IR 05000333/1999006
| ML20210U311 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 08/12/1999 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20210U266 | List: |
| References | |
| 50-333-99-06, NUDOCS 9908200056 | |
| Download: ML20210U311 (20) | |
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U.S. NUCLEAR REGULATORY COMMISSION REGION 1 Docket No.:
50-333 License No.:
DPR-59 Report No.:
99-06 Licensee:
New York Power Authority Facility:
James A. FitzPatrick Nuclear Power Plant Location:
Post Office Box 41 Scriba, New York 13093 Dates:
June 1 to July 17,1999 Inspectors:
R. A. Rasmussen, Senior Resident inspector R. A. Skokowski, Resident inspector B. S. Norris, Resident inspector T. A. Moslak, Radiation Specialist E. H. Gray, Sr. Reactor Inspector
. L. S. Cheung, Reactor Inspector Approved by:
J. F. Rogge. Chief.
Projects Bratich 2 Division of Reactor Projects l
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9908200056 990E12 PDR ADOCK 050C9333 PM
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SUMMARY OF FINDINGS James A. FitzPatrick Nuclear Power Plant NRC Inspection Report 50-333/99-06 The report covered a seven week period of resident inspection, and the results of announced inspections by a regional radiation safety inspector and a regional engineering inspector.
Inspection findings were assessed according to potential risk significance, and were assigned colors of GREEN, WHITE, YELLOW, or RED. GREEN findings are indicative of issues that,
- while not necessarily desirable, represent little risk to safety. WHITE findings would indicate
- issues with some increased risk to safety, and which may require additional NRC inspections.
' YELLOW findings would be indicative of more serious issues with higher potential risk to safe performance and would require the NRC to take additional actions. RED findings represent an unacceptable loss of margin to safety and would result in the NRC taking significant actions that
. could include ordering the plant shut down. The findings, considered in total with other inspection findings and perfe mance indicators, will be used to determine overall plant performance.
Mitigating Systems Green. The failure of the circulating lube oil pump for the "A" emergency diesel
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generator (EDG), and a subsequent relay failure during the post-maintenance test were evaluated for overall plant risk. These equipment failures, which resulted in emergency diesel generator inoperability, were determined to be green using the significance determination process. (Section 1R03.1)
Green. Mechanics altered the design of a safety bus control power fuse block
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and did not document the non-conformance. The fuse block manufacturer required grease on the fuse block contacts to prevent a loss of function due to corrosion. This grease was omitted during the assembly process and the omission was not entered into the corrective action system for resolution. The failure of this fuse clip could have resulted in a loss of one of the two plant safety electrical supply busses. This issue was considered green in the significance determination process because it did not have an immediate impact on equipment performance. The failure to initiate a deficiency report was contrary to station procedures, which require a DER to be initiated for conditions adverse to
quality, and was a violation of NRC requirements. (NCV 50-333/99-06-01)
(Section 1R03.2)
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Green. The inspectors identified approximately 25 minor discrepancies during a
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walkdown of the HPIC system. The large number of discrepancies co-existing on a single safety system represents a lapse in control of the system configuration and a violation of NRC requirements. However, because the discrepancies did not impact equipment operability the issue was a green finding in the significance determination process. Furthermore, the inspectors noted that it took the licensee an excessive amount of time, approximately two weeks, to enter most of ii
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Summary of Findings (cont'd)
the discrepancies into their corrective action program. (NCV 50-333/99-06-02)
(Section 1R04)
Barrier integrity Green. The inspectors identified a problem in a NYPA administrative procedure
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which resulted in operators not adhering to written operating procedures. This administrative procedure resulted in a misunderstanding by the licensed operators of the requirements of their licenses with regard to procedure compliance and of the requirements of 10CFR50.54(x). Operators not complying with plant procedures could have resulted in the inoperability of plant safety systems. This potential inoperability of plant safety systems was a green issue in the significance determination process. This issue was previously identified and was not adequately resolved by the licensee. The failure to take appropriate corrective actions following an NRC-identified deficiency is a violation of 10CFR50, Appendix B, Criterion XVI, " Corrective Action." (NCV 50-333/99-06-03) (Section 1R11)
-Occupational Radiation Safety Green. The actual collective dose for the control rod changeout, performed
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during the 1998 refueling outage, exceeded the projected dose by greater than 50%. The initial dose projection only addressed ancillary tasks and did not include the dose (approximately 5 person-rem) for removing and installing the CRDs. (Section 2OS2).
