IR 05000271/2009007
| ML100221157 | |
| Person / Time | |
|---|---|
| Site: | Vermont Yankee File:NorthStar Vermont Yankee icon.png |
| Issue date: | 01/22/2010 |
| From: | Doerflein L Engineering Region 1 Branch 2 |
| To: | Michael Colomb Entergy Nuclear Operations |
| References | |
| IR-09-007 | |
| Download: ML100221157 (19) | |
Text
December 10, 2009
SUBJECT:
VERMONT YANKEE NUCLEAR POWER STATION - NRC EVALUATION OF CHANGES, TESTS, AND EXPERIMENTS AND PERMANENT MODIFICATIONS TEAM INSPECTION REPORT 05000271/2009007
Dear Mr. Colomb:
On December 10, 2009, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at the Vermont Yankee Nuclear Power Station. The enclosed inspection report documents the inspection results, which were discussed on December 10,2009, with Mr. C. Wamser and other members of your staff..
The inspection examined activities conducted under your license as they relate to safety and compliance with the Commission's 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 of significance were identified.
In accordance with 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 document system (ADAMS). ADAMS is Elccessible from the NRC Web site at http://www.nrc.gov/reading-rm/adams.html(the Public Electronic Reading Room).
Sincerely
.
IRAJ Lawrence T. Doerflein, Chief Engineering Branch 2 Division of Reactor Safety Docket No.
50-271 License No.
Enclosure:
Inspection Report No. 05000271/2009007 wI Attachment: Supplemental Information
REGION I==
Docket No.:
50~271 License No.:
DPR~28 Report No.:
05000271/2009007 Licensee:
Entergy Nuclear Operations, Inc.
Facility:
Vermont Yankee Nuclear Power Station Location:
Vernon, Vermont 05354-9766 Inspection Period:
November 16 December 10, 2009 Inspectors:
F. Arner, Senior Reactor Inspector, Division of Reactor Safety (DRS).
Team Leader J. Richmond, Senior Reactor Inspector, DRS D. Orr, Senior Reactor Inspector, DRS Approved By:
Lawrence T. Doerflein, Chief Engineering Branch 2 Division of Reactor Safety i
Enclosure
SUMMARY OF FINDINGS
IR 05000271/2009007; 11/16/2009 - 12110/2009; Vermont Yankee Nuclear Power Station;
Engineering Specialist Plant Modifications Inspection.
This report covers a two week on-site inspection period of the evaluations of changes, tests, or experiments and permanent plant modifications. The inspection was conducted by three region based engineering inspectors. No findings of significance were identified. The NRC's program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649, "Reactor Oversight Process," Re\\l1sion 4, dated December 2006.
A.
NRC*ldentified and Self-Revealing Findin'9§
No findings of significance were identified.
Licensee-Identified Violations
None.
ii
.1
REPORT DETAILS
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity 1 R17 Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications (IP 71111.17)
Evaluations of Changes. Tests, or Experiments {28 samples}
a. Inspection Scope
The team reviewed two safety evaluations 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 10 CFR 50.59 requirements. In addition, the team evaluated whether Entergy had been required to obtain NRC approval prior to implementing the change. The team interviewed plant staff and reviewed supporting information including calculations, analyses, design change documentation, procedures, the UFSAR, the Technical SpeCifications (TSs),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 a sample of twenty six 10 CFR 50.59 screenings. and applicability determinations for which Entergy had concluded that no safety evaluation was required. These reviews were performed to assess whether Entergy's threshold for performing safety evaluations was consistent with 10 CFR 50.59. The sample included design changes, calculations, procedure Changes, and setpoint changes.
The team reviewed the safety evaluations that Entergy had performed during the time period covered by this inspection (Le., since the last modifications inspection). 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 Entergy's administrative procedures used to control the screening, preparation, review. and approval of safety evaluations to the guidance in NEI 96-07 to determine whether those procedures adequately implemented the requirements of 10 CFR 50.59. The reviewed safety evaluations, screenings, and applicability determinations are listed in the attachment.
b. Findings
No findings of significance were identified.
