IR 05000333/1998007
| ML20198L415 | |
| Person / Time | |
|---|---|
| Site: | FitzPatrick  |
| Issue date: | 12/17/1998 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20198L408 | List: |
| References | |
| 50-333-98-07, 50-333-98-7, NUDOCS 9901050011 | |
| Download: ML20198L415 (42) | |
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U.S. NUCLEAR REGULATORY COMMISSION
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REGION I
Docket No: 50-333 License No: DPR-59 i
l Report No: 98-07 l
I l Licenses: New York Power Authority l -.
Facility:' James A. FitzPatrick Nuclear Power Plant l
l Location: Post Office Box 41 I Scriba, New York 13093 l Dates: October 5,1998 - November 22,1998
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< Inspectors: R. Rasmussen, Senior Resident inspector ;
R. Fernandes, Resident inspector ,
B. Norris, Resident inspector i C. Sisco, Operations Engineer T. Moslak, Radiation Specialist R. Nimitz, Senior Radiation Specialist R. Barkley, Project Engineer .
T. Burns, Reactor Engineer '
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Approved by: J. Rogge, Chief l. Projects Branch 2 Division of Reactor Projects l
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9901050011 981217 PDR ADOCK 05000333 G PM
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l EXECUTIVE SUMMARY James A. FitzPatrick Nuclear Power Plant NRC inspection Report 50-333/98-07 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covered a seven week period of resident
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l inspection and the results of announced inspections in the areas of occupational exposure controlinspection, operator licensing, and inservice inspection by a region based specialists.
Operations
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The failure of the temporary torus dewatering hose was evaluated and corrective actions l were implemented. The failure to perform the procedure as written caused the over !
pressurization of the hose. This non-repetitive, licensee-identified and corrected violation is '
being treated as a Non-Cited violation. (NCV 50-333/98-07-01)(Section 01.2)
The implementation of the torus dewatering system temporary modification was poor due l to an inadequate engineering review, inadequate plant operating review committee (PORC) l review, inadequate project management review, insufficient detail in the temporary modification documentation, and inadequate operations involvement. Identified deficiencies were adequately addressed prior to the utilization of the dewatering system.
(Section O2.1) !
l Maintenance in general, maintenance and surveillance activities were adequately conducted. However, observations during the period indicated severalinstances of poor work practices in the field that the inspectors considered weak. (Section M1.1)
Pre-refueling maintenance activities specified in the station procedures for the special lifting equipment were completed satisfactorily. The inspectors noted good interaction between the maintenance and engineering staffs for resolving questions. The omission of labels and NDE of the lifting eyes were determined to be minor concerns and were properly addressed by the licensee. (Section M1.2)
Fuel moves were well controlled and completed without incident. Communications on the refuel bridge were generally good, especially between the NYPA operators. Classroom training for refueling operations was presented in a manner that was conducive to the learning process. Overall, the inspectors considered the refueling operations to be well coordinated. (Section M1.3)
The reactor head became stuck due to the lift rig being installed out of the usual orientation. The evolution of freeing the stuck reactor head was performed well with good participation by engineering, in support of maintenance. During the period with the head l stuck, outage management and operations were diligent in maintaining appropriate plant l conditions. However, the issue of the totalload applied to the lifting rig was not considered by NYPA and is considered a weakness in the engineering evaluation. (Section M2.1)
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Executive Summary (cont'd)
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l The inspector concluded that inservice inspection had been performed acceptably and included acceptable ASME program coverage, qualified personnel, approved procedures, proper implementation, appropriate examination documentation, and NYPA oversight. The inspections performed were thorough and of sufficient extent to determine component i integrity. (Section M2.2)
The inspector concluded that the welding of the construction access opening in the torus had weak control of welding process variables that could have enabled the weld to have a ;
heat input in excess of the qualification. N'rPA resolved the concern by means of an l acceptable post-weld qualification. The inspector reviewed the nondestructive tests performed on the reactor vessel closure studs, and concluded that NYPA's determination that no damage had occurred to the reactor vessel closure studs was accurate. (Section M2.3)
The procedure for outage risk assessment provided good guidance to the work planners for verifying that safety functions were maintained available. The inspectors reviewed the risk assessments on a routine basis, including the verification of available plant equipment against the status sheets, and identified no discrepancies. However, many station personnel did not understand the definition of risk Condition Yellow; this was due to a poor definition in the procedure and the lack of a defined training lesson plan. (Section M3.1)
The identification of the missed surveillance testing of the pressure isolation valves was a result of a self assessment of the Inservice Testing Program. This non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited violation. (NCV 50-333/98-07-03)(Section M8.2)
Enaineerina An inadequate engineering change notice allowed the installation of electrical cables to the main steam safety relief valve solenoids in the primary containment without the use of conduit. This installation was contrary to the final safety analysis report, General Electric requirements, and was a violation of NRC requirements (VIO 50-333/98-07-04).
Additionally, poor performance by personnel during the installation, and the lack of questioning attitude by personnel involved with oversight of the installation, resulted in the cables having an excessive length. An operability determination performed by NYPA concluded that the safety relief valves were operable.
Work on the emergency core cooling systems suction strainer project was properly controlled and the modification was adequately supervised by NYPA. Proper industrial safety, radiological safety, and foreign material exclusion practices were observed during work activities. Modification documents appeared comprehensive and the safety evaluation for the high pressure coolant injection and reactor core isolation cooling sucticn strainer replacements met the requirements of 10 CFR 50.59. (Section E2.1)
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i Executive Summary (cont'd)
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The inspectors considered the reload analysis, as submitted in the Core Operating Limits
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Report, to be acceptable and met the requirements of the Final Safety Analysis Report, the Technical Specifications, and the vendor design documents. (Section E 2.2)
2 The identification of an improper valva lineup during testing that bypassed the pressure
] suppression capability of the torus suppression pool was a result of a through review of a
similar industry event. This non-repetitive, licensee-identified and corrected violation is i
being treated as a Non-Cited violation. (NCV 50 333/98-07-05)(Section E8.1)
Plant Sucoort Overall radiological controls for outage work activities were generally effective in i minimizing dose and controlling contamination. ALARA program requirements with respect to the torus strainer modification were appropriately established and implemented.
Detailed procedures, extensive pre-job planning, comprehensive pre-job briefings, and close
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supervisory oversight were effective in addressing the changing radiological conditions in the torus in preparation for replacing the emergency core cooling systems : trainers.
