IR 05000336/2008004
| ML083170068 | |
| Person / Time | |
|---|---|
| Site: | Millstone  |
| Issue date: | 11/10/2008 |
| From: | Diane Jackson NRC/RGN-I/DRP/PB5 |
| To: | Christian D Dominion Resources |
| jackson d 610-337-5306 | |
| References | |
| IR-08-004 | |
| Download: ML083170068 (45) | |
Text
SUBJECT:
MILLSTONE POWER STATION - NRC INTEGRATED INSPECTION REPORT 05000336/2008004 AND 05000423/2008004
Dear Mr. Christian:
On September 30, 2008, the U.S. Nuclear Regulatory Commission (NRC) completed an inspection at your Millstone Power Station Unit 2 and Unit 3. The enclosed inspection report documents the inspection results, which were discussed on October 8, 2008, with Mr. A.
J. Jordan, 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. The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.
This report documents two self-revealing findings of very low safety significance (Green). One of these findings was determined to be a violation of NRC requirements. However, because of its very low safety significance and because the finding has been entered into your corrective action program, the NRC is treating this finding as a non-cited violation (NCV) consistent with Section VI.A.1 of the NRC 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 Nuclear Regulatory Commission, ATTN.: Document Control Desk, Washington DC 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement; and the NRC Senior Resident Inspector at Millstone.
In accordance with Title 10 of the Code of Federal Regulations (CFR) Part 2.390 of the NRCs 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 NRCs document 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/ Original Signed By:
Donald E. Jackson, Chief Projects Branch 5 Division of Reactor Projects
Docket Nos. 50-336, 50-423 License Nos. DPR-65, NPF-49 Enclosure:
Inspection Report No. 05000336/2008004 and 05000423/2008004
w/ Attachment A: Supplemental Information
Attachment B: TI 172 Documentation Questions for Millstone Unit 2
cc w/encl:
A. Jordan, Site Vice President, Millstone Station C. Funderburk, Director, Nuclear Licensing and Operations Support W. Bartron, Supervisor, Station Licensing J. Spence, Manager Nuclear Training L. Cuoco, Senior Counsel C. Brinkman, Manager, Washington Nuclear Operations J. Roy, Director of Operations, Massachusetts Municipal Wholesale Electric Company First Selectmen, Town of Waterford B. Sheehan, Co-Chair, NEAC E. Woollacott, Co-Chair, NEAC E. Wilds, Jr., Ph.D, Director, State of Connecticut SLO Designee J. Buckingham, Department of Public Utility Control C. Meek-Gallagher, Commissioner, Suffolk County, Department of Environment and Energy V. Minei, P.E., Director, Suffolk County Health Department, Division of Environmental Quality R. Shadis, New England Coalition Staff S. Comley, We The People D. Katz, Citizens Awareness Network (CAN)
R. Bassilakis, CAN J. M. Block, Attorney, CAN P. Eddy, Electric Division, Department of Public Service, State of New York P. Tonko, President and CEO, New York State Energy Research and Development Authority J. Spath, SLO Designee, New York State Energy Research and Development Authority N. Burton, Esq.
SUMMARY OF FINDINGS
IR 05000336/2008-004, 05000423/2008-004; 07/01/2008 - 09/30/2008; Millstone Power Station
Unit 2 and Unit 3; Problem Identification and Resolution.
The report covered a three-month period of inspection by resident and region-based inspectors.
Two Green findings were identified, one of which was determined to be a non-cited violation (NCV). The significance of most findings is indicated by their color (Green, White, Yellow, Red)using Inspection Manual Chapter (IMC) 0609, Significance Determination Process. Findings for which the significance determination process (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.
NRC-Identified and Self-Revealing Findings
Cornerstone: Initiating Events
- Green.
A self-revealing finding of very low safety significance (Green) was identified for Dominions failure to identify the correct internal trim package (cage) for the Millstone Unit 2 feedwater heater level control valves (2-HD-103A/B). Specifically, on multiple occasions, Dominion personnel had the opportunity to initiate a condition report to document discrepancies associated with cage assemblies. Most recently, the wrong cage was installed in 2-HD-103A, which resulted in level oscillations in the 2A feedwater heater, necessitating a manual reactor trip. Dominion entered this issue into their corrective action program (CR-08-07451) and installed the correct internal trim package in valve 2-HD-103A.
This finding was more than minor because it was associated with the Human Performance Attribute of the Initiating Events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. The inspectors conducted a Phase 1 screening, in accordance with IMC 0609, Significance Determination Process, and determined that the finding was of very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available. The inspectors determined that this finding had a cross cutting aspect in the area of Problem Identification and Resolution,
Corrective Action Program, because Dominion did not identify the issue completely, accurately, and in a timely manner. P.1(a) (Section 40A3.1)
- Green.
A self-revealing, Green, non-cited violation (NCV) of 10 CFR 50, Appendix B,
Criterion XVI, Corrective Action, was identified for Dominions failure to take effective corrective actions to prevent lifting of a steam generator safety valve following a simultaneous reactor and turbine trip from full power at Unit 2, as described in the Unit 2 Final Safety Analysis Report. Specifically, a momentary power loss to the VR-11 and VR-21 120V power supplies caused a delay in the generation of the quick open signal to the condenser steam dump valves and atmospheric dump valves, resulting in the lifting of the safety valve. Dominion entered this issue into their corrective action program (CR-08-07476) and changed the power supply to the quick open signal inputs to the steam dumps and atmospheric dump valves to a vital power supply.
This finding was more than minor because it affected the Equipment Performance Attribute of the Initiating Events cornerstone and affected the cornerstone objective to limit the likelihood of those events that upset plant stability. The inspectors conducted a Phase 1 screening, in accordance with IMC 0609, Significance Determination Process and determined that this finding was of very low safety significance (Green).
Specifically, the finding did not contribute to the likelihood of a primary loss of coolant accident, did not contribute to both the likelihood of a reactor trip and the unavailability of mitigating equipment, and did not increase the likelihood of a fire or internal/external flood. The inspectors determined that this finding had a cross cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not take appropriate corrective action to address the unnecessary lifting of the safety valve in a timely manner, commensurate with its safety significance and complexity. P.1(d) (Section 40A3.2)
Licensee-Identified Violations
None.
REPORT DETAILS
Summary of Plant Status
Units 2 & 3 operated at or near 100 percent power throughout the inspection period with the following exception. Unit 2 started the inspection period in mode 3 following the June 28, 2008, reactor trip (See Section 4OA3). Unit 2 returned to 100 percent power on July 2,
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection
Impending Adverse Weather Conditions Inspection
a.
