IR 05000244/1999004
| ML17265A724 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 07/29/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17265A723 | List: |
| References | |
| 50-244-99-04, NUDOCS 9908030134 | |
| Download: ML17265A724 (24) | |
Text
U.S. NUCLEAR REGULATORYCOMMISSION
REGION I
License No.
DPR-18 Report No.
50-244/99-04 Docket No.
50-244 Licensee:
Facility Name:
Location:
Rochester Gas and Electric Corporation (RG8 E)
R. E. Ginna Nuclear Power Plant 1503 Lake Road Ontario, New York 14519 Inspection Period:
Inspectors May 17, 1999 through June 27, 1999 P. D. Drysdale, Senior Resident Inspector C. C. Osterholtz, Resident Inspector T. A. Moslak; Radiation Specialist Approved by:
M. G. Evans, Chief Projects Branch
Division of Reactor. Projects
'7908030i34 9'70729 PDR ADQCK 05000244 PDR
EXECUTIVE SUMMARY R. E. Ginna Nuclear Power Plant NRC Inspection Report 50-244/99-04 This integrated inspection included aspects of licensee operations, engineering, maintenance,
.
and plant support.
The report covers a 6-week period of resident inspection, and it includes the results of an announced in-office inspection by a regional specialist in radiation protection requirements.
~Oerationa Operations personnel responded well to an unanticipated automatic withdrawal of control rods during maintenance on the nuclear instrument current comparator drawer N-38. Work package preparation and review for this work activity was poor, as it did not anticipate the need for control rods to be placed in manual. Additionally, Instrumentation and Control technicians exhibited a knowledge weakness by indicating that the comparator drawer work would not impact control rod motion. (04.1)
Maintenance Observed maintenance activities were accomplished in accordance with procedural requirements.
The licensee's post maintenance testing was adequate to demonstrate the operability of equipment prior to it's return to service.
Test procedures contained'adequate details for accomplishing test requirements.
Testing was performed by knowledgeable
.personnel, and test instrumentation was properly calibrated.
(M1.1)
Offsite power circuit 767 cable replacement was successfully performed in a timely manner.
However, the licensee demonstrated poor work coordination when activities outside the protected area, which could adversely impact the availability of the in-service offsite power circuit, were not suspended until after work on circuit 767 was commenced and after being brought to the licensee's attention by the inspectors.
(M2.1)
The licensee took proper actions following discovery of the failed A-emergency diesel generator output breaker to troubleshoot, identify, and correct the breaker failure mechanism.
The licensee's broader corrective actions to review all breaker maintenance procedures for adequacy with the vendor was considered a good initiative. (M2.2)
Encnineering The licensee has made progress to address program weaknesses for total instrument uncertainty calculations on Improved Technical Specification related instruments.
(E8.1)
Licensee management-identified, after the fact, a series of unplanned exposure events that were the result of deficiencies in the implementation of radiological controls, contrary to Improved
'
Executive Summary (cont'd)
In accordance with the established corrective action process, the licensee conducted a root cause assessment and completed (or planned) appropriate corrective actions to prevent a recurrence.
However, the corrective actions following the individual events were not effective, in that they did not prevent recurrence of the subsequent unplanned exposure events.
In addition, the overall series of events represented an indifference to proper radiological controls and precautions on the part of the individual radiation workers, radiation protection technicians, and direct supervision involved in the events.
The violation of Improved Technical Specifications was non-cited.
(R8.1)
TABLEOF CONTENTS EXECUTIVESUMMARY..
TABLEOF CONTENTS
.IV I. Operations..
Conduct of Operations
.
01.1 General Comments..
01.2 Summary of Plant Status
Operator Knowledge and Performance 04.1 Inadvertent Rod Motion During l&CMaintenance...
.. 2
......3
..... 3
3
. 3 II. Maintenance..
M1 Conduct of Maintenance
.
M1.1 Maintenance and Surveillance Testing Activities M2 Maintenance and Material Condition of Facilities and Equipment......
