IR 05000244/2004002
| ML041170041 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 04/23/2004 |
| From: | James Trapp Reactor Projects Branch 1 |
| To: | Mecredy R Rochester Gas & Electric Corp |
| References | |
| IR-04-002 | |
| Download: ML041170041 (36) | |
Text
April 23, 2004
SUBJECT:
R. E. GINNA NUCLEAR POWER PLANT - NRC INTEGRATED INSPECTION REPORT 05000244/2004002
Dear Dr. Mecredy:
On March 31, 2004, the US Nuclear Regulatory Commission (NRC) completed an inspection at your R. E. Ginna facility. The enclosed integrated inspection report documents the inspection findings, which were discussed on April 14, 2004, with Mr. Joe Widay and other members of your staff.
The inspection examined activities conducted under your license as they relate to safety and compliance with the Commissions rules and regulations and with the conditions of your license.
The inspectors reviewed selected procedures and records, observed activities, and interviewed personnel.
This report documents two findings of very low safety significance (Green). Both of these findings were determined to involve a violation of NRC requirements. However, because of their very low safety significance, and because they have been entered into your corrective action program, the NRC is treating these issues as non-cited violations, in accordance with Section VI.A.1 of the NRCs Enforcement Policy. If you deny the non-cited violations noted 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, D.C. 20555-0001; with copies to the Regional Administrator, Region I; the Director, Office of Enforcement, United States Nuclear Regulatory Commission, Washington, D.C. 20555-0001; and the NRC Resident Inspector at the Ginna facility.
Since the terrorist attacks on September 11, 2001, the NRC has issued five Orders and several threat advisories to licensees of commercial power reactors to strengthen licensee capabilities, improve security force readiness, and enhance controls over access authorization. In addition to applicable baseline inspections, the NRC issued Temporary Instruction (TI) 2515/148,
"Inspection of Nuclear Reactor Safeguards Interim Compensatory Measures," and its subsequent revision, to audit and inspect licensee implementation of the interim compensatory measures required by the order. Phase one of TI 2515/148 was completed at all commercial power nuclear power plants during calendar year 2002 and the remaining inspection activities for Ginna were completed in August 2003. The NRC will continue to monitor overall safeguards and security controls at Ginna.
Dr. Robert In accordance with 10 CFR 2.390 of the NRCs 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 management system (ADAMS). ADAMS is accessible from the NRC Website at http://www.nrc.gov/reading-rm/adams.html (the Public Electronic Reading Room).
Sincerely,
/RA/
James M. Trapp, Chief Projects Branch 1 Division of Reactor Projects Docket No. 50-244 License No. DPR-18
Enclosure:
Inspection Report 50-244/04-02 w/ Attachment: Supplemental Information
REGION I==
Docket No:
50-244 License No:
DPR-18 Report No:
50-244/04-02 Licensee:
Rochester Gas and Electric Corporation (RG&E)
Facility:
R. E. Ginna Nuclear Power Plant Location:
1503 Lake Road Ontario, New York 14519 Dates:
January 1, 2004 - March 31, 2004 Inspectors:
K. Kolaczyk, Senior Resident Inspector M. Marshfield, Resident Inspector P. Frechette, Physical Security Inspector E. Knutson, Resident Inspector, Nine Mile Point J. McFadden, Health Physicist N. Perry, Senior Project Engineer D. Silk, Senior Emergency Preparedness Inspector Approved by:
James M. Trapp, Chief Projects Branch 1 Division of Reactor Projects
Enclosure ii
SUMMARY OF FINDINGS
Summary of Findings (contd)
iii IR 05000244/2004002; 01/01/2004 - 03/31/2004; R. E. Ginna Nuclear Power Plant; Operator Performance During Non Routine Events, Post-Maintenance Testing.
The report covered a 3-month period of inspection by resident inspectors and announced inspections by regional specialists. This inspection identified two Green findings that were also non-cited violations (NCVs). The significance of most findings is indicated by their color (Green, White, Yellow, Red) using Inspection Manual Chapter (IMC) 0609, Significance Determination Process (SDP). Findings for which the SDP does not apply may be Green or be assigned a severity level after NRC management review. The NRCs program for overseeing the safe operation of commercial nuclear power reactors is described in NUREG-1649,
Reactor Oversight Process, Revision 3, dated July 2000.
NRC-Identified and Self-Revealing Findings
Cornerstone: Emergency Preparedness
- Green.
The inspectors identified a Green non-cited violation (NCV) of 10 CFR 50.47(b)(2) when after the declaration of an Unusual Event (UE) on February 16, 2004,
RG&E did not augment the shift crew with a 30-minute Radiation Protection Technician (RPT) responder in a timely manner. The shift crew delayed notification of this responder for 30 minutes. Once the notification was initiated, only one RPT responded to the site, and arrived 62 minutes after the UE declaration was made, instead of the required 30 minutes.
This finding is associated with the Emergency Response Organization Readiness attribute of the emergency preparedness (EP) cornerstone. It is greater than minor because it impacts the objective to ensure that RG&E is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The EP Significance Determination Process (SDP) was used to assess this performance. (Section 1R14)
Cornerstone: Containment Barriers:
- Green.
The inspectors identified a Green NCV of 10 CFR 50, Appendix B, Criterion XVI when they identified that RG&E did not implement effective corrective actions to ensure that supports for valves in intermediate building sample hood area were properly installed. Degraded supports in the sample hood area had previously been identified by the NRC in November 2001. A subsequent inspection of the area by RG&E personnel identified other seismic-related deficiencies, one of which rendered a containment penetration inoperable.