.2 Permanent Plant Modifications (10 samples)
.2.1 Service Water Pump (SWP) P-7-IC Motor Replacement
a. Inspection Scope
The team reviewed a modification (Engineering Change (EC) 13723) that replaced the
'C' service water pump (P-7-1C) motor with a refurbished and rewound motor removed from the 'B' service water pump during a previous motor replacement activity. The refurbished motor also included ventilation, lubrication, pump coupling, and motor lead modifications. Entergy implemented the modification because the existing 'C' SWP motor was original plant equipment and the replacement was intended to improve service water system reliability. The motor was rewound per Entergy specification SPEC-06-0008-V, Service Water Pump Motor, Rev. 3. The team conducted the review to verify that the design bases, licensing bases and performance capability of the 'C' SWP had not been degraded by the modification. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R17.1 of this report.
The team reviewed Final Motor Report, FMR09P0910, for the refurbished motor to ensure the installed motor maintained the intended characteristics and capabilities specified in Entergy specification SPEC-06-0008-V. The team verified that Entergy maintained proper electrical bus and breaker coordination as defined in VYS-040, Protection and Coordination of Electrical Systems, Rev. 4, and appropriately adjusted.
breaker relay setpoints based on the new motor electrical characteristics. The team verified that Entergy evaluated the impact of the new 'c' SWP motor on emergency diesel generator loading. In addition, the 'C' SWP capacity test results were reviewed to verify pump performance capability was maintained and baseline in-service test (1ST)values were properly established. The team walked down the 'c' SWP to identify abnormal conditions. The team also discussed the modification and design basis with design and system engineers to assess the adequacy of the modification. Additional documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
. 2.2 Reactor Building Closed Cooling Water (RBCCW) Pump Auto Start Time Delay Setpolnt Change
a. Inspection Scope
The team reviewed a modification (EC 3393) that reduced the time delay for auto starting RBCCW pumps when a loss-of-coolant accident (LOCA) signal is received concurrent with a loss-of-offsite power (LOOP) signal. Entergy determined in calculation VYC 2058, Drywell Cooler Response to a Simultaneous LOCA and LOOP Event, that the RBCCW
- 3 pumps must start within 73 seconds of the LOCAILOOP event to ensure containment integrity was not adversely affected and the RBCCW system would not experience water hammer. Thirteen seconds was allowed for emergency diesel generator (EDG) output breaker closure and a subsequent 60 seconds was allowed for RBCCW pump motor breaker closure. The time delay measured during historical integrated emergency core cooling system (ECCS) testing was not consistently within the required time interval.
Entergy had previously reviewed the issue within the corrective action program (CRw VTY-2004-01409 and CR-VTY-2007-02261) and determined that the actual time delay recorded during ECCS testing compared to as-left calibration was due to voltage dips experienced during the dynamic ECCS testing compared to stable voltage during calibration. The conclusion was that the reduced voltage could cause the relay to reset or the timing interval to increase. To offset this effect and ensure the RBCCW pumps start within 73 seconds of the LOCAILOOP event, this modification reduced the time delay calibration setpoint to 47 seconds {after EDG breaker closure} to ensure the RBCCW pumps would auto start within 73 seconds of the LOCAILOOP event. The team conducted the review to verify that the design bases, license basis, and performance capability of RBBCW as well as the EDG:s were not degraded by the modification. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R17.1 of this report.
The team verified that Entergy calibrated the time delay relays in accordance with vendor instructions and at the new setpoint. The team also reviewed the latest as-found setpoints during ECCS integrated testing and verified the setpo/nts were within 60 seconds after EDG beaker closure. The team verified that Entergy appropriately considered the potential to impact the transient performance of the EDG and its ability to maintain ECCS loads energized with the reduced RBCCW pump time delay set point.
The team also discussed the modification and design basis with design engineers to assess the adequacy of the modification. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identifieci.
.2.3 High Pressure Coolant Injection (HPCI) System Steam Line Temperature Isolation Time
Delay Relay Modification
a. Inspection Scope
The team reviewed a modification (EC 2012) that replaced a pneumatic style time delay relay with an electronic time delay relay. The time delay relays by design allowed the HPCI system to continue to operate for 30 minutes during a small break loss-of...coolant accident (LOCA) if the HPCI system steam leak detection was tripped by a steam leak not from the HPCI steam line. The previous pneumatic style relays would remain in one position for 18 months and would only operate during HPCI system logic testing at each refuel outage. The extended static condition of the pneumatic internal components was determined by Entergy to adversely affect the relay time delay. Entergy selected the electronic style time delay relays as a corrective action (CR-VTY-2007-1078) for long term resolution of the setpoint drift. The team conducted the review to verify that the design bases, license basis, and performance capability of the HPCI system and steam line isolation were not degraded by the modification. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R 17.1 of this report.