Inspection Scope (1 Sample)
The inspectors reviewed the sites readiness for impending adverse weather conditions from tropical storm Gustav, specifically high winds and rain, to determine if they were taking adequate precautions in accordance with Dominion=s procedures. The inspectors reviewed applicable Dominion procedures, walked down the intake structures, fire pump house, and site flood protection barriers to verify that flood protection equipment and structures were being maintained. The inspectors walked down the yard areas to verify that storm drains were clear and that materials were properly secured for the impending severe weather. The inspectors also interviewed shift managers, security, and maintenance personnel to verify that the departments were implementing severe weather preparations and to discuss potential issues identified during the walkdowns.
Documents reviewed during the inspection are listed in Attachment A.
b. Findings
No findings of significance were identified.
1R04 Equipment Alignment
.1 Partial System Walkdowns
a.
Inspection Scope (2 Samples)
The inspectors performed two partial system walkdowns during this inspection period.
The inspectors conducted a walkdown of each system to determine if the critical portions of the selected systems were aligned, in accordance with the procedures, and to identify any discrepancies that may have had an effect on operability. The walkdowns included selected switch and valve position checks, and verification of electrical power to critical components. Finally, the inspectors evaluated elements such as material condition, housekeeping, and component labeling. Documents reviewed during the inspection are listed in Attachment A. The following systems were reviewed based on their risk significance for the given plant configuration:
Unit 2
- B Emergency Diesel Generator (EDG) while the A EDG was out-of-service (OOS) for scheduled maintenance.
Unit 3
- B Train of Component Cooling Water System while the A Reactor Plant Component Cooling Water (RPCCW) Heat Exchanger (HX) was OOS for cleaning.
b. Findings
No findings of significance were identified.
.2 Complete System Walkdown
a.
Inspection Scope (1 Sample)
The inspectors completed a detailed review of the alignment and condition of the Unit 2 safety-related 125 VDC system. The inspectors conducted a walkdown of the system to determine whether critical portions, such as breakers and switches, were aligned in accordance with procedures and to identify any discrepancies that may have had an adverse effect on operability.
The inspectors also conducted a review of outstanding maintenance work orders to determine if the deficiencies significantly affected the system function. In addition, the inspectors reviewed the system health report and Condition Report (CR) database to determine whether equipment problems were being identified and appropriately resolved. Documents reviewed during the inspection are listed in Attachment A.
b. Findings
No findings of significance were identified.
1R05 Fire Protection
Annual Fire Drill Observation (71111.05A)
a.
Inspection Scope (1 Sample)
The inspectors observed personnel performance during a fire brigade drill on September 11, 2008, to evaluate the readiness of station personnel to fight fires. The drill simulated a fire in the Unit 3 East Electrical Room, in Battery Charger 3BYS*Charger 3. The inspectors observed the fire brigade members use of protective clothing, turnout gear, and self-contained breathing apparatus when entering the fire area. The inspectors also observed the fire fighting equipment brought to the fire scene to evaluate whether sufficient equipment was available to effectively control and extinguish the simulated fire. The inspectors evaluated whether the permanent plant fire hose lines were capable of reaching the fire area and whether hose usage was adequately simulated. The inspectors observed the fire fighting directions and communications between fire brigade members. The inspectors also evaluated whether the pre-planned drill scenario was followed and observed the post drill critique to evaluate if the drill objectives were satisfied and that any drill weaknesses were discussed. Documents reviewed during the inspection are listed in Attachment A.
b. Findings
No findings of significance were identified.
1R06 Flood Protection Measures
Internal Flooding Inspection
a.
Inspection Scope (1 Sample)
The inspectors reviewed the flood protection measures for equipment in the Unit 2 A Engineered Safety Feature (ESF) Room. The inspectors evaluated Dominions protection of safety-related systems from internal flooding conditions. The inspectors performed a walkdown of the area, interviewed the system engineer, reviewed the internal flooding evaluation and calculation, and verified that preventive maintenance (PM) was being performed on critical flood mitigation equipment to ensure that as-found equipment and conditions remained consistent with those indicated in the design basis and flooding evaluation documents. Documents reviewed during the inspection are listed in Attachment A.
b. Findings
No findings of significance were identified.
1R07 Heat Sink Performance
a.
Inspection Scope (1 Sample)
The inspectors observed the as-found condition of the Unit 2 reactor building component cooling water (RBCCW) HX after it was opened to verify that any adverse fouling concerns were appropriately addressed. The inspectors reviewed the results against the acceptance criteria in the procedure to determine whether all acceptance criteria had been satisfied. The inspectors also reviewed the Updated Final Safety Analysis Report (UFSAR) to ensure that HX inspection results were consistent with the design basis. Documents reviewed during the inspection are listed in Attachment A.
b. Findings
No findings of significance were identified.
1R11 Licensed Operator Requalification Program
Resident Inspector Quarterly Review (71111.11Q)
a.
Inspection Scope (2 Samples)
The inspectors observed simulator-based licensed operator requalification training for Unit 2 on July 23, 2008, and for Unit 3 on September 10, 2008. The inspectors evaluated crew performance in the areas of clarity and formality of communications; ability to take timely actions; prioritization, interpretation, and verification of alarms; procedure use; control board manipulations; oversight and direction from supervisors; and command and control. Crew performance in these areas was compared to Dominion management expectations and guidelines as presented in OP-MP-100-1000, AMillstone Operations Guidance and Reference Document.@ The inspectors compared simulator configurations with actual control board configurations. The inspectors also observed Dominion evaluators discuss identified weaknesses with the crew and/or individual crew members, as appropriate. Documents reviewed during the inspection are listed in Attachment A.
b. Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control
a.
Inspection Scope (6 Samples)
The inspectors evaluated online risk management for emergent and planned activities.
The inspectors reviewed maintenance risk evaluations, work schedules, and control room logs to determine if concurrent planned and emergent maintenance or surveillance activities adversely affected the plant risk already incurred with OOS components. The inspectors evaluated whether Dominion took the necessary steps to control work activities, minimize the probability of initiating events, and maintain the functional capability of mitigating systems. The inspectors assessed Dominion=s risk management actions during plant walkdowns. Documents reviewed during the inspection are listed in A. The inspectors reviewed the conduct and adequacy of risk assessments for the following maintenance and testing activities:
Unit 2
- Yellow risk associated with planned maintenance on B High Pressure Safety Injection (HPSI) header stop valve on July 3, 2008; and
- Troubleshooting and repair activities associated with the number 2 feedwater regulating valve (CR 08-08173) on July 21 and 22, 2008.
