IR 05000293/2013009
| ML13350A072 | |
| Person / Time | |
|---|---|
| Site: | Pilgrim |
| Issue date: | 12/12/2013 |
| From: | Paul Krohn Engineering Region 1 Branch 2 |
| To: | Dent J Entergy Nuclear Operations |
| References | |
| IR-13-009 | |
| Download: ML13350A072 (22) | |
Text
UNITED STATES ber 12, 2013
SUBJECT:
PILGRIM NUCLEAR POWER STATION - NRC EVALUATION OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT PLANT MODIFICATIONS TEAM INSPECTION REPORT 05000293/2013009 (REVISED SUBJECT)
Dear Mr. Dent:
On November 21, 2013, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Pilgrim Nuclear Power Station (PNPS). The enclosed inspection report documents the inspection results, which were discussed on November 21, 2013, 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.
In conducting the inspection, the team reviewed selected procedures, calculations and records, observed activities, and interviewed station personnel.
The NRC inspectors did not identify any findings or violations of more than minor significance.
In accordance with Title 10 of the Code of Federal Regulations 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 NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRC's document system, 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/
Paul G. Krohn, Chief Engineering Branch 2 Division of Reactor Safety Docket No: 50-293 License No: DPR-35
ML13350A072 Non-Sensitive Publicly Available SUNSI Review Sensitive Non-Publicly Available OFFICE RI/DRS RI/DRP RI/DRS NAME Spindale/via telecon RMcKinley PKrohn DATE 12/11/13 12/12/13 12/12/13
Enclosure:
Inspection Report 05000293/2013009 w/Attachment: Supplemental Information
REGION I==
Docket No: 50-293 License No: DPR-35 Report No: 05000293/2013009 Licensee: Entergy Nuclear Operations, Inc.
Facility: Pilgrim Nuclear Power Station (PNPS)
Location: Plymouth, MA 02360 Inspection Period: November 4 - 21, 2013 Inspectors: S. Pindale, Senior Reactor Inspector, Division of Reactor Safety (DRS),
Team Leader F. Arner, Senior Reactor Inspector, DRS K. Young, Senior Reactor Inspector, DRS Approved By: Paul G. Krohn, Chief Engineering Branch 2 Division of Reactor Safety i Enclosure
SUMMARY OF FINDINGS
IR 05000293/2013009; 11/4/13 - 11/21/13; Pilgrim Nuclear Power Station; Engineering
Specialist Plant Modifications Inspection.
This report covers a 2-week inspection of the evaluations of changes, tests, or experiments and permanent plant modifications. The inspection was conducted by three region based engineering inspectors. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 4, dated December 2006.
No findings were identified.
ii
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R17 Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications
.1 Evaluations of Changes, Tests, or Experiments (21 samples)
a. Inspection Scope
The team reviewed one safety evaluation to determine whether the changes to the facility or procedures, as described in the Updated Final Safety Analysis Report (UFSAR), had been reviewed and documented in accordance with Title 10 of the Code of Federal Regulations (CFR) Part 50.59 requirements. In addition, the team evaluated whether Entergy had been required to obtain U.S. Nuclear Regulatory Commission (NRC) approval prior to implementing the changes. The team interviewed plant staff and reviewed supporting information including calculations, analyses, design change documentation, procedures, the UFSAR, the Technical Specifications, and plant drawings to assess the adequacy of the safety evaluation. The team compared the safety evaluation and supporting documents to the guidance and methods provided in Nuclear Energy Institute (NEI) 96-07, Guidelines for 10 CFR 50.59 Evaluations, as endorsed by NRC Regulatory Guide 1.187, Guidance for Implementation of 10 CFR 50.59, Changes, Tests, and Experiments, to determine the adequacy of the safety evaluation.
The team also reviewed twenty 10 CFR 50.59 screenings for which Entergy had concluded that a safety evaluation was not required to be performed. These reviews were performed to assess whether Entergy's threshold for performing safety evaluations was consistent with 10 CFR 50.59. The samples included design changes, calculations, and procedure changes.
The team reviewed the safety evaluation that Entergy had performed and approved during the time period covered by this inspection (i.e., since the last plant modifications inspection) not previously reviewed by NRC inspectors. The screenings and applicability determinations were selected based on the safety significance, risk significance, and complexity of the change to the facility.