Other Activities None. On two occasions non-conforming conditions were not incorporated into
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the corrective action program. Additionally, corrective actions taken to correct the procedure compliance issue identified during the January hydrogen fire were inadequate. The specifics of these issues are discussed in sections 1R03.2, 1R04, and 1R11 of this report. As previously documented, two of these issues represented violations of NRC requirements. (Section 40A1)
None. Two discrepancies were identified through the review of licensee
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submitted performance indicator (PI) data. One error, identified by the NRC, was in not reporting a plant power change under the " Unplanned Power Changes per 7000 Critical Hours' Pl. The other error, identified by NYPA, was in not reporting an occurrence under the " Occupational Exposure Control Effectiveness" Pl. In each case, the error did not result in a change of indicator color and was corrected in a subsequent submittal. This item remains unresolved while the
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NRC evaluates errors in the Pl data submittal. (URI 50-333/99 06-04) (Section 40A2)
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Report Details SUMMARY OF PLANT STATUS The inspection penod began with the unit at full power. On June 9,1999, power was reduced to I
approximately 87 percent in preparation for a reactor shutdown due to a loss of control power to the 10500 safety bus. The control power problem was resolved within a few hours and the unit was returned to full power. On July 3,1999, reactor power was reduced and the main generator
- was removed from service for a planned outage to repair an electrical connection on one of the transformer output 345 KV lines. Following the repairs, reactor power was increased to approximately 80 percent and then subsequently reduced to take the generator out of service again. This was required to repair a motor operated disconnect that failed to fully close following the previous repair outage. The reactor was retumed to full power on July 5. On July 7, power
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was reduced to 65 percent due to the failure of a condensate system pump motor. The motor was replaced and the unit retumed to full power on July 8th. The reactor was shutdown on July 13th due to a steam leak on a feedwater piping drain. The leak was repaired and the reactor restarted on July.17th. The reactor was at full power on Ju!y 18th.
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REACTOR SAFETY Comerstones: Initiating Events, Mitigating Systems, Barrier Integrity 1R01 Adverse Weather a.
Insoection Scooe The inspectors reviewed conditions and actions being taken in response to periods of hot weather.
. b.
Observations and Findinas There were no findings identified and documented during this inspection.
1R02 Chanoes to License Conditions a.
Insoection Scooe Nine nuclear safety evaluations from the past two years were reviewed to determine whether the associated design changes resulted in more than a minimal increase in risk without prior NRC approval. NYPA's identification and resolution of problems related to 10 CFR 50.59 safety evaluations were also examined.
Safety Evaluations Reviewed
'A)
For Mitigation Systems:
1.
- SE-98-40, Temporary Deenergization of Drywell Cooling MOV Power Supplies, Revision 0, dated August 17,1998, and Revision 1, dated April 27,1999; 2.
SE-99-08, Update of FSAR to Reflect 1973 Replacement of LPCI Pump Low Flow Switches - Range Change, Revision 0, dated April 6,1999;
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3.
SE-98-34, Emergency Service Water System Check Valve Replacement, Revision 0, dated September 15,1998; 4.
- SE-98-13, Residual Heat Removal and Core Spray Suppression Pool Suction Strainer Replacement, Revision 2, dated November 24,1998; 5.
- SE-98-25, High Pressure Coolant injection and Reactor Core Cooling Suppression Pool Suction Strainer Replacement, Revision 1, dated November 24,1998.
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- All three safety evaluations were related to a strainer modification.
B)
For Barrier Integrity:
1.
SE-98-51, inboard Main Steam isolation Valves NAMCO connectors, Revision 0, dated December 14,1998; 2.
SE-98-04, Replacement of PCIS Display on 90-4 Panel, Revision 1, dated April 1,' 1998; 3.
SE-99-01, Altemate Hydrogen Supply, Revision 1, dated February 24,1999.
b.
Observations and Findinas There were no findings identified and documented during this inspection.
1R03 EmeroentWork
.1 Unolanned Emeroency Diesel Generator inoperability a.