M2.1 Replacement of Offsite Power Circuit 767 Cables
.
M2.2 Failure of Emergency Diesel Generator Output Breaker to Safeguards Bus 14 M8 Miscellaneous Maintenance Issues M8.1 (Closed) Violation 50-244/97-12-01: Failure to Follow Administrative Procedure Requirements
.......... 5 III. Engineering
.
E8 Miscellaneous Engineering Issues E8.1 (Updated) Inspector Follow-Up Item 50-244/98-13-01: Total Instrument Uncertainty Calculations Needed for Improved Technical Specification (ITS) Compliance................................
6
.6 IV. Plant Support..
R8 Miscellaneous Radiological Protection &Chemistry Issues R8.1 Unplanned Exposure Incidents.................
..7
~
0
~
V. Management Meetings..
X1 Exit Meeting Summary
..10
ATTACHMENTS Attachment 1 - Partial List of Persons Contacted
- Inspection Procedures Used
- Items Opened, Closed, and Discussed
- List of Acronyms Used
Re ort Details
Conduct of Operations'01.1 General Comments Ins ectionProcedure IP 71707 The inspectors observed plant operations to verify that the facilitywas operated safely and in accordance with licensee procedures and regulatory requirements.
This review included tours of the accessible areas of the facility. The inspectors conducted ongoing verifications of engineered safeguard feature (ESF) system operability, verifications of proper control room and shift staffing, verification that the plant was operated in conformance with the Improved Technical Specifications (ITS) and appropriate action statements were implemented for out-of-service equipment, and verification that logs and records accurately identified equipment status or deficiencies.
01.2 Summa of Plant Status The plant was at approximately 98.5 percent (%) reactor power at the beginning of the inspection period, with a 1.5% margin from full power being maintained while overpower and overtemperature delta temperature evaluations were being completed.
On May 17, 1999, the licensee placed offsite power in a 100/0 lineup on circuit 751, while circuit 767 was taken out-of-service for replacement of its underground cable (see section M2.1).
Offsite power was returned to a 50/50 line-up on May 21, 1999.
On May 22, 1999, power was reduced to approximately 58% to repair switchyard isolation switch 90813 that had been operating at an elevated temperature.
The isolation switch was repaired and the plant returned to 98.5% power the same day. On May 26, 1999, the plant returned to approximately 100% power following sub-component replacement and re-calibration of overpower and overtemperature delta temperature modules.
On May 27, 1999, an inadvertent automatic outward rod motion of four steps occurred during instrumentation and controls (IBC) staff maintenance on nuclear instrument current comparator drawer N-38 (see section 04.1).
On May 29, 1999, power was reduced to approximately 46% to repair switchyard isolation switch 90811 (which was operating at an elevated temperature) and a leaking 5B feedwater heater drain valve, and to perform main turbine stop valve testing. These work activities were successfully completed and the plant returned to 100% power on May, 30, 1999. The plant remained at full power through the end of the inspection period.
Topical headings such as 01, M8, etc., are used in accordance with the NRC standardized reactor inspection report outline. Individual reports are not expected to address all outline topic Operator Knowledge and Performance
'4.1 Inadvertent Rod Motion Durin IRC Maintenance S
(71707)
The inspectors reviewed operations response to an inadvertent automatic withdrawal of control rods.
b.
Observations and Findin s On May 27, 1999, an inadvertent automatic control rod withdrawal occurred during maintenance on the nuclear instrument (Nl) rate comparator drawer N-38. I8C technicians removed a cable from the N-38 drawer to replace some light bulbs, and the rod control system sensed a change in power and generated a rods out demand signal.
Control rods stepped out four steps, at which time control room operators properly placed rods in manual and entered abnormal procedure AP-RCC.1, "Continuous Control Rod Withdrawal/Insertion," and restored normal rod position.
The licensee subsequently discovered that the procedure used for the maintenance activity had not indicated that manual rod control was necessary.