In addition, the team compared Entergys administrative procedures used to control the screening, preparation, review, and approval of safety evaluations to the guidance in NEI 96-07, Revision 1, to evaluate whether those procedures adequately implemented the requirements of 10 CFR 50.59. The reviewed safety evaluation and screenings are listed in the Attachment.
b. Findings
No findings were identified.
.2 Permanent Plant Modifications (11 samples)
.2.1 Evaluation of Residual Heat Removal System Hydraulic Analysis to Resolve Operator
Manual Action Compliance for Appendix R Scenarios
a. Inspection Scope
The team reviewed Engineering Change (EC) 26711, which was initiated to address operator manual action compliance applicable to restoring power to the 480 volt alternating current (Vac) Swing Bus 6 in the event of fire in either the A (Upper) or B (Lower) Switchgear Rooms. The intent of the EC was to verify that the residual heat removal (RHR) hydraulic analysis could support the conclusion that remote operator action to isolate low pressure coolant injection (LPCI) valves was not required to align the RHR system to operate in the torus cooling mode for decay heat removal. This analysis verified that further opening the valve in the torus cooling line removed the previous requirement to perform operator manual actions to re-energize the 480 Vac breakers for the LPCI injection valves in order to close them. The EC revised procedures on aligning and operating torus cooling for the scenario where the LPCI injection valves were open and could not be closed.
The team reviewed the analysis to evaluate whether the design inputs and outputs were technically reasonable. The team reviewed the procedure revisions to verify that the design bases, licensing bases, and performance capability of the containment and equipment in containment necessary for safe shutdown had not been degraded. The team reviewed the impact of increased flowrates in torus cooling on RHR pump net-positive-suction-head as well as the increased potential for vibration loading on the RHR heat exchanger tubes to ensure the design bases were maintained. The team reviewed the hydraulic analysis to verify that, with maximum torus cooling flowrates established, the RHR system pressure would not divert flow away from the torus cooling mode of operation. The team walked down the torus cooling return valve to evaluate the material condition of the valve to support the procedure revisions. The 10 CFR 50.59 process applicability determination associated with this EC was also reviewed as described in Section 1R17.1 of this report. Documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.2 Revised Service Life for Inboard Main Steam Isolation Valve Limit Switches
a. Inspection Scope
The team reviewed EC 36109, which determined a new service life for the equipment qualification (EQ) files for the four limit switches associated with each of the four inboard main steam isolation valves (MSIV), AO-203-1A/B/C/D. The 16 limit switches are connected to a local junction box via a cable/connector. The switches are environmentally qualified to the requirements of 10 CFR 50.49. Each MSIV has two reactor protection system (RPS) limit switches which provide input to the RPS when the MSIV position is less than or equal to 90 percent of full open. Two other limit switches provide valve position indication.
The team reviewed the revised analyses to verify that the design and licensing bases and performance capability of the main steam system had not been degraded. The team reviewed actual measured temperature data and the service life calculations to ensure that the temperature used as a design input was conservative and had accounted for any potential heat transfer from the MSIVs to the limit switches.
Additionally, the service life was reviewed to ensure that the dose rate enveloped the tested value. The team verified that the applicable EQ files had been appropriately revised. A review of condition reports (CR) was performed to evaluate whether there were any reliability or performance issues associated with the limit switches.
Additionally, the 10 CFR 50.59 process applicability determination associated with this EC was also reviewed as described in Section 1R17.1 of this report. Documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.3 Replacement of High Pressure Coolant Injection Turbine Exhaust Check Valve
a. Inspection Scope
The team reviewed EC 33917, which replaced the 2301-74 lift check valve with a swing check valve designed to operate over the entire range of conditions expected in the high pressure coolant injection (HPCI) turbine exhaust line. This included the design attribute to reliably seat without leakage when the valves downstream pressure exceeds its upstream pressure. Additionally, a butterfly valve (23-HO-321) was installed downstream of the check valve to facilitate isolation of the check valve from the torus to allow 10 CFR Part 50, Appendix J, local leak rate testing (LLRT). This eliminated the requirement to use a spectacle flange at the torus room wall to isolate the system for this testing.
The team reviewed the modification to verify that the design and licensing bases and performance capability of the HPCI system had not been degraded by the modification, including the containment isolation function. The team reviewed drawings and procedures to verify that they were revised in accordance with the modification package instructions. The team reviewed post-modification test results to confirm that the pressure loss of the new components in the HPCI turbine exhaust line was conservative with respect to the revised calculation and to ensure the pressure in the exhaust lines was maintained within design limitations. Additionally, post-modification test results were reviewed to ensure that the check valve met LLRT test acceptance criteria for leak tightness. A review of CRs was performed to evaluate whether there were any reliability or performance issues associated with the new check valve. Additionally, the 10 CFR 50.59 screening determination associated with the EC was reviewed as described in Section 1R17.1 of this report. Documents reviewed are listed in the
.
b. Findings
No findings were identified.