This finding is associated with the Design Control attribute of the barrier integrity cornerstone. It is greater than minor because it affected the objective of maintaining containment integrity during seismic events. The issue is of very low safety significance because it did not represent a degradation of the radiological barrier function provided for the control room, auxiliary building, or the spent fuel or standby gas treatment
Summary of Findings (contd)
iv system. The finding did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere. Additionally, the finding did not represent an actual open pathway in the physical integrity of reactor containment or an actual reduction of the atmospheric pressure control function of the reactor containment.
(Section 1R19)
Licensee-Identified Violations
None.
REPORT DETAILS
Summary of Plant Status
Ginna operated at 100 percent power for the entire report period.
REACTOR SAFETY
Cornerstones: Initiating Events, Mitigating Systems, and Barrier Integrity
1R01 Adverse Weather Protection
a. Inspection Scope
During the week of January 4, 2004, the Ginna site experienced unseasonably cold weather with daytime high temperatures in the single digits. On January 9, 2004, temperatures in some areas of the auxiliary building were less than 45 degrees F.
During this week, while touring the plant, the inspectors paid particular attention to the temperatures in plant areas that contained equipment and systems that were susceptible to cold temperatures. Areas of focus were the battery rooms, intake structure, auxiliary building, and standby auxiliary feedwater pump room. The inspectors verified that temperatures in these room did not decease below the values outlined in the plant Updated Final Safety Analysis Report (UFSAR).
b. Findings
No findings of significance were identified.
1R04 Equipment Alignment
a. Inspection Scope
Partial System Walkdowns. (71111.04Q - 4 samples)
On two separate occasions, the inspectors completed walkdowns of the A and B diesel generators when the opposite diesel generator was out of service for maintenance activities. The conditions of the A and B diesel generators were examined, because of their high risk significance. The inspection reviewed the alignment of system valves and electrical breakers to ensure proper in-service or standby configurations as described in plant procedures and drawings. During the walkdown, the inspectors evaluated material conditions and general housekeeping of the system and adjacent spaces. The inspectors also verified that operations personnel were following plant Technical Specifications (TS).
The A and B motor-driven auxiliary feedwater trains were walked down while maintenance was being conducted on the turbine-driven auxiliary feedwater system.
These trains were examined because of their high risk significance. The inspection reviewed the alignment of the system valves and electrical breakers to ensure proper in-service and standby configurations were in place during the maintenance as described in plant procedures and drawings. The inspectors evaluated material conditions and general housekeeping of the system and adjacent spaces. The inspectors verified that operations personnel were following plant TS.
The safety injection (SI) A and C trains were walked down while the B SI pump was out of service for planned maintenance. These trains were examined because of their high risk significance. The inspection reviewed the alignment of the train valves and electrical breakers to ensure proper in-service and standby configurations were in place during maintenance as described in plant procedures and drawings. The material conditions and general housekeeping of the trains and adjacent spaces were examined as part of the inspection. The inspectors verified that operations personnel were following plant TS.
Complete System Walkdown. (71111.04S - 1 sample)
The inspectors conducted a detailed walkdown of the alignment and condition of the standby auxiliary feedwater system. The standby auxiliary feedwater system was chosen because of the important role it would play to provide makeup water to the steam generators if a high energy line break occurred in the intermediate building. In addition to verifying proper system alignment as required by plant TS, the plant (UFSAR), and RG&E procedures and drawings, the inspector reviewed system maintenance and action reports. None of the action reports or maintenance work orders indicated the performance/ reliability of the system had declined.
b. Findings
No findings of significance were identified.
1R05 Fire Protection
a. Inspection Scope
Using the Ginna Fire Protection program documents as a guide, the inspectors performed walkdowns of the following fire areas to determine if there was adequate control of transient combustibles and ignition sources. The material condition of fire protection systems, equipment and features, and the material condition of fire barriers were also inspected against industry standards. In addition, the passive fire protection features were inspected, including the ventilation system fire dampers, structural steel fire proofing, and electrical penetration seals. The following plant areas were inspected:
- Air Handling Room
- Relay Room
- Charging Pump Room
- Battery Room A
- Battery Room B
- Standby Auxiliary Feedwater Pump Room
- Screenhouse b
Findings No findings of significance were identified.
1R06 Flood Protection Measures
a. Inspection Scope
To evaluate RG&Es internal flood protection measures for the auxiliary building sub-basement area, the inspectors reviewed the Ginna UFSAR, and Probabilistic Safety Assessment. The inspectors toured the sub-basement, inspected the Residual Heat Removal (RHR) pipe trench area, and reviewed what actions RG&E has done to date to correct ongoing water leakage from the pipe trench into the sub-basement area. RG&E documented this leakage in Action Report (AR) 2004-0010, Ground Water in RHR Sub-basement. Leakage into the sub-basement from the pipe trench was a concern, since it could result in a common cause failure of the RHR pumps. The inspectors discussed the concerns with engineering personnel, and verified that drain piping and configurations were as detailed on Piping and Instrument Drawings (P&ID) 33013-1272 sheet 2 and 33013-1270 sheet 1.
b. Findings
No findings of significance were identified.
1R07 Heat Sink Performance
a. Inspection Scope
The inspector reviewed RG&Es periodic maintenance, testing, and inspection records for the spent fuel pool B heat exchanger. This review was done to determine if RG&E had reasonable assurance that the heat exchangers heat transfer capability would be adequate to meet its design heat removal requirements during plant operations.
As a part of the review, the inspector observed and discussed the test results of the eddy current inspection with the system engineer for the service water system. The inspector reviewed the service water program document, Service Water System Reliability Optimization Program, and applicable sections of the plant UFSAR.
b. Findings
No findings of significance were identified.
