IR 05000336/2013007
| ML13087A587 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 03/28/2013 |
| From: | Paul Krohn Engineering Region 1 Branch 2 |
| To: | Heacock D Dominion Resources |
| References | |
| IR-13-007 | |
| Download: ML13087A587 (28) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION rch 28, 2013
SUBJECT:
MILLSTONE POWER STATION - NRC EVALUATION OF CHANGES, TESTS, OR EXPERIMENTS AND PERMANENT MODIFICATIONS TEAM INSPECTION REPORT 05000336/2013007 AND 05000423/2013007
Dear Mr. Heacock:
On February 14, 2013, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Millstone Power Station, Units 2 and 3. The enclosed inspection report documents the inspection results, which were preliminarily discussed on February 14, 2013, with Mr. S. Scace, Site Vice President, 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.
This report documents one NRC-identified finding of very low safety significance (Green). This finding was determined to involve a violation of NRC requirements. However, because of the very low safety significance and because the finding was entered into your corrective action program, the NRC is treating the finding as a non-cited violation (NCV), consistent with Section 2.3.2 of the NRC's Enforcement Policy. If you contest the NCV in this report, 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, D.C. 20555-0001, with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, U.S. Nuclear Regulatory Commission, Washington, D.C.
20555-0001; and the NRC Resident Inspector at the Millstone Power Station. In addition, if you disagree with the cross-cutting aspect of the finding in this report, you should provide a response within 30 days of the date of this inspection report, with the basis for your disagreement, to the Regional Administrator, Region I, and the NRC Resident Inspector at the Millstone Power Station.
In accordance with Title 10 of the Code of Federal Regulations (10 CFR) Part 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.: 50-336, 50-423 License Nos.: DPR-65, NPF-49
Enclosure:
Inspection Report No. 05000336/2013007 and 05000423/2013007 w/Attachment: Supplemental Information
REGION I==
Docket Nos.: 50-336, 50-423 License Nos.: DPR-65, NPF-49 Report Nos.: 05000336/2013007; 05000423/2013007 Licensee: Dominion Nuclear Connecticut, Inc.
Facility: Millstone Power Station, Units 2 and 3 Location: P. O. Box 128 Waterford, CT 06385 Inspection Period: January 28 through February 14, 2013 Inspectors: M. Orr, Reactor Inspector, Division of Reactor Safety (DRS),
Team Leader F. Arner, Senior Reactor Inspector, DRS S. Pindale, Senior Reactor Inspector, DRS J. Richmond, Senior Reactor Inspector, DRS A. Bolger, Reactor Engineer, Division of Reactor Projects (in-training)
Approved By: Paul G. Krohn, Chief Engineering Branch 2 Division of Reactor Safety Enclosure
SUMMARY OF FINDINGS
IR 05000336/2013007 and 05000423/2013007; 1/28/13 - 2/14/13; Millstone Power Station,
Units 2 and 3; Evaluations of Changes, Tests, or Experiments and Permanent Plant Modifications.
This report covers a two week inspection of the evaluations of changes, tests, or experiments and permanent plant modifications. The inspection was conducted by four region based engineering inspectors. One finding of very low safety significance (Green) was identified, which was considered to be a non-cited violation. The significance of most findings is indicated by a color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609,
Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. 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.
Inspector Identified and Self-Revealing Findings
Cornerstone: Mitigating Systems
- Green.
The team identified a finding of very low safety significance involving a non-cited violation (NCV) of Title 10 of the Code of Federal Regulations (10 CFR) Part 50,
Appendix B, Criterion III, "Design Control," in that Dominion did not verify that Unit 3 safety-related motor control center (MCC) starters had adequate control voltage to operate under all design conditions. Specifically, Dominion did not use the minimum voltage that would be available at Unit 3 MCCs during the most limiting block starting of large electrical loads during a Unit 3 loss of coolant accident (LOCA) as the design input for the minimum voltage under which an MCC starter was required to operate, to ensure that the starter's contactor would close when Unit 2 off-site power is cross-tied to Unit 3.
