IR 05000245/2008004
Download: ML083170068
Text
November 10. 2008
Mr. David Christian Sr. Vice President and Chief Nuclear Officer Dominion Resources 5000 Dominion Boulevard Glenn Allen, VA 23060-6711
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 Commission's 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 NRC's
"Rules of Practice," a copy of this letter, its enclosure, and your response (if any) will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRC's document system (ADAMS). ADAMS is 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
.........................................................................................................3
REPORT DETAILS
.....................................................................................................................1
REACTOR SAFETY
...........................................................................................................1
1R01 Adverse Weather Protection
..........................................................................................1
1R04 Equipment Alignment
.....................................................................................................1
1R05 Fire Protection
................................................................................................................2
1R06 Flood Protection Measures
............................................................................................3
1R07 Heat Sink Performance
...............................................................................................3
1R11 Licensed Operator Requalification Program
..................................................................4
1R13 Maintenance Risk Assessments and Emergent Work Control
......................................4
1R15 Operability Evaluations
..................................................................................................5
1R18 Plant Modifications
.........................................................................................................6
1R19 Post-Maintenance Testing
.............................................................................................6
1R20 Refueling and Other Outage Activities
...........................................................................7
1R22 Surveillance Testing
.......................................................................................................8
RADIATION SAFETY
.........................................................................................................9 2OS1 Access to Radiological Significant Areas (71121.01).......................................................................9 2OS2 ALARA Planning and Controls (71121.02).....................................................................................10 2PS2 Radioactive Material Processing and Transportation (71122.02)..................................................12
a. Inspection Scope
(6 Samples)..........................................................................................12
OTHER ACTIVITIES
[OA]................................................................................................14
4OA1 Performance Indicator (PI) Verification
.............................................................................14
4OA2 Identification and Resolution of Problems
........................................................................15
4OA3 Followup of Events and Notices of Enforcement Discretion
.............................................16
4OA5 Other Activities...............................................................................................................................21 4OA6 Meetings, including Exit..................................................................................................................23 ATTACHMENT A:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
......................................................................................................1
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
..........................................................1 LIST OF ACRONYMS................................................................................................................5
Enclosure 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 NRC's program for overseeing
the safe operation of commercial nuclear power reactors is described in NUREG-1649,
"Reactor Oversight Process," Revision 4, dated December 2006.
A. NRC-Identified and Self-Revealing Findings Cornerstone: Initiating Events * Green. A self-revealing finding of very low safety significance (Green) was identified for Dominion's 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 Dominion's 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
Enclosure 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)
B. Licensee-Identified Violations None.
Enclosure 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, 2008.
1. REACTOR SAFETY Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity 1R01 Adverse Weather Protection (71111.01) Impending Adverse Weather Conditions Inspection
a. Inspection Scope (1 Sample) The inspectors reviewed the site's 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 (71111.04)
.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
Enclosure 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 (71111.04S) 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 (71111.05)
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,
Enclosure 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 (71111.06)
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 Dominion's
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 (71111.07A)
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.
Enclosure b. Findings No findings of significance were identified. 1R11 Licensed Operator Requalification Program (71111.11) 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 (71111.13) 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 Attachment 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
Enclosure * 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 (71111.15) 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; * OD MP2-020-08, through wall leak in weld on SW piping to "B" EDG flow element FE-6397; and * 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;
Enclosure * 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 (71111.18)
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 (71111.19) 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
Enclosure 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 * Surveillance Procedure (SP) 2613A-001, "Periodic Diesel Generator (DG) 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 (71111.20)
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.
Enclosure 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 (71111.22) 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 * SP 2613L, "Periodic DG Slow Start Operability Test, Facility 2 (Loaded Run)," 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 3636.7, "SW Pump 3SWP*P1D Operational Readiness Test," Revision 014-09, on August 19, 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.
Enclosure
b Findings No findings of significance were identified.
2. 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 licensee's 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.
Enclosure
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 licensee's 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.
Enclosure 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 site's 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 licensee's 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 licensee's 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.
Enclosure 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.
Enclosure 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 CR's 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.
Enclosure
b. Findings: No findings of significance were identified. 4. OTHER ACTIVITIES [OA] 4OA1 Performance Indicator (PI) Verification (71151) 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 HPSI System, 4th Quarter 2007 through 2nd Quarter 2008; * MSPI AFW System, 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 Emergency AC Power Systems, 4th Quarter 2007 through 2nd Quarter 2008; * MSPI HPSI System, 4th Quarter 2007 through 2nd Quarter 2008; * MSPI AFW System, 4th Quarter 2007 through 2nd Quarter 2008; * MSPI RHS, 4th Quarter 2007 through 2nd Quarter 2008; and * MSPI Support Cooling Water System, 4th Quarter 2007 through 2nd Quarter 2008.
b. Findings No findings of significance were identified.