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Prompt actions were taken to identify the source and limit the spread of contamination in
the reactor bUlding following an anomalous ventilation condition, immediate measures
, appeared effactive to prevent recurrences. Long term corrective actions appropriately addressed the suspected causes. (Section R2.1)
The licensee identified an unplanned exposure event that was the result of deficiencies in
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the application of radiological controls, including failure to recognize and properly assess a precursor event of a similar nature. In accordance with the established corrective action
- process, the licensee conducted a thorough and comprehensive root cause assessment, and planned and completed corrective actions designed to prevent recurrence appeared acceptable. This finding was identified as a Non-Cited violation. (NCV 50-333/98-07-06)
(Section R4.1)
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Quality Assurance audits /surveillances and third party evaluations effectively identified l 1 factors that could degrade radiological control program performance. Findings were l promptly communicated to the workforce at shift turnovers and planning meetings to expeditiously improve performance. (Section R7.1)
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TABLE OF CONTENTS
EX ECUTIV E S U M M A R Y . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii
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TAB LE O F CO NT ENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . v Summary of Plant Status ............................................1 1. O p e r a t i o n s . . . . . . . . . . . . . . . .. . . . . . . . . . . . . ........................ 1 O1 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01.1 G eneral Comm e n ts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01.2 Separated Hose During Torus Dewatering . . . . . . . . . . . . . . . . . . 1 02 Operational Status of Facilities and Equipment ................... 2 O2.1 Torus Modification Implementation ......................2 O2.2 Verification of Shutdown Safety System Instrumentation and Equipment (71707) .................................4 07 Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 07.1 Review of INPO Report (71707) . . . . . . . . . . . . . . . . . . . . . . . . . 4 08 Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 08.1 (Closed) Violation (50-333/98-02-01) . . . . . . . . . . . . . . . . . . . . . 4 08.2 (Closed) Violation (50-333/98-02-03) . . . . . . . . . . . . . . . . . . . . . 5 08.3 (Closed) Violation (50-333/98-02-02) . . . . . . . . . . . . . . . . . . . . . 5 08.4 (Closed) Violations (50-333/98-02-06and 07) . . . . . . . . . . . . . . . 5 11. M a i n t e n a n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 M1.1 G eneral Com m e nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 M1.2 Inspection of Pre-Refueling Activities . . . . . . . . . . . . . . . . . . . . . 7 M1.3 Refueling Operations ................................8 M2 Maintenance and Material Condition of Facilities and Equipment . . . . . . . 9 M2.1 Removal of Stuck Reactor Vessel Head ................... 9 M2.2 Inservice Inspection Activities . . . . . . . . . . . . . . . . . . . . . . . . . 10 M2.3 Material Concerns . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 M3 Maintenance Procedures and Documentation . . . . . . . . . . . . . . . . . . . 13 M 3.1 Outage Risk Assessment ...........................13 M8 Miscellaneous Maintenance issues . . . . . . . . . . . . . . . . . . . ....... 16 M8.1 (Closed) Licensee Event Report 50-333/98002-00. . . . . . . . . . . 15 M8.2 (Closed) Licensee Event Report 50-333/98006-00. . . . . . . . . . . 16 Ill . Engine e ring . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 E1 Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 E 1.1 G e ne ral Com ments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 E1.2 Implementation of Main Steam Safety Relief Valve Modifications .17 E2 Engineering Support of Facilities and Equipment ................. 19 E2.1 Emergency Core Cooling System Suction Strainer Modification . . 19 E2.2 Evaluation of the Reactor Core Reload Analysis . . . . . . . . . . . . . 21 E8 Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . 22 v
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Table of Contents (cont'd)
E8.1' (Closed) Licensee Event Report 50-333/97010-00. . . . . . . . . . . 22
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E8.2 (Closed) Violation 5 0-3 3 3/9 7 02-01 . . . . . . . . . . . . . . . . . . . . . 2 3 E8.3 (Closed) Unresolved item 50-333/97-05-02 . . . . . . . . . . . . . . . 23 1 i
I V. Pla n t S u p p o rt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 4 R1 Radiological Protection and Chemistry (RP&C) Controls ............24 R1.1 Implementation of the Radiation Protection Program .........24 R2 Status of RP&C Facilities and Equipment. . . . . . . . . . . . . . . . . . . . . . . 27 R2.1 Reactor Building Contamination incidents . . . . . . . . . . . . . . . . . 27 R4 . Staff Knowledge and Performance in RP&C . . . . . . . . . . . . . . . . . . . . 28 R4.1 Unplanned Exposure incident of September 17,1998 ........ 28 R7 Quality Assurance in RP&C Activities . . . . . . . . . . . . . . . . . . . . . . . . . 30 -
R7.1 Review of Outage Related QA Reports . . . . . . . . . . . . . . . . . . . 30 V. Ma nagement Meetings . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 ,
X1 Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 31 !
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ATTACHMENTS Attachment 1 - Partial List of Persons Contacted-Inspection Procedures Used i-Items Opened, Closed, and Discussed j
- List of Acronyms Used !
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Report Details Summarv of Plant Status The unit began the inspection period at 100 percent reactor power. On October 13,1998, reactor power was reduced to 70 percent due to problems with a feedwater heater level controller. On October 14,1998, power was further reduced to 65 percent due to increased feedwater heater level control problems. On October 16,1998, the reactor was shutdown for refueling outage number 13 (RFO13). The reactor was placed on shutdown I cooling on October 17,1998, wher> temperature was less than 212 d* grees. The reactor remained shutdown in a refueling outage throughout the remainder of the period.
- 1. Operations 01 Conduct of Operations 01.1 General Comments (71707) ,
l Using NRC Inspection Procedure 71707, the resident inspectors conducted frequent reviews of ongoinq plant operations. The reviews included tours of accessible and normally inaccessible areas, verification of engineered safety features (ESF) system l operability, verification of adequate control room and shift staffing, verification that j the unit was operated in conformance with Technical Specifications (TS), ;
observations of the reactor plant shutdown, observations of infrequently performed surveillance tests, and verification that logs and records accurately identified equipment status or deficiencies. In general, the conduct of operations was l
professional and safety-conscious; specific events and noteworthy observations are detailed in the sections below.
l 01.2 Seoarated Hose Durina Torus Dewaterina l
a. Inspection Scoce (37551. 71707)
During the final phase of the torus dewatering a temporary hose was over pressurized, causing a hose fitting to separate from the hose. The inspector reviewed the NYPA analysis of this event.
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l During the torus sludge removal and decontamination process a temporary hose connection separated while the rig was in use. The hose connection was located in the torus room at a temporary manifold connected to the radwaste system. The separation of the hose did not result in injury or the spread of contamination; however, an operator was sprayed with potentially contaminated water and a temporary electrical supply panel was sprayed and shorted out.
NYPA initiated a deficiency and event report (DER) and evaluated the event. The primary cause identified for the hose failure was improper operation of the system.
Temporary operating procedure, TOP 289, " Torus De-watering Operational Procedure," directed the pump to be secured prior to closing the shutoff valve.
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However, operators shut the valvo prior to securing the pump, which resulted in the l pump operating at shtrJ,ff head and the failure of the hose ezsembly. The l sequencing error was caused by poor communications between the control room and the pump operator.
As part of the corrective actions for this event, the procedure was revised to emphasize the required sequencing of the pump and valve operations. Personnel were briefed on the occurrence, and the use of three point communications was emphasized.
A second problem was the failure of the hose assembly. The hose was connected to the fitting with hose clamps. The shutoff pressure of the pump was determined to be approximately 80 psi, and the working pressure of the hose was rated to 150 psi. Although NYPA did not evaluate why the clamps failed to hold, the clamps were replaced by a higher pressure design.
The inspector reviewed the event critique and corrective actions, and concluded that the corrective actions adequately addressed the issues. The failure to perform the procedure as written is a violation of NRC requirements. This non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited violation (NCV), consistent with section Vll.B.1 of the NRC Enforcement Poliev. (NCV 50-333/98-07-01)
c. Conclusions in general, the conduct of operators during the torus dewatering was professional and safety-conscious. However, during the final phase of torus dewatering, the failure to perform the procedure as written caused the over pressurization of the hose. The failure of the temporary torus dewatering hose was evaluated and corrective actions were implemented. This non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited violation. (NCV 50-333/96-07-01)
02 Operational Status of Facilities and Equipment O2.1 Torus Modification imolementation a. Insoection Scoon (71707)
The inspector reviewed the temporary system installed by the New York Power Authority (NYPA) to drain the torus water for the strainer modification project. The inspector also observed portions of the torus draining evolution. Documents reviewed included temporary operating procedure (TOP) -289, temporary modification 96-059, and nuclear safety evaluation (NSE) JAF-SE-98-035.
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b. Observations and Findinas The temporary torus dewatering system was installed to drain the torus for the j installation of the new emergency core cooling systedi suction strainers. The dewatering rig was designed to transfer water from tiie torus, through a temporary i filtration unit, to a 450,000 gallon temporary storage' tank. The temporary dewatering rig was installed as a temporary modificat' ion, and a nuclear safety l evaluation (NSE) was performed to assure that the tefnporary modification did not involve an unreviewed safety question.
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l The inspectors walked down the temporary modification, reviewed the NSE, and reviewed the temporary operating procedure. The inspectors identified that several I items stated in the safety evaluation were not implerranted in the temporary )
modification or the procedure. For example, the NSE. assumed the installed filtration l
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system would reduce the activity in the water; however, there was no procedural requirement to sample the water to assure the NSE assumptions were met.
Another example involved the protection of plant safety systems from spray due to a postulated failed hose. The NSE stated the spray down of residual heat removal (RHR) system components was unlikely due to the insitallation of plastic barriers. l However, a walkdown of the system identified areas that were not adequately protected. Additionally, a leak during system pressura testing wetted an RHR pump. -
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The licensee acknowledged the inspectors concerns und initiated an effort to review the temporary modification. The licensee identified hoses between the reactor building and the filtration skid installed adjacent to the building. The NSE specifically stated that hoses outside of the building would incorporate a double containment provision. The licensee replaced the hoses with welded steel pipe.
The NRC and licensee identified issues were adequatsly addressed prior to performing the torus pump down.
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The licensee issued a deficiency and event report (DQl) to document the issues j associated with the implementation of the temporary modification. The DER
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identified an inadequate engineering review, inadequate plant operating review committee (PORC) review, inadequate project management review, insufficient detail in the temporary modification documentation, and inadequate operations involvement as contributing causes. Corrective actions for the DER were appropriately expanded beyond this temporary modification and included: (1) a quality assurance review of other temporary modifications and NSE's for this outage, and (2) a review of the temporary modification process. Although the conditions identified indicated poor performance, this was not a violation of NRC requirements.