Unit 3
- Planned work activities associated with the replacement of the main generator voltage regulator cables on July 1, 2008;
- Yellow risk associated with an A Service Water (SW) valve stroke surveillance and an installed jumper to support Recirculation Spray System HX flush on July 9, 2008;
- Planned electrical work on the 34A 4160V breaker coincident with high trip risk switchyard work on August 13, 2008; and
- Operational decision making regarding chlorides in the jacket cooling water for the A EDG (CR 08-01209) on August 12 through 15, 2008.
b. Findings
No findings of significance were identified.
1R15 Operability Evaluations
a.
Inspection Scope (7 Samples)
The inspectors reviewed seven operability determinations (ODs). The inspectors evaluated the ODs against the guidance in NRC Regulatory Issue Summary 2005-20, Revision to Guidance Formerly Contained in NRC Generic Letter 91-18, AInformation to Licensees Regarding Two NRC Inspection Manual Sections on Resolution of Degraded and Nonconforming Conditions and on Operability.@ The inspectors discussed the conditions with operators, system engineers, and design engineers, as necessary.
Documents reviewed during the inspection are listed in Attachment A. The inspectors reviewed the adequacy of the following evaluations of degraded or non-conforming conditions:
Unit 2
- OD MP2-017-08, VR 11 and VR 21 reasonable assurance of safety following manual reactor trip for loss of feedwater pumps;
- OD MP2-018-08, CR-08-07767, and reasonable assurance of continued operability, for instrument air valve 27.1 failing a stroke time test;
- CR-08-106992, auxiliary feedwater (AFW) pump room failed fire barrier inspection criteria.
Unit 3
- OD MP3-005-08, main steam valve building high temperature during steam line break motor operated valve operability;
- CR-107561, potential for water relief through pressurizer safety valves from a control room fire; and
- OD MP3-000-185, through wall leak on 3/4 flange downstream of 3SWP*V729.
b. Findings
No findings of significance were identified.
1R18 Plant Modifications
a.
Inspection Scope (1 Sample)
The inspectors performed walkdowns of selected plant systems and components to assess the adequacy of the plant modification. The inspectors interviewed plant staff and reviewed applicable documents, including procedures, calculations, modification packages, engineering evaluations, drawings, corrective action program documents, the UFSAR, and Technical Specifications (TS). The inspectors reviewed the modification to determine if selected attributes (component safety classification, energy requirements supplied by supporting systems, seismic qualification, instrument setpoints, uncertainty calculations, electrical coordination, electrical loads analysis, and equipment environmental qualification) were consistent with the design and licensing bases.
Design assumptions were reviewed to determine if they were technically appropriate and consistent with the UFSAR. For this modification, the inspectors reviewed the 10 CFR 50.59 screenings or safety evaluations, as described in Section 1R02 of this report. The inspectors also verified that procedures, calculations, and the UFSAR were properly updated with revised design information. In addition, the inspectors verified that the as-built configuration was accurately reflected in the design documentation and that post-modification testing was adequate to ensure that the structures, systems, and components would function properly. Documents reviewed during the inspection are listed in Attachment A. The inspectors reviewed the following plant modification:
Unit 2
- Design Modification (DM) 2-00-0233-08, Design Change to Modify Supplies for Various Facility 1 Reactor Regulating System (RRS) Circuits.
b. Findings
No findings of significance were identified.
1R19 Post-Maintenance Testing
a.
Inspection Scope (8 Samples)
The inspectors reviewed post-maintenance test (PMT) activities to determine whether the PMTs adequately demonstrated that the safety-related function of the equipment was satisfied, given the scope of the work specified, and that operability of the system was restored. In addition, the inspectors evaluated the applicable test acceptance criteria to evaluate consistency with the associated design and licensing bases, as well as TS requirements. The inspectors also evaluated whether conditions adverse to quality were entered into the corrective action program for resolution. Documents reviewed during the inspection are listed in Attachment A. The following maintenance activities and PMTs were evaluated:
Unit 2
Operability Test, Facility 1 (Fast Start, Loaded Run), Revision 20, Change 5, following a maintenance outage on the A EDG;
- SP 2401FA, Reactor Protection System Channel A High Power Trip Test, Revision 4, Change 4, following replacement of potentiometer after a High Power trip alarm was received in the control room;
- Work Order (WO) M2-08-08177 and WO M2-08-08228, regarding replacement of the current to pneumatic controller and valve positioner for the number 2 feedwater regulating valve; and
- Operating Procedure (OP) 2346C, B EDG, Revision 1, Change 1, and SP 2613L-001, Periodic DG Slow Start Operability Test, Facility 2 (Loaded Run), Revision 3, Change 3, following HX maintenance on the B EDG.
Unit 3
- M3 08 02334, Replace Mechanical Seal on Turbine Driven Feedwater Pump (TDAFWP);
- SP 3622.3-001, TDAFWP Operational Readiness Test following repair/PM of three system valves;
- SP 3630D.2-001, Charging Cooling Pump (CCP) Operational Readiness Test - Train B, Revision 008-02; and
- SP 3444A01-001R, Steam Generator (SG) Water Level Channel 1 Calibration - Rack Instrumentation.
b. Findings
No findings of significance were identified.
1R20 Refueling and Other Outage Activities
Millstone Unit 2 Forced Outage
a.
Inspection Scope (1 Sample)
Dominion entered a forced outage following a manual trip of the reactor following a loss of both steam generator feed pumps (SGFP) on June 28, 2008 (See Section 40A3).
The inspectors evaluated the outage plan and outage activities to confirm that Dominion had appropriately considered risk, had developed risk reduction and plant configuration control methods, had adhered to licensee and TS requirements, and had identified the cause of the scram and had taken appropriate corrective action prior to the start-up.
The inspectors observed portions of the reactor start-up and power ascension activities.
The inspectors verified that conditions adverse to quality identified during the outage were entered into the corrective action program for resolution. Documents reviewed during the inspection are listed in Attachment A.
b. Findings
No findings of significance were identified.
1R22 Surveillance Testing
a.
Inspection Scope (7 Samples)
The inspectors reviewed surveillance activities to determine whether the testing adequately demonstrated equipment operational readiness and the ability to perform the intended safety-related function. The inspectors attended pre-job briefings, reviewed selected prerequisites and precautions to determine if they were met, and observed the tests to determine whether they were performed in accordance with the procedural steps. Additionally, the inspectors reviewed the applicable test acceptance criteria to evaluate consistency with associated design bases, licensing bases, and TS requirements and that the applicable acceptance criteria were satisfied. The inspectors also evaluated whether conditions adverse to quality were entered into the corrective action program for resolution. Documents reviewed during the inspection are listed in A. The following surveillance activities were evaluated:
Unit 2
- SP 2610BO-002, TDAFP and Recirculation Check Valve In-Service Testing (IST),
Revision 000-04, and SP2610BO-004, AFP Turbine Trip Throttle Valve Exercise Test, Revision 000-00, on July 24, 2008;
- SP 2624A, A EDG Train B Starting Air Valves IST, Revision 002-01, and SP 2613K-001, Periodic DG Slow Start Operability Test, Facility 1 (Loaded Run),
Revision 003-03, on August 6, 2008; and
Revision 003-03 on August 21, 2008.