1R11 Licensed Operator Requalification Program
a. Inspection Scope
On January 26, 2004, the inspectors observed a licensed operator requalification training simulator scenario. The test observed was scenario ECA1112-12, LOCA Outside Containment. The inspectors reviewed the critical tasks associated with the scenario, observed the operators performance, and observed the post-evaluation critique. The crew did not properly diagnose the scenario, and as a result, had to be retrained before assuming shift duties in accordance with the Ginna licensed operator training program. Through conversations with training personnel, the inspector verified that the operators received the requisite refresher training prior to returning to shift duties. The inspectors also reviewed and verified compliance with Ginna procedure OTG-2.2, Simulator Examination Instructions.
b. Findings
No findings of significance were identified.
1R12 Maintenance Effectiveness
a. Inspection Scope
The inspectors evaluated RG&Es work practices and follow-up corrective actions for selected system, structure, or component (SSC) issues to assess the effectiveness of RG&Es maintenance activities. The inspectors reviewed the performance history of those SSCs and assessed RG&Es extent of condition determinations for those issues with potential common cause or generic implications to evaluate the adequacy of RG&Es corrective actions. The inspectors reviewed RG&Es problem identification and resolution actions for these issues to evaluate whether RG&E had appropriately monitored, evaluated, and dispositioned the issues in accordance with RG&E procedures and the requirements of 10 CFR 50.65, Requirements for Monitoring the Effectiveness of Maintenance. In addition, the inspectors reviewed selected SSC classification, performance criteria and goals, and RG&Es corrective actions that were taken or planned, to verify whether the actions were reasonable and appropriate. The following equipment issues were reviewed:
- The Overpower/ Differential Temperature Alarm on the main control board occasionally alarms repeatedly, reportedly due to hot leg streaming in the reactor coolant system at the outlet from the reactor. Actions are being taken to minimize nuisance alarms, while still ensuring proper operator response.
- The Engineered Safety Features Actuation System has been initiated at a greater frequency than the current plant level performance criteria of two in two cycles. The system was moved to maintenance rule A1 status, and will be monitored to see if plant procedure changes with regard to operation of the 751 line at various times during start-up and high wind conditions will reduce the challenges to this system.
b. Findings
No findings of significance were identified.
1R13 Maintenance Risk Assessments and Emergent Work Control
a. Inspection Scope
The inspectors evaluated the effectiveness of RG&Es maintenance risk assessments required by paragraph a(4) of 10 CFR 50.65. This inspection included discussions with control room operators and scheduling department personnel regarding the use of RG&Es online risk monitoring software. The inspectors reviewed equipment tracking documentation and daily work schedules, and performed plant tours to gain reasonable assurance that actual plant configuration matched the assessed configuration.
Additionally, the inspectors verified that RG&Es risk management actions, for both planned and/ or emergent work, were consistent with those described in procedure IP-PSH-2, "Integrated Work Schedule Risk Management." Risk assessments for the following out-of-service systems, structures, and/ or components were reviewed:
- Unplanned maintenance on the B diesel generator conducted on January 26, 2004, due to low lubricating oil temperature caused by a failed coil in the heating circuit.
- Repair of charging pump C, outlet relief valve isolation valve 279C, which was repaired on February 2, 2004. This work required a freeze seal and was conducted as a significant infrequently performed evolution.
- Installation of the conduit in support of the ventilation system permanent modification for the control room required significant scaffolding work in both battery rooms and the relay room. Control of the work, as well as establishment of protected system functions during work in the opposite battery room was monitored and observed.
- Planned maintenance on the A condensate storage tank (CST) that began on March 29, 2004. The maintenance involved draining the tank to facilitate installation of a new diaphragm that will be used to limit the amount of air entrainment in the CST water.
b. Findings
No findings of significance were identified.
==1R14 Operator Performance During Non-routine Plant Evolutions and Events (71111.14 - 1 sample)a.
==
Inspection Scope On February 16, 2004 at 1:40 a.m., the service air compressor failed because of a phase to phase short in an electrical contact located in the compressors motor controller. Because the failure caused a cover on the compressor controller to be dislodged, operators determined an electrically-induced explosion probably had occurred. At 2:25 a.m., operators declared an Unusual Event (UE) and activated the site emergency response plan in accordance with the Ginna emergency plan implementing procedures (EPIPs). At 2:58 a.m., Ginna operators informed the NRC operations officer of the event.
There was no warning of the impending compressor failure. Operators were alerted to the problem when they received an "instrument air trouble alarm" on the main control board. At the time of the event, the service air compressor was supplying both the instrument and service air systems. Upon loss of the service air compressor, the "A" and "B" instrument air compressors automatically started.
Ginna has a substantial number of compressed air sources that are all cross-connected.
In the turbine building, there are three instrument and one service air compressors.
Outside of the turbine building is a diesel-driven air compressor. Both the "C" instrument air compressor and the service air compressor, can supply all of the service air and instrument air loads.
Operator response to the event included removing electrical power to the service air compressor, and starting the C instrument air compressor. The A and B compressors were then returned to their normal standby status. Ginna terminated the UE at 4:20 a.m.
Upon notification of the UE, the inspector responded to the site and verified the failure did not adversely affect plant operations.
b. Findings
Introduction.
A Green Non Cited Violation (NCV) of 10 CFR 50.47(b)(2) was identified when the inspectors noted after the declaration of the UE on February 16, 2004, RG&E did not augment the shift crew with a 30-minute radiation protection technician (RPT)responder in a timely manner. The shift crew delayed notification of this responder for 30 minutes. Once the notification was initiated, only one RPT responded to the site.