In response, Dominion entered the issue into their corrective action program and issued an Operations Standing Order to ensure that the off-site electrical distribution system would not be placed in a configuration that would allow a lower minimum voltage than what was previously analyzed for the MCC starters until the issue was resolved.
The finding was more than minor because it was similar to Example 3.j of NRC Inspection Manual Chapter (IMC) 0612, Appendix E, "Examples of Minor Issues," because without verification that the components would operate at the lowest potential voltage possible, the team had reasonable doubt with the operability of the associated components. In addition, the finding was associated with the Design Control attribute of the Mitigating Systems cornerstone and affected the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e., core damage). Using IMC 0609, Appendix A, "The Significance Determination Process for Findings At-Power," Exhibit 2, "Mitigating Systems Screening Questions," a Region I Senior Reactor Analyst (SRA) conducted a detailed risk evaluation. Since the ability of the MCC starters to function under the worst case conditions could not be verified during the inspection period, a detailed risk evaluation was determined to be appropriate. Results of the evaluation demonstrated that the initiating event frequency was substantially below 1E-6, and therefore, the SRA concluded the finding to be of very low safety significance (Green).
ii
This finding had a cross-cutting aspect in the area of Human Performance, Decision Making, because, in the design of a Unit 3 480 volts alternating current (VAC) MCC starter modification, Dominion did not use a conservative or bounding value as a design input for the minimum voltage under which a component might be required to operate.
[IMC 0310, Aspect H.1(b) (1R17.2.1)iii
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 (28 samples)
a. Inspection Scope
The team reviewed eight 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 Title 10 of the Code of Federal Regulations (10 CFR) Part 50.59 requirements. In addition, the team evaluated whether Dominion 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 (TS), 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 10 CFR 50.59 screenings for which Dominion had concluded that a safety evaluation was not required to be performed.
These reviews were performed to assess whether Dominion'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 Dominion 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 Dominions 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 and screenings are listed in the
.
b. Findings
No findings were identified.
.2 Permanent Plant Modifications (12 samples)
.2.1 Unit 3 480VAC Motor Control Center Starter Replacement Project
a. Inspection Scope
The team reviewed modification MP3-2011-01065, "Motor Starter Replacement Project,"
which replaced 480VAC MCC Gould Series 5600 unitized starters with new starters that were specifically designed and manufactured for Dominion by Nuclear Logistics Inc.
(NLI). This facility change was performed because of limited availability of spare parts for the original Gould starters and an increase in component failure rates. This modification did not change any existing design function, and was intended to be a like-in-kind style change, with only slight adjustments to breaker and fuse coordination values.
The team reviewed the modification to determine whether the design basis, licensing basis, or performance capability of the 480VAC electrical distribution system had been degraded by the modification. The team assessed Dominion's technical evaluations and design details, including calculations, analysis, design specifications, vendor factory acceptance testing, installation specifications, and interviewed maintenance technicians and engineering personnel to determine whether the MCC starters would function in accordance with the modification's assumptions, and with design and licensing requirements. Drawings and calculations were reviewed to verify that they were properly updated to reflect the post-modification design and operation. The team also reviewed completed work orders to assess whether installation activities were performed as specified by the modification's design. The PMT results were reviewed to verify whether the acceptance criteria had been met. In addition, the team walked down selected starters, observed pre-installation testing and checks, and visually inspected a new starter on the work bench to independently evaluate material conditions and configuration control with the approved design. A review of condition reports was performed to determine whether there were any reliability or performance issues associated with 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. Documents reviewed are listed in the Attachment.
b. Findings
Introduction:
The team identified a finding of very low safety significance (Green)involving a non-cited violation of 10 CFR 50, Appendix B, Criterion III, "Design Control,"
in that Dominion did not verify that Unit 3 safety-related 480VAC MCC starters had adequate control voltage to operate under all design conditions. Specifically, Dominion did not use the minimum voltage that would be available at Unit 3 MCCs during the most limiting block starting of large electrical loads during a Unit 3 LOCA as the design input for the minimum voltage under which an MCC starter was required to operate, to ensure that the starter's contactor would close when Unit 2 off-site power is cross-tied to Unit 3.