The licensee determined that the maintenance procedure used on May 27 was last used while the reactor was shut down and the N-38 drawer was de-energized.
The inspector learned that a pre-job brief was conducted between the operators and IBC technicians, where operators specifically asked the technicians ifthe planned work would affect control rod motion. The technicians indicated it would not, since no such restriction was identified in the approved maintenance procedure.
The licensee initiated an ACTION report (99-0957) to further review the issues related to the planning and review of the subject work package and the adequacy of the 18C technicians'nowledge of nuclear instrumentation operation.
Conclusions Operations personnel responded well to a May 27, 1999, unanticipated automatic withdrawal of control rods during maintenance on the nuclear instrument current comparator drawer N-38. Work package preparation and review for this work activity was poor, as it did not anticipate the need for control rods to be placed in manual
~
Additionally, ILC technicians exhibited a knowledge weakness by indicating that the Nl work would not impact control rod motio II. Maintenance M1 Conduct of Maintenance M1 ~ 1 Maintenance and Surveillance Testin Activities Ins ection Sco e and Findin s (62707 and 61726)
The inspectors observed portions of plant maintenance activities to verify: the correct parts and tools were used; applicable industry Codes and ITS requirements were satisfied; and that equipment operability was appropriately demonstrated upon completion of the maintenance.
Selected surveillance tests were observed to verify:
approved test procedures were in use; procedure details were adequate; test instrumentation was properly calibrated; and test results satisfied acceptance criteria or were properly dispositioned.
The following maintenance and surveillance activities were observed:
Circuit 767 Cable Replacement Reactor Protection System Channel 3 Overpower Delta Temperature Math Module Calibration Inspection of the A-Containment Spray pump DB-50 circuit breaker Reactor Protection System Channel 3 Overpower Delta Temperature Bistable Reconfiguration and Testing Troubleshooting A-Emergency Diesel Generator DB-75 Supply Breaker to Bus 14 Following Test Failure (see section M2.2)
Periodic Test (PT)-16Q-B, "B-MotorDrive AuxiliaryFeedwater Pump" PT-2.1Q, "Safety Injection System Quarterly Test" PT-12.1, "Emergency Diesel Generator A" b.
Conclusions Observed maintenance activities were accomplished in accordance with procedural requirements.
The licensee's post-maintenance testing was adequate to demonstrate the operability of equipment prior to it's return to service.
Test procedures contained--
adequate details for accomplishing test requirements.
Testing was performed by knowledgeable personnel, and test instrumentation was properly calibrated.
M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Re lacement of Offsite Power Circuit 767 Cables TO7)
The inspectors reviewed the licensee's activities to replace underground cabling for offsite power circuit 76 Observations and Findin s The licensee decided to replace the underground cabling for both offsite power circuits 767 and 751 after an age related splice failure occurred in circuit 751 on November 20, 1998 (see IR 50-244/98-12).
The cable replacement for circuit 751 was successfully completed February 11, 1999 (see IR 50-244/99-01).
During this inspection period, the licensee maintained an offsite power lineup in a 100%/0% configuration on offsite circuit 751 from May 17, 1999 to May 21, 1999 during the 767 cable replacement.
The offsite power lineup was returned to a 50%/50% lineup on both offsite circuits following completion of the 767 cable replacement on May 21, 1999.
On May 17, 1999, the inspectors noted that several activities were ongoing outside the protected area that could have adversely impacted circuit 751. Specifically, a backhoe was excavating along Slocum Road, near the power poles carrying the transmission lines of circuit 751, and optical cable was being hung on the same power poles that carry circuit 751.
Upon being made aware of these activities, the licensee suspended that work until the circuit 767 cable replacement was completed and the plant was restored to the normal 50%/50% offsite power configuration.
Conclusions Offsite power circuit 767 cable replacement was successfully performed in a timely manner.
However, the licensee demonstrated poor work coordination when activities outside the protected area, which could adversely impact the availability of the in-service offsite power circuit, were not suspended until after work on circuit 767 was commenced and after being brought to the licensee's attention by the inspectors.