(The function of the RPT is to perform on-site surveys.) He arrived 62 minutes after the UE declaration was made instead of the required 30 minutes.
Description.
The shift supervisor declared a UE at 2:25 a.m. EPIP 1-5 Notifications directed RG&E personnel to implement Attachment 1, Unusual Event Notifications.
The first step in this attachment directed the user to select a computer icon to activate the 30-minute RPT responders. However, the control room crew implementing this procedure hesitated and questioned the need to notify an RPT for a non-radiological event even though it was required by EPIP 1-5. Accordingly, the shift crew skipped the first step and made notifications to New York State and local counties, which were completed by 2:37 a.m. At about 2:55 a.m., it was decided to initiate notification of the RPTs.
Although there are 17 RPTs that were available to fill this 30-minute responder position, only one RPT responded to the notification, and he arrived at the site guard house at 3:27 a.m. During previous unannounced call-in drills conducted during waking hours, and during the August 14, 2003, event, more than one RPT responded to the site.
However for this notification, RG&E indicated that many of the RPTs were not in the same room as their pager or had it on "vibrate" and thus were not awakened to respond.
To activate the 30-minute RPT responders pagers, a computer icon at the shift RPTs desk must be selected to activate the notification process. Although the Emergency Preparedness (EP) staff understood that either the shift RPT or the control room operators could activate the icon, EPIP 1-5 did not specifically state who was to activate this process. Further, EPIP 1-5 did not require the RPT to immediately implement any procedure upon a declaration of a UE or for an unplanned reactor trip. Instead the control room had to inform the RPT of the need for a call out. (In a May 23, 2003, letter to the NRC, RG&E committed to have 30-minute responders come to the site for a UE declaration or an unplanned reactor trip.) When reviewing this event, RG&E determined that their current call out process for RPT technicians was less than desirable since it required informing the shift RPT of a reactor trip, and the tech must remember to activate the icon for the response process. RG&E noted that a contributing factor to the late notification was the fact that after the requirement to call in an RPT was implemented at the end of July 2003, there was limited opportunity available to practice and reinforce this requirement due to an outage and the holiday season.
RG&Es search of the lesson plans for the shift RPTs and the shift communicators did not find any revisions to the lesson plans for implementing the new requirement. The change to the procedure was covered by reactor physics and auxiliary operator instructors in the simulator during scenarios that involved classified events. However, the simulator computers were not configured with a training icon that could be selected to simulate the notification process.
Analysis.
The untimely notification affects the EP cornerstone. It was determined to be more than minor because it is associated with the emergency response organization readiness and performance attributes of the EP cornerstone, and impacts the objective to ensure that RG&E is capable of implementing adequate measures to protect the health and safety of the public in the event of a radiological emergency. The EP Significance Determination Process (SDP) was used to assess this performance. Sheet 1 of the EP SDP (Failure to Comply) was applied for failure to comply with an NRC requirement and the issue was determined to be Green. Specifically, because of the programmatic issues stated above, the 30-minute RPT responded in an untimely manner during non-waking hours. This was a failure to comply with 10 CFR 50.47(b)(2)which states in part that "timely augmentation of response capabilities is available... "
This RG&E performance issue indicated a problem with planning standard 10 CFR 50.47(b)(2). This is not a risk-significant planning standard. Based upon the examples provided in Section 4.2 of Appendix B of Manual Chapter 0609, Significance Determination Process, the programmatic deficiencies did not result in more than one response function being unmet. (The 30-minute RPT only fulfills the function of performing in-plant surveys.) Therefore, this issue does not result in a planning standard function failure and thus is a finding of Green significance.
Enforcement.
10 CFR 50.47(b)(2) states, in part, that "timely augmentation of response capabilities is available... " Also, Section 5.2 of the Ginna Station Emergency Response Plan states that an "RP or Rad/Chem qualified technician will be activated to respond within 30 minutes of the declaration of an Unusual Event..." Contrary to the above, due to programmatic lapses in procedural clarity, training, and expectations for responders, an RPT responded in an untimely manner following the declaration of a UE at 2:25 a.m. on February 16, 2004. The RPTs were notified at 2:55 a.m. and the first (and only) responder arrived at the site at 3:27 a.m. This issue is not subject to traditional enforcement. For this self-revealing issue, RG&E took prompt corrective actions and documented these actions in their corrective action program under AR 2004-0509, 30 Minute Radiation Protection Response. According to the EP SDP, a failure to comply with planning standard 10 CFR 50.47(b)(2) that results in less than two functions being not filled is determined to be of Green significance. This violation is being treated as an NCV consistent with Section VI.A of the NRC Enforcement Policy:
NCV 05000244/2004-01, Failure to Correctly Implement the Site Emergency Plan.
1R15 Operability Evaluations
a. Inspection Scope
The inspectors reviewed operability determinations to verify that the operability of systems important to safety was properly established, that the affected components or systems remained capable of performing their intended safety functions, and that no unrecognized increase in plant or public risk occurred. In addition, the inspectors reviewed the following operability evaluations to determine if system operability was properly justified in accordance with IP-CAP-1.1, Technical Evaluation for Current Operability and Past Operability Determination Worksheet:
- Action Report (AR) 2004-0174, Packing Gland on MOV-9704B is Uneven
- AR 2004-0529, AOV-508 Stroke Time is Outside Code Limit
- Temporary Cooling Medium to Control Room A/C Water Chillers IAW T-36.4
- AR 2004-0882, Valve 1817, C Safety Injection Pump Suction Relief, Excess Vibration
- AR 2004-0903, Some Installed Splices Do Not Meet the Tested Configuration
b. Findings
No findings of significance were identified.
