IR 05000286/2001013
| ML020810692 | |
| Person / Time | |
|---|---|
| Site: | Indian Point  |
| Issue date: | 03/22/2002 |
| From: | Eselgroth P Division Reactor Projects I |
| To: | Barrett R Entergy Nuclear Operations |
| References | |
| IR-01-013 | |
| Download: ML020810692 (22) | |
Text
rch 22, 2002
SUBJECT:
INDIAN POINT 3 NUCLEAR POWER PLANT - NRC INSPECTION REPORT NO. 50-286/01-13
Dear Mr. Barrett:
On February 16, 2002, the NRC completed an inspection at the Indian Point 3 nuclear power plant. The enclosed report presents the results of that inspection. The results were discussed on March 13, 2002, with you and other members of your staff.
The inspection was an examination of activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations, and with the conditions of your license. Within these areas, the inspection consisted of a selected examination of procedures and representative records, observations of activities, and interviews with personnel.
No findings of significance were identified.
Immediately following the terrorist attacks on the World Trade Center and the Pentagon, the NRC issued an advisory recommending that nuclear power plant licensees go to the highest level of security, and all promptly did so. With continued uncertainty about the possibility of additional terrorist activities, the Nation's nuclear power plants remain at the highest level of security and the NRC continues to monitor the situation. This advisory was followed by additional advisories, and although the specific actions are not releasable to the public, they generally include increased patrols, augmented security forces and capabilities, additional security posts, heightened coordination with law enforcement and military authorities, and more limited access of personnel and vehicles to the sites. The NRC has conducted various audits of your response to these advisories and your ability to respond to terrorist attacks with the capabilities of the current design basis threat. On February 25, 2002, the NRC issued an Order to all nuclear power plant licensees, requiring them to take certain additional interim compensatory measures to address the generalized high-level threat environment. With the issuance of the Order, we will evaluate Entergy Nuclear Northeast compliance with these interim requirements.
Robert In accordance with 10 CFR 2.790 of the NRC's "Rules of Practice," a copy of this letter and its enclosure will be available electronically for public inspection in the NRC Public Document Room or from the Publicly Available Records (PARS) component of the NRCs document system (ADAMS). ADAMS is accessible from the NRC Web site at http://www.nrc.gov/reading-rm.html (the Public Electronic Reading Room).
Sincerely,
/RA/
Peter W. Eselgroth, Chief Projects Branch 2 Division of Reactor Projects Docket No.50-286 License No. DPR-64 Enclosure: Inspection Report No. 50-286/01-13 Attachment 1: Supplemental Information
Robert
SUMMARY OF FINDINGS
IR 05000286-01-013, on 12/30/2001 - 02/16/2002, Entergy Nuclear Northeast, Indian Point 3
Nuclear Power Plant. Resident inspection report, radiation protection.
The inspection was conducted by resident and regional inspectors. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described at its Reactor Oversight Process website at http://www.nrc.gov/NRR/OVERSIGHT/index.html A. Inspector Identified Findings None
Licensee Identified Violations
None ii
Report Details SUMMARY OF
PLANT STATUS
The reactor operated at full power for all of the inspection period. There were no equipment performance issues that had a significant impact on plant operations.
REACTOR SAFETY
(Cornerstones: Initiating Events, Mitigating Systems, Barrier Integrity, Emergency Preparedness )
1R04 Equipment Alignment
a. Inspection Scope
(71111.04)
- On January 4, 2002, the inspectors completed a partial walkdown of the safety injection (SI) system to verify the availability of the 31 and 33 SI pumps while surveillance test 3PT-Q116B, 32 Safety Injection Pump Functional Test was performed on the 32 SI pump. The inspectors used check-off list COL-SI-1, Safety Injection System, to verify the proper alignment of the 31 and 33 safety injection flow trains. In addition, the inspectors compared the check-off list to drawing 9321-F-27503, Flow Diagram Safety Injection System, to confirm the correct valve positions for the 31 and 33 flow trains.