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l The implementation of the torus dewatering system temporary modification was l
poor due to an inadequate engineering review, inadequate plant operating review committee (PORC) review, inadequate project management review, insufficient
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detail in the temporary modification documentation, and inadequate operations involvement. Identified deficiencies were adequately addressed prior to the utilization of the dewatering system.
02.2 Verification of Shutdown Safety System Instrumentation and Eouioment (71707)
As part of the review of the FitzPatrick risk assessment process (see Section M3.1 )
of this inspection report), the inspectors reviewed the availability of safety system instrumentation and equipment. In addition to a detailed review of the main control room panels, the inspectors reviewed the completed operations surveillance test ST-40X, " Daily Shutdown Surveillance and instrument Check," for the period November 15 to 21,1998. The inspectors identified no discrepancies during the !
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07 Quality Assurance in Operations l 07.1 Review of INPO Report (71707)
On October 8,1998, the Institute of Nuclear Power Operations (INPO) issued the !
interim report of their evaluation of the FitzPatrick station. The on-site eva!uation i was conducted August 17-28,1998. There were no findings discussed in the INPO i report that the NRC was not already aware of. No additional NRC inspection is i required. !
08 Miscellaneous Operations issues 08.1 (Closed) Violation (50-333/98-02-01): Failure to Carry Out the Actions of the Correct Procedure Durina the use of Emeraency Operatina Procedures (92901)
The violation was due to the shift operators failure to correctly utilize the Emergency Operating Procedures (EOPs) following a plant scram. Specifically, operators entered EOP-3, " Failure to Scram," which directed that oerators execute EP-3 * Backup Control Rod Insertion." Instead of entering EP-3, the operators incorrectly entered and implernented Abnormal Operating Procedure (AOP)-1
" Reactor Scram."
The inspector observed routine simulator training of an operating crew during their training cycle. The training consisted in part, of the use of the EOPs. Specifically, the operating crew was presented with the procedural requirement to enter EP-3
" Backup Control Rod Insertion," and they demonstrated their proiiciency in l executing the appropriate rod insertion methodology. The inspector concluded the j operating crew demonstrated proficiency in the use of EOP-3 and the corrective l
actions taken by NYPA were adequate to address the violation.
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l 5 08.2 (Closed) Violation (50-333/98-02-03): An Operator Aide Was Not Adeauate Concernino Plant Ooerations Durina Dearaded Flow Conditions (92901)
l The violation occurred based on an inadequate procedure, AOP-8, " Loss of Flow," l in that the power-to-flow map in the procedure differed from the maps of Operator Aid # 24 and Reactor Analyst Procedure RAP 7.3.16, " Plant Power Changes."
The inspector reviewed AOP-8, " Loss of Coolant Flow," RAP-7.3.16, " Plant Power Changes," and Operator Aid #24, and noted the power-to-flow maps that were contained in each were consistent. Based on the consistent power-to-flow maps i reviewed, the inspector concluded the corrective actions taken by NYPA were l adequate to address the violation.
08.3 (Closed) Violation (50-333/98-02-02): A Plant Procedure Was Not Adeauate Concernino Assioned Duties of the On-Shift Licensed Operator (92901)
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The violation occurred in that an administrative procedure was inades quote since it permitted reactor operators to direct the activities of other licent.ed individuals. A Senior Reactor Operators license is required to direct the activi',es of licensed individuals.
The inspector reviewed the revision made to plant procedure AP-12.03,
" Administration of Operations," and determined the duties and responsibilities of the shift operator were sufficiently clarified to assure that only licene d Senior Reactor Operators direct the activities of licensed individuals. In addition, a Final Safety Analysis Report (FSAR) change was initiated to clarify section 13.2, " Organizational Structure and Responsibilities," concerning the duties and responsibilities of the on shift operator. The inspector concluded the corrective actions taken by NYPA were adequate to address the violation.
08.4 (Closed) Violations (50-333/98-02-06and 07): Exclusion of Low Power or Shutdown Conditions From the Annual Operatina Test, and Exclusion of the Emeraency Plan From the Annual Operatina Test (92901)
The violations occurred due to the failure of NYPA to sample, during the annual operating test, low power and shutdown plant operations, and the emergency plan.
The inspector reviewed Training Procedure (TP-5.07), " Licensed Operator Requalification Examination Development and Administration," Revision 5. Based on this review, the inspector determined that low power and shutdown plant operations are required to be a part of the items sampled for possible inclusion in the annual operating examination. Also, the inspector determined that a sampling of all Senior Reactor Operators were required to be evaluated in their implementation of the Emergency Plan. The inspector concluded the corrective l actions taken by NYPA were adequate to close the violations, i
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ll. Maintenance M1 Conduct of Maintenance M1.1 General Comments I
a. Inspection Scope (61726,62707)
The resident inspectors periodically observed various maintencnce activities and l surveillance tests. As part of the observations, the inspectors evaluated the activities with respect to the requirements of the Maintenance Rule, as detailed in 10CFR50.65.
b. Observations and Findinas in general, maintenance and surveillance activities were adequately conducted. l However, observations during the period indicated severalinstances of poor work practices in the field that the inspectors considered weak. Although not considered violations of NRC requirements, the following examphs were discussed with appropriate levels of NYPA management throughout the period.
Personnel stood on a safety related motor operated valve instead of installing appropriate scaffolding.
Personnel testing a scram discharge volume drain valve utilized a 0-300 pound per square inch (psi) gage to verify test pressures in the 0-30 psi range.
Personnel, during troubleshooting, wrote technical information onto scratch paper and later transcribed the information rather than utilizing the troubleshooting )
procedure log directly. I i
Personnel conducting surface preparation of a residual heat removal system weld had an unapproved copy of the work request that was printed prior to the final revision and release of the document.
Several work areas were observed which exhibited particularly poor housekeeping practices. Examples included both radiological and fire safety concerns. l The inspectors reviewed procedures and observed all or portions of the following maintenance / surveillance activities: I AP-10.09 Outage Risk Assessment j ST-40X Daily Shutdown Surveillance and Instrument Check AOP-44 Dropped Fuel Assembly l
RAP-7.1.04 C Neutron Instrumentation Monitoring During in-Core Fuel
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Handling
. WR 98-02981 Repair Reactor Building Exhaust Fire Damper ST-20A Rod Worth Minimizer Functional Test
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t- 7 ST-50 - SRM [ Source Range Monitoring) Functional Test RAP-7.1.04B Spiral Offload /Onload Refueling Procedure
.WR 97-6474 Emergency Diesel Generator "B" Governor Actuator .
Replacement ST-398-X78 "B" Main Steam Line Combined Leakage Test ST-21 N Feedwater Pump Turbine Trip Test MP-093.06 Emergency Diesel Generator Woodward Governor Actuator Maintenance WR 98-02672' Repair Scram Discharge Vo'ume Drain isolation Valve WR 96-04458 Replace Emergency Service Water Keep Fill Check Valves WR 98-01805 RHR Piping Weld Inspection MP-029.01 Repair of Main Steam isolation Valves c. Conclusions ,
In general, maintenance and surveillance activities were adequately conducted. !
However, observations during the period indicated several instances of poor work practices in the field that the inspectors considered weak.
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M1.2 Inspection of Pre-Refuelina Activities a. Insoection Scope (60710)
The inspectors reviewed the pre-refueling maintenance activities, as specified in i station procedures, to determine if equipment checkout has been satisfactorily I (- completed prior to the disassembly of the reactor for the refueling outage. The inspector conducted interviews, reviewed work records and station procedures to access the maintenance activities.
b. Observations and Findinas Pre-refueling outage inspections of the rigging and special equipment used for i
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reactor disassembly was performed in accordance with maintenance procedure MP 4.01, " Disassembly of Reactor Vessel for Refueling (ISI)." The inspection of the equipment was directed utilizing the normal work control process with the quality l assurance staff performing the non-destructive examinations (NDE). The inspector l reviewed several work requests and NDE records for the equipment and found them
! to be complete. The inspector noted a good questioning attitude by the ( maintenance staff with regards to requesting clarification from engineering to !
L identify the " critical welds" on the handling equipment. The engineering response ]
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was thorough and provided additional information to the quality assurance i inspectors performing the NDE.