Enclosure 4OA2 Identification and Resolution of Problems (71152) .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 Dominion's 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
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
Enclosure 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 (71153)
.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
Enclosure RSST. Operators entered EOP 2525, "Standard Post Trip Actions;" and transitioned to
EOP 2526, "Reactor Trip Recovery."
The inspectors reviewed Dominion's 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 Dominion's 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 Dominion's 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.
Dominion's 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.
Enclosure 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 Dominion's 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 NRC's 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 Dominion's 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
Enclosure 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.
Dominion's 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 Dominion's 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 NRC's 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
Enclosure 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 Dominion's 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.
Enclosure
.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 Reactor Trip 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 (60855) 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
Enclosure 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 vendor's specifications and gathered information in accordance with
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.
Enclosure 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 ATTACHMENT B: TI 172 DOCUMENTATION QUESTIONS FOR MILLSTONE UNIT 3
A-1Attachment 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
J. Laine, Manager, Radiation Protection/Chemistry
J. Langan Manager, Nuclear Oversight
P. Luckey Manager, Emergency Preparedness
R. MacManus Director, Engineering
M. O'Connor 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/LER-2008-001-00 LER Failure of Eight Main Steam Safety Valves To Lift within the Acceptance Criteria
- A-2
- 05000336/LER-2008-003-00 LER Failed Pilot Wire Causes Reactor Trip
- 05000336/LER-2008-004-00 LER Reactor Trip due to a Loss of Normal Power Event
- 05000336/LER-2008-005-00 LER Feedwater Heater Level Oscillation and Manual Reactor Trip
LIST OF DOCUMENTS REVIEWED
Section 1R01: Adverse Weather Protection
- AOP 2560, Storms, High Winds and Tides, Revision 010-04
- AOP 3569, Severe Weather Conditions, Revision 016-00
- COP 200.6, Storms and Other Hazardous Phenomena (Preparation and Recovery), Revision 002-01
- SP 2665, Building Flood Gate Inspections, Revision 005-01
Section 1R04: Equipment Alignment
- EOP-2540, Station Blackout Operations, Revision 011
- OP 2315, Vital 125VDC Electrical Switchgear Room Cooling System, Revision 000-06
- OP 2345CO, 125 Bolt DC Station Battery, Revision 000-00
- OP 3330A-002, RPCCW (Common), Revision 007-02
- OPS Form 3330A-1, RPCCW Main Boards, Revision 4
- OPS Form 3330A-16, RPCCW (Train B), Revision 6
- OPS Form 3330A-32, RPCCW System Electrical Lineup, Revision 6
- SP 2736G, Battery Charger Capacity Test, Revision 012-02
- SP 3630A.3-001, RPCCW Train B Valve Verification, Revision 006-01
- SP 750, Battery Inspections, Revision 002-01
- SP 760, Battery Discharge Test, Revision 001-01
Section 1R05: Fire Protection Millstone Unit 2 Fire Hazards Analysis, Revision 9
Section 1R06: Flood Protection Measures 98-ENG-02411-C2,
- MP2 Evaluation of Flooding Outside Containment, Revision 1 M2 03 09785, Disassemble and Inspect Check Valve Internals
- M2 03 09786, Disassemble and Inspect Check Valve Internals
- M2 07 00191, Inspection and Cleaning of "A" Safeguards Room Sump
- S2-EV-98-0252, Internal Flooding Effects on the "A" ESF Room, Revision 0
- TE M2-EV-98-0194, Internal Flooding Effects on the "A" ESF Room, Revision 0
- W2-517-1070-RE, MP2 Internal Flooding Evaluation, Revision 0
Section 1R07: Heat Sink Performance
- EN 31084, "Operating Strategy for Service Water System at Millstone Unit 3", Revision 007 M2-07-05118
- M2-07-08381
- M2-08-01358
Section 1R11: Licensed Operator Requalification Program Unit 2
- LORT, Operational Exercise #1 (S08401)
- A-3
Section 1R15: Operability Evaluations
- CR-08-01281
- CR-08-02062
- CR-107561
- CR 107590
- NUCENG-08-032, "Service Water System Hydraulic Analysis Associated with Flange Defect at Valve 3SWP*729" dated August 30, 2008
- Performance of MOV Stem Lubricants at Elevated Temperature
Section 1R18: Plant Modifications Unit 2
- UFSAR DCN: DM2-00-0233-08
Section 1R19: Post Maintenance Testing
- EN 31084, 3CCE*E1B SW Cooled Heat Exchangers Inspection Form, Revision 007, Performed 7/23/08 M3 03 12061, 3CHS*V048 Has Boric Acid Buildup
- M3 03 12171, Charging Pump Aux Lube Oil Pump Calibration of Timer TCH3LP02
- M3 04 10818, B charging Pump Calibration of Relays and Meters
- M3 05 07086, B Charging Pump Cooler 3 Yr PM
- M3 06 10889, 3CCE*P1B Has Minor Seal Leak
- M3 06 11656, 3CCE*AOV26B 10 Yr PM
- M3 06 11658, 3CCE*AOV30B 10 Yr PM
- M3 07 02661, TDAFWP Turb Exh Silencer Drain 10 Year PM.