- On January 17, 2002, the inspectors completed a partial walkdown of the auxiliary feedwater (AFW) system to verify the availability of the 31 AFW pump while the 32 AFW pump was out of service for functional testing. The inspectors used check-off list COL-FW-2, Auxiliary Feedwater System, and system operating procedure SOP-FW-004, Auxiliary Feedwater System Operation, to verify the proper alignment of equipment in the 31 AFW flow train.
b. Findings
No findings of significance were identified.
1R05 Fire Protection
a. Inspection Scope
Temporary Instruction (TI) 2515/146, Hydrogen Storage Locations On January 9-10, 2002, the inspectors toured the IP3 hydrogen storage areas to evaluate the existence of potential fire hazards, to verify compliance with applicable National Fire Protection Association (NFPA) and NRC codes and commitments regarding hydrogen storage locations at nuclear power plants, and to ensure that licensee commitments regarding hydrogen storage were being followed. Storage of flammable material poses an industrial hazard that must be addressed in accordance with NFPA requirements. The inspectors also verified that there was greater than the required minimum of 50 feet of separation between the hydrogen storage location and ventilation intake openings, and between the hydrogen storage location and risk significant structures, systems, and components. This Temporary Instruction is complete.
The inspectors also conducted fire protection tours in the fire zones listed below to observe if the licensee was controlling transient combustibles in accordance with fire protection procedure FP-9 Control of Combustibles; to ensure that the licensee was controlling ignition sources in accordance with FP-8, Controlling of Ignition Sources; to ensure that the licensee had provided the fire protection equipment as specified in Pre-Fire Plans (PFPs) listed below; and to assess the general material condition of the fire protection equipment and fire protection barriers.
- On February 10, 2002, the inspectors used PFP-14, Volume Control Tank -
Primary Auxiliary Building to tour the chemical and volume control system (CVCS) volume control tank and valve areas in the Auxiliary Feedwater Building (PAB). The inspectors evaluated the existence of potential fire hazards, and verified that fire protection equipment was staged appropriately. The inspectors also observed the condition and routing of the Appendix R alternate shutdown cables running though this fire zone.
- On February 10, 2002, the inspectors used PFP-12, Charging Pumps - Primary Auxiliary Building to tour the charging pumps area on the 55 foot elevation of the PAB. The inspectors evaluated the existence of potential fire hazards and verified that fire protection equipment was staged appropriately. The inspectors also observed the condition and routing of the Appendix R alternate shutdown cables running though this fire zone.
- On February 13, 2002, the inspectors toured Fire Zones 3, 4, 8A, 9A, 12A, 13A, and 14A on the 15 foot elevation of the PAB. These areas contained safety-related systems and equipment necessary for a plant shutdown, and also Appendix R alternate shutdown equipment. The inspectors noted that some of the equipment in these areas was not accurately reflected on the area diagram in PFP-05, General Floor Plan - Primary Auxiliary Building, 15'-0" Elevation.
These items were discussed with the cognizant fire protection engineer for resolution.
b. Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessment and Emergent Work
a. Inspection Scope
(71111.13)
The inspectors reviewed the maintenance risk assessments and corrective maintenance work request (WR) packages for the following emergent work, and discussed the deficient conditions with cognizant personnel (system engineers, maintenance technicians, etc.):
WR 01-02441-00; Secondary Seal Leakage Indicated by Seal Tank Level Drop:
On January 9, 2002, the licensee took the 33 charging pump out of service for a planned seal replacement (WR 01-02441-00). In addition to this activity, the licensee planned to replace a Westinghouse W-2" switch that would cause the 31 emergency diesel generator (EDG) to be inoperable for approximately 2.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />. The inspectors reviewed the licensees risk assessment for this equipment configuration, and reviewed surveillance procedure 3PT-W019, which was used to verify the availability of the offsite power supply to the 480 volt safeguards buses.