However, in the review of the critical load path for various lifts, the inspector
! questioned the licensee on the lack of NDE on the built in lifting eyes in the shield
! plugc. The licensee subsequently determined that it was prudent to inspect the ,
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lifting eyes and performed the inspections. In addition, the inspector noted that the l lifting equipment lacked identification labels. American National Standard institute f
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(ANSI), N14.6,1978, section 5.1.5, discusses identification of special lifting devices for the purpose of tracking the equipment's intended use and history. The licensee acknowledged the concem and opened a tracking item to review the requirement. The omission of labels and NDE of the lifting eyes were determined to be minor concerns and were properly addressed by the licensee.
During pre-refueling activities the inspectors noted an inconsistency in knowledge between station personne', engineering and maintenance, on the practice of lowering the reactor building crane main hoist into the spent fuel pool. The inspectors were concerned with the potential for internal corrosion of the hoist as a result of water intrusion. The licensee acknowledged the concern and subsequently I determined that the hoist had been lowered into the spent fuel pool on a few, rare I occasions. The licensee performed a boroscopic inspection of the hoist internals l and performed an analysis of the grease. The inspection results were satisfactory l with no evidence of water contamination or large rust particles. The licensee was i reviewing long term actions to ensure proper maintenance of the hoist following I water borne evolutions, c. Conclusions
Pre-refueling maintenance activities specified in the station procedures for the
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special lifting equipment were completed satisfactorily. The inspectors noted good interaction between the maintenance and engineering staffs for resolving questions.
The omission of labels and NDE of the lifting eyes were determined to be minor concerns and were properly addressed by the licensee.
M1.3 Refuelina Operations a. Inspection Scope (60710)
During the period, the inspectors observed portions of defueling and refueling operations from the refuel bridge and monitoring of the process from the control room. The inspector also observed classroom training of some of the individuals that were to be involved in operating the refueling bridge.
b. Observations and Findinas The inspectors verified that the refueling bridge interlocks had been tested, that sufficient source range monitors were operable, and that there was a licensed fuel handling senior reactor operator (FHSRO) on the bridge when fuel moves were in progress.
The fuel moves were well controlled and completed without incident. The i
communications on the refuel bridge were generally good, especially between the
! NYPA operators; the contract personnel were sometimes lax, but the FHSRO usually corrected them on-the-spot. The communications between the control room and L the refuel bridge operator were also noted as being good. The inspectors noted l
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l During observations of classroom training for refueling operations, the inspector l
noted that the instructors were very well prepared for the class, and the instructional material was of a high quality. The information provided to the students was presented in a manner that was conducive to the learning process.
c. Conclusions Fuel moves were well controlled and completed without incident. Communications on the refuel bridge were generally good, especially between the NYPA operators.
Classroom training for refueling operations was presented in a manner that was conducive to the learning process. Overall, the inspectors considered the refueling operations to be well coordinated.
M2 Maintenance and Material Condition of Facilities and Equipment M 2.1 Removal of Stuck Reactor Vessel Head a. Inspection Scope (62707)
During the removal of the reactor vessel head, the head became stuck and required the use of hydraulic jacks to complete the removal. The inspector reviewed the procedures, the engineering analysis, and observed portions of the activities to free the head, b. Observations and Findinas On October 19,1998, the reactor vessel head was being removed as part of the planned refueling outage. While lifting the 64-ton, two-foot-thick reactor vessel head, the head tilted slightly and became stuck on the hold down studs. Removal ;
of the head required that the head be lifted straight up and maintained level to clear the studs, in the initial attempt to free the head, mechanics used two hydraulic jacks to apply 24 tons of force to the low side of the head. The initial pressure was based on an existing calculation for supporting the reactor head in a storage stand. The work request was revised to allow jacking per engineering direction. The jacks were set on the flange outside of the sealing surface and between bolt holes. To avoid damage to the surfaces, u e jacks were set on approximately s6 inch square buy two inch thick aluminum pads. The top of the jacks were set against aluminum columns of approximately 3.5 inches in diameter. This effort did not free the stuck head.
Engineering reviews and calculations were performed to allow applying up to 400 tons of force to the head using four hydraulic jacks. The jacks were supported the same as the previous attempt. On October 20, the head was unstuck by the
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application of approximately 200 tons of force, and was removed without further incident.
The stud thread protectors and guides had some minor indications of scraping and were evaluated for repair or replacement. Based on visualinspections of the studs, NYPA determined that the stuck head did not cause any damage.
During the period of time that the head was stuck, the reactor vessel was flooded to just below the reactor flange and the decay heat removal system was utilized to cool the reactor. The outage management group performed a special review of the plant conditions and systems available to support reactor cooling. One train of low pressure coolant injection (LPCI) was added to the list of protected systems as an additional precaution. Operators were briefed on available contingencies in the event of a loss of shutdown cooling.
NYPA performed an evaluation of the head lift rig and identified that improper orientation of the lift rig caused the head to tilt slightly and become stuck. The four-point lift rig was previously marked for proper orientation; however the marks were obscured and the procedure did not specifically address indexing.
Because the reactor building crane did not have a load cell, the NRC questioned the total load applied to the lifting rig in the process of sticking the head. NYPA conceded that they had no way of knowing the actualload applied to the lift rig.
NYPA reviewed the capacities of the lifting rig and determined the lifting rig turnbuckles were the weakest components. A visual inspection was performed and the rig was determined to be satisfactory prior to lifting the head. The inspector considered the failure of NYPA to consider the possible over stress to the lift rig a weakness.
c. Conclusions The reactor head became stuck due to the lift rig being installed out of the usual orientation. The evolution of freeing the stuck reactor head was performed well with good participation by engineering, in support of maintenance. During the period with the head stuck, outage management and operations were diligent in maintaining appropriate plant conditions. However, the issue of the totalload applied to the lifting rig was not considered by NYPA and is considered a weakness in the engineering evaluation.
M2.2 Inservice inspection Activities a. Insoection Scope (73753)
i The inspector reviewed plans and schedules for the current inservice inspection (ISI)
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interval (first outage, first period, third interval) to verify compliance with the l
requirements of ASME Section XI,1989 Edition, no addenda ar:d 10 CFR 50.55a(g). Specific areas inspected included ASME Section XI ISI program
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coverage, qualifications and certifications of the non destructive examination (NDE)
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1 personnel, ISI NDE procedures and results, and New York Power Authority (NYPA)
' overr,ight of NDE contractors. The inspector observed selected NDE activities, ,
including remote visual examination of the in-vessel core spray piping (loop A), all l
tee boxes, all bracket support welds (loop B), the access hole cover plate weld at i 180 degrees and visual examination results of selected shroud vertical welds. In
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addition, the inspector observed the magnetic particle (MT) examination of one RHR i pipe to fitting weld ano the ultrasonic examination (UT) of one core spray pipe to fitting weld.
NYPA had NDE contractors perform ISI and in-vessel visual inspection (IVVI) l examinations and provided an oversight which involved review and approval of qualifications, procedures, test results, and monitoring and independent reexamination of selected tests.
b. Observations and findinas The inspector found the ISl work activities to be performed acceptably. The ISI procedures being used were approved by the ISI contractor and NYPA and were in accordance with the ASME Code requirements. The work was found to be thorough and of sufficient extent to determine component integrity. The inspector reviewed the ultrasonic and magnetic particle test procedures and found them to be adequatt, for the NDE tasks performed. The inspector found the inspection implementation consistent with the approved procedures, and the procedures were being used during the test activities. The personnel qualifications records for three NDE examiners were reviewed and found to be in compliance with the ASME Code requirements. The inspector evaluated oversight of contractor NDE activities by review of the NDE surveillance activity reports and oversight checklists, which documented appropriate NYPA involvement to monitor and verify NDE contractor compliance to applicable codes, procedures, and drawings.
The inspector reviewed surveillance data and documentation and found them to be in accordance with the ISI procedures and ASME Code requirements. NDT personnel performing inspections had properly identified and recorded indications and, where applicable, had performed further exploration of surface indications to determine relevance. NDE personnel appropriately retrieved prior examination records to confirm the presence of subsurface indications detected during the ultrasonic examination. The trackin0 cf ISI examination results indicated that the ISI program was in compliance with the ASME Code,Section XI for the specified period.
