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

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.

P.1(d)

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

TI 2515/176.

The inspector assessment and information gathered while completing this TI was

discussed with licensee personnel. This information was forwarded to the Office of

Nuclear Reactor Regulation for further review and evaluation.

b. Findings

No findings of significance were identified.

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

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

AOP 2502C, Loss of Vital 4.16 kV Bus 24C, Revision 004-04

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 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

NUREG/CR-6750

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

OP 2209H, Revision 01 DSC Loading ISFSI

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 1.4.1, Revision 7, ALARA Reviews and Reports

RPM 1.4.2, Revision 2, ALARA Engineering Controls

RPM 1.6.4, Revision 4, Siemens Electronic Dosimetry System

RPM 2.1.1, Revision 7, Issuance and Control of RWPs

RPM 2.1.2, Revision 2, ALARA Interface with the RWP Process

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 5.2.3, Revision 3, ALARA Program and Policy

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

SP 3670.4, Revision 21, Routine PM's

RadWaste System Health Reports 2336A, Unit 2 Station Sumps and Drains

3335B-1, Unit 3 Reactor Plant Aerated Drains (Contaminated)

Process Control Program

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

25203-28105, Logic Diagram - Anticipated Transient Without Scram (A.T.W.S.), Sheet 36, Revision 4

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-FAP1.3, "Master Equipment List (MEL) and Bill of Material (BOM) Systems, Revision 002

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

PZR PORV, 3R11-015, 03-X-5650-D-T, weld

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)

PQR 675 Manual shielded metal arc welding of P8 to P8 (Alloy 304)

PQR 770R1 Machine gas tungsten arc welding of P3 to P43 (vessel nozzle)

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

AC Alternating Current

ADA [[]]

MS Agencywide Documents Access and Management System

AFW Auxiliary Feedwater

ALA [[]]

RA As Low As is Reasonably Achievable

AOP Abnormal Operating Procedure

AS [[]]

ME American Society of Mechanical Engineers

ASP Auxiliary Shutdown Panel

BOM Bills of Materials

CCE Charging Cooling Pump

CCP Component 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

DSC Dry Shielded Canister

EC [[]]

CS Emergency Core Cooling System

EDG Emergency Diesel Generator

ESF Engineered Safety Feature

FIN Finding

FS [[]]

AR Final Safety Analysis Report

FWH Feedwater Heater

HL Hot Leg

HP [[]]

SI High Pressure Safety Injection

HRA High Radiation Areas

HX Heat Exchanger

IMC Inspection Manual Chapter

ISF [[]]

SI Independent Spent Fuel Storage Installation

ISI Inservice Inspection

IST In Service Testing

LER Licensee Event Reports

LH [[]]

RA Locked High Radiation Areas

LNP Loss of Normal Power

LO [[]]

CA Loss of Coolant Accident

LSA Low Specific Activity

MP2 Millstone Unit 2

mrem millirem

MRP Materials Reliability Program

MR [[]]

RF Millstone Radwaste Reduction Facility

MS [[]]

IP Mechanical Stress Improvement Process

MS [[]]

PI Mitigating System Performance Indication

MS [[]]

SV Main Steam Safety Valve

NCV Non-Cited Violation

NEI Nuclear Energy Institute

NRC Nuclear Regulatory Commission

NRR Nuclear Reactor Regulation

NS [[]]

ST Normal System Station Transformer

NUHO [[]]

MS Nuclear Horizontal Modular Storage

OD Operability Determinations

ODM Operational Decision Making

OOS Out Of Service

OP Operating Procedure

PA [[]]

RS Publicly Available Records System

PCP Process Control Program

PDI Performance Demonstration Initiative

PI Performance Indicator

PI&R Problem Identification and Resolution

PM Preventive Maintenance

PMT Post Maintenance Testing

PO [[]]

RV Power Operated Relief Valve

PWR Pressurized Water Reactor

A-7RCA Radiologically Controlled Area RCE Root Cause Evaluation

RCP Reactor Coolant Pump

RCS Reactor Coolant System

RFO Refuel Outage

RHR Residual Heat Removal

RHS Residual Heat Removal System

RBC [[]]

CW Reactor Building Component Cooling Water

RPC [[]]

CW Reactor Plant Component Cooling Water

RPV Reactor Pressure Valve

RRS Reactor Regulating System

RS [[]]

ST Reserve Station Service Transformer

RWP Radiological Work Permit

SDP Significance Determination Process

SG Steam Generator

SG [[]]

FP Steam Generator Feed Pumps

SI Stress Improvement

SP Surveillance Procedure

SW Service Water

TDAF [[]]

WP Turbine Driven Auxiliary Feedwater Pump

TI Temporary Instruction

TS Technical Specification

UE Unusual Event

UFS [[]]

AR Updated Final Safety Analysis Report

UT Ultrasonic Test

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 baseline inspection requirements of

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

MRP -139 baseline inspection requirement, Nuclear Reactor Regulation(

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

UT inspected from the inside diameter using an automatic, phased array,

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

MRP -139, Section 5.1 for unmitigated welds or mechanical stress improved welds and consistent with

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

ASME Section

XI , Appendix

VIII. Procedures and personnel were qualified in the

PDI program for the manual phased array ultrasonic examination of weld overlays on similar

and dissimilar metal welds.

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

ASME Code welding requirements and consistent 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

ASME Code requirements (Section

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

ASME Section

IX and personnel performing examination of the

B-3completed weld overlays were qualified in accordance with

ASME Section

XI , Appendix

VIII and

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

PDI -qualified

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

ASME Code, Section

XI , Appendix

VIII was performed prior to and after the application of the

SI?

A N/A

Qd4. Does the

SI qualification report address limiting flaw sizes that may be found during pre-

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

MRP -139 Inservice inspection (

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

MRP -139

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

ASME Section

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

MRP -139

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

MRP -139

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

UT examination performed in the Spring of 2007(3R11

RFO) can be

credited as meeting the requirements for the visual examination required by MRP-139.

This automatic

PDI qualified

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

NEI or

NRC. [[]]