Failure of Emer enc Diesel Generator Out ut Breaker to Safe uards Bus 14 The inspectors reviewed the licensee's response to a failure of the A-emergency diesel generator (A-EDG) supply breaker to safeguards Bus 14.
Observations and Findin s On June 8, 1999, the A-EDG supply breaker to Bus 14 did not close on demand from the control room while running the periodic test (monthly). The test was aborted and the licensee commenced troubleshooting activities. The breaker was attached to monitoring equipment while still racked-in to the switchgear cubicle and in the test position. The breaker was successfully cycled twenty consecutive times.
However, the licensee noted some sluggishness when resetting the tripper bar following breaker closure.
On June 9, the breaker was removed from the cubicle and inspected on the bench in the electric shop.
The licensee observed that gaps on end supports for the tripper bar were not uniform, hindering free movement of the tripper bar. The end supports were realigned and the tripper bar was observed to move freely. The breaker was re-installed and successfully tested on June 9, 199 Initial licensee belief that the A-EDG may have been inoperable since the last successful performance of the PT (approximately one month earlier) could not be validated.
This was based upon the as-found condition of the breaker tripper bar and the inability to repeat the on-demand breaker closure failure. Consequently, the licensee concluded that the A-EDG was operable until June 8, 1999 (time of discovery) when the output breaker failed on-demand.
Further, this A-EDG failure was not reportable per 10 CFR 50.72 and 50.73 and the June 10, 1999 notification (Event No. 35811) was retracted on June 30, 1999.
The inspector determined that the B-EDG supply breaker was inspected and the tripper bar was observed to be in the fuliyreset position. Therefore, the licensee determined that no common mode failure existed.
The licensee's corrective actions to prevent recurrence included system operating procedure and maintenance procedure changes to visually verify proper tripper bar alignment and to verify that the tripper bar freely resets following an opening (trip) of the breaker.
In addition, a note will be added to the PT to have the system engineering staff and/or electricians locally observe breaker performance, when practicable, to monitor for any anomalies.
The licensee has also contracted Westinghouse to review breaker maintenance procedures to verify that proper tests and inspections are performed.
Conclusions The licensee took proper actions following discovery of the failed A-emergency diesel generator output breaker to troubleshoot, identify, and correct the breaker failure mechanism.
The licensee's broader corrective actions to review all breaker maintenance procedures for adequacy with the vendor was considered a good initiative.
Miscellaneous Nlalntenance Issues Closed Violation 50-244/97-12-01: Failure to Follow Administrative Procedure Re uirements During the 1996 and 1997 refueling outages, mechanical maintenance personnel improperly entered out-of-specification test data into a preventive maintenance procedure (M-15.1M) for the delta peak firing pressure on the B-emergency diesel generator (B-EDG), and certified the data as satisfactory.
The licensee subsequently returned the B-EDG to service without consulting with maintenance management on the out-of-specification data, and also without initiating an ACTION Report to enter the degraded condition into the corrective action program.
This represented a failure to comply with the licensee's administrative requirements contained in nuclear directive ND-MAI,"Maintenance," administrative procedure A-1603.5, 'Work Order Execution,"
and maintenance procedure M-15.1M, "Aor B Diesel Generator Mechanical Inspection and Maintenance." It was also a violation of 10 CFR 50, Appendix B, Criterion V,
"Instructions, Procedures, and Drawings." Although site engineering had consulted with the equipment manufacturer (ALCO) and with an engine performance analyst before concluding that the B-EDG was operable, no formal action was taken to document the
'xtent of the degraded condition or to develop a plan to restore the engine to within the
" manufacturer's delta peak pressure specifications.
In December 1997, the licensee initiated ACTION Report 97-2095 to address improper documentation of test data in the maintenance work package.
The ACTION Report's resolution included a revision to procedure M-15.1M to remove the instructions and data sheets for the peak firing pressures and temperatures, and delta peak pressure.