During the review of remote in vessel inspection results, the inspector identified an apparent anomaly on loop A core spray piping. The indicated location was re-examined by NYPA and contractor NDE personnel, who judged that the indication was not relevant; however, NYPA initiated an action commitment tracking form to identify the area for examination on RO 1<+. The inspector also noted that the location of one defect in the core shroud was indicated as partially traversing base metal and weld SH5 ID, but was in fact located at the intersection of horizontal weld SHS ID and vertical weld SV6B ID. The inspector requested NYPA review the
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! 12 video tape to confirm the location. NDE personnel reviewed the crack location and confirmed it to be approximately in the center of vertical weld SV6B ID with propagation into the horizontal weld SHS ID. The location of this crack was properly recorded in the final nondestructive test record, c. Conclusions The inspector concluded that inservice inspection had been performed acceptably and included acceptable ASME program coverage, qualified personnel, approved procedures, proper implementation, appropriate examination documentation, and NYPA oversight. The inspections performed were thorough and of sufficient extent to determine component integrity.
M2.3 Material Concerns a. Inspection scope The inspector reviewed those activities associated with the cutting of a temporary construction opening into the torus for personnel and equipment movement for the suppression pool suction strainer replacements. The activities inspected were specific to the removal and planned reinstallation (by welding) of the segment removed. The inspector reviewed the welding procedure to be used and the supporting procedure qualification records, filler metal test reports and installation sequence instructions.
Also, as a result of the jamming of the reactor vessel head against the head flange studs during vessel disassembly at the beginning of this outage, the inspector reviewed the steps taken to assure the integrity of the vessel closure studs, vessel head lifting lugs and the vessel head lifting rig (lifting eyes, plate welds and lifting pins).
b. Observations and findinas The inspector found that the cutting of the construction opening in the torus had been well planned and executed. The inspector examined the torus location where the cut had been made and found the area to be clean, orderly and maintained in a manner intended to preserve the material removed with provisions made to protect the access opening. The inspector concluded that effective steps had been taken to minimize the potential for distortion or damage to the removed material. NYPA intended to reinstall the removed material for closure of the access hole.
The inspector reviewed the weld procedure specification (wps) used to close the access hole at the completion of strainer installation. During post-welding review
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the inspector determined that the welding variables (amperage, voltage and travel speed) specified in the wps for the flux cored (fcaw) portion of the procedure could l result in exceeding the heat input levels qualified in the procedure qualification record (pqr). (Heat input is an essential variable when welding material where notch toughness requirements have been invoked for the base material. This is the case
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for all welding of the torus proper or attachments thereto.) Though NYPA believed that it was unlikely that the actual welding had exceeded the qualified heat input, there were insufficient records to confirm this. Subsequently, NPYA welded samples with the welding variables in excess of the permitted extremes, and destructive evaluation determined that the material parameters remained acceptable.
As such, the technical concern was resolved, and the weak control of welding process variables represented a violation of minor significance not subject to formal enforcement action.
The inspector reviewed the deviation event report (DER) which was initiated to track activities undertaken to free the vessel head from the interference with the closure studs. Examination results and documentation were reviewed and found to support the conclusion that the integrity of the closure studs and the vessel head lifting device had not been damaged. The critique of the event concluded that the head lifting device was not symmetric and was the likely cause of the side loading l condition that resulted in the binding of the head on the closure studs.
c. Conclusions The inspector concluded that the welding of the construction access opening in the torus had weak control of welding process variables that could have enabled the weld to have a heat input in excess of the qualification. NYPA resolved the concern uy means of an acceptable post-weld qualification. The inspector reviewed the nondestructive tests performed on the reactor vessel closure studs, and concluded that NYPA's determination that no damage had occurred to the reactor vessel closure studs was accurate.
M3 Maintenance Procedures and Documentation M3.1 putaae Risk Assessment a. Insoection Scope (62707)
As a continuing part of the refueling outage, each day NYPA performs an assessment of the potential risk related for several safety functions. The assessment considers available plant equipment and planned maintenance activities.
The inspectors reviewed several daily outage risk assessments, the related contingencies and safety evaluations, the controlling procedure, the technical specifications, and the associated training material. The inspectors interviewed severalindividuals as to their interpretation of the risk assessment and also
- discussed the risk assessment process with NYPA management.
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b. Observations and Findinas During major outages, the plant safety systems and functions are in unusual configurations and the operators are challenged to ensure that plant safety is maintained, in response to historical events during shutdown and low power l
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conditions, both the NRC and nuclear licensees reviewed the events and issued several recommendations en how to minimize challenges during these conditions.
Two of those publications are:
NUREG 1449, " Shutdown and Low-Power Operations at Commercial Nuclear Power Plants in the United States," issued by the NRC; and NEl 91-06, " Guidelines for industry Actions to Assess Shutdown Management,"
issued by the Nuclear Energy Institute The NYPA procedure for this process is Administrative Procedure (AP-10.09),
" Outage Risk Assessment." The procedure was based on the above publications.
The inspectors reviewed AP-10.09, and generally considered the procedure to provide good guidance for developing and communicating the risk associated with the plant equipment that is out of service for maintenance or testing. The procedure clearly defined the minimum requirements for each of the five safety functions: (1) decay heat removal, (2) inventory control, (3) reactivity, (4)
containment, and (5) electrical distribution. Each of the safety functions was classified as Condition Green, Condition Yellow, or Condition Red. As necessary, contingency plans were developed for safety functions that were other than Condition Green. The procedure defined the three condition levels, with Condition Green having all required systems and/or equipment being available. Condition Red was defined as having less that the required systems available, with a contingency plan developed and in effect. However, Condition Yellow was not clearly defined in the procedure, using the term " preferred" systems vice the terminology of
" required" systems used for red and green. The inspecton; considered the vagueness of this definition to be a weakness in the procedure.
Interviews with operations and planning personnel revealed that there was not a clear understanding of what Condition Yellow implied. The inspectors discussed this issue with the Supervisor, Operations Training, and learned that a formallesson plan had not been developed for the risk assessment process instead, a copy of the procedure was highlighted for use by the instructors. Due to the inconsistent understanding of Condition Yellow, the inspectors considered the lack of a detailed lesson plan to be a weakness in the training program.
Subsequent discussions related to these concerns with the Operations and Planning Managers resulted in a commitment to review the risk assessment procedure and the associated training, and to revise them as appropriate.
The inspectors routinely reviewed the safety system function risk assessments, including applicable contingency plans. During control room tours, the inspectors verified that the plant equipment listed on the system status sheets were, in fact, available; no discrepancies were identified.
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l The procedure for outage risk assessment provided good guidance to the work l planners for verifying that safety functions were maintained available. The inspectors reviewed the risk assessments on a routine basis, including the l verification of available plant equipment against the status sheets, and identified no discrepancies. However, many station personnel did not understand the definition i of risk Condition Yellow; this was due to a poor definition in the procedure and the l lack of a defined training lesson plan.
M8 Miscellaneous Maintenance issues '
M8.1 (Closed) Licensee Event Report 50-333/98002-00: Safety Relief Valve Setooint Drift in March 1998, during a review of test data, NYPA identified that two ofety relie,'
valves (SRVs) exceeded the technical specification (TS) allowable tolerance for as-found relief setpoint. The inspectors conducted an in-office review of the Licensee Event Report (LER), the Deficiency and Event Report (DER) and Action / Commitment Tracking System (ACTS) items, and the TSs. In addition, the inspectors discussed the issue with the responsible maintenance engineer.
The FitzPatrick TSs, Section 3.6.E.1, requires at least nine of the eleven SRVs to be operable. One of the associated TS required surveillance tests (4.6.E.1) states that at least five of the SRVs will be tested every 24 months, with all tasted being within a 48 month cycle. During a forced outage in December 1997, four of the SRVs were replaced with a different pilot assembly. The removed valves were sent to an independent laboratory for testing; on March 11,1998, the test results were reported to FitzPatrick. Two of the four valves exceeded the "as-found" lift setpoint tolerance of 1145 pounds per square inch gage (psig) plus/minus 3 percent (%).
NYPA assumed that the two SRVs had been inoperable since their installation during the previous refueling outage, and a review of plant records showed that no other SRVs were inoperable during that time frame. Therefore, since nine SRVs were operable during that period, there was no violation of TS Section 3.6.E.1.