The team verified that Entergy reviewed all aspects of the electronic time delay relay operation and its potential impact on the HPCI system steam isolation control system.
The team verified that the time delay as Itaft calibration setpoint for the new style relays was appropriately determined in accordance with Vermont Yankee Instrument Uncertainty and Setpoint Design Guide, Rev. 6.1, and maintained technical specification (TS) setpoint requirements. The team reviewed vendor test reports and specifications, and verified the relays were appropriate for the application. The team also determined that the specifications were correctly translated into setpoint calculations. The team reviewed post-modification calibration data and ensured as-left setpoints were accurately established. The team verified that OP 4361, HPCI System Isolation AlB Logic Functional/Calibration Test, Rev. 33, was revised for EC 2012. The team performed a walkdown of HPCI system steam line isolation time delay relays and observed the seismic installation and setpoint dials to ensure consistency with the design change instructions. The team also discussed the modification and design basis with design engineers to assess the adequacy of the modification. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
. 2.4 Reactor Core Isolation Cooling (RCIC) System Steam Line Temperature Isolation Time Delay Relay Modification
a. Inspection Scope
The team reviewed a modification (EC 2470) that replaced a pneumatic style time delay relay with an electronic time delay relay. The time delay relays by design allowed RCIC to continue to operate for 30 minutes for core cooling and makeup if the RCIC steam leak detection was tripped by a steam leak not from the RCIC steam line. The previous pneumatic style relays would remain in one position for 18 months and would only operate during RCIC logic testing at each refuel outage. The extended static condition of the pneumatic internal components was determined by Entergy to adversely affect the relay time delay. Entergy selected the electronic style time delay relays as a corrective action (CR-VTY-2007-1078) for long term resolution of the setpoint drift. The team conducted the review to verify that the design bases, license basis, and performance capability of RCIC and stearn line isolation were not degraded by the modification. The 10 CFR 50.59 screening determination associated with this modification was also reviewed as described in section 1 R17.1 of this report.
Encl.osure The team verified that Entergy reviewed all aspects of the time delay relay operation and its potential impact on the RCIC steam isolation control system. The team verified that the time delay as-left calibration setpoint 'for the new style relays was appropriately determined in accordance with Vermont Yankee Instrument Uncertainty and Setpolnt Design Guide, Rev. 6.1, and maintained TS setpolnt requirements. The team reviewed vendor test reports and specifications, and verified the electronic relays were appropriate for the application and that the relay specifications were correctly translated Into setpoint calculations. The team reviewed post-modification calibration data and ensured as-left setpoints were accurately established. The team verified that OP 4368, Reactor Core Isolation Cooling (RCIC) System AlB Isolation Logic Functional/Calibration Test, Rev. 39, was appropriately revised. The team performed a walkdown of RCIC steam line isolation time delay relays and observed the seismic installation and setpolnt dials to ensure consistency with the design change instructions. The team also discussed the modification and design basis with design engineers to assess the adequacy of the modification. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
. 2.5 Emergency Diesel Generator (EDG) Fuel Oil Storage Tank (FOST) Level Requirement Changes for Ultra Low Sulfur Diesel (ULSD) Fuel Oil
a. Inspection Scope
The team reviewed a modification (EC 1888) that established a new FOST level requirement to ensure the minimum volurne of EDG fuel oil required by technical speCifications was maintained. A new FOST level requirement was necessary because Environmental Protection Agency (EPA) regulations required a shift to ULSD from low sulfur diesel and the energy content per gallon of ULSD is less. The minimum fuel oil supply requirement is based on operating one EDG at the continuous rating of 2750 kilowatts (kW) for seven days. Entergy completed this modification to ensure that the FOST low level alarm setpoint was appropriately adjusted and that the administrative limits established in plant operating procedures ensured the TS required seven day supply was maintained. The team conducted the review to verify that the design bases and performance capability of the EDGs were maintained. Entergy's process applicability determination with this modification was also reviewed as described in section 1 R17.1 of this report.
The team reviewed calculation VYC -1404, Emergency Diesel Generator Fuel Oil Usage and Storage Capacity, Rev. 2, to verify that Entergy conservatively determined the new FOST level requirement and incorporated industry data for the ULSD energy content.
