Reactor Coolant Pump 1-1-1 Seal Cavity Vent, into specifications, drawings, procedures, and instructions.
Description:
The Davis-Besse Reactor Coolant System (RCS) uses four RCPs. The RCPs are shaft-sealed with a seal cartridge assembly that consists of three mechanical face-type sealing stages. Each sealing stage has a cavity vent line. The vent lines are classified as American Society of Mechanical Engineers (ASME)Section III, Class 2 piping (3/4 inch pipe). The lines form part of the Reactor Coolant Pressure Boundary (RCPB) and are safety-related. Typically, piping that is part of the RCPB is required to be classified as ASME Class 1 piping per 10 CFR 50.55a(c)(1). However, the RCP seal cavity vent lines at Davis-Besse are classified as ASME Class 2 in accordance with 10 CFR 50.55a(c)(2) because in the event of a postulated failure of the lines, the reactor can be shut down and cooled down in an orderly manner assuming seal injection is maintained by the Makeup System.
On August 6, 2012, and later on August 27, 2013, the licensee submitted Licensee Event Reports (LERs) to the NRC which described two different leaks from welds on the first stage seal cavity vent line for RCP 1-2. The LERs identified high-cycle fatigue as the failure mechanism for the welds. As part of the planned corrective actions to address the high-cycle fatigue, the licensee developed Engineering Change Package (ECP) 12-0785, Install Flexible Hose on RCP Seal 1st, 2nd, 3rd Stage Vents, Seal Injection and Controlled Bleedoff Lines, to replace a section of each of the RCP seal cavity vent lines with flexible hoses, among other things. The modification was implemented during the sites 18th refueling outage which began on February 1, 2014, and ended on May 6, 2014. A total of 12 flexible hoses were installed, three on each of the four RCPs. One end of each flexible hose was connected directly to its corresponding RCP, while the other end of the flexible hose was connected to the existing vent piping.
The licensee procured the flexible hoses to meet the same design requirements as the original ASME Section III, Class 2 piping. The flexible hoses were designed, analyzed, and fabricated by a vendor. The vendor provided the licensee with the flexible hose design calculations, and these were incorporated into new pipe stress calculations for each RCP developed in support of ECP 12-0785. Through a review of the pipe stress calculation related to RCP 1-1-1, SP-274-I, Pipe Stress
Analysis:
Reactor Coolant Pump 1-1-1 Seal Cavity Vent, and discussions with the licensee, the inspectors identified two design control issues related to the flexible hoses.
The first issue identified by the inspectors was a non-conformance between the analyzed configuration of the flexible hoses by the vendor, and the installed configuration of the flexible hoses. The vendor analyzed the flexible hoses as being fixed (anchored) at the ends which attached to the existing piping, as documented in 5 of calculation SP-274-I. However, the inspectors identified that two-way restraints, instead of fixed restraints, were used to support the piping near those flexible hose ends. Fixed restraints would prevent displacement and rotation of the flexible hoses and piping in all directions. The installed two-way restraints allow the pipe and flexible hoses to displace in the axial direction, and also allow rotation in all directions.
Since the installed configuration of the flexible hoses and piping allows axial displacement, the inspectors questioned the qualification of the flexible hoses and piping under design basis loads, such as earthquakes. Specifically, the inspectors questioned whether the analysis for the flexible hoses had considered the additional seismic displacement loads resulting from the non-conformance of the installed configuration.
The licensees pipe stress analysis, including the vendors flexible hose analysis, did not account for the non-conformance.