WR 94-01648-03; Modify Battery 31 Fuse Holder and Install New Fuses per DCP 00-3-018 The replacement fuse clips associated with the 31 station battery modification (DCP 00-3-018) did not arrive onsite in sufficient time to meet the original modification schedule.
The licensee rescheduled the fuse clip replacement and performed the installation on January 18, 2002. The inspectors reviewed the licensees schedule change sheet and risk assessment for the revised replacement date, and evaluated the impact to the work already scheduled for that week.
WR 02-00847-00; Troubleshooting Isolation Valve Seal Water System (IVSWS) Check Valve Leakage On January 21, 2002, the licensee observed that the IVSWS header to containment isolation valves in the steam generator blowdown (SGBD) and SGBD sample lines had been over-pressurized (DER 02-00160). The pressure gage (PI-6200) for that header was pressurized beyond its maximum indication range of 160 psig, and the relief valve (IV-1487) for that header did not lift at its 150 psig setpoint (discussed further in Section
1R19 ). The licensee considered that one of the eight check valves in the IVSWS
header was back-leaking and causing the over-pressure condition. The licensee subsequently developed a work plan to troubleshoot the problem by isolating the entire IVSWS header from the containment isolation valves, and selectively venting each check valve to identify the source of the leakage. The inspectors reviewed the licensees risk assessment for this work, and evaluated the impact of removing the entire IVSWS header from service on other planned work and work in progress.
Isolating the IVSWS header required entering a Technical Specification 7-day limiting condition for operation (LCO 3.6.9.A), while the header was out of service.
b. Findings
No findings of significance were identified.
1R14 Personnel Performance During Non-Routine Plant Evolutions and Events
a. Inspection Scope
(71111.14)
On January 18, 2002, the inspectors observed the replacement of the 31 battery charger fuse clip replacement as a non-routine evolution that had the potential to cause a plant transient (loss of a direct current (DC) electrical bus). The 31 battery charger fuse clip replacement, part of the 31 station battery replacement modification (See Section 1R19), was classified as a special evolution by the licensee due to the proximity of the 31 battery charger fuse clip to energized electrical wires. The inspectors observed the maintenance technicians actions to determine if they were in accordance with applicable station procedures and training.
b. Findings
No findings of significance were identified.
1R15 Operability Evaluations
a. Inspection Scope
(71111.15)
The inspector reviewed various DERs on degraded or non-conforming conditions that raised questions on equipment operability. The inspectors reviewed the resulting operability determinations (OD) for technical adequacy, whether or not continued operability was warranted, and to what extent other existing degraded systems adversely impacted the affected system or compensatory actions.
The following DERs and ODs were evaluated:
- DER 02-00141 (January 16, 2002) documented that there was an inadequate supply of lube oil on site for seven days of continuous operation of all three emergency diesel generators (EDGs). 110 gallons were currently available on site, but the three EDGs would consume 412 gallons in seven days according to the EDG manufacturers published oil consumption rate at full load.
OD 02-01, EDG Lube Oil Supply. The licensee determined that the available 110 gallons would be sufficient for approximately 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> of EDG operation under design basis conditions. In addition, the lube oil contained in the sump of the Appendix R diesel generator was the same Category I lube oil as the EDGs, and was recently replaced on December 16, 2001. The Appendix R diesel sump contained approximately 385 gallons, which could be removed and transferred to the EDGs during the first 45 hours5.208333e-4 days <br />0.0125 hours <br />7.440476e-5 weeks <br />1.71225e-5 months <br /> of operation. Since there is no scenario in the IP3 licensing basis requiring simultaneous operation of the EDGs and the Appendix R diesel, there would be sufficient oil (495 gallons) for EDG operation to last for seven days. (The licensee subsequently purchased approximately 5500 gallons of new lube oil that was delivered to the site in January 2002).