The team verified that operating procedures, alarm response procedures, and operator rounds provided sufficient guidance to ensure the minimum FOST seven day supply was maintained. The team reviewed as-left calibration data and verified the new FOST low level alarm setpoint was accurately established. The team discussed the modification and design basis with design engineers and system engineers to assess the adequacy of the modification. The te.am also discussed NRC Information Notice 2006-22, New Ultra-Low-Sulfur Diesel Fuel Oil Could Adversely Impact Diesel Engine Performance, and Entergy's internal operating experience with system engineers to assess Entergy's implementation of ULSD fuel oil. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
. 2.6 'B' Emergency Diesel Generator (EOG) Service Water (SW) Piping Upgrade
a. Inspection Scope
The team reviewed a design change (EC 3138) that modified service water (SW) piping to the '8' emergency diesel generator (EDG) to determine whether it was technically adequate. The modification replaced sections of the 8-inch supply and return header pipe, removed and eliminated a SW supply pressure control valve, replaced the supply isolation gate valve with a ball valve, and replaced several carbon steel instrument isolation valves and Instrument lines with stainless steel valves and lines. The piping was replaced because of pinhole leaks due to corrosion. The team evaluated the changes to ensure that the design bases, licensing bases, and performance capability of the EDG and SW systems had not been adversely affected by the modification. In addition, the 10 CFR 50.59 applicability determination associated with this modification was reviewed, as described In section 1 R17.1 of this report.
The team assessed Entergy's actions following the pipe replacement to verify that the SW and EOG systems had been fully restored to service. The team assessed selected design inputs and attributes to ensure that they were consistent with the design and licensing bases. These design inputs and attributes included seismic and stress analysis, material susceptibility to microbiological induced corrosion (MIC), and material susceptibility to galvanic corrosion. In addition, the team reviewed associated drawings to verify that they had been properly updElted to incorporate the changes to the SW system. The team performed a walkdown of the replaced piping to identify abnormal conditions, and discussed the modification with system and design engineers to assess the adequacy of the modification. The dC1cuments reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
. 2.7 4 kV Bus 3-4 Switchgear Cubicle Remote Racking Upgrade
a. Inspection Scope
The team reviewed a design change (EC 2270) that modified the 3-4 Bus 4 kV switchgear to determine whether it was technically adequate. The modification to the switchgear cubicles allowed remote operation of the breaker racking motor with the cubicle door closed. Entergy performed the modification to improve worker safety, by allowing operators to rack-in or rack-out 4 kV breakers without being in close proximity to the breaker cubicle. The team evaluated the changes to verify that the design bases, licensing bases, and performance capability of the switchgear had not been adversely affected by the modification. In addition, the 10 CFR 50.59 applicability determination associated with this modification was reviewed, as described in section 1 R17.1 of this report.
The team assessed Entergy's actions following the switchgear modification to verify that the 4 kV system had been fully restored to service. The team assessed selected design inputs and attributes to ensure that they were consistent with the design and licensing bases. These attributes included seismic: qualification, component operating methodology, and operator training. In addition, the team reviewed associated operating procedures to ensure that they had been properly updated to incorporate the changes to the switchgear. The team performed a walkdown of the affected switchgear to identify abnormal conditions, and discLissed the modification with plant operators and design engineers to assess the adequacy of the modification. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
. 2.8 Emergency Diesel Generator Load Profile AnalYSis
a. Inspection Scope
The team reviewed an engineering chan~le (EC 10519) that revised calculation VTC-836, Diesel Generator Loading, to determine whether it was technically adequate.
Entergy revised the calculation to correct some non-conservative loading assumptions and reduce other overly conservative assumptions, to reflect a more accurate EDG loading analysis. The change also resultt9d in a revision to the EDG loading tables in the Updated Final Safety Analysis Report (UFSAR), and to operating procedure op 4126, Diesel Generator Surveillance. Tht~ team evaluated the changes to ensure that the revised calculation was consistent with assumptions in the plant's design and licensing bases. In addition, the 10 CFR 50.59 screening determination associated with the engineering change was reviewed, as; described in section 1R17.1 of this report.
The team evaluated the calculation to determine whether the assumptions were appropriate and valid, and to determine the accuracy and acceptability of the analysis.
Specifically, design inputs such as load s,equencing, motor efficiency at other than 100%
rated load, and motor starting currents WE~re reviewed to determine whether they were conservative and consistent with the design and licensing bases. In addition, the team reviewed OP-4126 to determine whether it had been properly updated to incorporate the changes to the loading analysis. The team reviewed condition reports associated with the calculation change to verify that the changes appropriately resolved the original identified issues. The team also discusse~d the calculation with system and design engineers to verify the design inputs useel in the revision were appropriate. The documents reviewed are listed in the attal~hment.
b. Findings
No findings of signi'ficance were identified.