NYPA determined the root cause for the recent failures to be the same as that reported in LER 95-06-01;specifically, corrosion between the pilot disc and seat, resulting in setpoint drift. For the risk analysis, although the two SRVs in question would have lifted at a setpoint higher than the allowed value, NYPA determined that the reactor pressure vessel emergency overpressure rating would not have been exceeded. Corrective actions included: (1) a continuing commitment to test all SRVs every refueling outage, and (2) incorporation of a BWR Owners Group modification to provide the SRVs with a pressure switch actuation during the next refueling outage The inspectors reviewed the LER, DER 98-0466, supporting documentation, and test data from the 1996 refueling outage; in addition, the inspectors discussed the failures of the SRVs with the responsible maintenance engineer. The root cause
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analysis, and corrective actions appear appropriate to prevent recurrence. The inspectors will review the test data from the current refueling outage when it becomes available. (IFl 50 333/98-07-02)
M8.2 (Closed) Licensee Event Reoort 50-333/98006-00: Missed Pressure Testina of
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Pressure Isolation Valves a. Insoection Scope (92700)
During a review of Inservice Testing (IST) program requirements, the licensee identified that the pressure isolation valves for the residual heat removal system, 10MOV-25A and B were not pressure tested as required by the ASME Code,Section XI. In addition, as part of the immediate corrective actions , the licensee identified another RHR system valve,10MOV-18, which was not tested within the last 24 months. The inspector reviewed the LER and the associated corrective actions, b. Observations and Findinas The licensee determined the cause to be inadequate implementation of other IST program requirements during implementation of Appendix J, Option B, leakage interval extensions for two of the valves and removal of the remaining valve from the type C leakage testing requirements. The valves were subsequently leak tested during the next shift with satisfactory test results. The inspector reviewed the actions taken by the operations staff in entering the TS limiting conditions for operation (LCO) and determined their actions to be appropriate. The licensee discovered the test omissions during the implementation of an IST self-assessment action plan, performed, in part, by the recognition of a high turn-over rate in their IST staffing. In addition the licensee performed a root cause analysis for the missed surveillance testing. The analysis summarized the causes to lack of thorough review of all code requirements, personnel turnover, and lack of an IST program basis document. The inspector concluded that the recommended corrective actions were extensive and properly captured the causes identified by the licensee.
This event was caused by a lack of thorough review of code requirements and is a viclation of NRC requirements. The safety significance of this error was low because there were indications and alarms available to the operators to detect valve i
ieakage, and the valves were found to meet the test acceptance criteria. This non-repetitive, licensee identified and corrected violation is being treated as a Non-Cited violation, consistent with section Vll.B.1 of the NRC Enforcement Policy. (NCV 50-333/98-07-03)
c. Conclusions The identification of the missed surveillance testing of the pressure isolation valves was a result of a self assessment of the Inservice Testing Program. This non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited violation. (NCV 50-333/98-07-03)
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E1 Conduct of Engineering E1.1 General Comments (37551)
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Using NRC Inspection Procedure 37551, the inspectors frequently reviewed design
! and system engineering activities and the support by the engineering organizations to plant activities. Specialist inspectors in this area used other procedures during l their reviews of engineering activities; these inspection procedures are listed, as applicable, for the respective sections of the inspection report.
l E1.2 Imolementation of Main Steam Safety Relief Valve Modifications a. Insoection Scoce (37551)
During a tour of the containment drywell, the NRC questioned the adequacy of the electrical cables connected to the main steam safety relief valve solenoid valves.
The inspector reviewed the NYPA response to this issue.
b. Observations and Findinas The inspector noted exposed electrical cables exiting a conduit and terminating at the main steam safety relief valve solenoid valves (MSSRVs). The solenoid valves allow remote operation of the MSSRVs and also control the automatic depressurization system (ADS) mode of operation of the ADS MSSRVs. The cables were notable due to their excessive length and disorderly installation. In some cases, up to fifty feet of cable was used to connect to the valve approximately four feet away. The excess cable was coiled, laid on pipes, or in some cases tied with a thin steel wire. The installation was similar at all eleven MSSRVs.
Modification DI-96-007, " Replacement of MSSRV 3-way Solenoid Valves," was conducted during a 1996 forced outage. Along with the replacement of the solenoid valves, the power supply cables were also replaced. The provision to install the cable without conduit was added to the design package by NYPA engineering, through an engineering change notice. The originalinstallation used conduit over the cables.
NYPA initiated a DER and performed an operability determination, and concluded the MSSRVs were operable, as installed. However, the review of the DER identified that the installation of exposed cable within containment was not in accordance
- with the final safety analysis report (FSAR) and a General Electric technical report, NEDO 10139. The FSAR, section 7.1.9, states that all cabling inside the primary containment is routed in conduit. The General Electric document required the installation of the ADS wiring in separate conduits to reduce the probability that a
, single event would prevent the opening of more than one relief valve.
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Additionally, the excessive cable length was not in accordance with the work instructions. The work requests required the cables to be cut to the proper length to reach the solenoid operated valves. Based on the engineering evaluation that the MSSRV's were operable, the failure to adequately perform the maintenance procedure is considered a violation of minor significance, and therefore is not subject to formal enforcement action. l As a corrective action, NYPA initiated work requests to cut the cables to proper length and place the cables in conduit. Additionally, an extent-of-condition review was done which identified additional exposed cables associated with the main steam isolation valves. An engineering evaluation to allow the existing l configuration for the main steam isolation valve cables was ongoing, and was listed i as a startup item.
The DER response also addressed corrective actions for the design engineering error. Actions included briefing of personnel, and a revision of electricalinstallation procedures to specify the requirement for conduit on electrical cables located within the primary containment.
10CFR50, Appendix B, Criteria Ill, " Design Control," requires the design basis for systems, structures, and components be correctly translated into specifications and instructions. The FSAR, section 7.1.9, states that all cabling inside the primary containment is routed in conduit. General Electric technical report, NEDO 10139,
" Compliance of Protection Systems to industry Criteria: General Electric BWR Nuclear Steam Supply System," dated June 1970, section 3.3.1.2, states that the installation of the ADS wiring is in separate conduits to reduce the probability that a single event would prevent the opering of more than one relief valve. Contrary to the above, modification DI-96-007," Replacement of MSSRV 3-way Solenoid Valves," conducted during a February 1996, forced outage installed cabling to the MSSRV solenoid valves without the use of conduit. This is a violation of NRC requirements. (VIO 50-333/98-07-04)
c. Conclusions An inadequate engineering change notice allowed the installation of electrical cables to the main steam safety relief valve solenoids in the primary containment without the use of conduit. This installation was contrary to the final safety analysis report, General Electric requirements, and was a violation of NRC requirements (VIO 50-333/98-07-04). Additionally, poor performance by personnel during the installation, and the lack of questioning attitude by personnel involved with oversight of the installation, resulted in the cables having an excessive length. An operability determination performed by NYPA concluded that the safety relief valves were operable.
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E2 Engineering Support of Facilities and Equipment E2.1 Emeroency Core Coolino System Suction Strainer Modification a. Insoection Scope (93803)
The inspectors reviewed key aspects of the modification to replace the emergency core cooling systems (ECCS) suction strainers with large, passive strainers in resporse to NRC Bulletin 96-03.
b. Observations and Findings The modification to the ECCS suction strainers involved the installation of six large, i stacked-disk shaped passive strainers fabricated by the PCI Corporation. The two largest strainers were multi-segmented units for the RHR suction lines; two smaller multi-segmented strainer units were being installed on the core spray suction line.
Although not required by the NRC Bulfetin, NYPA elected to replace the stainers on the high pressure coolant injection (HPCI) and reactor core isolation cooling (RCIC)
suction lines with single strainer units, which have significantly larger surface area than the previous strainers. In support of this work, an access hole was cut in the torus for personnel safety reasons, and the torus was dewatered and cleaned.
The inspector reviewed the following documents regarding this modification:
The ECCS Suction Strainer Project Organization Chart
The September 2,1998, Readiness Review Meeting on the status of the ECCS suction strainer replacement project
Minor Modification Package No. M1-97-131, Rev. O, regarding the replacement of mineral wool piping insulation in the drywell
Safety Evaluation 98-025, Rev. O, regarding the replacement of the torus suction strainers for the HPCI and RCIC systems
Modification F1-98-100, Rev. O, regarding the replacement of the torus suction strainers for the HPCI and RCIC systems
Preliminary Torus Water Storage Tank Suction Piping Leak Action Plan, JPRJ-APL-98-001, Rev. O
Foreign Material Exclusion Plan for Modification #F1-97-031 Regarding the documents reviewed, the inspector found the modification documents to be comprehensive and the safety evaluation to meet the requirements of 10 CFR 50.59. However, the inspector questioned whether confirmatory inspections of the drywell would be made to verify that no other mineral wool existed in the drywell, since the new strainers are not sized to handle a mineral wool debris loading; no evidence of a proactive inspection plan was provided. Subsequent discussions with
, engineering revealed that such inspections would be conducted in response to
- multiple DERs, which indicated that sections of instrument lines noted as insulated per mod M1-97-131, were either uninsulated or were insulated with materials other than those prescribed. None of the insulation materials noted were mineral wool and engineering analysis indicated most of the uninsulated lines did not require
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insulation. A tour of the drywell by the inspector confirmed the in-progress replacement of the most sizeable sections of mineral woolinsulation; no remaining mineral wool was visible during this tour.