This issue was entered into the licensees Corrective Action Program (CAP) as CR-2014-17626, 2014 NRC MOD/50.59 Inspection - Evaluation of RCP Flex Hose Installed Configuration and CR-2014-17983, 2014 NRC MOD/50.59 Inspection -
Additional NRC Questions/Concerns on RCP Flex Hoses, which addressed the non-conformance between the analyzed and installed configuration of the flexible hoses and the piping. The licensee determined that the non-conformance could increase the seismic fatigue of the flexible hoses, based on discussions with the vendor and a review of the pipe stress calculations. However, because additional margin existed in the flexible hose fatigue analysis, the licensee concluded that the flexible hoses remained operable with reasonable expectation. The additional margin in the fatigue analysis had been incorporated into the thermal fatigue portion which evaluated a conservative RCS heat up and cool down fatigue life of 1550 cycles versus the actual RCS limit of 240 cycles. The licensees planned corrective action for the issue at the conclusion of the inspection was to obtain a revised calculation from the vendor and revise their pipe stress calculations to address the as-built configuration of the flexible hoses and piping.
The second issue identified by the inspectors for the flexible hoses is related to the installation fatigue evaluated by the vendor in the flexible hose design analysis. The design calculations for the flexible hoses evaluated an installation fatigue of five cycles, as documented in Attachment 5 of calculation SP-274-I. An installation cycle consists of one removal and re-installation of the flexible hoses. Therefore, in order for the design basis installation fatigue evaluation to remain valid, the licensee was required to control and limit the flexible hose installation cycles. During the inspection, the inspectors determined that the licensee did not have any controls in place to limit the installation cycles for the flexible hoses. The flexible hoses would, at a minimum, undergo an installation cycle during each RCP motor or seal replacement. The RCP seals are typically replaced every 8 years.
This issue was entered into the licensees Corrective Action Program (CAP) as CR-2014-17319, 2014 NRC MOD/50.59 Inspection - Tracking of Installation Cycle Life for RCP Flex Hoses, which recognized that the flexible hoses had an installation limit of five cycles, and that the installation cycles were not being controlled because there was no tracking mechanism for these cycles. Currently, the flexible hoses have only accumulated one installation cycle because they were installed during the 2014 refueling outage and have not been removed since. Therefore, the flexible hoses have not exceeded the installation fatigue analysis limit, and there is reasonable expectation that all RCP seal cavity vent flexible hoses are operable. The licensees planned corrective actions at the conclusion of the inspection for this issue were to either track the flexible hose installation cycles and replace the hoses prior to exceeding their installation cycle limit, or to revise the fatigue analysis to increase the number of installation cycles to the point where tracking of the cycles would not be required.
Analysis:
The inspectors determined the licensees failure to install and control the RCP seal cavity vent flexible hoses in accordance with the design basis analysis was contrary to 10 CFR Part 50, Appendix B, Criterion III, Design Control, and was a performance deficiency. The performance deficiency was determined to be more than minor because, if left uncorrected, it would have the potential to lead to a more significant safety concern. Specifically, the failure to install and control the flexible hoses in accordance with the design basis analysis could lead to failure of the hoses due to operation beyond their analyzed limits.
The inspectors determined the finding could be evaluated using the Significance Determination Process (SDP) in accordance with IMC 0609, Significance Determination Process, Attachment 0609.04, Initial Characterization of Findings, issued on June 19, 2012. Specifically, the inspectors used IMC 0609 Appendix A SDP for Findings At-Power, issued June 19, 2012, Exhibit 1, Initiating Events Screening Questions, to screen the finding. The Initiating Events Screening Questions were used instead of the Barrier Integrity Screening Questions because RCS Boundary issues are evaluated under the Initiating Events cornerstone unless they are related to pressurized thermal shock, as discussed in Section 5.0 of IMC 0609 Appendix A. The finding screened as of very low safety significance (Green) because the inspectors answered No to all of the screening questions in Subsection A, LOCA Initiators, of Exhibit 1.
Specifically, the finding could not result in exceeding the RCS leak rate for a small Loss of Coolant Accident (LOCA) after a reasonable assessment of degradation, and it could not have likely affected other systems used to mitigate a LOCA resulting in a total loss of their function after a reasonable assessment of degradation. As discussed above, the flexible hoses remained operable and are not expected to fail due to the lack of design controls. In addition, the size of the lines is such that in the event of a postulated failure of a single line the reactor can be shut down and cooled down in an orderly manner assuming seal injection is maintained by the Makeup System.