Unit 3
- SP 3604A.2-001, 3CHS*P3B Operational Readiness Test (Two charging Pumps Aligned for Service), Revision 015-07, on July 24, 2008;
- SP 31005A, Moderator Temperature Coefficient and Power Coefficient Measurements, Power Exchange Method, Revision 002, on August 21, 2008; and
- SP 3441A21, PRN42 Analog Channel Op Test, Revision 003-05, on September 2, 2008.
b Findings
No findings of significance were identified.
RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS1 Access to Radiological Significant Areas (71121.01)a.
Inspection Scope (6 Samples)
During the period September 8 through 11, 2008, the inspectors conducted the following activities to verify that the licensee was properly implementing physical, administrative, and engineering controls for access to locked High Radiation Areas (HRA), and other radiological controlled areas (RCA) during normal power operations, and that workers were adhering to these controls when working in these areas. Implementation of these controls was reviewed against the criteria contained in 10 CFR 20, relevant Millstone Unit 2 and Unit 3 TS, and the licensees procedures. Documents reviewed during the inspection are listed in Attachment A. This activity represents the completion of six samples relative to this inspection area.
Plant Walk down and Radiological Work Permit (RWP) Reviews
1. The inspectors identified plant areas where radiologically significant work activities
were being performed. These activities included entering the Unit 3 containment building during power operations to perform routine maintenance activities. The inspectors reviewed the applicable RWPs for these activities to determine if the radiological controls were acceptable, attended the pre-job briefing, and reviewed the electronic dosimeter dose/dose rate alarm setpoints to determine if the setpoints were consistent with plant policy.
2. The inspectors determined that there were no current RWPs for airborne
radioactivity areas with the potential for individual worker internal exposures to exceed 50 mrem. During 2008, there were no internal dose assessments for any actual internal exposures that reached the reporting threshold of greater than 10 mrem Committed Effective Dose Equivalent (CEDE).
3. The inspectors also reviewed data contained in dose/dose rate alarm reports and
determined that no exposure exceeded site administrative, regulatory, or performance indicator criteria.
Problem Identification and Resolution
4. A review of Nuclear Oversight Department field observation reports was conducted
to determine if dose intensive tasks were being independently evaluated to assess procedural compliance and identification of problems related to implementing radiological controls.
5. CRs associated with radiation protection control access were reviewed and
discussed with the licensee staff to determine if the follow-up activities were being conducted in an effective and timely manner, commensurate with their safety significance.
High Radiation Area and Very High Radiation Area Controls
6. Procedures for controlling access to Locked High Radiation Areas (LHRA) and Very
High Radiation Areas (VHRA) were reviewed to determine if the administrative and physical controls were adequate. The inspectors attended a pre-job briefing for a Unit 3 containment building entry, a LHRA during power operations, to determine if procedural controls were implemented. These procedural controls included discussions of work site radiological conditions, roles/responsibilities of team members, emergency actions, and responses to electronic dosimeter alarms.
b. Findings
No findings of significance were identified.
2OS2 ALARA Planning and Controls (71121.02)a.
Inspection Scope (8 Samples)
During the period September 8 through 11, 2008, the inspectors conducted the following activities to verify that the licensee was properly implementing operational, engineering, and administrative controls to maintain personnel exposure as low as is reasonably achievable (ALARA) for past activities performed during the spring refueling outage (2R18) and during current power operations. Also reviewed were the preparations being made for the fall 2008 (3R12) refueling outage. Implementation of these controls was reviewed against the criteria contained in 10 CFR 20, applicable industry standards, and the licensees procedures. Documents reviewed during the inspection are listed in A. This activity represents the completion of eight samples relative to this inspection area.
Radiological Work Planning
The inspectors reviewed pertinent information regarding cumulative exposure history, current exposure trends, and ongoing activities to assess past performance during the spring refueling outage (2R18) and preparations to meet the dose challenges for the fall 2008 (3R12) outage.
1. The inspectors reviewed the exposure data for tasks performed during 2008 and
compared actual exposure with forecasted estimates. Included in this review were the tasks performed during the Unit 2 (2R18) outage, on-line tasks performed for both operating units, and dry cask loading/storage operations.
2. The inspectors evaluated the departmental interfaces between radiation protection,
operations, maintenance crafts, and engineering to identify missing ALARA program elements and interface problems. The evaluation was accomplished by reviewing recent ALARA Council meeting minutes and outage challenge board minutes, post-job ALARA Reviews, departmental dose summaries, attending 3R12 pre-outage challenge boards (for valve preventative maintenance and radiation protection technician activities), and interviewing the ALARA coordinator. The inspectors also reviewed the sites ALARA Strategic Plan that identifies areas for further improving radiological controls.
Verification of Dose Estimates
4. The inspectors reviewed the assumptions and basis for the annual 2008 site
collective exposure projections for routine power operations and 2R18 refueling outage activities, and compared the estimated dose with the actual dose received by workers. The inspectors also reviewed the dose projections for the upcoming 3R12 refueling outage.
5. The inspectors reviewed the licensees procedures associated with monitoring and
re-evaluating dose estimates when the forecasted cumulative exposure for tasks differed from the actual dose received. The inspectors reviewed the dose/dose rate alarm reports and exposure data for selected individuals to confirm that no individual exposure exceeded the regulatory limit or met the performance indicator reporting guideline.
Jobs-In-Progress
6. The inspectors reviewed the ongoing radiation work permits, attended a pre-job
briefing for a Unit 3 containment building entry and attended a site morning plant status/work planning meeting to determine if radiological controls were clearly communicated to affected departments.
7. The inspectors reviewed 3R12 ALARA Reviews/Radiation Work Permits for dose
intensive activities that are expected to exceed five person-rem, including operational and radiation protection department support activities, refueling, boric acid inspection/mitigation, and SG inspections/repairs.
Problem Identification and Resolution (PI&R)
8. The inspectors reviewed elements of the licensees corrective action program
related to implementing the ALARA program to determine if problems were being entered into the program for timely resolution. Eighteen CRs related to controlling individual personnel exposure and programmatic ALARA challenges were reviewed.
b. Findings
No findings of significance were identified.
Cornerstone: Public Radiation Safety
2PS2 Radioactive Material Processing and Transportation (71122.02)a.