.2.4 A Emergency Diesel Generator Governor Control System Upgrade
a. Inspection Scope
The team reviewed modification EC 05974 that replaced the governor control system for the A emergency diesel generator (EDG). Entergy implemented the modification because the previous governor control system for the A EDG was no longer supported by the original equipment vendor, and the equipment obsolescence issues challenged the reliability of the EDG. The modification included installation of new specific governor control system components including the governor/actuator, the load sharing and speed control unit, and the digital reference unit to improve control system performance, minimize component failures, and avoid EDG load swings during testing.
The team reviewed the modification to verify that the design and licensing bases had not been degraded by the A EDG modification. The team interviewed the responsible engineers and reviewed associated evaluations to verify that the modified configuration was consistent with existing design assumptions. The team reviewed associated drawings, operating and maintenance procedures, calculations, and training documents to ensure they had been properly updated to incorporate the changes as a result of this modification. The team also reviewed CRs and EDG system health reports to determine if there were reliability or performance issues that may have resulted from the modification. In addition, post-modification testing and completed surveillance testing were reviewed to verify proper operation of the governor control system. The team performed a walkdown of the accessible components of the governor control system to identify abnormal conditions. The 10 CFR 50.59 screening determination associated with this modification was reviewed as described in Section 1R17.1 of this report. The documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.5 Replacement of High Pressure Coolant Injection Flow Controller and Square
Root Converter Located on an Auxiliary Shutdown Panel
a. Inspection Scope
The team reviewed modification EC 31306 that replaced the HPCI flow controller (FIC-2340-2) and square root converter (SQRT-2340-20) located on the alternate shutdown panel (C158). Entergy implemented the modification because the previous installed equipment had a history of marginal performance/failures and had caused unplanned availability that impacted NRC performance indicators. Additionally, the controller and square root converter were no longer produced or supported by the vendor. The replaced components provided the same function as the previous installed components, in that, they allowed control of HPCI pump flow from the auxiliary shutdown panel to permit shutdown of the plant from outside the control room.
The team reviewed the modification to verify that the design and licensing bases had not been degraded by the HPCI flow controller and square root converter modification. The team interviewed the responsible engineers and reviewed associated evaluations to verify that the modified configuration was consistent with existing design assumptions.
The team reviewed associated drawings, operating procedures, maintenance procedures, calculations, and training documents to ensure they had been properly updated to incorporate the changes as a result of the modification. The team also reviewed CRs and HPCI system health reports to determine if there were reliability or performance issues that may have resulted from the modification. In addition, post-modification testing and completed surveillance testing were reviewed to verify proper operation of the flow controller and square root converter from the alternate shutdown panel. The team performed a walkdown of the flow controller, square root converter, and the C158 alternate shutdown panel to identify abnormal conditions. The 10 CFR 50.59 screening determination associated with this modification was reviewed as described in Section 1R17.1 of this report. The documents reviewed are listed in the
.
b. Findings
No findings were identified.
.2.6 Replacement of the Residual Heat Removal Containment Spray Flow Indicators on
Control Room Panel
a. Inspection Scope
The team reviewed modification EC 41830 that replaced the RHR containment spray flow indicators (FI-1040-11B and FI-1040-12B) on control room panel C903. The indicators are used to monitor flow rates between 0 and 6,000 gallons per minute when the RHR system is operated in the containment spray mode. Entergy implemented the modification because the previously installed RHR flow indicators experienced a number of failures and required equivalent replacements to improve reliability and ensure technical specification requirements are met when RHR is placed in the containment spray mode of operation. The modification was installed to improve the reliability of the indicators and the RHR system.
The team reviewed the modification to verify that the design and licensing bases had not been degraded by the RHR containment spray flow indicator modification. The team interviewed the responsible engineers and reviewed associated evaluations to verify that the modified configuration was consistent with existing design assumptions. The team reviewed associated drawings, operating procedures, maintenance procedures, and training documents to ensure they had been properly updated to incorporate the changes as a result of the modification. The team also reviewed CRs and RHR system health reports to determine if there were reliability or performance issues that may have resulted from the modification. In addition, post-modification testing and completed surveillance testing were reviewed to verify proper operation of the containment flow indicators. The team performed a walkdown of the RHR containment spray flow indicator components and the C903 control room panel to identify abnormal conditions.