Inspection Scope The inspectors reviewed the licensee's actions associated with the failure of the "A"
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emergency diesel generator (EDG) circulating lube oil pump and a subsequent reverse power trip of the EDG, which occurred during post-maintenance testing for the lube pump replacement.
b.
Observations and Findinas The failure of the circulating lube oil pump for the "A" emergency diesel generator (EDG),
and a subsequent relay failure during the post-maintenance test were evaluated for overall plant risk. These equipment failures were determined to be green using the significance determination process.
i On June 8,1999, operators received an "EDG 'A' engine trouble or shutdown" annunciator in the control room. Upon investigation, operators found that the "A" EDG circulating lube pump motor thermal overloads had tripped -The Shift Manager (SM)
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declared EDG "A" inoperable, entered the applicable limiting condition for operations (LCO), and tested the remaining EDG as required by technical specifications (TS).
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Troubleshooting identified that the circulating lube oil pump motor was difficult to turn.
As a result, the licensee replaced the motor. Subsequently, the licensee determined that
. the output shaft bearing was bad.
After verifying proper operation of the newly installed motor, the licensee tested the EDG
"A" operation in accordance with Surveillance Test Procedure ST-9BA, "EDG A and C Full Load Test and ESW [ emergency service water) Pump Test." During the performance of this test, EDG "A" tripped on reverse power. NYPA wrote a deviation / event report (DER) to address the problem, and initiated troubleshooting.
While troubleshooting, NYPA was able to recreate the problem, and subsequently identified the cause to be the over-travel of an auxiliary switch within the EDG "A" and EDG "C" tie breder. As a result, the load balance portion of the EDG control circuit was disabled, this allowed sufficient circulating currents between EDG "A" and "C" to be developed, which caused the reverse power trip.
NYPA replaced the auxiliary relay and, on June 11,1999, performed a satisfactory post-maintenance test of the relay and surveillance of the EDGs. Based on related problems with similar applications of this type of auxiliary switch, NYPA was evaluating long term corrective actions to either upgrade or eliminate these switches.
Prior to the reverse power trip of EDG "A," the last successful surveillance test of the EDG was May 27,1999. Since there was the potential that the EDG was incapable of performing its function for at least one-half the time since the last successful test, the inspectors evaluated the risk associated with having the "A" EDG inoperable during that time. The inspectors considered the other equipment unavailable during the period, and success paths for a loss-of offsite-power (LOOP) at the FitzPatrick Station as described in the licensee's Individual Plant Examination (IPE), and concluded that the increase in risk was minimal (Green).
.2 Loose Fuse for Safety Bus Control Power a.
Inspection Scope The inspectors observed NYPA's response to an electrical problem which resulted in the inoperability of a safety bus.
b.
Observations and Findinas Mechanics altered the design of a safety bus control power fuse block and did not document the non-conformance. The fuse block manufacturer required grease on the fuse block contacts to prevent a loss of function due to corrosion. This grease was omitted during the assembly process and the omission was not entered into the corrective action system for resolution. The failure of this fuse clip could have resulted in a loss of one of the two plant safety electrical supply busses. This issue was considered green in the significance determination process because it did not have an immediate impact on equipment performance.
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On June 9,1999, NYPA identified a loss of control power to the 10500 safety bus. As a result of this problem, NYPA declared the safety bus inoperable and entered a limiting condition for operation (LCO) which required a plant shutdown within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. During
' troubleshooting of the loss of control power, NYPA discovered a loose fuse block assembly in the control power circuit to the 10500 safety bus. NYPA utilized information from the manufacturer of the fuse block to bend and align the electrical connections within the fuse block. Following the reinstallation, NYPA declared the safety bus operable and exited the shutdown LCO.
During the inspection, NYPA observed black grease on the contact areas of the fuse block assembly. With the support of maintenance engineering, NYPA elected to reassemble the fuse block assembly without grease. However, a deficiency report or other formal tracking mechanism was not utilized to track this potential non-conformance and operability issue.