- M3 07 02699, FWA*V030 Inadequate for Isolating FWA*P2
- M3 07 12118, 3CHS*V590 Has Boric Acid on Packing Gland Area
- M3 07 12119, 3CHS*V849 Has Boric Acid on Packing Gland Area
- M3 07 12561, MP3 'B' Charging Pump Auxiliary Lube Oil Pump Exhibits Elevated Vibration
- M3 07 16300, 3CCE*TV37B AOV Flowscan Test Identified High Unseating Load
- M3 08 01313, Turbine Driven SGAFW Pump Steam Isolation PM
- M3 08 04455, Small Oil Leak at Discharge Pipe of 3CHS*P7B
- M3 08 07609, D SG Narrow Range Level Spiking Low
- MP 3704A-303, Preventive Maintenance Technique for Terry Turbine Trip Throttle Valve Linkage, Revision 002-01, Performed 7/21/08.
- SP 3616A.1-002, Stroke Time and Failure Mode Test of 3MSS*AOV31A, B, D #MSS*AOV65; Stroke Time Test of 3MSS*MOV17A, B and D, Revision 008, Performed 7/22/08
- SP 3622.3-001, TDAFWP Operational Readiness Test, Revision 014, performed 4/26/08
- SP 3622.3-001, TDAFWP Operational Readiness Test, Revision 014, performed 2/4/08
- SP 3622.3-001, TDAFWP Operational Readiness Test, Revision 014, performed 2/5/08
- SP 3622.3-001, TDAFWP Operational Readiness Test, Revision 014, performed 11/16/07
- SP 3622.8-008, Manual Cycling of TDAFWP Suction Header Isolation Valves, Revision 000-02, Performed 7/21/08
- SP 3630D.3-004, 3CCE*TV37B Failure Test, Revision 000-04, Performed 7/24/08
- SP 3630D.3-005, Train B CCE Valve Stroke Time Test, Revision 000, Performed 7/24/08
- SP 3646A.8, Slave Relay Testing - Train A, Revision 022-02
- SP 3646A.8-021, AFW Pump Start S941 - Relay K641, Slave Relay Actuation Test - Train A, Revision 001, Performed 7/22/08
- SP 3646A.9-021, AFW Pump Start S941 - Relay K641, Slave Relay Actuation Test - Train B, Revision 002, Performed 7/22/08
- A-4 Sections 2OS1/2OS2:
- Access to Radiologically Significant Areas/ALARA Planning and Controls Procedures
- COP 200.12, Revision 1, Interim Administrative Controls for Systems and Equipment to be Retired-In-Place
- MP-03-DCC-GDL03, Revision 1, Retired-In-Place Equipment
- RP-AA-201, Revision 0, Access Controls for High and Very High Radiation Areas
- RP-AA-202, Revision 0, Radiological Posting
- RP-AA-220, Revision 0, Radiological Survey Scheduling
- RP-AA-221, Revision 0, Radiological Survey Records
- RP-AA-230, Revision 0, Personnel Contamination Monitoring and Decontamination RPM 1.3.13, Revision 8, Bioassay Sampling and Analysis
- RPM 1.3.14, Revision 7, Personnel Dose Calculations and Assessments
- RPM 1.3.8, Revision 8, Criteria for Dosimetry Issue
- RPM 1.6.4, Revision 4, Siemens Electronic Dosimetry System
- RPM 2.1.3, Revision 2, Identification and Control of High Radiological Risk Work
- RPM 2.4.2, Revision 14, Radiological Control of Material and Vehicles RPM 2.5.9, Revision 01, Dry Shielded Canister (DSC) Surveys ISFSI
- RPM 4.8.9, Revision 9, Source Checking of Health Physics Instruments
- RPM 5.2.2, Revision 10, Basic Radiation Worker Responsibilities
- RPM-GDL-008, Revision 0, Electronic Dosimeter Alarm Set Points
- RW 46000, Revision 8, Shipment of Radioactive Materials - General Guidelines
- RW 46001, Revision 7, Shipment of Radioactive Materials-Empty Packaging
- RW 46004, Revision 9, Shipment of Radioactive Materials-Low Specific Activity
- RW 46016, Revision 7, Shipment of Radioactive Waste - Waste Processing Facility
- RW 46030, Revision 3, Radioactive Material Storage Areas
- RW 46041, Revision 5, Compliance with 10
- CFR 61 - Waste Classification
- RW 46047, Revision 5, Radioactive Material Shipment Surveys
- RW 46052, Revision 4, Packaging Dry Active Waste
- RW 46053, Revision 5, Packaging Radioactive Waste Filters
- SP 2669A-003, Revision 22, Unit 2 Plant Equipment Operator Rounds
- RadWaste System Health Reports 2336A, Unit 2 Station Sumps and Drains
- 3335B-1, Unit 3 Reactor Plant Aerated Drains (Contaminated)
- MP-24-RWQA-PRG, Revision 1, Radioactive Waste PCP Implementation
- MP-27-RW-PRG, Revision 0, Radioactive Waste PCP
- Shipping Manifests Shipment Number 07-005, Dewatered Resin, LSA II, Type A
