IR 05000272/2016007
| ML16256A310 | |
| Person / Time | |
|---|---|
| Site: | Salem  |
| Issue date: | 09/12/2016 |
| From: | Paul Krohn Engineering Region 1 Branch 2 |
| To: | Sena P Public Service Enterprise Group |
| Krohn P | |
| References | |
| IR 2016007 | |
| Download: ML16256A310 (19) | |
Text
ber 12, 2016
SUBJECT:
SALEM NUCLEAR GENERATING STATION, UNITS 1 AND 2 - EVALUATION OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT MODIFICATIONS TEAM INSPECTION REPORT 05000272/2016007 AND 05000311/2016007
Dear Mr. Sena:
On August 5, 2016, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Salem Nuclear Generating Station, Units 1 and 2. The enclosed inspection report documents the inspection results, which were discussed on August 5, 2016, with Mr. Eric Carr, Site Vice President (Acting), 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.
Based on the results of this inspection, no findings were identified.
In accordance with Title 10 of the Code of Federal Regulations (10 CFR) 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 (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/
Paul G. Krohn, Chief Engineering Branch 2 Division of Reactor Safety Docket Nos. 50-272 and 50-311 License Nos. DPR-70 and DPR-75
Enclosure:
Inspection Report 05000272/2016007 and 05000311/2016007 w/Attachment: Supplemental Information
REGION I==
Docket Nos. 50-272 and 50-311 License Nos. DPR-70 and DPR-75 Report Nos. 05000272/2016007 and 05000311/2016007 Licensee: PSEG Nuclear LLC (PSEG)
Facility: Salem Nuclear Generating Station, Units 1 and 2 Location: Hancocks Bridge, NJ Inspection Period: July 18 through August 5, 2016 Inspectors: J. Kulp, Senior Reactor Inspector, Team Leader J. Brand, Reactor Inspector C. Jones, Reactor Inspector Approved By: Paul G. Krohn, Chief Engineering Branch 2 Division of Reactor Safety Enclosure
SUMMARY OF FINDINGS
Inspection Report 05000272/2016007 and 05000311/2016007; 07/18/2016 - 08/05/2016; Salem
Nuclear Generating Station (Salem), Units 1 and 2; 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 U.S. Nuclear Regulatory Commissions (NRCs) program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, Reactor Oversight Process, Revision 6, dated July 2016.
No findings were identified.
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 (29 samples)
a. Inspection Scope
The team reviewed seven safety evaluations to evaluate 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 Salem had been required to obtain 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, technical specifications, and plant drawings to assess the adequacy of the safety evaluations. The team compared the safety evaluations 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 evaluations.
The team also reviewed 22 samples of 10 CFR 50.59 screenings for which Salem had concluded that a safety evaluation was not required to be performed. These reviews were performed to assess whether Salem's threshold for performing safety evaluations was consistent with 10 CFR 50.59. The sample included design changes, calculations, and procedure changes.
The team reviewed the safety evaluations that Salem 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 Salems administrative procedures used to control the screening, preparation, review, and approval of safety evaluations to the guidance in NEI 96-07 to evaluate whether those procedures adequately implemented the requirements of 10 CFR 50.59. The reviewed safety evaluations and screenings are listed in the
.
b. Findings
No findings were identified.
.2 Permanent Plant Modifications (10 samples)
.2.1 Unit 1 Auxiliary Building External Flood Seal Enhancement - Room 15417
a. Inspection Scope
The team reviewed modification design change package (DCP) 80017999 which installed internal conduit seals to control minor ground water leakage into the 84 elevation of the Unit 1 Auxiliary Building through three conduits located in the containment spray pump area (Room 15417). The three conduits are 3 diameter conduits and contain the 4160 V power cables to safety-related service water pumps 15 and 16, and a 460 V power cable to a non-vital motor control center at the Salem service water intake structure.
The team reviewed the modification to determine if the design, licensing bases, and performance capability of the service water pumps and the containment spray pump area had been degraded by the modification. The team reviewed calculations, licensee design documents, submerged cables program test reports, and associated maintenance work orders to determine if the changes were appropriately implemented.