The data sheets for these parameters were moved to the periodic surveillance tests for the EDGs and were to be used during engine diagnostic tests.
The licensee also wrote a training work request and developed specific training for shop personnel on how to assure that out-of-specification data is properly addressed and evaluated.
The resolution ofACTION Report 97-2095 documented the licensee's plan for further technical review and analysis of the out-of-spec data with the ALCO owners group and the EDG diagnostics consultant to RGLE. These actions were completed by March 1, 1998.
The licensee formally responded to the violation on March 13, 1998, and indicated that maintenance personnel inappropriately accepted the'test data based upon the apparent acceptability of the data by the engineering staff. The licensee also reviewed the applicable nuclear directives and administrative procedures with all maintenance personnel, especially the requirements in procedure A-1603.5 for contacting maintenance supervision when unexpected or out-of-specification conditions were encountered during maintenance.
This training was completed on October 1, 1998.
The inspectors concluded that the licensee's actions to address this violation were adequate.
The inspectors have not identified any other instances where the licensee certified out-of-specification data as acceptable without a formal evaluation.
This violation is closed.
III. En ineerin E8 Miscellaneous Engineering Issues E8.1 U dated Ins ector Follow-U Item 50-244/98-13-01: Total Instrument Uncertaint Calculations Needed for Im roved Technical S ecification ITS Com liance ln December 1998, the inspectors identified that as-found out-of-specification conditions in some ITS-related plant instruments may not have been evaluated for operability in a timely manner.
Plant instruments listed in the ITS had total instrument [loop] uncertainty (TIU) calculations available to support operability evaluations.
However, other instruments needed to perform ITS surveillance tests, but not specifically listed in ITS, did not always have supporting TIU calculations.
For example, the containment spray system sodium hydroxide (NaOH) tank level instrument and the containment hydrogen recombiner pressure instruments did not have TIU calculations available to evaluate operability when these instruments were found out of their specified tolerance range during recent calibrations.
The inspectors initiated this follow-up item to review the
licensee's actions to investigate past instances of this problem and to review the licensee's assessment of their program for TIU calculations.
Inspector review determined that the licensee initiated Technical Staff Request (TSR)99-001 to develop uncertainty calculations for those instruments that are relied upon to verify ITS limits, but that'were not listed in the ITS. Additional licensee actions included:
1) requiring all ITS-related instruments found out-of-specification have an ACTION Report generated and an operability evaluation performed, whether or not the instrument already had a TIU calculation; 2) holding strategy meetings to develop a comprehensive plan to determine which ITS-related instruments need a TIU calculation; 3) hiring a technical consultant to review the existing TIU calculations, and to provid'e a cross reference to all ITS surveillances (including specific instruments used); 4) developing software to trend as-found out-of-specification data for ITS-related instruments over the past ten years; and, 5) obtaining a formal independent assessment from the consultant on the TIU program.
Overall, the licensee has made progress to address program weaknesses for TIU calculations on ITS-related instruments.
This Inspector Follow-up Item remains open.
IV. Plant Su ort R8.1 miscellaneous Radiological Protection &Chemistry Issues Un lanned Ex osure Incidents An in-office review was completed of ACTION Report No. 99-0664 and subsequent telephone calls were made with the licensee to discuss associated issues.
ACTION Report 99-0664 documented a series of incidents, occurring on April 6 and 7, 1999, in
. which workers wearing electronic dosimeters (ED) received unplanned exposures as a result of not hearing or seeing ED alarms when working in the reactor cavity, a High Radiation Area (HRA). The adequacy of the cause analysis and corrective actions that resulted from these incidents were evaluated.
Observations and Findin s During the period from 4:30 p.m. on April6, 1999, to 6:00 a.m. on April7, 1999, three incidents occurred in which workers received unplanned exposures while performing activities in the reactor cavity. These unplanned exposures were the result of the workers not being able to see or hear their ED alarming when the integrated dose setpoint was exceeded and poor Radiation Protection staff oversight of radiological work activities.