1R19 Post-Maintenance Testing
a. Inspection Scope
The inspectors reviewed post-maintenance tests for the following work orders (WO) to verify that RG&E appropriately demonstrated the components ability to perform their intended safety function as described in the plant UFSAR.
- WO 2003-3341 Set Up Spare Breaker to be Used as a Substitute for 52/CSP1B (Containment Spray Pump B Bkr)
- WO 2004-040 Packing Leak on North Side of Packing Gland
- WO 2003-3041, Boric Acid Leak Between V-956H and 956E
- WO 2003-2064, Inspect, Clean Lubricate D/G Room A Backdraft Dampers
- WO 2003-2198, Change Grease in Limitorque Valve 9701A and Retest
b. Findings
Introduction.
A Green NCV of 10 CFR 50, Appendix B, Criterion XVI was identified by the inspectors when they noticed that RG&E did not implement effective corrective action to ensure air-operated sample isolation valves located in the intermediate building sample hood area were adequately supported. An inadequately supported valve in this area was identified by the NRC in November 2001.
Description.
The pressurizer liquid sample line is located in the intermediate building sample hood area (a location in the plant which contains a substantial number of air-operated sample isolation valves.) While observing RG&E personnel performing a post-maintenance leak test on a section of tubing for the pressurizer liquid sample line as outlined in WO 2003-3041, the inspector noted that several sections of tubing to the valves in the area appeared to be missing supports or were not well-supported. The inspector discussed this observation with RG&E personnel who performed a detailed walkdown of the area. During this walkdown, RG&E personnel identified numerous other support-related discrepancies, one of which was missing bolts on a support for the pressurizer liquid sample line between valves 966B and 956E, which rendered that containment penetration inoperable.
The inspector noted that the adequacy of the seismic supports in this area had already been identified as an issue by the NRC in November 2001. During a Problem Identification and Resolution inspection, the NRC identified that valve AOV-966C was not properly supported.
Analysis.
The performance deficiency in this event was a failure of RG&E to implement effective corrective action following the November 2001 NRC inspection and thoroughly examine other supports in the intermediate building sample hood area - which is an approximately four by 14 foot enclosure. This finding is greater than minor because it is associated with the Design Control (Structural Integrity) attribute of the barrier integrity cornerstone, and it could adversely affect the cornerstone objective because containment integrity may be adversely impacted following a significant seismic event.
In accordance with Manual Chapter 0609, Appendix A, Significance Determination of Reactor Inspection Findings for At-Power Situations, the inspectors conducted an SDP Phase 1 screening and determined that the finding is of very low safety significance (Green). Using the Reactor Safety SDP, this finding screened directly to Green since the deficiency did not represent a degradation of the radiological barrier function provided for the control room, auxiliary building, or the spent fuel or standby gas treatment system. The finding did not represent a degradation of the barrier function of the control room against smoke or a toxic atmosphere. Additionally, the finding did not represent an actual open pathway in the physical integrity of reactor containment or an actual reduction of the atmospheric pressure control function of the reactor containment.
Enforcement.
10 CFR 50, Appendix B, Criterion XVI, Corrective Action states, in part, that measures shall be established to assure that conditions adverse to quality be promptly identified and corrected. Contrary to 10 CFR 50 Appendix B, Criterion XVI, RG&E did not implement effective corrective action to ensure the supports for the remaining AOVs in the intermediate building sample hood area were correctly installed and maintained, following a November 2001 NRC inspection that identified AOV 966C was not adequately supported, until additional degraded supports in the area were identified by the NRC. However, because the issue was of very low safety significance, and has been entered into the RG&E corrective action program as Action Report 2004-0420, Support Needs Bolts, this issue is being treated as a NCV, consistent with section V1.A of the NRC Enforcement Policy: NCV 05000244/2004002-01, Seismic Support Not Operable.
1R22 Surveillance Testing
a. Inspection Scope
The inspectors witnessed the performance and/ or reviewed test data for the following surveillance tests that are associated with selected risk-significant systems, structures, and components (SSCs) to verify that TS were followed, and that acceptance criteria were properly specified. The inspectors also verified that proper test conditions were established as specified in the procedures, that no equipment preconditioning activities occurred, and that acceptance criteria had been met.
- M-1306, Ginna Station Material Condition Inspection Program, performed on January 2, 2004.
- CME-38-12-STABATTMON, Station Battery Monthly Performance Tests, performed on January 5, 2004.
- PT-36Q-D, Standby Auxiliary Feedwater Pump D - Quarterly, performed on January 8, 2004.
- PT-12.1, Emergency Diesel Generator A, performed on January 13, 2004.
- PT-32B, Reactor Trip Breaker Testing - B Train, performed on January 16, 2004.
b. Findings
No findings of significance were identified.
1R23 Temporary Plant Modifications
a. Inspection Scope
The following temporary modifications were reviewed by the inspectors to verify they were installed in conformance with the instructions contained in procedure IP-DES-3, Temporary Modifications:
- 2003-0024, RTD-1 Terminal Block Changes in Rack R1 for Failed T-hot Resistance Temperature Detector RTD TE-401A
- 2004-0002, 1B MSR Second Phase Drain Thermocouple Well Leak Repair
b. Findings
No findings of significance were identified.