Description:
Modification design specification SP-M3-EE-2425 required NLI to utilize starter components that were designed to operate at 85% to 110% of rated voltage. The MCC design rated voltage of 480VAC corresponded to 120VAC for control power inside each starter, based on the control power transformer rating. Therefore, the required operating voltage range for the NLI starter contactors was 102 to 132 VAC.
The modification stated that MP3-ENG-ETAP-04125E3, "Electrical Distribution System Analysis," showed that the Millstone Unit 3 degraded grid voltage of 422 volts at an MCC was the bounding value in the load flow analysis. Dominion staff stated the value of 422 volts was based on off-site power at the minimum allowable voltage limit during a Unit 3 loss of coolant accident (LOCA). As a result, Dominion calculated the minimum pick-up voltage required for contactor operation based on 422 VAC at the input to the starter's control power transformer (e.g., at the MCC) and worst case control wiring circuit resistances. Dominion's design specification required the NLI starters to pick-up and operate at minimum MCC voltage without hesitation, in order to be considered qualified.
For National Electrical Manufacturers Association (NEMA) size 1 starters, the calculated minimum pick-up voltage was 87VAC, and for NEMA size 2, 3, and 4 starters, the calculated minimum pick-up voltage was 90VAC.
NLI's design utilized Square D contactors to meet Dominion's design specification of operation at 85% to 110% of rated voltage. Square D, the original equipment manufacturer, designed and manufactured the contactors to operate at 85% to 110% of rated voltage, which was greater than the minimum available voltage during a postulated accident. To resolve this issue, Dominion required NLI to qualify the new starters and contactors by test, to satisfy the safety design needs of the electrical distribution system.
NLI determined by test that NEMA size 2, 3, and 4 starters, with contactors nominally rated 120VAC, could reliably pick-up at 90VAC. But for a size 1 starter, with a contactor rated at 120VAC, the minimum pick-up voltage was 96VAC. NLI subsequently selected a size 1 starter contactor rated at 110-115 VAC, which was able to satisfy the qualification requirement for pick-up at 87VAC. NLI performed factory acceptance testing to demonstrate that each contactor could pick-up at a coil voltage of 87VAC or 90VAC, respectively, and also verified by test that each starter operated as designed with a minimum applied voltage of 422 VAC. Dominion's post modification testing also verified that each contactor could pick-up at a minimum coil voltage of 90VAC or 87VAC, respectively.
The team reviewed MP3-ENG-ETAP-04125E3 and identified that 422 VAC was not the bounding worst case minimum voltage. Specifically, MP3-ENG-ETAP-04125E3 BE-2, "Start Large CDA Loads, MOV Inrush Load, NSST Feed, Tie 34B-24E, 345KV," showed a minimum voltage of 418 to 421 VAC at 8 different safety-related MCCs. This scenario analyzed Unit 3 electrical distribution when the Unit 2 off-site power circuit was cross-tied to the Unit 3 off-site power feed during a postulated Unit 3 LOCA, with Unit 3 off-site power at the minimum allowable value. The analysis predicted that the MCC voltages would drop to less than 422 VAC during the 2 to 5 second period of block load starting of large electrical loads (e.g., emergency core cooling systems). Dominion reviewed other electrical distribution scenarios and determined that 422 VAC was a bounding worst case value for all other analyzed conditions.
The team determined that the off-site power cross-tie was not part of either Unit's original design, and had been installed by a modification in 2000. The team noted that since 422 VAC was the bounding worst case value for all other analyzed electrical loading scenarios, that prior to the cross-tie modification, 422 VAC would have been the correct bounding value. The team concluded that, in 2000, the cross-tie modification had reduced the calculated worst case available voltage from 422 to 418 VAC, which had not been reflected in Dominion's design specifications for the Unit 3 MCC starter modification.