- DER 02-00215 (January 22, 2002) documented that the calibration records for the over-pressure delta temperature (OPDT) and over-temperature delta temperature (OTDT) dynamic compensators had been lost. The calibrations were performed in March 2001 and the compensators were installed during the 2001 refueling outage.
OD 02-02, Shop Calibrations for Over-pressure Delta-temperature and Over-temperature Delta-temperature Setpoint Calculators. The licensee recovered electronic records of the work request packages which indicated that the calibrations were satisfactorily performed in the Instrumentation and Controls (I&C) shop prior to installation. These records also indicated that the work packages were reviewed by an I&C supervisor and the work control department.
The work packages contained steps which required an independent verification that the instruments were calibrated prior to installation.
- DER 02-00228 (January 16, 2002) documented that door seals in the 31 fan cooler unit (FCU) failed an inspection following their recent replacement. The seals in the doors between the fan and cooling coils and to the high efficiency particulate (HEPA) filters had some material missing and contained gaps probably caused by bent doors.
OD 02-03, Evaluation of 31 FCU Door Seals. The licensee inspected the 31 FCU door seals and noted that more material was present, and with fewer gaps, than before the seals were replaced. In addition, the 31 FCU was last tested in May 2001, and demonstrated its ability to pass more than the rated total air flow of 34,000 cubic feet per minute (cfm), with 7% more than the design-basis minimum of 8000 cubic feet per minute bypass flow through its filtration section (moisture separator, HEPA and carbon filter assembly). Since the test was performed before the door seals were replaced, the licensee concluded that the new seals would enhance FCU air flow. On January 24, 2002, the licensee inspected the door seals of the other four FCUs inside containment and concluded that they were in satisfactory condition.
b. Findings
No findings of significance were identified.
1R17 Permanent Plant Modification
a. Inspection Scope
(71111.17)
The inspectors reviewed the design change package (DCP 00-3-018) for replacement of the 31 and 32 Station Batteries to verify that the design bases, licensing bases, and performance capability had not been degraded through modification; and to verify that the performance of this modification while at power did not place the plant in an unsafe condition. DCP 00-03-018 consisted of removal of the old battery cells, modifying the existing battery racks, installing the new battery cells, and replacing the 31 and 32 battery charger fuse clips. In order to replace the batteries while the plant was at power, temporary batteries were installed and separately tied to the 31 or 32 battery circuits to provide a continuous supply of DC power to safety-related equipment and instruments.
The inspectors also observed portions of the installation and post-installation testing of the 31 temporary battery, installation of the new 31 station battery, the post-modification testing of the new 31 station battery, and replacement of the 31 battery charger fuse clip.
b. Findings
No findings of significance were identified.
1R19 Post Maintenance Testing
a. Inspection Scope
(71111.19)
The inspectors reviewed post-maintenance test procedures and associated testing activities to assess whether 1) the effect of testing in the plant had been adequately addressed by control room personnel, 2) testing was adequate for maintenance performed, 3) acceptance criteria were clear and adequately demonstrated operational readiness consistent with design and licensing documents, 4) test instrumentation had current calibrations, range, and accuracy for the application, and 5) test equipment was removed following testing. The following surveillance activities were evaluated:
WR 94-01648-03: Modify 31 Battery Fuse Holder and Install New Fuses per DCP 00-3-018 On January 18, 2002, the inspectors observed the replacement of the 31 battery charger fuse clip and the post replacement determination that declared the fuse clip operable. The inspectors verified that the declaration of operability was appropriate and the functional capability of the battery charger was maintained by the replacement fuse clip.