Inspection Scope (6 Samples)
During the period August 11 through 14, 2008, the inspectors conducted the following activities to verify that the licensee=s radioactive processing and transportation programs complied with the requirements of 10 CFR 20, 61, and 71; and Department of Transportation (DOT) regulations contained in 49 CFR 170-189. Documents reviewed during the inspection are listed in Attachment A.
Radioactive Waste System Walkdown
The inspectors walked down accessible portions of the Unit 2 and Unit 3 radioactive liquid and solid waste collection/processing systems with the cognizant system engineer. The inspectors evaluated if the systems and facilities were consistent with the descriptions contained in the UFSAR and the Process Control Program (PCP),evaluated the general material conditions of the systems and facilities, and identified any changes made to the systems. In addition, the inspectors and the supervisor of Radioactive Material Controls visually inspected the radwaste storage areas located within the site protected area, including Warehouse Number 9, the Millstone Radwaste Reduction Facility (MRRF), Condensate Polishing Facility, and outdoor staging areas.
Stored material inventories were reviewed for these areas.
The inspectors discussed with the radioactive waste systems engineer the status of non-operational abandoned/retired-in-place radioactive waste processing equipment, and the administrative and physical controls for various components in these systems.
The inspectors evaluated any recent changes made to radwaste processing systems and their potential impact on routine plant operations.
The inspectors also reviewed the current processes for transferring radioactive resin and sludge to shipping containers and subsequent resin sampling and de-watering.
Waste Characterization and Classification
The inspection included selective review of the waste characterization and the classification program for regulatory compliance, including:
- the radio-chemical analytical results for samples taken from various radioactive waste streams, including spent resins, dry active waste, and mechanical filters;
- the development of scaling factors for hard-to-detect radio-nuclides from the radio-chemical data;
- methods and practices to detect changes in waste streams; and
- characterization and classification of waste relative to 10 CFR 61.56 and to determine DOT shipment subtype per 49 CFR 173.
Shipment Preparation
The inspection included a review of radioactive waste program documents and shipment preparation procedures, and in-progress activities for regulatory compliance, including:
- review of certificates of compliance for in-use shipping casks;
- verification of appropriate NRC (or agreement state) license authorization for shipment recipients for six shipments listed in the shipping records section;
- verification that training was provided, in accordance with NRC Bulletin 79-19, and 49 CFR 172, Subpart H, to appropriate personnel directly responsible for classifying, handling, and shipping radioactive materials;
- review of the 2007 Radioactive Effluent Release Report;
- review of radiological survey data for various spent resin liners and mechanical filters;
- review of radioactive material inventories for material staged on site; and
- review of shipping logs for 2006, 2007, and 2008 (to August 11, 2008).
Shipping Records
The inspectors selected and reviewed records associated with six non-excepted shipments of radioactive materials made since the last inspection of this area. The shipments were numbers08-087, 07-005,08-002, 08-046,07-089, and 07-096. The following aspects of the radioactive waste packaging and shipping activities were reviewed for these shipments:
- implementation of applicable shipping requirements including proper completion of the uniform manifests;
- implementation of specifications in applicable certificates of compliance for the approved shipping casks including limits on package contents;
- classification of radioactive materials relative to 10 CFR 61.55 and 49 CFR 173;
- labeling of containers;
- placarding of transport vehicles;
- radiation and contamination surveys of packages;
- conduct of vehicle checks;
- providing of driver emergency instructions;
- completion of shipping papers; and
- notification of shipment arrival at the receiving site.
Problem Identification and Resolution
The inspectors reviewed seventeen CRs and two Nuclear Oversight Audit Reports (07-06 and 06-08) relating to radioactive material processing and shipment. Through this review, the inspectors assessed the licensee=s threshold for identifying problems, and the promptness and effectiveness of the resulting corrective actions. This review was conducted against the criteria contained in 10 CFR 20.1101, TS and the licensee=s procedures. Documents reviewed during the inspection are listed in Attachment A.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
[OA]
4OA1 Performance Indicator (PI) Verification
Cornerstone: Mitigating Systems
a.
Inspection Scope (10 Samples)
The inspectors reviewed Dominion submittals for the PIs listed below to verify the accuracy of the data reported during that period. The PI definitions and guidance contained in Nuclear Energy Institute (NEI) 99-02 were used to verify the basis for reporting each data element. The inspectors reviewed portions of the operations logs, monthly operating reports, and Licensee Event Reports (LERs) and discussed the methods for compiling and reporting the PIs with cognizant licensing and engineering personnel. Documents reviewed during the inspection are listed in Attachment A.
Unit 2
- Mitigating System Performance Indication (MSPI) Emergency Alternating Current (AC) Power Systems, 4th Quarter 2007 through 2nd Quarter 2008;
- MSPI Residual Heat Removal System (RHS), 4th Quarter 2007 through 2nd Quarter 2008; and
- MSPI Support Cooling Water System, 4th Quarter 2007 through 2nd Quarter 2008.
Unit 3
- MSPI Support Cooling Water System, 4th Quarter 2007 through 2nd Quarter 2008.
b. Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems
.1 Review of Items Entered into the Corrective Action Program
a. Inspection Scope
As required by Inspection Procedure 71152, "Identification and Resolution of Problems,"
and in order to help identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into Dominion's corrective action program. This was accomplished by reviewing the description of each new CR and attending daily management review committee meetings. Documents reviewed during the inspection are listed in Attachment A.
b. Findings
No findings of significance were identified.
.2 Annual Sample - Root Cause Evaluations for Unit 2 Reactor Trips
a.
Inspection Scope (1 Sample)
The inspectors assessed Dominions Root Cause Evaluations (RCE), RCE-M-08-06119, Unit 2 Trip Due to a Loss of Load, RCE M-08-06-06209 Millstone 2 Reactor Trip and Unusual Event (PU1) Following Loss of Offsite Power, and RCE Millstone 2 Feedwater Heater (FWH) Level Oscillation and Manual Reactor Trip CR-08-07451 to determine whether Dominion had adequately identified the root causes, the contributing causes, and implemented corrective actions to prevent recurrence. RCE-M-08-06119 was performed as a result of a Millstone Unit 2 trip on May 22, 2008, RCE M-08-06209 for a Millstone Unit 2 trip on May 24, 2008, and RCE M-08-07451 for a manual reactor trip on June 28, 2008. Documents reviewed during the inspection are listed in Attachment A.
b. Findings and Observations
No findings of significance were identified.
On May 22, 2008, a lightning strike caused an electrical disturbance on one of the offsite power lines. This resulted in the Unit 2 switchyard breakers opening and remaining open because of protective relaying, thus causing a unit trip. Unit 3 was unaffected by the grid disturbance.