Through follow up with the manufacturer, NYPA determined that grease was required to prevent long term corrosion of the contacts. However, a deficiency report had still not been generated to document the non-conformance. The inspectors questioned the lack of a deficiency report and were informed that the maintenance personnel intended to document the issue of the grease in the deficiency report associated with the fuse block failure. The inspectors considered this inappropriate because the grease issue was not directly tied to the failure evaluation which was the primary focus of the original deficiency report. Additionally, without an individual deficiency report, there was not a mechanism to trigger an operability evaluation for the current condition of the safety bus.
The inspectors brought the issue to the attention of NYPA management, and a deficiency report ( DER 99-997) was initiated to track the deficiency. NYPA determined the fuse block was still operable, but was working to evaluate an appropriate schedule for correcting the deficiency.10CFR50, Appendix B, Criterion 16, " Corrective Action",
requires that measures shall be established to assure conditions adverse to quality are promptly identified and corrected. Administrative procedure AP-03.02, " Deviation and Event Reporting," was developed by NYPA to meet this requirement. The failure to initiate a deficiency report was contrary to the guidance of administrative procedure AP-03.02, which requires a DER to be initiated for conditions adverse to quality, and was a violation of NRC requirements. This violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as DER 99-1238. (NCV 50-333/99-06-01)
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- Other Emeroent Work a.
Insoection Scope in addition to those items discussed above, the inspectors reviewed actions taken in response to a failure of the reactor core isolation cooling (RCIC) system steam isolation valve.
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b.
Observations and Findinas-There were no findings identified and documented during this inspection.
1R04 Eauioment Alianments a.
Inspection Scope The inspectore performed a complete system walkdown (visual Inspection) of the high pressure coolant injection (HPCI) system. The inspectors also performed a partial system walkdown of the reactor core isolation cooling (RCIC) system.
b.
Observations and Findirgtg The inspectors identified approximately 25 minor configuration discrepancies during a walkdown of the HPIC system. The large number of discrepancies co-existing on a single safety system represents a lapse in control of the system configuration and a violation of NRC requirements. However, because the discrepancies did not impact equipment operability the issue was a green finding in the significance determination process.
The inspectors performed a walkdown of the accessible portions of the HPCI system to compare plant drawings and Procedure OP-15, "High Pressure Coolant injection,"
Revision 47 with actual component positions and configuration. The inspectors identified approximately 25 minor discrepancies between, the drawings, actual confguration and the equipment lineup provided in OP-15. These discrepancies were discussed with the system engineer and the Technical Services Manager, and included:
1.
'a drain valve specified by the drawing and procedure valve lineup as locked closed was found without a locking device; 2.
the position of two lube oil filter bleed valves as specified by the drawing was different from the position specified by the procedure valve lineup; 3.
vent and drain valve caps installed in the plant were not included in the procedure valve lineup; 4.
a vent valve and a test valve specified by the drawings as capped were found without caps; 5.
the description of the circuit breakers provided in Procedure OP-15 did not match the description provided on the breaker labels; and
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other minor discrepancies between the installed configuration and the drawings.
The failure to adequately control the configuration of the HPCI system is a violation of Code of Federal Regulations (CFR) Title 10 Part 50 Appendix B, Criterion V,
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- Instructions, Procedures, and Drawings." This violation is being treated as a Non-Cited Violation, consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's corrective action program as DERs 99-01099,01171,01186. (NCV 50-333/99-06 4 2)
The identified discrepancies were ultimately entered into the corrective action program, however the inspectors noted that it took the licensee an excessive amount of time, approximately two weeks, to enter most of the discrepancies. NYPA's Senior Management Team acknowledged that the timeliness failed to meet their expectations.
1R05 Fire Protection a.
Inspection Scope During tours of the reactor and turbine buildings, the inspectors performed assessments of fire protection issues.
b.
Observations and Findinas There were no safety significant findings identified and documented during these inspections.
1R09 Inservice Testina a.
Inspection Scope The inspectors observed surveillance testing of the emergency service water system pumps, b.
Observations and Findinas There were no findings identified and documented during this inspection.
1R11 Licensed Ooerator Reaualification a.
Inspection Scooe The inspectors observed a shift-crew of licensed operators during simulator training.
The inspectors also observed the instructor critique of the crew's performance and reviewed the two scenarios, b.