Cornerstone: Emergency Preparedness
1EP4 Emergency Action Level and Emergency Plan Changes
a. Inspection Scope
An in-office inspection that reviewed recent changes to emergency plan implementing procedures was conducted on March 3-5, 2004. A thorough review was conducted for documents related to the risk significant planning standards (RSPS) and a general review was completed for non-RSPS documents. The review verified that the changes satisfied the standards of 10 CFR 50.54(q), 10 CFR 50.47(b), the requirements of 10 CFR 50 Appendix E, the intent of NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response Plans and Preparedness in Support of Nuclear Power Plants," and verified that the changes did not decrease the effectiveness of the plan. These changes are subject to future NRC inspections to ensure that as a result of these changes the emergency plan continues to meet NRC regulations.
b. Findings
No findings of significance were identified.
1EP6 Drill Evaluation
a. Inspection Scope
On January 26, 2004, the inspectors observed a licensed operator simulator training scenario that included a limited test of the Ginna emergency response plan. Scenario ECA1112-12, LOCA Outside of Containment, was observed. During the exercise, the crew did not classify the event in a timely manner, and as a result, this drill was counted as a failure in the Ginna Drill/Exercise Performance performance indicator.
b. Findings
No findings of significance were identified.
RADIATION SAFETY
Cornerstone: Occupational Radiation Safety
2OS1 Access Control to Radiologically Significant Areas (71121.01 - 7 samples)
a. Inspection Scope
The inspector reviewed radiological work activities and practices, and procedural implementation during observations and tours of the facilities, and inspected procedures, records, and other program documents to evaluate the effectiveness of Ginnas access controls to radiologically significant areas. This inspection activity represents the completion of seven
- (7) samples relative to this inspection area (i.e.,
inspection procedure sections 02.01, 02.03.a, b, c, and d, and 02.05.a and b) in partial fulfillment of the annual inspection requirements.
Inspection Planning (02.01)
The inspector verified that there were no Performance Indicator events for the Occupational Exposure Cornerstone which required follow-up.
Problem Identification and Resolution (02.03.a, b, c, and d)
During this inspection, the inspector reviewed RG&Es self-assessment activities for any results related to the access control program since the last inspection. The intent of this review was to determine if identified problems were entered into the corrective action program for resolution. The inspector also reviewed corrective action reports related to access controls and included in this review any high radiation area radiological events that have occurred since the last inspection in this area. The inspector discussed the corrective action reports with several members of the radiological protection staff to determine that the follow-up activities were being conducted in an effective and timely manner commensurate with their importance to safety and risk. The self-assessments, conducted since the last inspection, did not cover health physics access controls.
However, the inspector noted that numerous Quality Assurance (QA) surveillance reports, performed since the last inspection in this area and during the last refueling outage, did address health physics access controls and the type of deficiencies identified in the corrective action reports. As noted above, there were no RG&E Performance Indicator events or documentation packages for the Occupational Exposure Cornerstone which required review.
High Risk Significant, High Dose Rate HRA and VHRA Controls (02.05.a and b)
During this week of inspection, the inspector met at various times with the Radiation Protection Manager and the Senior Health Physicist for Radiation Protection (RP)
Operations and discussed the controls and procedures for high-dose-rate high radiation areas (HRAs) and for very high radiation areas (VHRAs). The inspector reviewed the subject procedures (as listed in the List of Documents Reviewed section) to verify that the level of worker protection was adequate.
Related Activities On February 24, 2004, the inspector observed Radiologically-Controlled Area (RCA)entries and exits being made by radiation workers at the primary 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 inspector toured various elevations in the intermediate, auxiliary, and contaminated storage buildings inside the primary RCA to verify the adequacy of the radiological controls which were being implemented. The inspector reviewed observed work activities for compliance with the radiation work permit (RWP) requirements. During these observations and tours, the inspector reviewed, for regulatory compliance, the posting, labeling, barricading, and level of radiological access control for locked high radiation areas (LHRAs), high radiation areas (HRAs), radiation and contamination areas, and radioactive material areas. Also, on February 24, the inspector examined satellite RCAs including sea/ land containers, the upper radioactive waste storage building, the dry radioactive waste storage area, the old steam generator storage building, a storage room for sealed packages of radioactive material, a warehouse used for storage of a spare reactor coolant pump, the lab for processing thermoluminescent dosimeters, and the radiation instrument calibration facility. On February 25, 26, and 27, the inspector observed the morning turnover meetings for the Health Physics (HP)technicians.
The inspector performed a selective examination of documents (as listed in the List of Documents Reviewed section) to evaluate the adequacy of radiological controls. The review in this area was against criteria contained in 10 CFR 19.12, 10 CFR 20 (Subparts D, F, G, H, I, and J), Technical Specifications, and procedures.
b. Findings
No findings of significance were identified.
2OS2 ALARA Planning and Controls (71121.02 - 2 samples)
a. Inspection Scope
The inspector reviewed the effectiveness of RG&Es program to maintain occupational radiation exposure as low as is reasonably achievable (ALARA). This inspection activity represents the completion of two
- (2) samples relative to this inspection area (i.e.,
inspection procedure sections 02.03.a and b) in partial fulfillment of the biennial inspection requirements.
Verification of Dose Estimates and Exposure Tracking Systems (02.03.a and b)
On February 26, 2004, the inspector met with the Senior Health Physicist, RP Operations, and the Lead ALARA Technician. During this meeting, the inspector reviewed the assumptions and basis for the current annual collective exposure estimate including that for normal operations. The inspector also reviewed the applicable ALARA procedures used to determine the methodology for estimating work-activity-specific exposures and the intended dose outcome. The inspector also reviewed RG&Es method for adjusting exposure estimates, or re-planning work, when unexpected changes in scope or emergent work are encountered.
Related Activities On February 25, 2004, the inspector observed a meeting of the station ALARA committee. Discussion topics included a review of the year-to-date dose, the dose estimate for 2004, and current dose challenges.