. 2.9 125 VDC Batterv A-1 Sizing and Short Circuit Analysis
a. Inspection Scope
The team reviewed an engineering change (EC 4716) that revised calculation VYC-2153, 125 VDC Battery A-1 Electrical System Calculation, to determine whether it was technically adequate. Entergy revised the calculation to correct some non conservative 125 VDC loading assumptions used to calculate the battery cell size and the battery load profile. The change also resulted in a revision to OP-4215, Main Station Battery Performance or Service Test: Entergy determined that no 10 CFR 50.59 applicability or screening determination was required for this engineering change because the calculation revision did not result in a physical facility change or a change.
to the facility's licensing basis documents that were controlled in accordance with 10 CFR 50.59. The team evaluated the change to ensure that the revised calculation was consistent with assumptions in the plant's design and licensing bases. The team evaluated the calculation to ensure that the assumptions were appropriate and valid, and to determine the accuracy and acceptability of the analysis. Specifically, deSign inputs such as load sequencing, load variations due to high pressure coblant injection (HPCI) system operation in pressure control mode, and battery cell specific gravity operating bands were reviewed to determine whether they were conservative and consistent with the design and licensing bases. In addition, the team reviewed OP-4215 to verify that it had been properly updated to incorporate the changes to the loading analysis. The team reviewed condition reports associated with the calculation change to verify that the changes appropriately resolved the original identified issues. The team also discussed the calculation with system and design engineers to verify the design inputs used in the change were appropriate. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
. 2.10 HPCI Torus Suction Line Vent Valve
a. Inspection Scope
The team reviewed a design change (EC 9528) that installed a vent valve on the HPCI torus suction line to the pump. During ultrasonic testing as a followup to NRC Generic Letter (GL) 2008-01, Managing Gas Accumulation in Emergency Core Cooling, Decay Heat Removal, and Containment Spray Systems, Entergy discovered the presence of a gas void in the 16 inch HPCI horizontal line from the torus. The objective of this engineering change was to provide the cclpability to vent any air which could accumulate in the HPCI suction piping from the torus. The configuration of the suction piping between the torus and the suction HPCI i:solation valves has an inverted U where air pockets have accumulated. The team evaluated the change to ensure that the design bases, licensing bases, and performance capability of the HPCI system had not been adversely affected by the modification. In addition, the 10 CFR 50.59 screen associated with the modification was reviewed, as described in section 1 R17.1 of this report.
The team assessed selected design inputs and attributes to determine whether they were consistent with the design and licensing bases. These design inputs and attributes included ASME Section XI Code requirements, and program impacts such as Appendix J local leak rate testing (LLRT) for the new vent valve in the pressure boundary. The team also reviewed the post modification test plan and results from the required pressure test and seat leakage test to ensure appropriate acceptance criteria had been applied. In addition, the team reviewed associated drawings such as isometrics, and piping and instrument diagrams to verify that they had been properly updated to incorporate the changes to the HPCI system suction piping. The team performed a walkdown of the installed vent valve to ensure consistency with the design change specifications. The team also discussed the modification, including the associated design criteria, with the responsible design engineer to assess the adequacy of the modification. The documents reviewed are listed in the attachment.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
40A2 Identification and Resolution of Problems (IP 71152)
a. Inspection Scope
The team reviewed a sample of condition reports (CRs) associated with 10 CFR 50.59 and plant modification issues to determine: 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 problem into the corrective action system.
The CRS reviewed are listed in the attachment.
b. Findings
No findings of significance were identified, 40A6 Meetings. including Exit The team presented the inspection results to Mr. Chris Wamser, General Manager of Plant Operations, and other members of Entergy's staff at an exit meeting on December 10, 2009. The team verified that this repclrt does not contain proprietary information.
A-'I ATTACHMENT
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Entergy Personnel
- S. Buckley, Civil Design Engineer
- J. Devincentis, Senior Licensing Engineer
- M. Flynn, Electrical Design Engineer
- D. Mannai, Licensing Manager
- R. Meister, Licensing
- J. Mully, Diesel System engineer
- B. Pelzer, Engineering Programs
- J. Rogers, Manager, Design Engineering
LIST OF ITEMS
OPENED, CLOSED AND DISCUSSED
None.