Therefore, the team concluded that Dominion had not used the most limiting design value to calculate the minimum control voltage under which a starter might be required to operate. As a result, the design adequacy of the NLI starters had not been verified for the worst case minimum voltage that could occur if Unit 2 off-site power was cross-tied to Unit 3, which was allowed by Dominion's operating procedures.
During the inspection period, Unit 2 off-site power was not cross-tied to Unit 3.
Dominion entered this issue into their corrective action program as condition reports (CR) 505581 and 506091, and issued Operations Standing Order SO-13-0 to not cross-tie Unit 2 off-site power with Unit 3 in the specific configuration of concern unless an operability determination was performed.
Analysis:
The team determined that the failure to use the most limiting design values in design calculations and analyses was a performance deficiency. Specifically, Dominion did not use the minimum voltage that would be available at Unit 3 MCCs during the block starting of large electrical loads under accident conditions as the design input for the minimum control voltage under which MCC starter components were required to operate, to ensure that the Unit 3 motor starter contactors would work under all design conditions.
The finding was more than minor because it was similar to Example 3.j of NRC Inspection Manual Chapter (IMC) 0612, Appendix E, "Examples of Minor Issues," which determined that calculation errors would be more than minor if, as a result of the errors, there was reasonable doubt of the operability of the component. For this issue, the team had a reasonable doubt of operability as to whether safety-related 480VAC MCC starters had adequate control voltage to operate under all design conditions. In addition, the finding was associated with the Design Control attribute of the Mitigating Systems cornerstone and affected the objective to ensure the availability, reliability, and capability of systems that respond to initiating events to prevent undesirable consequences (i.e.,
core damage). Traditional enforcement does not apply because the issue did not have any actual safety consequences or potential for impacting the NRC's regulatory function, and was not the result of any willful violation of NRC requirements.
Using IMC 0609, Appendix A, "The Significance Determination Process for Findings At-Power," Exhibit 2, "Mitigating Systems Screening Questions," a Region I Senior Reactor Analyst (SRA) conducted a detailed risk evaluation. Since the ability of the MCC starters to function under the worst case conditions could not be verified during the inspection period, a detailed risk evaluation was determined to be appropriate. Specifically, the condition warranting evaluation included a Unit 3 large break loss of coolant event coupled with a degraded electrical grid, along with the presence of Unit 2 off-site power cross-connected to Unit 3 (a configuration made possible through a year 2000 modification). Dominion stated that the off-site power cross-tie had not been utilized during the current Unit 3 operating cycle. Results of the evaluation demonstrated that the initiating event frequency was substantially below 1E-6, and therefore, the SRA concluded the finding to be of very low safety significance (Green).
This finding had a cross-cutting aspect in the area of Human Performance, Decision Making, because, in the design of a Unit 3 480VAC MCC starter modification, Dominion did not use a conservative or bounding value as a design input for the minimum voltage under which a component might be required to operate. [IMC 0310, Aspect H.1(b)
Enforcement:
10 CFR 50, Appendix B, Criterion III, "Design Control," requires, in part, that design control measures shall provide for verifying the adequacy of design, such as by design reviews, or the use of alternate or simplified calculational methods.
Contrary to the above, from 2011 until February 14, 2013, Dominion had not properly verified that Unit 3 safety-related MCC motor starters had adequate control voltage to operate under all design conditions. Specifically, Dominions MP3-ENG-ETAT-04125E3 load flow analysis had calculated a minimum voltage of 418 to 421 VAC at 8 safety-related MCCs when Unit 2 off-site power was cross-tied to Unit 3, coincident with a Unit 3 LOCA. However, Dominions safety-related modification, MP3-2011-01065, had specified a minimum operating voltage of 422 VAC for the starters. Because this violation was of very low safety significance (Green) and was entered into Dominion's corrective action program (CRs 505581 and 506091), this violation is being treated as an NCV, consistent with Section 2.3.2 of the NRC's Enforcement Policy.