Observations and Findinas
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-The inspectors identified a problem in a NYPA administrative procedure which resulted in operators not adhering to written operating procedures. This administrative procedure resulted in a misunderstanding by the licenced operators of the requirements of their licenses with regard to procedure compliance and of the requirements of
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10CFR50.54(x). Operators not complying with plant procedures could have resulted in the inoperability of plant safety systems. This potentialinoperability of plant safety l
systems was a green issue in the significance determination process. Furthermore, this issue was previously identified and was not adequately resolved by the licensee.
. During a simulator scenario, the inspectors noted an instance where the CRS appeared to deviate from the EOP steps. Specifically, during implementation of EOP-5,
" Secondary Containment Control," the CRS directed the main steam isolation valves
. (MSIVs) to be shut before the step was reached in EOP-5. The instructors did not
. question this action until challenged by the inspectors; the instructors stated that procedures allowed the operators to depart from procedures.
Based on similar concerns identified during the NRC special inspection of the January 1999 hydrogen fire, the inspectors reviewed the administrative procedures to see if the
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procedures had been corrected. (Reference NRC Inspection Report 50-333/99-02, Section 01.1)
1.
AP-02.06, " Procedure Use and Adherence," Revision 11, continues to have two distinct steps which address deviating from procedures. Step 7.4 discusses the l
latitude allowed by 10CFR50.54(x) with respect to departing from license l
conditions or Technical Specifications in an emergency. But Step 7.3, which allows operations personnel to depart from procedures is also still in the procedure. As such, and as demonstrated in the simulator, operators stated that procedure compliance was not required.
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AP-12.03, " Administration of Operations," Revision 16, Step 8.12.1, states
" Procedure compliance is mandatory,"
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kP-12.03, Step 8.12.6, states "If... a transient or accident sequence begins and entry conditions for EOPs are met, plant operators shall operate the plant according to the EOPs.
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to resolve the issue by revising the procedure to remove Step 7.3 and training the station personnel as to the intent and use of 10CFR50.54(x). The inspectors considered l
the proposed actions to be adequate. Nonetheless, the failure to appropriately revise
AP-02.06 following an NRC-identified deficiency is a violation of 10CFR50, Appendix B, Criterion XVI " Corrective Action " This violation is being treated as a Non-Cited Violation (NCV), consistent with Appendix C of the NRC Enforcement Policy. This violation is in the licensee's correction action program as DER 99-99-0118. (NCV 50-333/99-06-03)
1R12 Maintenance Rule implementat'on a.
Inspection Scope
The inspectors assessed the licensee's implementation of the Maintenance Rule, Title 10 of the Code of Federal Regulations Part 50.65 (10 CFR 50.65), for the reactor core
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isolation cooling (RCIC) system following a failure of the RCIC turbine steam supply outboard isolation on June 22,1999, for the HPCI system as part of the system walkdown described in Section 1R04 of this report, and for the condensate system following the failure of a condensate system pump motor.
b.-
Observations and Findinos There were no findings identified and documented during these inspections.
1R13 Maintenance Work Prioritization a.
Insoection Scooe The inspectors reviewed maintenance work prioritization issues associated with NYPA's preparations for a forced outage. The forced outage was to repair an electrical connection from the output transformer to the 345kv transmission line.
b.
Observations and Findinos
- There were no findings identified and documented during this inspection.
1R14 Nonroutine Plant Evolutions a.
Inspection Scone The inspectors observed control room activities performed in response to an inoperable safety bus and an inoperable emergency diesel generator lube oil circulating pump.
b.
Observations and Findinos There were no findings identified and documented during these inspections.
1R15 Ooerability Evaluations a.
Inspection Scope
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The inspectors reviewed operability determinations associated with three issues identified during the period.. The operability determinations reviewed involved:
EQ analyses for equipment within the drywell the operability of an emergency diesel generator with a noisy circulating lube oil Pump operation with an inoperable containment isolation valve
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Observations and Findinas There were no findings identified and documented during these inspections.
1R16 - Ooerator Work-Arounds a.
Inspection Scooe The inspectors reviewed an operator work-around involving manual actions required to maintain pressure control on the containment atmosphere dilution system nitrogen supply. The inspectors also reviewed the status of the HPCI system to assure no operator work-arounds were present.
b.
-Observations and Findinas There were no findings identified and documented during these inspections.