The inspectors determined that this root cause evaluation was detailed. The identified root and contributing causes were reasonable. Dominion identified corrective actions to prevent recurrence, which appeared appropriate. The extent of condition review for Unit 2 and Unit 3 was adequate.
On May 24, 2008, during Unit 2 reactor startup, a loss of normal power event was experienced resulting in a Unit 2 trip. The loss of normal power was caused when supply breakers for 4160 volt and 6900 volt busses from the reserve station service transformer (RSST) unexpectedly opened. A reactor trip signal was initiated on reactor coolant pump (RCP) low speed and low reactor coolant flow. The Unit 2 trip resulted in the declaration of an unusual event (UE).
The inspectors determined that the root cause evaluation was detailed. The inspectors determined that the most probable cause and the contributing causes were reasonable.
Corrective actions to prevent recurrence appeared appropriate. The extent of condition review concerning Unit 3 was adequate.
On June 28, 2008, operations personnel were conducting main turbine combined intercept valve testing when the level in the #2A feedwater heater began to oscillate.
The oscillations caused reduced heater drain flow that resulted in an automatic trip of the main feedwater pumps due to low suction pressure. This caused the operators to initiate a manual reactor trip.
The inspectors determined that the root cause evaluation was detailed. The root cause evaluation of the Unit 2 trip on June 28, 2008 was reasonable and Dominion identified corrective actions to prevent recurrence.
.3 Annual Sample - Evaluation of Unit 3 Service Water Strainer Issues
a.
Inspection Scope (1 Sample)
The inspectors performed a focused review of the actions taken and planned in response to a number of Unit 3 Service Water (SW) strainer septum issues. The review included events that occurred from December 2003 to August 2008. The inspectors reviewed causal evaluations contained in the associated CRs, the maintenance rule evaluation, corrective actions taken, ongoing troubleshooting efforts, and planned corrective actions. The inspectors also interviewed personnel and performed a plant walkdown of the Unit 3 SW strainers. Documents reviewed during the inspection are listed in Attachment A.
b. Findings and Observations
No findings of significance were identified.
4OA3 Followup of Events and Notices of Enforcement Discretion
.1 Unit 2 Reactor Trip - Loss of Feedwater
a. Inspection Scope
On June 28, 2008, Unit 2 operators manually tripped the reactor, as required, following the loss of both steam generator feedwater pumps (SGFPs). The SGFPs automatically tripped on low suction pressure due to the isolation of the feedwater heaters during main turbine combined intercept valve testing. Following the reactor trip, off-site power automatically swapped from the Normal System Station Transformer (NSST) to the RSST. Operators entered EOP 2525, Standard Post Trip Actions; and transitioned to EOP 2526, Reactor Trip Recovery.
The inspectors reviewed Dominions event review team report, which determined the cause of the trip to be from loss of both SGFPs due to low suction pressure. The low suction pressure resulted from feedwater heater level oscillations during the combined intercept valve testing. The inspectors reviewed Dominions RCE report, which identified the root cause to be an incorrect internal trim package (cage) in valve 2-HD-103A, the 1A feedwater heater level control valve during the refueling outage.
Documents reviewed during the inspection are listed in Attachment A.
b. Findings
Introduction:
A self-revealing finding of very low safety significance (Green) was identified for Dominions failure to identify the correct internal trim package (cage) for the Millstone Unit 2 feedwater heater level control valves (2-HD-103A/B). Specifically, Dominion repeatedly failed to identify that the wrong internal trim package had been incorporated into Millstone documents for valves 2-HD-103A/B.
Description:
On June 28, 2008, Millstone Unit 2 was conducting SP 2651M, Combined Intermediate Valves Operability Test. Water level in feedwater heater 2A began oscillating. The amplitudes of the oscillations increased, resulting in a reduction of heater drain flow to the Steam Generator Feedwater Pumps (SGFPs) leading to a low suction pressure trip of the SGFPs. Operations personnel manually tripped the reactor in response to the loss of main feedwater.
Dominions root cause investigation determined that an incorrect type of cage had been installed in valve 2-HD-103A during the April 2008 refueling outage. Specifically, an equal percentage cage was installed instead of a linear response cage. The root cause investigation determined that the Bill of Materials (BOM) for the valve listed the incorrect style cage. The root cause investigation also determined that Dominion had several opportunities to identify the correct cage. In 2002, the pneumatic control system on the feedwater drains was replaced with a digital system. As part of the modification, the cages on valves 2-HD-103A and 2-HD-103B were changed to the linear response style.
However, it was not until the BOM Upgrade project in January 2005 that the correct cage stock code was entered into the BOM.
In May 2006, a planner ordered the wrong style cage for 2-HD-103B, even though the BOM listed the correct style cage. In November 2006, during installation for valve 2-HD-103B, maintenance identified that the equal percentage cage was the wrong part and installed the linear cage. However, maintenance did not write a CR to document that the incorrect cage was issued to the field. In February 2008, the BOM group incorrectly changed the BOM to the old style cage for valves 2-HD-103A/B, based on the 2006 work order. No CR was generated to identify what the BOM group believed was an error in the BOM. In April 2008, during installation of a new cage for valve 2-HD-103A, maintenance identified that the new cage was different from the installed cage, but installed the cage that had been issued to them. Again, no CR was written to document the discrepancy.
During the plant start-up in May 2008, system engineering identified that the valve positions for 2-HD-103A and 2-HD-103B were different for the same power level.
Investigation determined that the wrong cage was installed in 2-HD-103A and a CR was generated. However, system engineering incorrectly concluded that the valve would be able to perform its design function without affecting plant operations. This assessment was noted in the operations department logs.
Analysis:
The inspectors determined that Dominions failure to identify the correct cage for the Millstone Unit 2 feedwater heater level control valves (2-HD-103A/B), as required by Millstone procedure MP-16-MMM, Organizational Effectiveness (Corrective Action Program, Operating Experience Program, Independent Safety Engineering Function)was a performance deficiency. Specifically, on multiple occasions, Dominion personnel had the opportunity to initiate a condition report to document discrepancies associated with cage assemblies. Most recently, the wrong cage was installed in 2-HD-103A, which resulted in level oscillations in the 2A feedwater heater, necessitating a manual reactor trip. Traditional enforcement does not apply because there were no actual safety consequences, impacts on the NRCs ability to perform its regulatory function, or willful aspects to the finding.
This finding was more than minor because it was associated with the Human Performance Attribute of the Initiating Events cornerstone and affected the cornerstone objective of limiting the likelihood of those events that upset plant stability and challenge critical safety functions during power operations. Specifically, Dominion installed the wrong cage assembly in valve 2-HD-103A, ultimately resulting in a reactor trip. The inspectors conducted a Phase 1 screening, in accordance with IMC 0609, Significance Determination Process, and determined that the finding is of very low safety significance (Green) because it did not contribute to both the likelihood of a reactor trip and the likelihood that mitigation equipment or functions would not be available.