(NCV 05000423/2013007-01, Failure to Verify 480VAC MCC Starters Had Adequate Control Voltage to Operate Under All Design Conditions)
.2.2 Evaluation of Unit 3 C&D Battery Charger Calibration Acceptance Criteria
a. Inspection Scope
The team reviewed engineering technical evaluation (ETE) MP-2012-1027, "C&D Battery Charger Calibration Acceptance Criteria," which assessed and approved acceptance criteria values for as-found test results for Unit 3 C&D battery chargers.
Prior to this ETE, the calibration procedure specified a desired range for various process parameters, but did not specify any acceptance criteria or limits for the as-found data.
The team reviewed the new test acceptance criteria to determine whether the design basis, licensing basis, or performance capability of a battery charger had been degraded by the change to the as-found criteria. The team assessed Dominion's technical evaluation to determine whether the battery chargers would function in accordance with design and licensing requirements, and vendor manual specifications. Maintenance procedures were reviewed to verify whether they were properly updated. A review of condition reports was performed to determine whether there were any reliability or performance issues associated with the change in test acceptance criteria. Additionally, the 10 CFR 50.59 applicability evaluation associated with this change 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 Replace Unit 2 Containment Electrical Penetration Assembly
a. Inspection Scope
The team reviewed equivalency evaluation MP2-2010-01095, "Containment Electrical Penetration Module Assembly Replacement," which replaced an existing Unit 2 containment electrical penetration assembly for the "B" reactor coolant pump motor power feed with a Conax assembly. The original penetration assembly required replacement because seal degradation had resulted in undesirable increases in local leak rate test values.
The team reviewed the facility change to determine whether the design basis, licensing basis, or performance capability of the containment system and electrical power distribution system had been degraded by the facility change. The team assessed Dominion's technical evaluations and design details, including installation specifications and post-installation testing, and interviewed engineering personnel to determine whether the containment and electrical distribution systems would function in accordance with the engineering assumptions, and with design and licensing requirements. Drawings and procedures were reviewed to verify that they had been properly updated to reflect the post-facility change design and operation. The team also reviewed completed work orders to assess whether installation activities were performed as specified by the equivalency evaluation, plant design basis, and vendor specifications. The PMT results were reviewed to verify that the acceptance criteria had been met. A review of condition reports was performed to determine whether there were any reliability or performance issues associated with the post-facility change configuration. Additionally, the equivalency evaluation associated with this facility change was reviewed as described in Section 1R17.1 of this report. Documents reviewed are listed in the attachment.
b. Findings
No findings were identified.
2.4 Critical Operator Action Time for Reactor Coolant Pump During a Small Break Loss-of Coolant Accident
a. Inspection Scope
The team reviewed Unit 3 evaluation ETE-NAF-2012-0051 that documented the basis for the time critical operator action for tripping the reactor coolant pumps (RCPs)following a postulated small break loss-of-coolant-accident (SBLOCA). The results of the evaluation provided the technical basis to correct (shorten) the time required to complete the action. Prior to this evaluation, and as per emergency procedures, the required time to trip the RCPs was 10 minutes. However, Dominions re-evaluation of vendor information concluded that the correct time was 5 minutes.
The team reviewed the evaluation to verify that the design and licensing bases had not been degraded. The team interviewed design engineers and reviewed associated evaluations and condition reports to verify that the new response time supported continued operability and event response assumptions. Operating and emergency procedures were reviewed to confirm that the appropriate changes were made. The team also reviewed the results of a recently performed SBLOCA simulator scenario, where operators were evaluated as completing this step within one minute of reaching the RCP trip criteria. In addition, 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 Service Water Flowrates/Assumptions for EN21203A/B and Technical Justification for 2-
SW-8.1A/B/C Full Flow Position
a. Inspection Scope
The team reviewed Unit 2 modification ETE-CME-2012-1010 that evaluated changes to the throttle position of service water (SW) valves 2-SW-8.1A/B/C, which are the reactor building closed cooling water heat exchanger outlet temperature control valves. This modification was performed because Dominion identified that the valves provided less than expected flow in the fully opened position due to the existence of an internally mounted deflector plate (flow limiter).