- A-5Shipment Number 07-089, Radioactive Source, Type B Shipment Number 07-096, Dewatered Resin, LSA II, Type A Shipment Number 08-002, Dewatered Resin, LSA II, Type A Shipment Number 08-046, Dewatered Resin, LSA II, Type A Shipment Number 08-087, Dewatered Resin, Low Specific Activity (LSA) II, Type A
Condition Reports
- 06-04490
- 06-06383
- 07-01078
- 07-01153
- 07-01234 07-02651 07-04873
- 07-07299
- 07-07386
- 07-09210 07-11152 07-11169
- 08-02895
- 08-02978
- 08-04127 08-04506 08-09172
- Access Controls/ALARA:
- 08-04343
- 08-04348
- 08-04857
- 08-04878
- 08-04905 08-04915 08-05043
- 08-05153
- 08-05267
- 08-05268 08-05386 08-05457
- 08-05457
- 08-05458
- 08-05459 08-05460 08-05549
- 08-07311
Drawings
- P & ID for Unit 2 liquid radwaste system (25203-26020)
- P & ID for Unit 3 radioactive liquid waste & aerated drains (25212-26906)
- Design Change Notices Modification of MP3 plant sampling system for secondary SG sampling points (M3-07002)
- RCS leak detection
- LT-9796 (DM2-00-0613-00)
- Temporary installation of aerated waste system sock filtration units (DM2-00-0018-08)
- Nuclear Oversight Audits/Assessments Audit 06-08, Radiological Protection & Process Control Program Audit 07-06, Radiation Protection, Process Control Program, Chemistry
- Nuclear Oversight Field Observation Reports 7589
- 27
- 7959
- 8101 8102 8109
- 08-023
- 08-024 08-025 08-026
- 08-028
- 08-030 08-032
Miscellaneous Documents
- 2007 Radioactive effluent release report Dose and Dose Rate Alarm Reports for April 2008 through August 2008
- Millstone Power Station, Dose and Source Term Reduction Strategic Plan, January 2008
- NF-AA-NSF-101, Revision 0, ISFSI Design and Licensing Basis Radioactive Material Shipping Logs for 2006, 2007, 2008
- Site HP High Rad C-Van Inventory Waste Services Liner Inventory Waste Services Material/Box Inventory
- A-2ALARA Reviews (3R12) 3-08-04, Refueling Activities
- 3-08-05, In-service Inspections
- 3-08-07, Boric Acid Response Team
- 3-08-09, Mechanical Maintenance
- 3-08-13, Scaffolding Installation/Removal
- 3-08-14, Insulation Removal/Installation
- 3-08-27, Operational Activities
- ALARA Council Meeting Minutes Meeting Dates: 04/03/2008, 04/18,19, 20/2008, 06/02/2008
- Challenge Board Meeting Handouts/Action Items for 3R12 Projects Boric Acid Response Team Containment Coordination & Cleanup Electrical Team
- I&C Team In-service Inspections Operations Team Refueling Team Sea Water Team Service Water Team Site Power Upgrade Project Snubber Inspections Split Pin Replacement Steam Generator Systems Team Valve Maintenance
- Section 4OA1 - Performance Indicator (PI) Verification Mitigating System Performance Index Millstone Unit 2, Revision 1 Mitigating System Performance Index Millstone Unit 3, Revision 2
- CR-08-05643
- CR-08-08710
Section 4OA2: Identification and Resolution of Problems Root Cause Evaluations
- RCE-M-08-06119, Unit 2 Trip Due to a Loss of Load
- RCE-M-08-06209, Millstone 2 Reactor Trip and Unusual Event (PU1) Following Loss of Offsite Power
- RCE-M-08-07451, Millstone 2 FWH Level Oscillation and Manual Reactor Trip
- CR-08-07451
- License Event Reports
- LER 2008-003-00, Failed Pilot Wire Causes Reactor Trip
- LER 2008-004-00, Reactor Trip Due to a Loss of Normal Power Event
- LER 2008-005-00, FWH Level Oscillation and Manual Reactor Trip
Procedures
- Implementation of 10CFR21: Reporting of Defects and Noncompliance,
- RAC-11, Revision-000-03 Quality Assurance Program Elements for Supply Chain Management,
- DNAP-1802, Revision 2
- Service Water Strainer Maintenance,
- MP 3750AA, Revision 003-05
- Work Order M3 06 03669
- M3 06 10050 M3 07 13479 M3 07 17849 M3 08 01033 M3 08 01034
- A-3 Condition Reports 06-02955
- 06-11755 07-01142 07-04858 07-07655 07-09904 08-00258 08-00894 08-06578 08-08133
- Design Change Notice
- DM3-00-0027-08, Install New Strainer Elements in All Four Unit 3 Service Water Housings
- DM3-00-0081-06, New Tubes (Strainer Baskets) for Service Water Strainer
- DM3-00-0102-06, Add Poro-Edge Tube Missing Weld Criteria to the Service Water Strainer VTM