The inspectors determined this finding had an associated cross-cutting aspect, Design Margins (H.6), in the Human Performance cross-cutting area. This corresponds to the apparent cause identified by the inspectors for the finding, the licensees failure to operate and maintain equipment within design margins, to carefully guard and change margins only through a systematic and rigorous process, and to place special attention to maintaining fission product barriers, defense-in-depth, and safety-related equipment.
Specifically, the licensee did not carefully guard and change the RCP seal cavity vent lines, which form part of the RCS fission product barrier, through a systematic and rigorous process. This led to the failure to adequately install and control the flexible hoses in accordance with the design basis analysis. [H.6]
Enforcement:
10 CFR Part 50, Appendix B, Criterion III, Design Control requires, in part, that measures be established to assure that applicable regulatory requirements and the design basis are correctly translated into specifications, drawings, procedures, and instructions.
Contrary to the above, since about May 6, 2014, the licensee failed to assure that applicable regulatory requirements and the design basis were correctly translated into specifications, drawings, procedures, and instructions. Specifically, the licensee failed to install and control the RCP seal cavity vent flexible hoses per the design basis analysis in calculation SP-274-I, Pipe Stress
Analysis:
Reactor Coolant Pump 1-1-1 Seal Cavity Vent.
The licensees immediate corrective actions taken to restore compliance were:
- (1) to verify through discussions with the vendor that the flexible hoses would be expected to perform as designed under design basis loads, and
- (2) to verify that the flexible hoses had not exceeded the installation fatigue analysis limit.
Because this violation was of very-low safety significance and was entered into the licensees CAP as CR-2014-17626, 2014 NRC MOD/50.59 Inspection - Evaluation of RCP Flex Hose Installed Configuration, CR-2014-17983, 2014 NRC MOD/50.59 Inspection - Additional NRC Questions/Concerns on RCP Flex Hoses, and CR-2014-17319, 2014 NRC MOD/50.59 Inspection - Tracking of Installation Cycle Life for RCP Flex Hoses, this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC Enforcement Policy. [NCV 05000346/2014007-01, Failure to Install and Control RCP Seal Cavity Vent Flexible Hoses Per Design Basis Analysis]
OTHER ACTIVITIES (OA)
4OA2 Problem Identification and Resolution
.1 Routine Review of Condition Reports
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 attachment to this report.
b. Findings
- (1) Failure to Use Most Limiting 4160 Volts Alternating Current (VAC) Bus Voltage in Design Calculations
Introduction:
The inspectors identified an unresolved item (URI) for the licensees failure to perform an analysis demonstrating that at the degraded voltage relay (DVR) set point specified in Technical Specifications (TSs) adequate voltage would be available to safety-related equipment to start and run during a design basis accident.
Description:
On September 5, 2014, the licensee was informed by Schulz Electric that High Pressure Injection (HPI) Pump 1-2 Motor will not meet all Purchase Order Requirements. Specifically, the supplied motor may not be capable of accelerating the driven load to normal operating speed within 6.5 seconds at a minimum starting voltage of 70 percent. The licensee issued a condition report and initiated Prompt Operability Determination (POD) 2014-13985, HPI Pump 1-2 Motor to evaluate the impact of the condition on the ability of the HPI Pump 1-2 Motor to perform its design function. The licensee concluded that HPI 1-2 Pump was operable but nonconforming and placed a note to operators that in certain electrical plant alignments the HPI 1-2 Pump should be declared inoperable.
On October 8, 2014, the licensee initiated CR- 2014-15452, Inconsistency in the Treatment of the Plant Restrictions in Operability Determinations. The corrective actions documented the review of POD 2014-13985 to determine if the recommended actions/instructions to operators should have been considered compensatory actions and therefore screened in the licensees 50.59 process. The licensee concluded that the instructions to operators only provided information and therefore not considered a compensatory action.