The first incident occurred on the afternoon of April6, 1999. A crew of three individuals entered the reactor cavity to verify that the reactor head was correctly aligned and to perform radiation surveys followingthe setting of the reactor head on the vessel and
cavity draindown. While performing these tasks, two individuals were unable to hear their EDs alarming due to the high noise level in the area.
In addition, they were unable to see the alarm because the ED was taped on the knee in such a way that the ED could not be read.
The Radiation Protection Technician (RPT) accompanying the workers observed shortly after entering the cavity that the area dose rates were higher than expected, but did not take any remedial actions.
Upon completing the tasks and exiting the cavity, the RPT observed the workers'Ds were alarming and directed the individuals to exit containment.
The alarming EDs were found to have exceeded the integrated'dose setpoint of 200 mrem (276mrem and 438 mrem).
The second incident occurred at about 2:00 a.m. on April7, 1999.
Following completion of a radiological survey of the cavity, the As Low As Reasonably Achievable (ALARA)
group established a new ED integrated dose setpoint of 400 mrem and a maximum cavity stay time of 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br />.
The Radiation Work Permit (RWP) was modified accordingly. A crew entered the cavity to decontaminate the area and worked for over two (2) hours, because the stay time was not monitored by the RPT. Upon leaving the area, two workers'Ds were identified to be alarming with indicated doses of 415 mrem and 459 mrem.
The third incident occurred at about 6:00 a.m. on April7, 1999. A crew entered the cavity to continue cleaning and removing equipment. A stay time of 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> was again established for the crew. As a result of the RPT providing ineffective timekeeping, the crew worked in the cavity for about two (2) hours. When the RPT entered the cavity to check personnel doses, four individuals'Ds were alarming with indicated doses of 400 mrem, 420 mrem, 437 mrem, and 506 mrem.
When day-shift (relief) management was informed of these three incidents, immediate actions were taken to prevent further occurrences and to evaluate the causes of the incidents.
Allcavity work was stopped, interviews were held with the cognizant individuals, and an ACTION Report was generated to initiate management actions to further evaluate this series of events.
Details of the individual events, with immediate corrective measures and lessons learned, were promptly communicated to the work force through department stand-down meetings, shift turnover briefings, and daily management meetings.
The radiation protection staff was counseled on management expectations regarding the proper use of EDs, monitoring of stay time, technician response to unexpected radiological conditions, and providing instructions to workers.
To address the known human performance deficiencies, broad-based and long-term corrective actions were being developed including: procedural changes mandating when RPTs must monitor stay times to ensure dose limits are not exceeded; increasing the number of RPTs trained in using the remote dosimetry monitoring system; and, revising ALARApackages to better define RPT responsibilities and communications with the ALARAgroup.
The ACTION Report No. 99-0664 and the associated corrective actions were discussed with the Manager, Radiation Protection and Chemistry and various staff members.
Based upon information provided from these discussions and supporting records, the
inspector concluded that these events were in violation of Improved Technical Specification (ITS) 5.7, "High Radiation Area," which states, in part, "Any individual or group of individuals permitted to enter such areas shall be provided with...a radiation monitoring device that continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received."
Implicitin 1TS 5.7 is that the use of such a device is sufficient to inform the worker that the established dose limitwas exceeded.
Contrary to this requirement, workers could not see or hear the electronic dosimeter alarms informing them that the integrated dose setpoint had been exceeded.
ITS 5.7 also states, in part, "An individual qualified in radiation protection procedures...
is responsible for providing positive control over the activities within the area and shall perform periodic radiation surveillance at the frequency specified... in the RWP."
Contrary to this requirement, positive control was not provided by the RPTs, in that, the workers'tay time, as specified on the radiation work permit and associated records, was not monitored.
The root cause analysis for these events was appropriately conducted and identified various program controls that were ineffective. The immediate and long-term corrective actions were determined to be comprehensive and appropriate to prevent a recurrence.
A human performance assessment was ongoing at the close of the inspection.