The team also performed independent field walkdowns and inspections of all three flood seals to determine if the seals configuration was in accordance with the installation requirements and to verify the condition of the seals. Additionally, the team interviewed design and system engineers to determine if the changes met design and licensing requirements. Finally, the 10 CFR 50.59 screening 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.
.2.2 Increase Unit 1 and Unit 2 Positive Displacement Pump Discharge Relief Valve CV141
Set Pressure to 2825 PSIG
a. Inspection Scope
The team reviewed modification DCP 80099537 which modified the pressure relief set point on the Unit 1 and Unit 2 charging system positive displacement pumps from 2735 psig to 2825 psig. The relief valves are in the chemical and volume control system (CVCS) and provide overpressure protection for the 13 and 23 charging pumps. This modification was implemented to address repeated relief valve failures due to valve wear caused by the lift set pressure being set close to the maximum operating pressure of the pumps.
The team reviewed the modification to determine if the design, licensing bases, and performance capability of the Unit 1 and Unit 2 charging pumps, the associated piping, and the CVCS had been degraded by the modification. The team reviewed calculations, licensee design documents, test reports, and associated maintenance work orders to determine if the changes were appropriately implemented. A review of corrective action program reports was performed to evaluate whether there were any reliability or performance issues associated with the Unit 1 and 2 CVCS. Finally, the team interviewed design and system engineers to determine if the changes met design and licensing requirements. The 10 CFR 50.59 screening associated with this modification was 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 Salem Motor-Operated Valve Motor Replacement - 21SJ44 and 24SJ54
a. Inspection Scope
The team reviewed modification DCP 80102513 which replaced magnesium rotor motors with aluminum rotor motors in the Limitorque actuators for motor-operated valves 21SJ44 and 24SJ54. The two valves are the 21 residual heat removal (RHR) pump containment sump suction isolation valve (21SJ44) and the safety injection accumulator isolation valve (24SJ54). These replacements are part of a systematic industry effort to replace all magnesium rotors due to rotor corrosion on the collector ring and end ring of magnesium rotors. The team focused the inspection on the penetration seals associated with this modification, and their capability to perform their safety-related function of a flood and fire barrier. The motor changes also required changing the thermal overload heater at the respective motor control centers.
The team reviewed the modification to determine if the design, licensing bases, and performance capability of the RHR system and the safety injection accumulator were not degraded by the modification. The team reviewed calculations, design documents, test reports, and associated maintenance work orders to determine if the changes were appropriately implemented. A review of corrective action program notifications was also performed to evaluate whether there were any reliability or performance issues associated with the RHR system and the safety injection accumulator. Additionally, the team interviewed the Salem motor-operated valve program owner, and design and system engineers to determine if the changes met design and licensing requirements.
Finally, the 10 CFR 50.59 screening associated with this modification 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 Unit 1 and Unit 2 Component Cooling Heat Exchanger and Diesel Cooling Controller
Replacement
a. Inspection Scope
The team reviewed modification DCP 80111212 which replaced existing Moore pneumatic cascading temperature and flow controllers and differential pressure controllers with pneumatic Foxboro controllers. The cascaded temperature and flow controllers were used to control service water flow through the 11, 21, and 22 component cooling water heat exchangers (CCHXs) and the differential pressure controllers were used to control service water flow through the 1A, 1B, 1C, 2A, 2B, and 2C emergency diesel generator (EDG) jacket water lubricating oil coolers. This modification was implemented because the existing Moore controllers were obsolete and failure of the controller could affect the operability of several safety-related components.
The team reviewed the modification to determine if the design, licensing bases, and performance capability of the CCHX and EDGs had been degraded by the modification.