WR 02-00515-01; IVSWS Check Valve Leak Tests and Relief Valve Setpoint Tests:
On February 8, 2002, the licensee replaced IVSWS check valve IV-1545 and pressure relief valve IV-1487. This emergent work was necessary after the licensee determined that back-leakage through IV-1545 (check valve to the 31 SGBD sample line) had over-pressurized the IVSWS header to the SGBD and sample lines, and that IV-1487 did not lift at its setpoint pressure. After removing these valves from the system, the licensee conducted bench testing and observed that IV-1545 leaked at 30 milliliters/hour when subjected to nominal SGBD pressure (750 psig). Also, IV-1487 lifted above its setpoint pressure (150 psig) at 220 psig. The licensee noted that the IVSWS check valves were expected to have zero leakage, and that 220 psig would not have over-pressurized that portion of the system (design pressure for the weakest component in the IVSWS header is 250 psig). Neither valve showed signs of internal degradation; however, IV-1487 did not have lockwire attached to its setpoint adjustment, as expected. Consequently, the licensee performed an extent-of-condition review of all other IVSWS relief valves to confirm the lockwire was installed. In addition, the licensee performed an inspection of IVSWS header pressure indicator PI-6200, which was over-pressurized from the back-leakage through IV-1545. No damage was apparent; however, the gage required calibration.
The replacement for relief valve IV-1487 was satisfactorily bench tested prior to installation in the IVSWS. Following replacement of both IV-1545 and IV-1487, the licensee observed zero leakage when they were placed in service.
WR 00-03283-17: Replace 34 Reactor Coolant Pump (RCP) Seal Return High Flow Transmitter FT-156B; On February 13, 2002, the licensee replaced the flow transmitter (FT-156B) for the seal water return line from the No. 1 seal of the 34 RCP. The replacement was deferred from the previous refueling outage, and was performed as a Special Evolution with the plant at power due to the increased risk of replacing this transmitter online. The inspectors reviewed the results of the pre-installation shop test and the post-installation leak check of the transmitter.
b. Findings
No findings of significance were identified.
1R22 Surveillance Testing
a. Inspection Scope
(71111.22)
The inspectors observed portions of the following surveillance tests and reviewed the surveillance test procedures to assess whether 1) the test pre-conditioned the component(s) tested, 2) the effects of testing were adequately addressed by control room operators, 3) the acceptance criteria demonstrated operational readiness consistent with design calculations and licensing documents, 4) the test equipment range and accuracy was adequate with proper calibration, 5) the test was performed in the proper sequence, and 6) test equipment was removed following testing and the system was properly restored to service.
- 3PT-Q058, 38 Back-up Service Water Pump Test, performed on January 17, 2002
- RE-CS-051, Total Reactor Coolant Specific Activity Determination, performed during the week of January 7, 2002
- 3PT-Q036, IST Stroke Test of Valves AC-MOV-822A & B and AC-751A & B; and 3PT-Q088, Component Cooling Pumps Functional Test, performed on February 11, 2002.
b. Findings
No findings of significance were identified.
RADIATION SAFETY
Cornerstone: Occupational Radiation Safety (OS)
2OS1 Access Control To Radiologically Significant Areas
a. Inspection Scope
(71121.01)
The inspectors reviewed the effectiveness of the licensees access controls to radiologically significant areas.
The inspectors toured the radiologically controlled areas (RCAs) including various elevations of the primary auxiliary and radioactive waste handling buildings, the 68-foot and 80-foot elevations of containment at power, the health physics (HP) counting and calibration facilities, and the HP access control point. During these walkdowns, the inspector observed and verified the appropriateness of the radiological safety controls in place for active radiological work permits (RWPs). Also, the inspectors reviewed the adequacy of the locking, posting, barricading, and labeling of radiation and high radiation areas, contamination areas, and radioactive material areas. The status of locked high radiation areas was also reviewed. The licensee implemented a new computerized HP access control system at the start of this year. The inspector observed activities at the main RCA access control point to verify compliance with requirements for RCA entry and exit, wearing of record dosimetry, and issuance and use of alarming electronic radiation dosimeters. The inspectors evaluated the effectiveness of pre-job radiation safety briefings for a containment entry at power on January 15 and 16, 2002. On January 16, 2002, the inspector made a containment entry at power and observed and verified the appropriateness of the radiological safety controls in place for RWP 02-0025. The work activity involved the F incore-detector drive on the 80-foot elevation and the B five-path selector on the 68-foot elevation.