Additionally, at the request of the inspector, the licensee re-evaluated the incident to determine whether the affected workers could have potentially received an exposure in
= excess of 10 CFR Part 20 limits. Based upon review of the licensee's subsequent evaluation, no substantial potential existed for a worker to receive a dose in excess of the regulatory limits.
This licensee identified and corrected Severity Level IVviolation is being treated as a Non-Cited Violation (NGV 50-244/99-04-04), consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as ACTION Report No. 99-0664.
Conclusions Licensee management identified, after the fact, a series of unplanned exposure events that were the result of deficiencies in the implementation of radiological controls, contrary to Improved Technical Specification 5.7.
In accordance with the established corrective action process, the licensee conducted a root cause assessment and completed (or planned) appropriate corrective actions to prevent a recurrence.
However, the corrective actions following the individual events were not effective, in that they did not prevent recurrence of the subsequent unplanned exposure events.
In addition, the overall series of events represented an indifference to proper radiological controls and precautions on the part of the individual radiation workers, radiation protection technicians, and direct supervision involved in the events.
The violation of Improved Technical Specifications was non-cite V. Mana ementMeetin s
X1 Exit Meeting Summary Afterthe inspection was concluded, the inspectors presented the results to members of licensee management on July 6, 1999. The licensee acknowledged the findings presented.
The licensee indicated that no materials examined during the inspection were considered proprietar ATTACHMENTI PARTIALLIST OF PERSONS CONTACTED Licensee T. Alexander P. Bamford G. Graus G. Hermes J. Hotchkiss G. Joss M. Lilley R. Marchianda T. Marlow J. Pascher T. Plantz R. Ploof P. Polfleit R. Popp J. Smith W. Thomson J. Wayland T. White J. Widay G. Wrobel Operational Review Manager Reactor Engineering Manager 18 C/Electrical Maintenance Manager Acting Primary Systems Engineering Manager Mechanical Maintenance Manager Results and Test Supervisor Quality Assurance Manager Nuclear Assessment Manager Nuclear Engineering Services Manager Electrical Systems Engineering Manager Maintenance Systems Manager Secondary Systems Engineering Manager Emergency Preparedness Manager Production Superintendent Maintenance Superintendent Radiological Protection 8 Chemistry Manager Scheduling Manager Operations Manager Plant Manager Nuclear Safety 8 Licensing Manager
Attachment I
IP 37551:
IP 40500:
IP 61726:
IP 62707:
IP 71707:
IP 71750:
IP 92901:
IP 92902:
IP 92903:
INSPECTION PROCEDURES USED Onsite Engineering Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing Problems Surveillance Observation Maintenance Observation Plant Operations Plant Support Follow-up - Operations Follow-up - Maintenance Follow-up - Engineering
~oened NCV Closed VIO NCV Discussed ITEMS OPENED, CLOSED, AND DISCUSSED 50-244/99-04-01 Unplanned Radiological Exposures 50-244/97-12-01 Failure to Follow Administrative Procedure Requirements 50-244/99-04-01 Unplanned Radiological Exposures'FI 50-244/98-13-01 TIU Calculations Needed for ITS Compliance
Attachment I
LIST OF ACRONYMS USED ALARA ED EDG ESF HRA IFI IR I8C
.
ITS LCO
'NCV NRC NRR PDR PT RCA RG8E RP RPT RP8C RWP TSR As Low As Reasonably Achievable Electronic Dosimeter Emergency Diesel Generator Engineered Safety Feature High Radiation Area Inspector Follow-up Item Inspection Report.
Instrumentation and Control Improved Technical Specification Limiting Condition for Operation Non-cited Violation Nuclear Regulatory Commission Nuclear Reactor Regulation Pubic Document Room Periodic Test Radiologically Controlled Area Rochester Gas and Electric Corporation Radiation Protection Radiation Protection Technician Radiological Protection and Chemistry Radiation Work Permit Total Instrument Uncertainty Technical Staff Request