After the torus water was transferred to the temporary storage tank, a smallleak developed in the inner tank liner at the piping connection. The preliminary torus water storage tank piping leak action plan was thorough and comprehensive, providing actions commensurate with the size of the leak experienced, inspector observations of the storage tank noted a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> watch on the tank contents and checks on the piping leak rate and tank temperature every 2 hours2.314815e-5 days <br />5.555556e-4 hours <br />3.306878e-6 weeks <br />7.61e-7 months <br />. Plots of the piping leak rate versus time indicated that the leak remained constant at a low flow rate; calculations also indicated that the tank contents would not freeze unless the tank continued to be used well past the end of November. 1 The Foreign Material Exclusion (FME) program provided an adequate method of ensuring FME control, provided a very thorough internalinspection of the suction strainers and the torus was performed prior to closecut. FME controls during the bulk of the construction efforts in the torus were limited to capping open penetrations.
The inspector toured the torus on two occasions to observe industrial safety practices, particularly during rigging evolutions, and to check the condition of interior coatings, preparations for piping cuts, and the installation of an additional seismic support on the HPCI suction line. No concerns with industrial safety l practices were noted. Frequent tours by NYPA industrial safety representatives were noted. Physical examination of selected strainer components was also conducted. Discussions with NYPA's coatings specialist revealed that instances of rust pitting observed on the torus inner wall were of limited depth and that it was progressing at a slow rate; therefore, no coatings repairs were planned for this outage. The use of a full scale mockup of one bay of the torus provided welders an excellent opportunity to perform cutting on the torus shell and rewelding of the plate removed, one of the most critical evolutions of this modification.
Observations of daily plant and ECCS project team status team meetings indicated that the project was being supervised and controlled by NYPA personnel and/or long-term NYPA contractors. Detailed outage plant configuration risk assessments with contingency actions were presented at daily plant meetings to ensure that system configuration changes posed by this modification and other plant activities did not adversely impact plant safety. Diccussions were also conducted with the ECCS suction strainer project manager regarding the cause for emergent work experienced during the implementation of the modification (i.e. torus water transfer system problems, torus water storage tank suction pipe leak, and as-low-as-reasonably-achievable (ALARA) concerns regarding the torch cutting of the RHR, I
Core Spray, HPCI and RCIC suction piping). NYPA was in the process of conducting a DER analysis of the water transfer system planning and
- implementation problems. The principle impact of this emergent work was a significant outage schedule delay.
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c. Conclusions I
Work on the emergency core cooling systems suction strainer project was properly j controlled and the modification was adequately supervised by NYPA. Proper ;
~ industrial safety, radiological safety, and foreign material exclusion practices were ;
observed during work activities. Modification documents appeared comprehensive l and the safety evaluation for the high pressure coolant injection and reactor core isolation. cooling suction strainer replacements met the requirements of 10 CFR L 50.59.
l l E2.2 Evaluation of the Reactor Core Reload Analysis l
a. Inspection Scooe (37551)
The inspectors reviewed the reactor core reload analysis for the current cycle, as
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summarized in the Core Operating Limits Report (COLR). The COLR was based on
.several design documents provided by the vendor, General Electric Nuclear Energy (GENE), and the FitzPatrick TS. The inspectors reviewed the basis documents, the COLR, and the associated safety evaluation; and discussed the COLR with the Supervisor, Reactor Engineering.
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b. Observations and Findinas l
As required by the FitzPatrick TS, Section 6.9(A)4, NYPA established the COLR for the upcoming cycle.- The COLR is the plant specific document that provides operating limits for the current reactor core reload cycle. Those limits are: (1)
- . average planar linear heat generation rate, (2) average power range monitor flow-I biased thermal power scram setpoint, (3) linear heat generation rate, (4) minimum i
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critical power ratio, (5) core flow adjustment factor, and (6) rod-block instrumentation setpoint. The inspectors verified that the information in the basis documents provided by GENE, listed below, was accurately translated into the COLR.
" Supplemental Reload Licensing Report for James A. FitzPatrick," Reload 13, Cycle 14, J11-03359SRL, Revision 1, Class I, dated October 1998
" Lattice Dependent MAPLHRG Report for James A. FitzPatrick," Reload 13, I
Cycle 14, J11-03359MAPI, Revision 0, Class Ill, dated October 1998 The inspectors reviewed the Nuclear Safety Evaluation (NSE 98-006, Revirion 0)
and the associated Minor Modification Package (MMP No, M1-97-30, Rev.sion O).
The NSE was consistent with the GENE reference documents, and adequalely addressed the reload design and related changes in the safety and thermal limits.
The COLR was verified to be consistent with the controlling administrative t
' procedure AP-12.05, " Control of Core Operating Limits Report." In additior?, the inspectors attended the Plant Operating Review Committee (PORC) meeting that
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reviewed and approved COLR. During that PORC meeting, the inspectors verified l' _ that there was a quorum of PORC members, as required by the TSs. The inspectors l
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reviewed the Final Safety Analysis Report (FSAR) and confirmed that the COLR was I consistent with the requirements of the FSAR. '
c. Conclusion The inspectors considered the reload analysis, as submitted in the Core Operating Limits Report, to be acceptable and met the requirements of the Final Safety Analysis Report, the Technical Specifications, and the vendor design documents.
E8 Miscellaneous Engineering issues E8.1 (Closed) Licensee Event Report 50-333/97010-00: Surveillance Testina of the Pressure Sucoression Chamber - Drvwell Vacuum Breakers Cottf,4 ave Resulted in a Partial Loss of the Primarv Containment Pressure Suporession ' safety Function a. Inspection Scoce (92700)
The inspector reviewed the LER and the associated corrective actions.
b. Observations and Findinas The condition reported in this LER involved bypassing the pressure suppression safety function of the torus during surveillance testing. This was accomplished by opening the drywell vent valves and the torus vent valves simultaneously to vent the torus to the drywell. The resulting flow path would have allowed the pressure spike due to a loss of coolant accident (LOCA) to bypass the torus suppression pool and over pressurize the containment.
The licensee identified and reported this condition during a review of industry events. Corrective actions included a review of surveillance tests for similar discrepancies, and revising related tests with an alternate test method. The inspector reviewed the revised test procedures and determined that the revised procedures corrected the condition.
This event was caused by an inadequate technical review of the associated plant procedures and was a violation of NRC requirements. The safety significance of l this error was low because the valve configuration was only established for venting and was not maintained for sustained periods of time. Additionally, the torus and drywell vent valves automatically close upon receiving a primary containment l isolation signal. This non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited violation, consistent with section VM.B.1 of the NRC
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Enforcement Policv. (NCV 50-333/98-07-05)
c. Conclusions I
The identification of an improper valve lineup during testing that bypassed the pressure suppression capability of the torus suppression pool was a result of a through review of a similar industry event. This non-repetitive, licensee-identified
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and corrected violation is being treated as a Non-Cited violation. (NCV 50-333/98-
-07-05)
E8.2 (Closed) Violation 50-333/97-02-01: Failure to Take Adeauate Corrective Actions Followino Previous Shutdown Coolina isolation Events Previous corrective actions for resolution of the long term problem of shutdown cooling isolation events were determined to be inadequate. The plant had a history of the shutdown cooling isolation valves going shut upon starting the residual heat removal pump in the shutdown cooling mode of operation. Previous corrective l actions included the modification of a pressure switch sensing line to reduce the j potential for air entering the instrument lines, this action did not eliminate the
l problem. In addition, engineering had also recommended installation of a high point vent in the piping system. This action was not taken by the licensee because of the air trap in the sensing line was believed to be the source of the problem and, as l
l such, the system high point vent was made a low prioritv. Recent corrective I actions taken by the licensee included installation of the high point vent during the current refueling outage and procedure changes to direct the amount of time the i system is to be filled and vented. Corrective actions were determined to be l complete.