The inspection included a review of the following RWPs and procedures to evaluate the adequacy of controls for access to radiologically controlled areas.
- RWP 02-0025, Flux Drive and Moveable Incore-detector System Work - While Reactor Critical
- RWP 02-0028, Containment entry - While Reactor Critical - Outside the Crane Wall
- AP-7, Radiation Protection Plan
- RE-REA-4-1, Radiation Work Permit
- RE-REA-4-16, Radiological Controls for Diving
- RE-ACC-5-2, Instructions to Control Point Personnel
- RE-SUR-6-3, Airborne Radioactivity Surveys, Posting, and Assessment
- RE-SUR-6-6, Health Physics Periodic Task Scheduling The inspectors reviewed the following six Deviation/Event Reports (DERs) and their associated Action Commitment Tracking System (ACTS) items for appropriateness of category, immediate correction actions, corrective actions to prevent recurrence, and for corrective action timeliness and effectiveness: DERs 01-04283, 01-04341, 01-04342, 01-04417, 01-04419, and 01-04445. These DERs were generated during the period of late October 2001 to early January 2002.
The review of the documents and activities listed above was against criteria contained in Title 10 of the Code of Federal Regulations (CFR) Parts 20.1201 (Occupational dose limits for adults), 20.1204 (Determination of internal exposure), 20.1208 (Dose equivalent to an embryo/fetus), Subpart F (Surveys and monitoring), 20.1601 (Control of access to high radiation areas), Subpart H (Respiratory protection and controls to restrict internal exposures in restricted areas), 20.1902 (Posting requirements), site Technical Specification 6.12 (High Radiation Area), and site procedures (identified above in this section).
b. Findings
No findings of significance were identified.
2OS2 ALARA Planning and Control
a. Inspection Scope
(71121.02)
The inspectors reviewed the effectiveness planning and control of ALARA (As Low As is Reasonably Achievable).
The inspectors reviewed selected individual exposure records for 2001. The inspectors also reviewed the following procedures, records, and documents for regulatory compliance and for adequacy of control of radiation exposure.
- AP-7, Radiation Protection Plan
- RE-REA-4-1, Radiation Work Permit
- 2002 Station person-rem estimates by department
- Fourth Quarter 2001 Review of Station ALARA Program, dated January 10, 2002
- 2001 End of Year Review of Station ALARA Program, dated January 10, 2002
- Station ALARA Committee Meeting Minutes for December 3, 2001 and December 19, 2001
- American Nuclear Insurers Nuclear Liability Insurance Inspection Report, dated December 17, 2001 The review was against criteria contained in 10 CFR 20.1101 (Radiation protection programs), in 10 CFR 20.1701(Use of process or other engineering controls), and in site procedures (identified above in this section).
b. Findings
No findings of significance were identified.
2OS3 Radiation Monitoring Instrumentation
a. Inspection Scope
(71121.03)
The inspectors reviewed the program for health physics instrumentation to determine the accuracy and operability of the instrumentation.
During plant tours, the inspectors reviewed field instrumentation utilized by health physics technicians and plant workers to measure radioactivity and radiation levels, including portable field survey instruments, hand-held contamination frisking instruments, and continuous air monitors. The inspectors conducted a review of the instruments observed in the toured areas, specifically verification of current calibration, of appropriate source checks, and of proper function.
On January 15, 2002, the inspectors witnessed part of the performance of the calibration procedure for an installed personnel monitor (Model IPM8, HP#11553) to verify procedural compliance. The inspectors also reviewed the provisions in place for electronic dosimeter alarm recognition when used in high noise areas in the RCA.
The inspectors evaluated the following procedures, records, and documents for regulatory compliance and adequacy.