The team reviewed the modification to verify that the design and licensing bases had not been degraded by the revised valve position. The team interviewed design engineers, and reviewed associated evaluations and condition reports to verify that the new valve position supported service water system flow requirements during design basis scenarios. The team reviewed associated technical evaluations to confirm the design basis assumptions remained valid. The team reviewed the associated post-evaluation flow testing results to determine whether sufficient service water flow was provided to the reactor building closed cooling water heat exchangers during bounding conditions (e.g., considering seasonal system demands) as well as to other components cooled by the service water system. The team also conducted a walkdown of the temperature control valves to confirm the new configuration and to assess the material condition of the equipment. 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 Evaluation of Unit 2 Emergency Diesel Generator Surveillance Test Duration
a. Inspection Scope
The team reviewed engineering technical evaluation ETE-MP-2011-1137 initiated to reduce the duration of monthly surveillance tests on Unit 2 Emergency Diesel Generators (EDGs). This evaluation was performed to improve Dominions adherence in meeting new U.S. Environmental Protection Agency (EPA) annual emission standards. This ETE provided an evaluation and justification to change surveillance test operational time at rated load from 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> to 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />.
The team reviewed the evaluation to verify that the design and licensing bases had not been degraded by the revised test run time. The team interviewed system engineers and reviewed equipment history, past failure analyses, and improvements that had been incorporated to trending and monitoring. The review included vendor and industry documentation for recommended loading and durations to ensure the technical adequacy of the determination. The team reviewed the associated operating and test procedures to ensure the revised information was properly translated into those documents. The team conducted walkdowns of the Unit 2 EDGs to assess material condition of the equipment and systems. Additionally, the 10 CFR 50.59 applicability review form associated with this evaluation 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 Safety Injection Accumulator Tank Alarm Setpoint Change
a. Inspection Scope
The team reviewed Unit 3 modification MP3-2011-01204 that revised the safety injection accumulator tank level alarm setpoint to compensate for an uncertainty introduced by an incorrect bellows orientation on the accumulator level transmitters. Specifically, the reference leg had its vent mounted such that some amount of water could accumulate in a small recess in the top of the bellows, which could create an instrument error and result in the indicated level reading lower than the actual accumulator level. The results of this evaluation resulted in lowering the high level alarm from 6940 to 6920 gallons.
The team reviewed the modification to verify that the design and licensing bases had not been degraded by the revised high level setpoint. The team interviewed the design engineer and reviewed associated evaluations and condition reports to verify that the new setpoint was consistent with existing design assumptions. The team reviewed the associated operating and calibration procedures to ensure the revised information was properly translated into procedures. 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 High Pressure Safety Injection Pump - Comprehensive In-service Test Acceptance
Criteria
a. Inspection Scope
The team reviewed Unit 2 evaluation 12-ENG-04355M2 that developed the instrument loop measurement uncertainties associated with the flow and pressure instruments used for the comprehensive pump test for the high pressure safety injection (HPSI) system pumps as per the In-Service Test (IST) Program. These uncertainties were used to develop the minimum pump IST acceptance criteria in order to ensure the HPSI pumps can deliver flows within the bounds of the accident analysis assumptions.
The team reviewed the evaluation to verify that the design and licensing bases had not been degraded by the calculated acceptance criteria. The team interviewed engineers and reviewed associated evaluations and condition reports to verify that the new comprehensive test acceptance criteria were sufficient to demonstrate that the HPSI pumps can perform their intended design function of providing cooling flow to the reactor during postulated accident scenarios. The team reviewed the associated technical evaluations to confirm the design basis assumptions and IST Program requirements remained valid. The team reviewed the associated historical and recent HPSI pump flow test results in order to verify that test results supported pump operability requirements.