The 10 CFR 50.59 screening determination associated with this modification was reviewed as described in Section 1R17.1 of this report. The documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.7 Revision to Calculation PS257, 125 Vdc Battery D40/Station Blackout
Battery Sizing Calculation
a. Inspection Scope
The team reviewed a revision (EC 42720) to calculation PS257, 125 Vdc Battery D40/Station Blackout (SBO) Battery Sizing Calculation, to update the battery load profile to more accurately reflect the actual load, load period, and load profile for the SBO battery (D40). Additionally, the calculation was revised to incorporate comments and observations from previous NRC inspection activities. The SBO battery provides reliable direct current control voltage for the SBO diesel generator engine circuits and the 4160 Vac switchgear A8 loads in a loss of all alternating current (AC) event. The battery is sized to meet the voltage and load requirements (profile) of the system, without battery charger support, for the duration of the load profile (SBO coping period).
However, the previous calculation did not reflect the required rigor to accurately address actual load, load periods, and load profile and needed to be updated to better characterize these attributes and more rigorously detail the voltage drop analysis.
The team evaluated the calculation revision to confirm that the systems design bases, licensing bases, and performance capability would not be adversely affected by the change. The team interviewed the responsible engineers and reviewed the calculation and associated analysis to verify that the assumptions used in the calculation were valid.
Additionally, the team reviewed the calculation to evaluate whether proper rigor had been applied to better address actual load, load periods, and load profile for the SBO battery. Entergy used the battery sizing methodology outlined in Institute of Electrical and Electronic Engineers (IEEE) 1115, IEEE Recommended Practice for Sizing Nickel-Cadmium Batteries for Stationary Applications, as guidance to perform the calculation enhancements. The team also reviewed the calculation to verify that the battery had sufficient size and margin and remained able to perform its function during a SBO event.
The team also reviewed applicable surveillance procedures and calculation methodology to verify their adequacy. The team reviewed CRs, the 1E 125 Vdc and 250 Vac battery system health reports, and completed surveillance procedures to determine if reliability or performance issues existed. The team verified that Entergy had plans in place to conduct a battery discharge test to validate the assumptions in the calculation. The team walked down accessible portions of the D40, SBO battery and associated components to identify abnormal conditions. The 10 CFR 50.59 screening determination associated with this revised calculation was also reviewed as described in Section
1R17 .1 of this report. The documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.8 Qualify Automatic Depressurization System Accumulator Check Valve Soft Seats for
Use during an Additional Cycle
a. Inspection Scope
The team reviewed modification EC 43954 that evaluated and increased the qualified life for the automatic depressurization system (ADS) accumulator check valve soft seats from 4 years to 6 years (one additional operating cycle). The check valves provide a path for essential instrument air/nitrogen supply valves to the safety relief valves and safety valves. The existing design is Ethylene Propylene Diene Monomer (EPDM),which is a synthetic rubber that is commonly used in soft seat and cable insulation applications, and has a relatively high resistance to radiation.
The team reviewed the modification to verify that the design and licensing bases of the ADS system had not been degraded by extending the service life of the check valve soft seats. The team reviewed Entergys evaluation of the worst-case radiation exposure and temperature profile, as well as nominal parameter assumptions in evaluating the environment in which the check valves operate. The team also confirmed that the analysis considered relevant operating experience in evaluating this change. The team reviewed plant specific history in order to evaluate whether adverse performance trends were apparent. The 10 CFR 50.59 screening determination associated with this modification was reviewed as described in Section 1R17.1 of this report. The documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.9 Revise Calculation to Include Motor-Operated Valve Loading during Torus Cooling
a. Inspection Scope
The team reviewed a revision (EC 35395) to calculation PS162, SBO Diesel Generator Loading, to update the electrical load profile for the SBO diesel generator to add motor-operated valve (MOV) loads that would be required to operate during the torus cooling mode of the RHR system. The electrical load for the four MOVs was previously and inadvertently omitted from the calculation. In incorporating this change, Entergy identified an additional change that would reduce the SBO diesel generator load.
Specifically, three air compressors, which are no longer in service, still had their electrical loads included in the calculation. The team evaluated this aspect of the change as well.