- RE-INS-7UE-7, Use of Nuclear Enterprises CM7A/CM11"
- RE-INS-7UG-6, Use of the Merlin Gerin WRM91 Wireless Remote Monitor Dosimetry System with Windows
- RE-INS-7CE-6, Calibration of the N.E. IPM 7/8 Installed Personnel Monitor
- IPM 7/8 calibration sheets for HP#11553, dated January 15, 2002
- Vendor Analysis Report for Dry Active Waste Smears (report dated May 24, 2001)
- Licensee Sample Report for Dry Active Waste Smears (sample date of February 16, 2001)
- Indian Point 3 Nuclide Mix Evaluation Report - 1998, TID-99-002
- Memo on counting room proportional counter calibration source, dated April 17, 2000 The review was against criteria contained in 10 CFR 20.1501, 10 CFR 20 Subpart H, site Technical Specifications, and site procedures.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES (OA)
4OA1 Performance Indicator Verification
a. Inspection Scope
(71151)
Reactor Coolant System Leakage The inspectors reviewed the reactor coolant system (RCS) identified and unidentified leakage data that the licensee submitted for the third and fourth quarters of year 2001 against the applicable criteria specified in Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Performance Indicator Guideline, Rev 1. The review included a sample of daily leakage calculations performed by the operators and verification that the data was accurate.
Reactor Coolant System Specific Activity The inspectors reviewed the licensees sample and analysis data used to report the RCS specific activity performance indicator for the third and fourth quarters 2001 against the applicable criteria specified in NEI 99-02, Rev 1.
b. Findings
No findings of significance were identified.
4OA6 Meetings
Exit Meeting Summary
On March 13, 2002, the inspectors presented the inspection results to Mr. R. Barrett and other Entergy staff members who acknowledged the inspection results presented. The inspectors confirmed with Entergy personnel that no materials evaluated during the inspection were considered proprietary.
ATTACHMENT 1
SUPPLEMENTAL INFORMATION
a. Key Points of Contact
- R. Barrett Vice President, Operations - IP3
J. Barry Senior Radiological Engineer
R. Burroni I&C Manager
R. Cavalieri Site Planning and Outage Scheduling Manager
- J. Comiotes Director, Safety Assurance
J. DeRoy General Manager of Plant Operations
R. Deschamps RES Department Manager
J. Donnelly Licensing Manager
M. Gillman Operations Manager
G. Hocking HP Supervisor
B. Kyler ALARA Specialist
R. LaVera Senior Radiological Engineer
K. Peters Corrective Actions and Assessment Manager
- M. Smith Director, IP3 Engineering
A. Vitale Maintenance Manager
R. Solano HP Supervisor
J. Stewart HP Supervisor
C. Welling Radiation Protection and Chemistry Manager
b. List of Items Opened, Closed, and Discussed
Opened
None
Closed
None
Opened/Closed
None
c. List of Acronyms
ACTS Action Commitment Tracking System
ADAMS Agency Document Access and Management System
AFW auxiliary feedwater system
ALARA As Low As Reasonably Achievable
CFR Code of Federal Regulations
COL check-off list
CVCS chemical and volume control system
DC direct current
DCP design change package
DER Deviation/Event Report
EDG emergency diesel generator
FP fire protection
FCU fan cooler unit
HP Health Physics
I&C Instrumentation and Controls
IPM Installed Personnel Monitor
IR inspection report
IVSWS isolation valve seal water system
LCO limiting condition for operation
NEI Nuclear Energy Institute
NFPA National Fire Protection Association
NRC Nuclear Regulatory Commission
OPDT Over-Pressure Delta Temperature
OS Occupational Radiation Safety
OTDT Over-Pressure Delta Temperature
PAB primary auxiliary building
PFP Pre-Fire Plan
PI performance indicator
PMT post-maintenance test
RCA Radiologically Controlled Area
RCP Reactor Coolant Pump
RWP Radiation Work Permit
SGBD steam generator blowdown
SI safety injection
SOP system operating procedure
SSCs structures, systems, and components
TS technical specifications
WR work request