The team also conducted a walkdown of the HPSI system pumps to assess the material condition of the equipment. Finally, the 10 CFR 50.59 screening determination associated with this evaluation 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 Unit 3 Refueling Water Storage Tank (RWST) Level Switch Upgrade
a. Inspection Scope
The team reviewed modification MP3-2011-01008 that replaced RWST level switches 3QSS*LS54A/B/C/D and 3QSS*LS56A/B/C/D. The modification was performed to install switches which were qualified to perform their design function with less instrument inaccuracy. The 3QSS*LS54A/B/C/D (LOW-LOW) switches actuate to send a trip signal to the associated residual heat removal (RHS) pumps. The trip function is in preparation for the manual alignment of the emergency core cooling systems (ECCS) for the containment sump recirculation mode of operation. The design feature minimizes the loss of water dedicated for the Quench Spray system along with maintaining adequate net-positive-suction-head (NPSH) for the charging and safety injection pumps during the swapover to cold leg recirculation. These switches also give a permissive for the starting of the associated containment recirculation pumps. The 3QSS*LS56A/B/C/D (EMPTY) switches actuate to trip the associated Quench Spray pumps and close valves to establish containment isolation. The increased accuracy of the switches was intended to allow credit for an increased volume in the RWST that is calculated to be available between the LOW-LOW and EMPTY level setpoints.
The team reviewed the modification to verify that the design bases, licensing bases, and performance capability of the RWST level switch functions had not been degraded. The team interviewed design engineers and reviewed calculations and evaluations to determine if the capability of the higher accuracy switches met design and licensing requirements. Additionally, the team reviewed post-modification testing results and associated maintenance work orders to confirm that the modification was appropriately implemented. Finally, the team reviewed the results of the modification to determine consistency between engineering calculations, licensing bases documents, actual field installation, and test procedure acceptance criteria. The 10 CFR 50.59 screening determination associated with this modification 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.
.2.10 Procedure Revisions to Address Isolation of Charging System Emergency Recirculation
Lines
a. Inspection Scope
The team reviewed modification MP3-2011-01006 that evaluated crediting operator action to isolate the Class 4 portion of the alternate minimum flow (AMF) line as necessary following a safety injection (SI) initiation signal. The design of the charging system is that when an SI signal is initiated, motor-operated valves in the AMF line automatically open to align the path to the RWST while closing the normal minimum recirculation path. This feature is to protect against pump full shut-off head operation in response to small break loss-of-coolant accident scenarios. The evaluation of the capability of isolation of this path was in response to Dominions identification that non-safety related, Class 4 piping cannot be credited in safety analysis calculations. The assumption was that a line break in this section of piping may impact design basis analysis parameters because of assumed water loss within the RWST and less available volume. This would lead to lower post-LOCA sump level and NPSH available for sump recirculation if a break was not identified and isolated in a timely manner.
The team reviewed the revisions made to emergency operating procedures (EOPs) EOP 35 ES-1.2, Post-LOCA Cooldown and Depressurization, and E-1, Loss of Reactor or Secondary Coolant, to verify that the actions were reasonable with respect to the assumed operator response times analyzed. The team reviewed the implementation of the proposed actions to verify that the AMF lines would also be isolated prior to transferring to containment sump recirculation to prevent the release of containment water to the RWST. The team reviewed the procedure revisions to verify if any ECCS actuation logic or interlock logic would be adversely affected by the crediting of these manual operator actions. The team reviewed the change to determine that the isolation of the AMF lines would only be executed during postulated LOCA conditions when the reactor coolant system pressures were well below those corresponding to the shutoff head of the charging pumps. Finally, the team performed a walkdown of the AMF lines to evaluated the material condition of the isolation valves.
b. Findings
No findings were identified.
.2.11 Evaluation of Unit 2 Motor-Driven Auxiliary Feedwater Pump (MDAFWP) Test Flow Rate
Limitations
a. Inspection Scope
The team reviewed engineering technical evaluation, ETE-CME-2012-1032, Determine M2 MDAFWP (M2P9A, M2P9B) Maximum Flow Rate Limit For SP 2610AS &
SP2610KS, which supported increasing the maximum allowable test flowrate limits.