Miscellaneous
- Lessons Learned from the MP3 Quarterly Service Water Strainer Boroscope Inspections, 5-15-2008 Millstone Unit 3 Maintenance Rule (a)(1) Evaluation for the Service Water System Operability Determination
- MP3-006-06 for CR 06-02955
Section 4OA3: Followup of Events and Notices of Enforcement Discretion
- CR-08-06117
- CR-08-06292
- CR-08-06328
- CR-08-07451
- CR-08-07476 Event Review Team Report, Millstone 2 Feedwater Heater Level Oscillation & Manual Reactor Trip,
- CR-08-07451 Event Review Team Report, Millstone 2 Loss of Load Event May 22, 2008,
- CR-08-06119
- Event Review Team Report, Millstone 2 Loss of Power Event May 24, 2008,
- CR-08-06209
- M2-05-05639
- MP-03-DCC-GDL06, "Processing an EMBUR", Revision 001
- MP-16-MMM, "Organizational Effectiveness (Corective Action Program, Operating Experience Program, Independent Safety Engineering Function), Revision 012-00 OD
- MP2-017-08, "SSC Affected by the Degraded or Non-Conforming Condition", Revision 0
- RCE M-08-07451, Millstone 2 Feedwater Heater Level Oscillation and Manual Reactor Trip CR-08-07451
Section 4OA5: Other Activities Procedures
- MP-PDI-UT-8 Revision C
- PDI Generic Ultrasonic Examination Procedure for Weld Overlaid Austenitic Pipe Welds
- SP 3646A.16, Revision 014-03, "Train B Loss of Power Test"
- SP 3646A.2, Revision 017-03, "Emergency Diesel Generator B Operability Test"
- SP 3646A.2, Revision 017-03, "Emergency Diesel Generator B Operability Test"
- SP3646A.15, Revision 016-03, "Train A Loss of Power Test"
- VPROC
- ENG 07-002R1 Generic Procedure for the Ultrasonic Examination of Weld Overlaid Similar and Dissimilar Metal Welds using
- PDI-UT-8 (WDI-STD-1007) VPROC
- ENG 07-003
- General Welding Standard (GWS-1), ASME Applications
- VPROC
- ENG 07-004
- Weld Material Control (WCP-3)
- A-4Examination Data Packages
- PZR 311-01-052, Pressurizer Spray, 03-X-5641-E-T, weld
- RCS-517-FW-12
- PZR 3R11-015, Pressurizer Safety "C", 03-X-5649-C-T, weld
- RCS-516-FW-5
- PZR A-3R11-015,
- Pressurizer Safety "A", 03-X-5644-A-T, weld
- RCS-516-FW-1
- PZR B-3R11-015,
- Pressurizer Safety, "B" nozzle, weld
- RCS-516-FW-3
- RCS-513-FW-1
Condition Reports
- 07-03865 R1, Reject indications in 1st weld layer,"B" safety nozzle pipe base metal 07-10682
- Examination Reports Millstone Unit 3 Safety Nozzle "A" SWOL Examination Coverage Summary Millstone Unit 3 Safety Nozzle "B" SWOL Examination Coverage Summary Millstone Unit 3 Safety Nozzle "C" SWOL Examination Coverage Summary Millstone Unit 3 PORV Nozzle SWOL Examination Coverage Summary
- 03-X-5551-X-T Surge Nozzle
- SWOL Examination Coverage Summary
- RCS-SL-FW-4
- 03-X-5641-E-T UT manual exam results of overlay on
- RCS-517-FW-12, PZR Spray
- Welding Procedures (WP) and Procedure Qualification Records (PQR)
- PQR 467R1 Manual gas tungsten arc welding of P43 to P43 (Alloy 600)
- V571, 662,665,666,670 & 9097 Welder Performance Qualification Records (Sample)
- VPROC
- ENG 07-006 ASME weld procedure P3-8/52-TB MC
- GTAW-N638
- VPROC
- ENG 07-007 Welding Procedure 43 MN GTAW/SMAW
- VPROC
- ENG 07-008 Welding Procedure 8-F43,
- MN-GTAW
- VPROC
- ENG 07-011 Welding Procedure 8-F43,
- MN-SMAW
Drawings
- 10017D86R1 Pressurizer Safety "A" Nozzle,
- RCS-516-FW-1
- 10017D89R1 Pressurizer PORV
- RCS-513-FW-1
- 10051C91R0 Spray Nozzle SWO (weld overlay) Design
- 10058C82R0 Pressurizer Safety /Relief Nozzle SWOL (weld overlay) Design
- 10058C83R0 Pressurizer Surge Nozzle SWOL (weld overlay) Design
- Completed Surveillance Procedures
- OP 2346A-004, Revision 023-00, "'A' DG Data Sheet" performed August 6, July 11, and June 11 2008
- OP 3346A-014, Revision 11, "EDG A Operating Log" completed July 16, June 17, and May 22, 2008
- OP 3346A-015, Revision 012-01, "EDG B Operating Log" completed July 2, Jun 11, and May 6, 2008 OP2346C-002, Revision 001-00, "'B' DG Data Sheet" performed August 20, July 24 and June 25, 2008
- SP 2613A-001, Revision 020-05, "Periodic DG Operability Test, Facility 1 (Fast Start Loaded Run)" completed July 11, 2008