The inspectors determined that this finding had a cross cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because Dominion did not identify the issue completely, accurately, and in a timely manner. P.1(a)
Enforcement:
No violation of regulatory requirements occurred, because the feedwater heating system is not safety-related. Because this finding does not involve a violation of regulatory requirements and has very low safety significance, it is identified as a finding.
Dominion entered this issue into their corrective action program (CR-08-07451) and installed the correct cage in valve 2-HD-103A.
(FIN 05000336/2008004-01, Failure to Identify the Correct Internal Trim Package for Valve 2-HD-103A Results in Reactor Trip)
.2 Failure to Prevent the Lifting of a Unit 2 Steam Generator Safety Valve
Introduction:
A Green, self-revealing, non-cited violation (NCV) of 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, was identified for Dominions failure to take effective corrective actions to prevent lifting of a steam generator safety valve following a simultaneous reactor and turbine trip at full power at Unit 2. Specifically, a momentary power loss to the VR-11 and VR-21 120V power supplies caused a delay in the generation of the quick open signal to the condenser steam dump valves and atmospheric dump valves, resulting in the lifting of the safety valve.
Description:
On May 22, 2008, Unit 2 steam generator safety valve 2-MS-247 lifted, following a reactor trip from 100% power. The safety valve lifted because of a delayed quick open signal to the condenser steam dump valves and atmospheric dump valves; the delay was caused by a post-trip momentary power loss of the VR-11 and VR-21 non-vital 120V power supplies. Dominion initiated CR-08-06117 and identified the cause to be the design of the regulating transformers and transfer switches to VR-11 and VR-21, but no specific design deficiency was highlighted.
On May 24, 2008, during the reactor startup, Unit 2 tripped due to a failed RSST.
Dominion noted that the power supply to VR-11 cycled several times between its normal and emergency power transformers following the trip. Dominion initiated CR-08-06320 and identified that not enough information was available to determine the exact cause of the VR-11 cycling. To prevent future cycling, Dominion implemented a modification that deenergized the normal power supply to VR-11 and VR-21, forcing the use of the emergency power supply only; both the normal and emergency supplies are from non-vital sources.
On June 28, 2008, Unit 2 tripped from 100% power due to a loss of feedwater. Safety valve 2-MS-247 lifted for the same reason as on May 22nd. Unit 2 FSAR, Section 7.4.5.2, states, The total steam dump and turbine bypass is sufficient to prevent lifting of the secondary steam safety valves following a simultaneous reactor and turbine trip at full power. The inspectors determined that the actions taken by Dominion as a result of the May 22nd and May 24th, trips did not correct the problem of a safety valve lifting following reactors trips from 100% power. The cycling of a safety valve resulting from full power trips results in an increased likelihood that the valves may not reseat properly, increasing the likelihood of an initiating event.
Dominions corrective actions included two design changes (DM2-00-0233-08 and DM2-00-0234-08) that moved several important loads from VR-11 and VR-21 to battery backed vital power supplies VA30 and VA40, respectively. Specifically, the design changes moved the reactor coolant system average temperature (RCS Tavg) inputs and the condenser vacuum signals required for the quick open logic from the non-vital VR-11 and VR-21 to VA30 and VA40.
Analysis:
The inspectors determined that Dominions failure to implement adequate corrective actions to prevent the unnecessary lifting of a steam generator safety valve following a simultaneous reactor and turbine trip at full power was a performance deficiency. Traditional enforcement does not apply because there were no actual safety consequences, impacts on the NRCs ability to perform its regulatory function, or willful aspects to the violation.
The finding was more than minor because it affected the Equipment Performance Attribute of the Initiating Events cornerstone and the cornerstone objective to limit the likelihood of those events that upset plant stability. The inspectors conducted a Phase 1 screening, in accordance with IMC 0609, Significance Determination Process and determined that this finding was of very low safety significance (Green). Specifically, the finding did not contribute to the likelihood of a primary loss of coolant accident, did not contribute to both the likelihood of a reactor trip and the unavailability of mitigating equipment, and did not increase the likelihood of a fire or internal/external flood.
The inspectors determined that this finding had a cross cutting aspect in the area of Problem Identification and Resolution, Corrective Action Program, because the licensee did not take appropriate corrective action to address the unnecessary lifting of the safety valve in a timely manner, commensurate with its safety significance and complexity.
Enforcement:
10 CFR 50 Appendix B, Criterion XVI, Corrective Action, requires that measures be established to assure that conditions adverse to quality, be promptly identified and corrected. Contrary to the above, from May 22, 2008 to June 28, 2008, Dominion failed to take prompt, adequate corrective action to prevent the lifting of a steam generator safety valve following a simultaneous reactor and turbine trip at full power, as described in the Unit 2 FSAR. Because this violation was determined to be of very low safety significance and has been entered into Dominions corrective action program (CR-08-07476), it is being treated as a non-cited violation (NCV), consistent with Section VI.A.1 of the NRC Enforcement Policy.
(NCV 05000336/2008004-02, Failure to Take Adequate Corrective Action to Prevent Lifting of a Steam Generator Safety Valve)
.3 (Closed) LER 05000336/2008001-00, Failure of Eight Main Steam Safety Valves to Lift
within the Acceptance Criteria
On April 3 and 4, 2008, with the plant at 100 percent power, eight main steam safety valves (MSSVs) failed to lift within the established (+/- 3 percent) acceptance criteria during a planned test. Dominion identified that six of the failures were the result of differences between two approved 10 CFR 50, Appendix B testing methods, the other two failures were due to a corrosive oxide locking action between surface layer materials to the disc-seat interface.
The inspectors reviewed this LER and associated CRs. No findings of significance were identified. This LER is closed.
.4 (Closed) LER 05000336/2008003-00, Failed Pilot Wire Causes Reactor Trip
On May 22, 2008, with Unit 2 at 100 percent power, the main turbine tripped after a lighting strike on an offsite 345 kV power line. The main turbine trip resulted in an automatic reactor trip. Dominion performed a RCE and determined that a main turbine to switchyard pilot wire had failed prior to the lighting strike; this pre-existing condition coupled with the lighting strike caused the pilot wire relay to act as an over-current protection device, which opened switchyard breakers in a scheme to protect the main generator.
This LER was reviewed as part of the inspection in Section 4OA2.2. No findings were identified. This LER is closed.