Dominion, within Condition Report (CR) 490868, had identified that higher pump flowrates occur when the surveillance test is performed with low steam generator pressures. Performing testing at low backpressures had created a challenge to remain below the previously evaluated 450 gallons per minute (gpm) limitation. This ETE provided an evaluation and justification to change the surveillance procedure upper flow limit to 590 gpm including a 10 minute time limit for operating a pump at a flowrate greater than 515 gpm.
The team reviewed calculation 98-CST-02644M2, MP2 CST LO-LO Alarm AFW Pump Net Positive Suction Head, Revision 0, to determine whether Dominion had adequately evaluated NPSH margin for the increased flowrate limitation established. This review was performed to determine if acceptable margins would remain for the individual MDAFWP tested at the maximum allowable flowrate. The team reviewed the brake horsepower (BHP) for the increased flowrate to ensure that the additional load on the motor would be acceptable and the estimated current drawn would remain below the overcurrent relay protection to prevent an inadvertent trip of the motor. The team also reviewed Dominions evaluation of the effect on bearing cooling with the decreased pump discharge head because the pump and motor bearing cooling is provided by diverting a small portion of the discharge from the 1st stage of the pump impeller through the bearing and back to the pump suction. The 10 CFR 50.59 applicability review form was also reviewed to ensure the technical adequacy of the determination. The documents reviewed are listed in the Attachment.
b. Findings
No findings were identified.
.2.12 Evaluation of Revised Coating Material for the Internal Coated Surfaces in the Unit 2
Condensate Storage Tank (T-40)
a. Inspection Scope
The team reviewed engineering evaluation ETE-MP-2012-1117 that was initiated to support work orders prepared to inspect and repair, as necessary, any degradation identified with the internals of the Unit 2 condensate storage tank (CST). The function of the CST is to provide water through the auxiliary feedwater system to the steam generators to remove heat if the main condenser is not available. The team reviewed the adequacy of the evaluation in determining that the proposed coating material would be suitable for the CST environment for corrosion protection. The team reviewed the chemical compatibility analysis to evaluate the suitability of the Bio-Dur 561 material for the CST internal application. This included the review of the product data sheet as well as the surface preparation and application requirement to ensure the engineering evaluation was consistent with the material product specifications.
The team reviewed the coating material evaluation to verify that the design and licensing bases for the Unit 2 CST would not be degraded by the approved proposed repair method. This review included Dominions analysis that repair to the original coated surfaces with the revised coating material would not create a potential source of debris.
The team interviewed Dominion engineers to discuss the associated evaluation and condition reports related to the Unit 2 CST. The 10 CFR 50.59 screening determination associated with this engineering evaluation 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 condition reports (CR) associated with 10 CFR 50.59 and plant modification issues to determine whether Dominion 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. S. Scace, Site Vice President, and other members of Dominion's staff at an exit meeting on February 14, 2013. The team returned the 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
- W. Adams, Plant Manager
- W. Bartron, Supervisor Nuclear Licensing
- G. Closius, Licensing Engineer
- D. Fredericks, Electrical Design Engineer
- G. Gardner, Senior Structural Design Engineer
- N. Jaycox, Sr. Mechanical Project Engineer
- R. Kasuga, System Engineer
- D. MacDonald, Design Engineering Manager
- J. Majewski, Primary Plant System Engineer
- R. Patel, Senior Electrical Design Engineer
- J. Plourde, System Engineer
- T. Rippel, Design Engineer
- J. Semancik, Director Nuclear Engineering
- E. Smith, System Engineer
- D. Tilton, Supervisor Primary Systems Engineering
NRC Personnel
- J. Ambrosini, Senior Resident Inspector - Millstone
- B. Haagensen, Resident Inspector - Millstone
- J. Krafty, Resident Inspector - Millstone
- P. Kaufman, Senior Reactor Inspector, Engineering Branch 1, DRS, Region 1
ITEMS OPENED, CLOSED AND DISCUSSED
Opened and Closed
05000423/2013007-01 NCV Failure to Verify 480VAC MCC Starters Had
Adequate Control Voltage to Operate Under All
Design Conditions (Section 1R17.2.1)