- SP 2613B-001, Revision 021-02, "Periodic DG Operability Test, Facility 2 (Fast Start, Loaded Run)" completed June 25, 2008
- A-5SP 2613J-001, Revision 002-02, "'B' Emergency DG Loss of Load Test" completed June 25, 2008
- SP 2613K-001, Revision 003-03, "Periodic DG Slow Start Operability Test, Facility 1 (Loaded Run)" completed August 6 and June 11, 2008 SP2613L-001, Revision 003-03, "Periodic DG Slow Start Operability Test, Facility 2 (Loaded Run)" completed August 20 and July 24, 2008
Miscellaneous
- 01-20-2006 Approval Letter from NRC approving relief from Code requirements for Millstone 3 for use of weld overlay for repair and use of the Performance Demonstration Initiative (PDI)
program for inspection as alternatives to the requirements of the ASME Code,Section XI. 05-03-2007 Approval Letter from NRC for use of
- IR-2-47 for dissimilar metal weld overlays as an alternative repair technique 07-494, Dominion Nuclear Connecticut, Inspection and Mitigation of Alloy 82/182Pressurizer Butt Welds, Results of Inspections 25203-ER-98-0056, "EDG Parallel Operation to NSST or RSST"
- IR-2-39R
- Relief Request pertaining to the repair and inspection of weld No. 03-X-5641-E-T (Pressurizer Spray Nozzle) only, 10-20-2005 M3-EV-07-0026R00 Control and Remediation Plan for Alloy 600
- Millstone Unit 3 Pressurizer Structural Weld Overlay Project Final Report
- NDE Examiner Qualifications - Various,
- PDI-UT-8 Manual Ultrasonic
- NL-033, Revision 4, "Millstone 3 Emergency Generator Loading & Starting kVA Calculation"
- PA-79-126-1027-E2, "MP2 EDG Loading Calculation"
- PDI-UT-8 (Table 1) Instrument Settings
- PDI-UT-8 (Table 2) Qualified ultrasonic instruments and associated essential I Instrument settings that have an impact on pulse tuning
- WDI-PJF-1303606-EPP-001R0, R1 and Amendment 1, Examination Program Plan for the Pre-Service Inspection of Pressurizer Nozzle Structural Weld Overlays at Millstone Unit 3
- WDI-PJF-1303692-EPP-001R0 and Amendment 1 Examination Program Plan for the In-Process UT of Pressurizer Nozzle Structural Weld Overlays
LIST OF ACRONYMS
- ADA [[]]
MS Agencywide Documents Access and Management System
- ALA [[]]
RA As Low As is Reasonably Achievable
- AS [[]]
ME American Society of Mechanical Engineers
ASP Auxiliary Shutdown Panel
BOM Bills of Materials
CCE Charging Cooling Pump
- CE [[]]
DE Committed Effective Dose Equivalent
CFR Code of Federal Regulations
CIV Combined Intercept Value
CL Cold Leg
CR Condition Report
DG Diesel Generator
DM Design Modifications
A-6DNC Dominion Nuclear Connecticut DOT Department of Transportation
DRP Division of Reactor Projects
DRS Division of Reactor Safety
- EC [[]]
CS Emergency Core Cooling System
EDG Emergency Diesel Generator
ESF Engineered Safety Feature
- FS [[]]
AR Final Safety Analysis Report
- HP [[]]
SI High Pressure Safety Injection
HX Heat Exchanger
- ISF [[]]
SI Independent Spent Fuel Storage Installation
ISI Inservice Inspection
IST In Service Testing
- LH [[]]
RA Locked High Radiation Areas
- LO [[]]
CA Loss of Coolant Accident
LSA Low Specific Activity
MP2 Millstone Unit 2
mrem millirem
- MR [[]]
- MS [[]]
- MS [[]]
- MS [[]]
NCV Non-Cited Violation
NEI Nuclear Energy Institute
NRC Nuclear Regulatory Commission
- NS [[]]
- NUHO [[]]
MS Nuclear Horizontal Modular Storage
ODM Operational Decision Making
OOS Out Of Service
- PA [[]]
RS Publicly Available Records System
PDI Performance Demonstration Initiative
PI Performance Indicator
PI&R Problem Identification and Resolution
PM Preventive Maintenance
- PO [[]]
RV Power Operated Relief Valve
PWR Pressurized Water Reactor
A-7RCA Radiologically Controlled Area RCE Root Cause Evaluation
RCP Reactor Coolant Pump
RFO Refuel Outage
- RBC [[]]
- RPC [[]]
CW Reactor Plant Component Cooling Water
RPV Reactor Pressure Valve
- RS [[]]
ST Reserve Station Service Transformer
RWP Radiological Work Permit
SDP Significance Determination Process
- SG [[]]
FP Steam Generator Feed Pumps
SI Stress Improvement
SP Surveillance Procedure