1R17 Permanent Plant Modifications
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Inspection Scooe The inspectors reviewed modification and test documentation associated with upgrades to the feedwater measurement system. This modification will allow more accurate measurement of feedwater system flow, which is a critical parameter used to determine reactor power. Review of this modification was incomplete due to emergent plant
- problems which prevented the testing and implementation of this modification. Testing associated with this modification required steady state power and fission product levels.
NYPA was in the process of rescheduling this modification.
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Observations and Findings There were no findings identified and documented during this inspection.
1R19 Postmaintenance Testina a.
Inspection Scope The inspectors reviewed post maintenance testing performed following replacement of the emergency diesel generator (EDG) circulating lube oil pump, the EDG control circuit relay, and the 10500 safety bus fuse clip troubleshooting.
b.
Observations and Findings There were no findings identified and documented during these inspections.
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1R22 Surveillance Testina a.
Insoection Scooe The inspectors reviewed surveillance testing of emergency diesel generators, the emergency service water system, and high pressure coolant injection.
b.
Observations and Findinas There were no findings identified and documented during these inspections.
Cornerstone: Emergency Preparedness [EP).
1EP3 Emeroency Resoonse Oraanization Auomentation
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Insoection Scooe The inspectors reviewed contingency plans developed by NYPA due to a large music festival which was scheduled in the immediate vicinity of the emergency response facility. The music festival would have restricted access to the offsite response facility.
b.
Observations and Findinas The inspectors concluded that the contingency plans were appropriate and that emergency response personnel were adequately appraised of the plans.
2.
RADIATION SAFETY Comerstone: Occupational Radiation Safety [OS]
2OS1 Access Control To Radioloaically Sionificant Areas a.
Insoection Scooe The administrative and physical controls for controlling access to radiologically significant i
areas and limiting worker dose were inspected. The inspectors toured the Reactor i
Building, Turbine Building, and Radioactive Waste Processing Building. The inspectors also performed independent surveys in plant areas and for various plant systems
including the attemate Decay Heat Removal system.
i A sample of five Deviation / Event Reports (DER), involving activities in high radiation areas, were reviewed that addressed incidents which were below the threshold for a performance indicator occurrence.
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Observations and Findinos j
There were no findings identified and documented during this inspection.
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' 2OS2 ALARA Plannina and Controls 1.
Insoection Scooe ALARA performance was reviewed for radiologically significant activities performed during 1998 and the Significance Determination Process (SDP) was used to evaluate the collective exposure data. The effectiveness of source term reduction efforts was reviewed. Records were reviewed regarding exposure results for declared pregnant workers, maintenance of self contained breathing apparatus (SCBA), and the training of control room personnel associated with using SCBAs.
2.
Observations and Findinos The actual collective dose for the control rod changeout, performed during the 1998
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refueling outage, exceeded the projected dose by greater than 50%. The initial dose projection only addressed ancillary tasks and did not include the dose (approximately 5 person-rem) for removing and installing the CRDs.
For 1998, the collective exposure for activities perform during the operating cycle and outages was 357 person-rem. The resulting three (3) year rolling average collective dose of 208 person-rem was greater than the BWR industry average of 240 person-rem, in part, due to the dose accrued (87.9 person-rem) in replacing emergency core cooling system (ECCS) strainers.
The collective dose for various dose intensive jobs was compared to the projected dose and the results were evaluated using the Significance Determination Process (SDP).
Included in this review were the torus modification project involving removal / installation of ECCS strainers, changeout of control rod drives (CRD), and activities performed by the Instrumentation & Control Department during 1998. Using the SDP, the dose accrued for CRD changeout (10.019 person-rem) represented a GREEN finding, in that, the actual dose exceeded the projected dose (4.800 person-rem) by more than 50%, the three year rolling average for FitzPatrick was greater than 240 person-rem, actual job dose was greater than 10 person-rem but less than 60 person-rem, and this finding represented a single occurrence meeting the SDP criteria. Shortcomings in accurately estimating job dosage for the CRD changeout resulted in part from a lack of a questioning attitude regarding the completeness of forecasted data. The initial dose projection only addressed ancillary tasks and did not include the dose (approximately 5 person-rom) for removing and installing the CRDs. The licensee initiated DER 99-01100 to resolve this discrepancy.