The inspector performed a selective examination of documents (as listed in the List of Documents Reviewed section) for regulatory compliance and for adequacy of control of radiation exposure. The review was against criteria contained in 10 CFR 20.1101 (Radiation protection programs), 10 CFR 20.1701 (Use of process or other engineering controls), and procedures.
b. Findings
No findings of significance were identified.
2OS3 Radiation Monitoring Instrumentation and Protective Equipment (71121.03 - 1 sample)
a. Inspection Scope
The inspector reviewed the program for health physics instrumentation to determine the accuracy and operability of the instrumentation. This inspection activity represents the completion of one
- (1) sample relative to this inspection area (i.e., inspection procedure section 02.04.a) in partial fulfillment of the biennial inspection requirements.
Problem Identification and Resolution (02.04.a)
The inspector reviewed RG&E self-assessments, audits, and ARs since the last inspection in this area to focus on radiological incidents that involved personnel-contamination-monitor alarms due to personnel internal exposures. The reviewed documents did not identify any recordable internal exposures.
Related Activities As mentioned in Section 2OS1, on February 24, 2004, the inspector examined the lab used for processing thermoluminescent dosimeters and the radiation instrument calibration facility.
The inspector performed a selective examination of documents (as listed in the List of Documents Reviewed section) for regulatory compliance and adequacy in this area.
The review was against criteria contained in 10 CFR 20.1501, 10 CFR 20 Subpart H, Technical Specifications, and procedures.
b. Findings
No findings of significance were identified.
OTHER ACTIVITIES
4OA1 Performance Indicator Verification
1. Initiating Events Cornerstone (71151 - 3 samples)
a. Inspection Scope
Using the criteria specified in Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Performance Indicator Guideline, Revision 2, the inspectors verified the completeness and accuracy of the following performance indicator data for unplanned scrams per 7,000 critical hours, scrams with loss of normal heat removal, and unplanned power changes per 7,000 critical hours for calendar year 2003. To verify the accuracy of the data, the inspector reviewed monthly operating reports, NRC inspection reports, and licensee event reports issued during calender year 2003.
b. Findings
No findings of significance were identified.
2. Safeguards Cornerstone (71151 - 3 samples)
a. Inspection Scope
The inspector performed a review of performance indicator (PI) data submitted by RG&E for the physical protection cornerstone. The review was conducted of RG&Es programs for gathering, processing, evaluating, and submitting data for the fitness-for-duty, personnel screening, and protected area security equipment PIs to verify these PIs had been properly reported as specified in Nuclear Energy Institute (NEI) 99-02, Regulatory Assessment Performance Indicator Guideline, Rev. 1 and Rev 2, to verify that all occurrences that met the NEI criteria were identified and reported as performance indicators.
The review included RG&Es tracking and trending reports, personnel interviews and security event reports for the PI data collected from the 1st quarter of 2003 through the 4th quarter of 2003. The inspector noted from RG&Es submittal that there were no reported failures to properly implement the requirements of 10 CFR 73 and 10 CFR 26 during the reporting period. This inspection activity represents the completion of three
- (3) samples relative to this inspection area; completing the annual inspection requirement.
b. Findings
No findings of significance were identified.
4OA2 Identification and Resolution of Problems
1. Main Steam Header Support System
a. Inspection Scope
In 1972, RG&E installed a temperature-compensated support system for the eight main steam safety valves at Ginna. The support system was intended to reduce high localized stresses, which Gilbert, the Ginna architect engineering firm, determined could be present in the steam header as a result of the blowing reaction of the valves.
Installation of the system was necessitated by an event that occurred at the Turkey Point Nuclear Station Unit 3, where three safety valves ruptured the main steam piping when they were actuated.
The support system consists of eight six-inch diameter schedule 80 pipe columns that are bolted at the base to a rigid support structure. At the top of each column is an anti-friction plate that abuts a similar plate that is attached to the base of each relief valve.
To ensure the columns expand vertically so the anti-friction plates remain in contact with each other, they are heated with steam from the main steam header. The vertical gap between the plates is monitored by an alarm system that measures the gap between the anti-friction plates, and the temperature differential between the steam header and column. If the gap between the plates becomes excessive or the temperature differential between the header and column exceeds a preset value, an alarm will annunciate in the control room and a local panel. The system is considered safety-significant, and the supports are examined in the Inservice Inspection (ISI) program.
Since startup from the October refuel outage, the inspectors noted several alarms on the monitoring system have been locked in. This condition was not investigated by RG&E until December 2003 when an action report (AR) was initiated by personnel in the operations department. Since initiation of that AR, RG&E personnel have initiated several other ARs that documented other discrepancies with the system. The inspector reviewed the following ARs that involved the steam header system to determine if they evaluated/ dispositioned the issue, considered extent of condition, and properly focused corrective actions to address the issue:
As part of this review, the inspector also interviewed several plant engineers and operations personnel.
b. Findings and Observations
After reviewing the corrective actions associated with the aforementioned ARs, the inspector determined that although RG&E personnel had addressed the immediate issues documented in the ARs, the inspector could not determine if the corrective actions were timely or commensurate with the importance of the system. This was due to the fact that there was limited system design information, and the information that did exist was silent regarding the relative importance of the system. For example, a June 5, 1972, system design report indicated that the supports were installed to ensure operability of the plant safety valves. This information would imply that if the support system was not functioning properly, the main steam safety valves may not be operable.
However, an RG&E design engineer informed the inspector that the supports were not required, since the piping adjacent to the main steam valves had been reinforced, which negated the need for the supports.