- TDAF [[]]
WP Turbine Driven Auxiliary Feedwater Pump
TI Temporary Instruction
TS Technical Specification
- UFS [[]]
AR Updated Final Safety Analysis Report
- VH [[]]
RA Very High Radiation Areas
WO Work Order
B-1ATTACHMENT B TI 2515/172 Documentation Questions for Millstone 3 Station Introduction:
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
(MRP)-139 regarding nondestructive examination and evaluation of certain dissimilar metal
welds in reactor coolant systems (RCS) containing Alloy 600/82/182. The TI requires
documentation of specific questions in an inspection report. The questions and responses are included in this Attachment "B".
In summary, Millstone Unit 3 has four 6 inch pressurizer safety/relief nozzles, one 14 inch surge
line nozzle, one 4 inch spray nozzle, four 29 inch reactor vessel hot leg (HL) outlet nozzles and
four 27.5 inch reactor vessel cold leg (CL) inlet nozzles. Millstone 3 has submitted a proposed
alternative to the
- AS [[]]
ME Code to allow the application of a preemptive full structural weld
overlay on the pressurizer surge, spray, and safety/relief line welds. The proposed alternative
(IR-2-39, Revision 1 submitted 10/20/2005 and IR-2-47, Revision 1 submitted 03/28/2007) were
approved on 01/20/2006 and 05/03/2007, respectively, by NRC Staff.
a. For MRP-139 baseline inspections:
Qa1. Have the baseline inspections been performed or are they scheduled to be performed in accordance with MRP-139 guidance?
A Yes. Baseline automatic ultrasonic test (UT-PDI qualified phased array inspections have been performed on the four reactor pressure vessel (RPV) hot leg and four cold leg
dissimilar metal butt welds in accordance with MRP-139 guidance during outage 3R11
(spring 2007). Also, a surface eddy current was performed of these welds at this time.
The licensee plans to perform further mitigation of the eight RPV inlet and outlet
nozzles welds by application of a weld "inlay" on the inside diameter of the nozzle weld
during outage 3R14 (2011). The three safeties, one relief and one surge line nozzle
were mitigated by application of a full structural weld overlay during this same outage
(spring 2007). The pressurizer spray nozzle was also mitigated by application of a full
structural weld overlay during the fall 2005 refuel outage. A baseline manual PDI-
qualified phased array UT inspection was completed during the spring 2007 outage on
the four safety/relief nozzles and the surge line nozzle. A baseline manual PDI-qualified
inspection was performed in the fall 2005 outage on the pressurizer spray nozzle
dissimilar metal butt weld.
Qa2. Is the licensee planning to take deviations from the
MRP-139? If so, what deviations are planned and what is the general
basis for the deviation? If inspectors determine that a licensee is planning to deviate
from any
NRR)
should be informed by email as soon as possible.
A No deviations have been taken. However, a deviation is planned for the refuel outage in the fall of 2008 from the requirement to perform a bare metal visual examination of the
four reactor vessel hot leg outlet and four cold leg inlet nozzle welds due to the inability
B-2to access the outside diameter. These eight welds were
PDI qualified technique in the spring 2007
refuel outage. In addition, an eddy current examination was performed of the inside
diameter of the weld surface.
b. For each examination inspected, was the activity:
Qb1. Performed in accordance with the examination guidelines in
NRC staff
relief request authorization for weld overlaid welds?
A Yes. The overlay activity on the six previously identified pressurizer welds were applied and examined in accordance with the examination guidelines of MRP-139 and the relief
request authorization. The relief request authorization permitted the application of a full
structural weld overlay with subsequent volumetric PDI qualified manual phased array
ultrasonic examination of the weld overlay. Mechanical stress improvement was not
used on any dissimilar weld.