.5 (Closed) LER 05000336/2008004-00, Reactor Trip to a Loss of Normal Power Event
On May 24, 2008, Unit 2 was in Mode 2 when an automatic reactor trip occurred following a loss of normal power (LNP) event. At the time of the LNP, a reactor startup was in progress and the reactor was critical with power below the point of adding heat.
The LNP was caused when the low-side supply breakers from the RSST to the 4160 volt and 6900 volt buses unexpectedly opened.
This LER was reviewed as part of the inspection in Section 4OA2.2. No findings were identified. This LER is closed.
.6 (Closed) LER 05000336/2008005-00, Feedwater Heater Level Oscillation and Manual
On June 28, 2008, with Unit 2 at 100% power and combined intercept valve testing in progress, operators manually tripped the reactor when both feedwater pumps tripped.
This LER was reviewed as part of the inspection associated with this event and is documented in Section 4OA3.2 of this report. This LER is closed.
4OA5 Other Activities
.1 Independent Spent Fuel Storage Installation
a. Inspection Scope
An independent spent fuel storage installation (ISFSI) inspection was conducted during the period September 8 through 11, 2008. Using Inspection Procedure 60855, the inspectors reviewed the ongoing maintenance and surveillance activities for the onsite storage of spent fuel. The ISFSI licensing basis documents and implementing procedures were reviewed as the standards for the inspection. The inspection consisted of observing the condition of the Nuclear Horizontal Modular Storage (NUHOMS)system; performing independent radiation surveys of the storage modules; examining environmental dosimeters; and review of the surveillance records, including air vent inspections and recent daily air vent outlet temperature readings.
b. Findings
No findings of significance were identified.
.2 TI 2515/172, RCS Dissimilar Metal Butt Welds
a. Inspection Scope
Temporary Instruction (TI) 2515/172 provides for confirmation that owners of pressurized-water reactors (PWR) have implemented the industry guidelines of the Materials Reliability Program-139 (MRP) regarding nondestructive examination and evaluation of certain dissimilar metal welds in reactor coolant systems containing Alloy 600/82/182. The TI requires documentation of specific questions in this inspection report. The questions and responses are included in Attachment B to this report.
In summary, Millstone Unit 3 has fourteen MRP-139 applicable Alloy 600/82/182 RCS welds. Those welds are:
- One 14 pressurizer surge line nozzle;
- One 4 pressurizer spray nozzle;
- Four 6 safety/relief nozzles (3 safety, one relief);
- Four 29 RCS hot leg (HL) reactor vessel outlet nozzles; and
- Four 27.5 RCS cold leg (CL) reactor vessel inlet nozzles.
Millstone 3 has submitted Alternative Request IR-2-39, Revision 1 (October 20, 2005),and Relief Request IR-2-47, Revision 1 (March 28, 2007), Use Of Weld Overlays As An Alternative Repair Technique and use of the Performance Demonstration Initiative (PDI)program for inspection, as alternatives to the requirements of the American Society of Mechanical Engineers Boiler and Pressure Vessel Code (ASME Code),Section XI.
These relief requests are applicable to the above welds with the exclusion of the eight RV inlet and outlet nozzles. The proposed alternatives (IR-2-39, Revision 1 and IR-2-47, Revision 1) were approved by NRC Staff on January 20, 2006 and May 3, 2007, respectively.
b. Findings
No findings of significance were identified.
.3 Implementation of Temporary Instruction (TI) 2515/176 - Emergency Diesel Generator
Technical Specification Surveillance Requirements Regarding Endurance and Margin Testing
a. Inspection Scope
The objective of TI 2515/176, Emergency Diesel Generator Technical Specification Surveillance Requirements Regarding Endurance and Margin Testing, is to gather information to assess the adequacy of nuclear power plant emergency diesel generator (EDG) endurance and margin testing as prescribed in plant-specific technical specifications (TS). The inspectors reviewed emergency diesel generator ratings, design basis event load calculations, surveillance testing requirements, and emergency diesel generator vendors specifications and gathered information in accordance with TI 2515/176.
The inspector assessment and information gathered while completing this TI was discussed with licensee personnel. This information was forwarded to the Office of Nuclear Reactor Regulation for further review and evaluation.
b. Findings
No findings of significance were identified.
4OA6 Meetings, including Exit
Exit Meeting Summary
On October 8, 2008, the resident inspectors presented the overall inspection results to Mr. A. J. Jordan, and members of his staff. The inspectors confirmed that no proprietary information was provided or examined during the inspection.
ATTACHMENT A:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
G. Auria
Nuclear Chemistry Supervisor
B. Bartron
Supervisor, Licensing
J. Cambell
Manager, Security
C. Chapin
Supervisor, Nuclear Shift Operations Unit 2
A. Chyra
Nuclear Engineer, PRA
T. Cleary
Licensing Engineer
G. Closius
Licensing Engineer
L. Crone
Supervisor, Nuclear Chemistry
C. Dempsey
Assistant Plant Manager
J. Dorosky
Health Physicist III
M. Finnegan
Supervisor, Health Physics, ISFSI
R. Griffin
Director, Nuclear Station Safety & Licensing
W. Gorman
Supervisor, Instrumentation & Control
J. Grogan
Assistant Plant Manager
C. Houska
I&C Technician
A. Jordan
Site Plant Manager
J. Kunze
Supervisor, Nuclear Operations Support
B. Krauth
Licensing, Nuclear Technology Specialist
Manager, Radiation Protection/Chemistry
J. Langan
Manager, Nuclear Oversight
P. Luckey
Manager, Emergency Preparedness
R. MacManus
Director, Engineering
M. OConnor
Manager, Engineering
A. Price
Site Vice President
M. Roche
Senior Nuclear Chemistry Technician
J. Semancik
Manager, Operations
A. Smith
System Engineer
S. Smith
Supervisor, Nuclear Shift Operations Unit 3
J. Spence
Manager, Training
S. Turowski
Supervisor, Health Physics Technical Services
C. Vournazos
IT Specialist, Meteorological Data
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened and Closed
- 05000336/2008004-01 FIN Failure to Identify the Correct Internal Trim Package for Valve 2-HD-103A Results in Reactor Trip (Section 4OA3.1)
- 05000336/2008004-02 NCV Failure to Take Adequate Corrective Action to Prevent Lifting of a Steam Generator Safety Valve (Section 4OA3.2)
Closed
- 05000336/2008001-00 LER Failure of Eight Main Steam Safety Valves To Lift within the Acceptance Criteria
- 05000336/2008003-00 LER Failed Pilot Wire Causes Reactor Trip
- 05000336/2008004-00 LER Reactor Trip due to a Loss of Normal Power Event
- 05000336/2008005-00 LER Feedwater Heater Level Oscillation and Manual Reactor Trip