During the review of POD 2014-13985, the inspectors were concerned that the licensee did not adequately address the operability of HPI Pump 1-2 Motor with respect to the DVR set point TS value of 3712 VAC. The inspectors reviewed NRC Regulatory Issue Summary (RIS) 2011-12, Revision 1, Adequacy of Station Electrical Distribution System Voltages; dated December 29, 2011, which was issued to clarify the NRC staffs technical position on existing regulatory requirements. Specifically, the RIS clarifies voltage studies necessary for DVR setting bases. The RIS states, in part, the limiting voltage at the bus monitored by the DVR can be calculated in terms of the voltage at the terminals of the most limiting safety-related component in the plant in all required operating conditions (such as starting and running). In addition, the RIS discusses that calculations of voltages at the terminals of all safety-related equipment with the voltage at the DVR monitored bus at the DVR dropout setting would ensure adequate voltage (starting and running) to all safety-related equipment.
The inspectors requested the licensees DVR set point analysis as described in RIS 2011-12. The licensee had an analysis that shows all of the safety-related loads would be able to run at steady-state at the TS DVR set point during design basis accidents. However, they do not have an analysis which shows all safety-related loads would start and run at the TS DVR set point.
The licensees position is that if a design basis accident were to occur and the 4160 VAC Bus (Bus C1/D1) was at the Degraded Voltage set point of 3712 VAC, because the time delay is set at 7 seconds, they would divorce from the offsite power source and power would be supplied by the Emergency Diesel Generators (EDGs).
Therefore, the licensee believes that they are not required to have an analysis which demonstrates that all required safety-related loads can start and run at the DVR set point, as described in RIS 2011-12, Revision 1.
Based on this information, the inspectors were concerned that the licensee does not have an analysis which demonstrates the safety-related loads could start and run at the TS DVR set point. The licensee captured the inspectors concerns in their CAP as CR 2014-17296, 2014 50.59 Inspection: Davis-Besse does not have an analysis to satisfy item 1 of RIS 2011-12, dated November 19, 2014. During the evaluation of CR 2014-17296 the licensee stated, An analysis has been performed and demonstrates that all loads receive adequate voltage to start and perform their intended function with the exception of some Motor Operated Valves (MOVs). In order to ensure the MOVs have adequate voltage to perform their function, a minimum voltage of 4070 VAC should be maintained on either 4160 VAC Bus (C1/D1). Since the minimum voltage of 4070 VAC is higher than the previous voltage of 3900 VAC, the licensee has added a compensatory action to monitor Bus C1/ D1 at the higher voltage. This issue is unresolved pending consultation with Nuclear Reactor Regulation to determine if the licensee is required to demonstrate that safety-related loads can start and run at the DVR TS set point.
4OA6 Management Meetings
.1 Interim Meeting Summary
On November 20, 2014, the inspectors presented the preliminary inspection results to Mr. K. Byrd and other members of the licensee staff. The licensee acknowledged the issues presented. The inspectors confirmed that none of the potential report input discussed was considered proprietary. The inspectors had outstanding questions that required additional review and a follow-up exit meeting.
.2 Exit Meeting Summary
On January 9, 2015, the inspectors presented the inspection results to Mr. R. Lieb 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.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee
- R. Lieb, Site Vice President
- J. Hook, Design Engineering Manager
- G. Michael, Design Engineering Supervisor
- G. Wolf, Regulatory Compliance Supervisor
- V. Wadsworth, Regulatory Compliance Specialist
U.S. Nuclear Regulatory Commission
- D. Kimble, Senior Resident Inspector
LIST OF ITEMS
OPENED, CLOSED AND DISCUSSED
Opened
- 05000346/2014007-01 NCV Failure to Install and Control RCP Seal Cavity Vent Flexible Hoses Per Design Basis Analysis (Section 1R17.2.b.(1))
- 05000346/2014007-02 URI Failure to Use Worst Case 4160 VAC Bus Voltage in Design Calculations (Section 4OA2.1.B(1))
Closed
- 05000346/2014007-01 NCV Failure to Install and Control RCP Seal Cavity Vent Flexible Hoses Per Design Basis Analysis (Section 1R17.2.b.(1))