20S4 Radiation Worker Performance a.
inspection Scope Jobs-in-progress were observed to evaluate the effectiveness of worker practices in keeping exposure as low as reasonably achievable (ALARA). Activities observed
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included, the transfer of spent resin to a storage container, the venting of the attemate Decay Heat Removal system in preparation for testing, removal of electrical connections from a solenoid operated valve, transfer of contaminated equipment from the refuel floor to the storage pit by overhead crane, and the removal of a nitrogen purge line from the traversing incore probe (TIP) machine.
b.
Observations and Findinos There were no findings identified and documented during this inspection.
4.
OTHER ACTIVITIES [0A]
Cross Cutting Elements 40A1 Identification and Resolution of Problems a.
Inspection Scope During the inspections documented in this report, the inspectors assessed the effectiveness of the licensee's programs for correcting and limiting non-compliance with NRC regulations.
b.
Observations and Findinas On two occasions non-conforming conditions were not incorporated into the corrective action program. Additionally, corrective actions taken to correct the procedure compliance issue identified during the January hydrogen fire were inadequate. The specifics of these issues are discussed in Sections 1R03.2,1R04, and 1R11 of this report. As previously documented, two of these issues represented violations of NRC requirements.
40A2 PlVerification
.1 Performance Indicator Data Submittal a.
Inspection Scooe As part of the pilot inspection program, historical performance indicator data was submitted. The data was evaluated, discussed during a public meeting, and inspection requirements were determined.
b.
Observations and Findinas The performance indicator data was in the green performance band, except a white performance threshold was exceeded for the " Unplanned Power Changes per 7000 Critical Hours" indicator during the fourth quarter of 1998, and the first quarter of 1999.
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We also note that this indicator has subsequently retumed to the green band with the submittal of the second quarter 1999 data. This indicator was discussed during the periodic performance review meeting on June 3,1999. Actions to improve performance in this area included the implementation of an equipment reliability performance improvement plan. As noted by this indicator returning to the green band, the NRC recognizes that actions are being taken to improve performance in this area. Therefore, the NRC has chosen to monitor activities through the baseline inspection program.
Detailed review of the performance indicator data was performed in two areas. The results of these detailed reviews are documented below. In each case, data submission errors were identified and corrected. The NRC is developing procedures to disposition inaccuracies when licensee data errors are identified. When the procedures are established these errors may receive further NRC review and consideration.
.2 Unolanned Power Channes per 7000 Critical Hours a.
Inspection Scope The inspectors reviewed performance indicator (PI) data submitted by NYPA and performed a sampling inspection in the area of unplanned power changes per 7000 critical hours. The inspectors reviewed data for the previous year.
b.
Observations and Findinas The NRC identified an unplanned power transient that was not included in the data submitted by NYPA. On July 8,1998, power was reduced by 32 percent due to a problem during testing of a main steam isolation valve. The valve was reopened and power was retumed to 100 percent the following shift. As a result of this finding, NYPA initiated a DER to document the occurrence and identify the cause of the error.
At the time of the initial performance indicator submittal, the unplanned transient performance indicator was white. This error, if properly reported, would not have resulted in a change of indicator status. NYPA reported a correction to the data in the June data submittal.
Exposure Control
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Inspection Scope.
The Occupational Exposure Control Effectiveness performance indicator (PI) data was reviewed for the past three years.
2.
Observation and Findings-In the May 1999 Pl report, the licensee initially identified four (4) incidents occurring within the past twelve (12) quarters that represented either 1) a failure to secure an area against unauthorized access or 2) a failure to provide a means of personnel dose
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monitoring or control as required by technical specifications. Subsequent to report submittal, the licensee identified and evaluated four (4) additional occurrences that could be potentially reportable. Following an evaluation of the historical records and interviews
' with cognizant individuals, the licensee determined that one of the four occurrences was reportable in that sufficient evidence was not available to conclude otherwise.
Accordingly, the licensee the July 1999 PI report corrected the information; changing the number of Pl occurrences from four (4) to five (5), with the performance rating remaining in the GREEN (normal regulatory response) band.