The team reviewed calculations, licensee design documents, test reports, and associated maintenance work orders to determine if the changes were appropriately implemented. A review of corrective action program notifications was also performed to evaluate whether there were any reliability or performance issues associated with the component cooling water system and associated CCHXs and the EDGs following the modification. Finally, the team interviewed design and system engineers to determine if the changes met design and licensing requirements. Additionally, the 10 CFR 50.59 screening associated with this modification was reviewed as described in section 1R17.1 of this report. Documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.5 Advanced Digital Feed Control System
a. Inspection Scope
The team reviewed modification DCP 80104782 that replaced an obsolete digital feedwater control system with a new advanced digital feedwater control system based on the Emerson Ovation distributed control system platform. The modification scope included installation of new processor racks, cabinets, work stations, and control room hardware for control and monitoring of steam generator pump speed, feedwater flow regulating valve position, and operation of steam system atmospheric relief valves.
Software changes included modifications to the gains for steam generator level control, the control algorithm for steam flow - feed flow mismatch, and the basis and method for feed pump control.
The team assessed whether the modification was consistent with the design and licensing bases and applicable system requirements. The team conducted interviews with the responsible system engineer, operations personnel, and training personnel.
Walkdowns were performed to examine the hardware installations in the control room, the control equipment room, and the training simulator. The team also reviewed results from post-modification testing of the software and affected components to verify testing demonstrated proper control and operation in accordance with defined acceptance criteria. The team also evaluated whether affected plant procedures, training, calculations, and drawings were properly updated. Finally, 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.
.2.6 Salem Unit 2 FLEX Electrical Connections
a. Inspection Scope
The team reviewed modification DCP 80110421 that installed new electrical connection points and associated distribution systems in order to implement diverse and flexible coping strategies (FLEX) for beyond-design-basis external events. The FLEX modifications were designed to maintain the station in a safe condition following an extended loss of alternating current (AC) power concurrent with a loss of normal access to the ultimate heat sink. The DCP provided electrical power modifications to maintain core cooling and heat removal, maintain reactor coolant system inventory control, maintain primary containment, and maintain spent fuel pool cooling during an extended loss of AC power event. Added items included a combination of permanently installed station power distribution equipment, permanently installed FLEX power distribution equipment, and specialized portable cables used to distribute power from the FLEX generators.
The team assessed whether the modification was consistent with the design and licensing bases and operational requirements. The team conducted interviews with responsible engineers and operations personnel, and walked down hardware installations in the auxiliary building, turbine building, and the outdoor Canyon area.
The team also evaluated whether affected plant procedures, calculations, and drawings were properly updated. Finally, 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 Changing Diesel Generator Area CARDOX from Auto to Manual
a. Inspection Scope
The team reviewed modification DCP 80115237 that reconfigured the actuation scheme for the carbon dioxide (CO2) fire suppression systems in the diesel generator areas. The change removed the existing automatic injection of CO2 in response to sensed fires and maintained the capability for manual system actuation. The modification addressed a concern that when in auto mode, the system represented a life safety hazard for people in the areas if the system were to inadvertently discharge via its automatic function.
Salem engineers determined the removal of the automatic system actuation conformed to applicable fire system standards and did not present an adverse impact to the capability to achieve safe plant shutdown.
The team assessed whether the modification was consistent with the design and licensing bases and fire protection program requirements. The team conducted interviews with a responsible engineer and walked down hardware installations in the diesel generator area vestibule, diesel generator local control rooms, and vicinity of the oil storage and transfer pumps. The team reviewed results from post-modification testing of the modified system to verify required functions performed correctly in response to manual actuations. The team also evaluated whether affected plant procedures, calculations, drawings, and license documents were properly updated.
Finally, 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.
.2.8 Pressurizer Heaters Power Supply Upgrade
a. Inspection Scope
The team reviewed modification DCP 80095831 that addressed an issue with nuisance breaker trips in circuits providing electrical power to the pressurizer heaters. The existing molded case circuit breakers were drawing more than 80 percent of rating and the panel ventilation system did not have sufficient capacity to prevent overheating problems. In addition, the design of components in the pressurizer heater panel required strengthening to provide adequate short circuit withstand capability for analyzed peak currents. The modification replaced the breakers with fused disconnect switches contained in a larger panel that was designed to alleviate buildup of excessive heat.
Internal panel components were braced to withstand the effects of analyzed fault currents.