At the close of the report period, RG&E was in the process of determining whether the steam support system has to be functioning properly to ensure the main steam safety valves are operable. Once this issue has been resolved by RG&E, the inspector can determine if the main steam safety valves remained operable, when RG&E personnel were attempting to address the conditions identified in the ARs. This issue is unresolved pending completion of the RG&E review and will be tracked as URI (50-244/2004-02-03).
2. Annual Sample Review
a. Inspection Scope
The inspector selected the following ten issues identified in the Corrective Action Program (CAP) for detailed review: ARs 2003-2339, -2365, -2447, -2452, -2454, -2514,
-2588, -2589, -2660, and 2004-0570). The issues were associated with violations of radiological boundaries, workers signed in on the wrong RWP, RO-20 survey meters not source-checked, an unexpected dose rate alarm, and an unheard battery alarm on an electronic dosimeter. On February 26, 2004, the inspector met with the radiation protection manager to discuss these issues.
b. Findings and Observations
No findings of significance were identified. The documented reports for the issues were reviewed to ensure that the full extent of the issues was identified, an appropriate evaluation was performed, and appropriate corrective actions were specified and prioritized.
3. Continuous Corrective Action Review by Residents
a. Inspection Scope
As required by Inspection Procedure 71152, "Identification and Resolution of Problems, and in order to identify repetitive equipment failures or specific human performance issues for follow-up, the inspectors performed a daily screening of items entered into the RG&E corrective action program. This activity was accomplished by reviewing hard copies of each condition report, attending daily screening meetings, and accessing RG&Es computerized database.
b. Findings
No findings of significance were identified.
4. Cross Reference to Cross Cutting Issues Documented Elsewhere
Section 1R19 of the report describes a finding that is related to the Problem Identification and Resolution cross-cutting area. Specifically, following a November 2001 NRC inspection that identified a valve which was not adequately seismically supported, RG&E did not examine adjacent valves. As a result, additional degraded conditions involving similar valves were not identified by RG&E personnel until they were prompted by NRC inspectors to examine other valves.
4OA3 Event Follow-up
1. Security Event Follow-Up
a. Inspection Scope
On March 30, 2004, the inspectors observed RG&Es contractor personnel respond to an event involving a suspicious object outside of the protected area. The response was timely and appropriate for the situation. Notifications were made as required and the situation was resolved. No further follow-up or reporting of this event is required.
b. Findings
No findings of significance were identified.
2. (Closed) LER 05000244/2003006, Emergency Diesel Generator Start Resulting From
Loss of Off-Site Power Circuit 751 On November 13, 2003, with the plant at approximately 100% power and offsite electrical power split between the safeguards busses, high winds caused the loss of off-site circuit 751 when a tree branch fell across the line. Because of the electrical lineup that was in effect at the time of the event, the loss of circuit 751 de-energized safeguards busses 16 and 17, which caused the B diesel generator to automatically start, and resupply power to the busses. The remaining offsite power line-circuit 767-remained powered throughout the event. Operators subsequently transferred the loads from busses 16 and 17 to circuit 767, and shut down the diesel generator. On November 15, operators restored the normal electrical lineup when busses 16 and 17 were transferred back to circuit 751. This Licensee Event Report (LER) was reviewed by the inspectors, and no findings of significance were identified. RG&E documented the event in AR 2003-3067, Loss of Circuit 751. This LER is closed.
4OA5 Other Activities
Spent Fuel Material Control and Accounting at Nuclear Power Plants (Temporary Instruction 2515/154)
a. Inspection Scope
Phase I and Phase II of the instruction were completed during this inspection period.
Appropriate documentation was provided to NRC management as required.
b. Findings
No findings of significance were identified.
4OA6 Meetings, Including Exit
Periodically during the course of this inspection, the inspectors met with Ginna representatives to discuss certain aspects of the inspection. For example, on March 26, 2004, NRC regional inspectors summarized the preliminary results of an inspection of the Ginna stations performance indicator program for the safeguards cornerstone.
On April 14, 2004, the resident inspectors summarized the contents of this inspection report to Mr. Widay, and other members of his staff, who acknowledged the findings.
The inspectors returned any proprietary items, and verified that no proprietary information is presented in this inspection report.
ATTACHMENT:
SUPPLEMENTAL INFORMATION
KEY POINTS OF CONTACT
Licensee personnel
P. Bamford
Operations Manager
R. Biedenbach
Safety/Fire Coordinator
M. Flaherty
Nuclear Safety & Licensing Manager
B. Flynn
Primary Systems and Reactor Engineering Manager
J. Hotchkiss
Mechanical Maintenance Manager
R. Marchionda
Nuclear Assessment Department Manager
B. Mecredy
Vice President Nuclear Operations
R. Ploof
Scheduling Manager
R. Popp
Production Superintendent
J. Smith
Maintenance Superintendent
T. White
Balance of Plant Systems Engineering Manager
J. Widay
VP, Plant Manager
LIST OF ITEMS OPENED, CLOSED, AND DISCUSSED
Opened
- 05000244/2004002-03 URI Operability of the Main Steam Safety Valves is Not Known When Alarms Occur on The Main Steam Compensated Support System
Opened and Closed
- 05000244/2004002-01 NCV Health Physics Technicians Did Not Respond to The Site as Required by The E-Plan During an Event
- 05000244/2004002-02 NCV Failure to Implement Effective Corrective Action to Resolve Seismic Support Issues in The Intermediate Building Sample Hood Area.
Closed
- 05000244/2003-006 LER Emergency Diesel Generator Start Resulting From Loss of Off-Site Power Circuit 751 (Section 4OA3)
Discussed
NONE