Qb2. Performed by qualified personnel? Briefly describe the personnel training/qualification process used by the licensee for this activity.
A Yes. The examinations were performed by personnel qualified to the requirements of
PDI program for the manual phased array ultrasonic examination of weld overlays on similar
Qb3. Performed such that deficiencies were identified, dispositioned, and resolved.
A Yes. Indications identified in the ultrasonic examination were evaluated for relevance, characterized and entered into the licensee's corrective action program for disposition
and resolution.
c. For each weld overlay inspected, was the activity:
Qc1. Performed in accordance with
NRC staff relief request authorizations? Has the licensee submitted a relief request and
obtained NRR staff authorization to install the weld overlays?
A Yes. The application of the weld overlays were performed in accordance with the
IX and XI) using qualified welding procedures and qualified
welders. Weld overlay of the six dissimilar metal welds was authorized by NRR in their
approval dated 01/20/2006 and 05/03/2007 for Millstone 3 to apply full structural weld
overlays on the surge, spray and four safety/relief nozzle welds.
Qc2. Performed by qualified personnel? (Briefly describe the personnel training/qualification process used by the licensee for this activity).
A Welders applying the structural weld overlay were qualified in accordance with the requirements of
IX and personnel performing examination of the
B-3completed weld overlays were qualified in accordance with
PDI qualified for manual phased array ultrasonic examination.
Qc3. Performed such that deficiencies were identified, dispositioned, and resolved?
A Yes. Indications identified as a result of the ultrasonic
UT examination were evaluated for relevance, characterized and entered into the licensee's corrective
action program for disposition and resolution.
d. For each mechanical stress improvement (SI) used by the licensee during the outage, was the activity performed in accordance with a documented qualification report for stress improvement processes and in accordance with demonstrated procedures? Specifically:
Qd1. Are the nozzle, weld, safe end, and pipe configurations, as applicable, consistent with the configuration addressed in the SI qualification report?
A N/A, the mechanical stress improvement process was not used.
Qd2. Does the SI qualification report address the location radial loading is applied, the applied load, and the effect that plastic deformation of the pipe configuration may have on the
ability to conduct volumetric examinations?
A N/A
Qd3. Do the licensee's inspection procedure records document that a volumetric examination per the
SI?
A N/A
Qd4. Does the
SI and post-SI inspections and that any flaws identified during the volumetric
examination are to be within the limiting flaw sizes established by the SI qualification
report?
A N/A
Qd5. Performed such that deficiencies were identified, dispositioned, and resolved?
A N/A
e. For the inservice inspection program:
Qe1. Has the licensee prepared an
ISI) program? If not, briefly summarize the licensee's basis for not having a documented program and when
the licensee plans to complete preparation of the program.
A Yes. The licensee has an
ISI program which is implemented through M3-EV-07-0026 Revision 00, June 22, 2007, Control and Remediation Plan for Alloy 600. This
B-4program is separate from the
XI ISI program. This program provides the basis to support management strategies needed to address technical operating
experience with all Alloy 600/82/182 pressure boundary butt welds including materials,
commitments, remediation, inspection, and regulatory requirements. These welds will
be included in the Risk-Informed ISI program upon completion of the remediation plan
for Millstone Unit 2 Alloy 600.
Qe2. In the
ISI Program, are the welds appropriately categorized in accordance with MRP-139? If any welds are not appropriately categorized, briefly explain the
discrepancies.
A Yes. All fourteen welds identified during this inspection are appropriately categorized in accordance with MRP-139.
Qe3. In the
ISI Program, are there in-service inspection frequencies, which may differ between the first and second 10-year intervals after the MRP-139 baseline
inspection, consistent with the in-service inspection frequencies called for by MRP-139?
A All MRP-139 applicable welds are scheduled either for mitigation and/or inspection prior to the end of the current 10-year inspection interval, which is April 2009.
Qe4. If any welds are categorized as H or I, briefly explain the licensee's basis for the categorization and the licensee's plans for addressing potential
- PWS [[]]
CC.
A There are no welds at Unit 3 that are categorized as H or I.
Qe5. If the licensee is planning to take deviations from the in-service inspection requirements of MRP-139, what are the deviations and what are the general bases for the deviations?
Was the NEI 03-08 process for filing deviations followed?
A The licensee currently plans to submit a relief request from the MRP-139 requirement for a visual inspection of the four hot leg reactor vessel outlet nozzles and four cold leg
reactor vessel inlet nozzles during the fall 2008 outage. The bases for this request will
be that the visual examination specified in MRP-139 cannot be performed due to
inaccessibility (shield blocks/insulation obstruction) from the outside diameter and, the normally scheduled
RFO) can be
credited as meeting the requirements for the visual examination required by MRP-139.
This automatic
UT examination will be performed from the nozzle inside
diameter in the fall of 2008. The licensee had not yet submitted this deviation request to
- NRC. [[]]