The team assessed whether the modification was consistent with the design and licensing bases and operational requirements. The team conducted interviews with the responsible engineers and walked down hardware installations in the electrical penetration room. The team also evaluated whether affected plant procedures, calculations, and drawings were properly updated. Results of post-modification monitoring and evaluations were evaluated to verify the design intent was met. Finally, 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 Salem Unit 2 FLEX Mechanical Modifications
a. Inspection Scope
The team reviewed modification DCP 80110419 that installed new mechanical connections for portable pumps in existing safety system piping in order to implement FLEX for beyond-design-basis external events. The FLEX modifications were designed to maintain the station in a safe condition following an extended loss of AC power concurrent with a loss of normal access to the ultimate heat sink. The DCP installed additional valves and piping modifications so that portable pumps could be rapidly installed in existing safety systems to maintain core cooling and heat removal capabilities, reactor coolant system inventory control capabilities, and spent fuel pool cooling during an extended loss of AC power event.
The team assessed whether the modification was consistent with the design and licensing bases and operational requirements. The team conducted interviews with responsible engineers and operations personnel, and walked down hardware installations in the auxiliary building and the turbine building. The team also evaluated whether affected plant procedures, calculations, and drawings were properly updated.
Drawings, procedures, and training lesson plans were reviewed to verify that they were properly updated to reflect the post-modification configuration. The team reviewed the completed American Society of Mechanical Engineers non-destructive testing performed following installation of the new piping in the existing systems.
Finally, 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.
.2.10 Revise Unit 1 Turbine Overspeed Protection Setpoint
a. Inspection Scope
The team reviewed DCP 80106289 which raised the digital electro-hydraulic control turbine overspeed protection setpoint from 103 percent to 108 percent in order to lower the systems sensitivity to spurious overspeed trips. The DCP also implemented a Woodward service bulletin to install jumper wires across all unused speed input terminals to eliminate the possibility of inadvertent turbine trips due to stray electromagnetic signals.
The team assessed whether the modification was consistent with the design and licensing bases and operational requirements. The team assessed Salem's technical evaluations and design details to evaluate whether the digital electro-hydraulic control system would function to protect the turbine as designed and as described in the UFSAR. The team reviewed industry studies and standards to assess whether the assumptions made in those studies regarding system testing and periodic non-destructive evaluation of the turbine assembly were addressed during the generation of the DCP. Drawings, procedures, and training lesson plans were reviewed to verify that they were properly updated to reflect the post-modification configuration. Additionally, 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 notifications associated with 10 CFR 50.59 and plant modification issues to evaluate whether PSEG 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 notifications written on issues identified during the inspection to verify adequate problem identification and incorporation of the issues into the corrective action system. The notifications reviewed are listed in the Attachment.
b. Findings
No findings were identified.
4OA6 Meetings, including Exit
The team presented the preliminary inspection results to Mr. Eric Carr, Site Vice President (Acting), and other members of Salem staff at a meeting on August 5, 2016.
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
- E. Carr, Site Vice President (Acting)
- K. Grover, Plant Manager
- J. Mallon, Site Compliance Director
- S. Marcos, Design Engineering Director
- M. Adair, Fire Protection
- R. Andersen, Equipment Operator
- R. Cordrey, Operations Support Manager
- H. Flores, Electrical Design Engineer
- E. Gallagher, Instructor
- K. Moore, Instructor
- R. Page, System Engineer
- D. Pino, Electrical Design Engineer
- D. Shaver, Training Coordinator
- M. Adair, Fire Protection Program Engineer
- S. Boesch, System Engineer
- E. Castillo, Design Engineer
- E. Ciemiewicz, Motor Operated Valves Engineer
- B. Cordre, Operations Support
- R. Denardi, Motor Operated Valves Engineer
- B. Down, Senior Design Engineer
- J. Dugan, Engineering Response Team Design Engineer
- M. Halterken, Maintenance Service Superintendent
- B. Henriksen, Engineering Response Team Branch Manager
- K. King, Senior Design Engineer
- D. Kolasinski, System Engineer
- J. Leopardi, Site Facilities Manager
- G. Morrison, Principal Design Engineer
LIST OF ITEMS
OPENED, CLOSED AND DISCUSSED
None