40A4 Other (Closed) LER 50-333/E3-04: Reactor Core Isolation Cooling (RCIC) Instrumentation Inoperable Due to incomplete Protective Tag Restoration. This LER was reviewed in NRC inspection report 50-333/99-04 and NCV 50-333/99-04-01 was issued. No further issues were identified and this LER is closed.
40A5 Meetinas The inspectors presented the inspection results to Mr. D. Lindsey and other members of licensee management on July 22,1999. The licensee acknowledged the findings presented. No proprietary information was identified.
On June 3,1999, the NRC conducted a meeting with the licensee to review plant i
performance. The meeting was held at the facility and was open for public observation.
On June 30,1999, the NRC conducted a meeting with the public to discuss the new NRC inspection program being piloted at the FitzPatrick plant. Members of the public and media were in attendance.
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ATTACHMENT 1 PARTIAL LIST OF PERSONS CONTACTED Licensee M. Anderson Respiratory Protection Supervisor N. Avrakotos Emergency Preparedness Coordinator G. Bregg Instrumentation & Control Manager G. Brownell Licensing Engineer P. Brozenich Operations Manager M. Colomb Site Executive Officer R. Converse General Manager Maintenance (Acting)
F. Edler Assistant Maintenance Manager J. Fitzgerald Construction Services Manager K.Hobbs Health Physics - General Supervisor i
D. Kieper General Manager-Maintenance D. Lindsey Plant Manager R.Locy Training Manager A. McKeen Radiological and Environmental Department Supervisor D. Morrison ALARA Technician W.O'Malley -
General Manager, Operations S. Pointon Radiological Protection - Central Planning K.Pushee ALARA Supervisor D. Ruddy Director, Design Engineering J. Solini Senior Quality Assurance Engineer G. Tasick Licensing Manager A. Zaremba General Manager Support Services
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Attachment 1 (cont'd)
ITEMS OPENED, CLOSED, AND DISCUSSED
' Opened None Opened and Closed NCV 50-333/99-06-01: The failure to initiate a deficiency report was contrary to station procedures, which require a DER to be initiated for conditions adverse to quality, and was a violation of NRC requirements.
, NCV 50-333/99-06-02: The failure to adequately control the configuration of the HPCI system is a violation of Code of Federal Regulations (CFR) Title 10 Part 50 Appendix B, Criterion V,
" Instructions, Procedures, and Drawings."
- NCV 50-333/99-06-03: The failure to appropriately revise AP-02.06 following an NRC-identifed
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deficiency is a violation of 10CFR50, Appendix B, Criterion XVI, " Corrective Action."
Closed LER 50-333/99-04: Reactor Core isolation Cooling (RCIC) Instrumentation Inoperable Due to incomplete Protechve Tag Restoration -
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Attachment 1 (cont'd)
LIST OF ACRONYMS USED
.ACT'
Action / Commitment Tracking ALARA As Low As Reasonably Achievable AOP Abnormal Operating Procedure CRD Control Rod Drive CRS Control Room Supervisor DER.
Deficiency and Event Report ECCS Emergency Core Cooling System EDG Emergency Diesel Generator EOP Emergency Operating Procedure ESF_
. Engineered Safety Feature HPCI High Pressure Coolant injection IPE
' individual Plant Examination ITS Improved Technical Specifications LCO Limiting Conditions for Operation LER Licensee Event Report LOCA Loss of Coolant Accident LOOP Loss of Offsite Power LPCI-Low Pressure Coolant injection MG Motor Generator MSIV Main Steam isolation Valve NRC Nuclear Regulatory Commission Pl Performance Indicator PLCO Potential Limiting Conditions for Operation QA QualityAssurance RCA-Radiologically Controlled Area RCIC Reactor Core Isolation Cooling RES.
R:dictogic ! and Environmental Services RHR Residual Heat Removal RO Reactor Operator RP Radiation Protection RP&C Radiological Protection and Chemistry RTID Radiological Technical Information Document RWP Radiation Work Permit SCBA Self Contained Breathing Apparatus SDP.
Significance Determination Process SGT'
. Standby Gas Treatment SM Shift Manager SRO Senior Reactor Operator STA-Shift Technical Advisor
. TIP '
Traversing incore Probe TS Technical Specifications.
TSP Temporary Shielding Packages UFSAR-Updated Final Safety Analysis Report UPS
' Uninterruptable Power Supply L