The team evaluated the calculation revision to confirm that the system design bases, licensing bases, and performance capability would not be affected by the change. The team interviewed the responsible engineers and reviewed the calculation and associated analysis to verify the assumptions used in the calculation were valid. Additionally, the team reviewed the calculation to verify that proper rigor had been applied to better address actual load, load periods, and load profile for the SBO diesel generator. The team also reviewed the calculation to verify that the SBO diesel generator load profile had sufficient margin and the diesel generator remained able to perform its function during a SBO event. The team also reviewed applicable procedures to confirm that the three air compressors would not be re-connected to the SBO diesel generator without the appropriate precautions and testing. The 10 CFR 50.59 screening determination associated with this revised calculation was also reviewed as described in Section
1R17 .1 of this report. The documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.10 Condensate Storage Tank Repair Coating Evaluation
a. Inspection Scope
The team reviewed modification EC 34597 that evaluated underwater coating repairs for the condensate storage tank (CST) internal surface. The evaluation was completed following the results of prior visual inspections of the internal surface of the CST.
Specifically, some of the coating degradation included weak spots, blistering, and perforations. The modification package evaluated the use of a new coating repair material that had been successfully tested with the existing CST coating material (for compatibility and acceptability).
The team reviewed the modification to verify that the design and licensing bases of the associated systems, including the HPCI and reactor core isolation cooling systems (both take suction from the CST) would not be degraded by the new coating material. The team confirmed that the new coating material met the appropriate quality standards. As the repair activity was contingent upon the results of new underwater inspections, the team reviewed the video results and associated report of the inspection. The team reviewed the CST wall thickness acceptance criteria to assess Entergys determination of acceptable degradation. The 10 CFR 50.59 screening determination associated with this modification was reviewed as described in Section 1R17.1 of this report. The documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.11 Revise the Overload Relay Heater Size in Motor Control Center Cubicle B1736 for
Drywell Unit Cooler VAC-206B1
a. Inspection Scope
The team reviewed modification EC 41711 that provided a new size overload heater in the overload relay for MCC B17, cubicle B1736. The modification increased the heater size and provided a slightly higher current range in order to provide additional margin when the drywell cooler is placed in service, while continuing to provide the necessary thermal protection for the drywell cooler motor.
The team reviewed the modification to verify that the design and licensing bases of the drywell cooling and electrical systems had not been degraded by the increased size for the overload heater. The team confirmed that the replacement heater met the appropriate quality standards. Performance issues with the previously installed overload heater were evaluated by the team by reviewing associated CRs and related evaluations. The team reviewed post-modification testing to verify proper operation and interaction with the existing system controls. The 10 CFR 50.59 screening determination associated with this modification was reviewed as described in Section 1R17.1 of this report. The documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
OTHER ACTIVITIES
4OA2 Identification and Resolution of Problems (IP 71152)
a. Inspection Scope
The team reviewed a sample of CRs associated with 10 CFR 50.59 and plant modification issues to evaluate whether Entergy was appropriately identifying, characterizing, and correcting problems associated with these areas, and whether the planned or completed corrective actions were appropriate. In addition, the team reviewed CRs written on issues identified during the inspection to verify adequate problem identification and incorporation of the issues into the corrective action system.
The CRs reviewed are listed in the Attachment.
b. Findings
No findings were identified.
4OA6 Meetings, including Exit
The team presented the inspection results to Mr. J. Dent, Site Vice President, and other members of Entergy staff at a meeting on November 21, 2013. The team returned proprietary information reviewed during the inspection and verified that this report does not contain proprietary information.
ATTACHMENT
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee Personnel
- B. Aheran, System Engineer
- A. Barrie, Design Engineer
- D. Berkland, Design Engineer
- J. Bonner, Design Engineer
- R. Byrne, Licensing
- P. Doody, Design Engineer
- J. Falconieri, Design Engineer
- P. Harizi, Design Engineer
- A. Madeiras, Design Engineer
- M. McClellan, Design Engineer
- F. McGinnis, Licensing Engineer
- R. Morris, System Engineer
- J. ODonnell, System Engineer
- S. Paranjape, Senior Staff Engineer
- G. Perry, Design Engineer
- D. Peyvan, Component Engineer
- B. Rancourt, Senior Lead Engineer
- K. Woods, Design Engineer
- J. Wytas, Design Engineer (contractor)
ITEMS OPENED, CLOSED AND DISCUSSED
None.