E8.3 (Closed) Unresolved item 50-333/97-05-02: Control Rod Drive Closure Bolt Crack Indications While performing a closeout review of the inservice inspection program (ISI)
following the previous refueling outage, the licensee identified that the bolts
, removed from the control rod drives (CRDs) had not been properly inspected. The l cause of the error was due, in part, to an improper maintenance procedure change i
that was not reviewed by the ISI program engineer. The bolts were subsequently ,
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retrieved from the waste facility and inspected. The issue remained unresolved l pending review of the licensee's ISI program with respect to procedure controls and a closure review of the licensee's supplemental acceptance criteria for the CRD closure bolts.
l During this inspection period, the inspector reviewed the current process for
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conducting maintenance on components within the scope of the licensee's ISI l
program. The inspector discussed the process with engineering, maintenance and l
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planning staff. The current work control process identified, as part of the planning process, work packages which required ISI engirseer review. This reduced the i potential for inspection being omitted as the result of maintenance procedure
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changes. In addition, the inspector reviewed the licensee's root cause analysis and corrective actions and found them to be appropriate. The inspector concluded that no violations of NRC requirements existed.
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ts. 24 IV. Plant Support
, Using NRC Inspection Procedure 71750, the resident inspectors routinely monitored l: ~ the performance of activities related to the areas of radiological controls, chemistry, p ' emergency preparedness, security, and fire protection. Minor deficiencies were l discussed with the appropriate management, significant observations are detailed ll . below. : Specialist inspectors in the same areas used other procedures during their l
reviews of plant support activities; these inspection procedures are listed, as applicable, for the respective sections of the inspection report.
I R1 ' Radiological Protection and Chemistry (RP&C) Controls l R1.1. limolementation of the Radiation Protection Proaram
'a. Insoection Scope (83750)
A review was performed of the external and internal exposure controls implemented during the refueling outage. Particular emphasis was placed on the radiological controls associated wkh preparing the torus for replacement of the ECCS strainers.
The inspection m, ;J torus dewatering, water processing and transfer, torus i
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desludging and decontamination, and shielding / contamination control measures applied to contaminated components within the torus. -
L The radiological controls applied to other outage-related activities were reviewed.
! through observation of work in radiologically controlled areas, examination of relevant records, including radiation work permits, surveys, as low as is reasonably
- achievable (ALARA) reviews, and discussions with cognizant individuals.
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. b. Observations and Findinas
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Torus Modification i
Overall planning and preparation for torus work activities was good. Similar I l modifications at other facilities were evaluated, experienced crews were used, and a l torus mock-up was used for training construction crews. An ALARA review L
(98-031) was prepared detailing the various aspects of the project requiring exposure controls. Separate plans were developed and attached to the ALARA review including the project's overall radiological control plan, miscellaneous items
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and contingencies in support of the project (RTID 8-98-010), a torus water pumpdown/ temporary storage plan, a torus ventilation plan, a torus 4- decontamination plan, and a torus shielding plan. These plans adequately
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addressed the radiological and engineering challenges the torus modification i
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Prior to beginning the draindown, a pre-job briefing was conducted detailing l' Individual responsibilities, contingencies, and the activities contained in the Temporary Operating Procedure (TOP-289). Participants were knowledgeable of
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1 procedural requirements and demonstrated a questioning attitude to address I contingencies.
Engineering controls were effectively used to remove contaminants from the water j prior to transferring and temporarily storing the processed water onsite. Minor leaks l were promptly identified and corrected. The double-containment temporary water storage tank held approximately 300,000 gallons of processed water, which contained a total activity of about 1.8 E-3 Curies, which was less than the Technical Specification limit of 10 Curies. Following processing, surveys of the transfer line and tank confirmed that dose rates were within acceptable limits.
Access to the water processing trailer was properly controlled, through locked !
gates. Postings were promptly changed during water processing to reflect increased dose rates on system components. Following completion of water ;
processing, the trailer was expeditiously moved to the Radwaste Building where l contaminated components were flushed and contaminated resin / filters transferred in i preparation for eventual disposal offsite.
An effective torus decontamination strategy was implemented. Beginning with the initial draindown to the reactor cavity, workers in collapsible boats scrubbed the bath tub ring and accessible surfaces in the torus. Following draindown, with less than a foot of water remaining, decontamination, using hoses, squeegees, and scrub pads, was performed in conjunction with desludging activities. Personnel performing these activities were appropriately equipped with vu:ner repellent apparal, wireless remote dosimetry, extremity dosimetry, te ex headsets, and respiratory protection. Air sampling was appropriately puformed.
On October 27,1998, all in-torus work was suspended following indications of high airborne activity; confirmatory samples were taken and analyzed. Upon determining the activity was naturally occurring radon, work resumed.
Subsequent to desludging, floors and walls were decontaminated using a Hot Z (pressurized, warm water) cleaning machine to further reduce contamination levels.
Surveys were performed to determine decontamination effectiveness and identify hot spots requiring more aggressive measures. Components difficult to decontaminate, (e.g., T-quenchers and downcomers), were wrapped in plastic, with lead blankets secured to lower dose rates.
Close supervisory oversight improved the efficiency of staging and installing shielding, scaffolding, and deck grating. Control of foreign materials, housekeeping, and overall industrial safety conditions improved as jobs progressed.
Contamination control measures were effectively used. Torus surfaces to be cut or welded were appropriately prepared using paint remover, instead of grinding, to limit generation of airborne radioactive contamination. Due to solvent vapor build-up in I the work areas, temporary ventilation systems (installed to remove airborne l
contamination and reduce heat stress) were optimized. Portable ventilation units were also used to reduce airborne contamination.
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Aporopriate personnel dosimetry was worn during work activities. Radiation monitoring devices were supplemented with additional dosimetry for workers -
entering areas exhibiting dose rate gradients.
A departure from otherwise proactively planned activities was the reactive effort
'taken to assure that torch cutting of highly contaminated RHR and Core Spray suction lines was completed from an ALARA perspective. The subsequent .
measures were effective in minimizing the spread of contamination.
Other Outaae Activities Cameras, wireless remote dosimetry, and radio communications equipment were i widely used to limit the number of individuals working in radiation areas and to ,
improve worker efficiency.
Low dose waiting areas and "No Lingering" areas were conspicuously posted, with !
workers observed generally complying with this guidance. Temporary shielding was extensively used for dose intensive tasks and to lower general area dose rates in high traffic areas. Line flushing and hydrolasing of primary system components was partially effective in lowering dose rates.
Pre-job ALARA briefings were, in some cases, videotaped to assure that detailed
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information was consistently communicated to a variety of workers. ALARA review 98-024, for removing the under vessel shootout steel and changing out control rod drives, was appropriately detailed identifying anticipated radiological conditions, dose projections, pre-job preparations, access controls, contingencies, and personnel monitoring requirements. The associated pre-job Radiation Work Permit (RWP) (98-0500) briefing conducted to support the (under vessel) steel removal on October 28,1998, was formalized. A briefing checklist was adhered to, with the most current survey data and measures to minimize dose clearly communicated to the workers. Overall conditions and resources available in the drywell quiet room
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for conducting briefings was conducive for effective communications.
Collective 9xposure goals were established and performance was closely tracked for outage tasks. The outage cumulative dose projection was about 343 person-rem with aggressive individual departmental goals established.
c. Conclusion Overall radiological controls for outage work activities were generally effective in minimizing dose and controlling contamination. ALARA program requirements with respect to the torus strainer modification were appropriately established and
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imralemented. Detailed procedures, extensive pre-job planning, comprehensive pre-
! Job briefings, and close supervisory oversight were effective in addressing the changing radiological conditions in the torus in preparation for replacing the emergency core cooling systems strainers.
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l R2 Status of RP&C Facilities and Equipment R2.1 Reactor Buildino Contamination incidents a. Scope (83750)
On October 21 and 23,1998, an increase in personnel contaminations occurred in the Reactor Building (DER 98-02657). A review was made to determine the adequacy of the actions taken to evaluate the cause and to correct this recurring i condition. '
b. Observations and Findinas
An increase in personnel contaminations, primarily foot contamination identified in the portal monitors, occurred on the subject dates resulting in prompt work stoppages until the contaminated areas on the 326' and 344' reactor building elevations could be identified and cleaned. An in-depth investigation was subsequently performed by Systems Engineering to establish what plant activities and systems contributed to this anomalous condition and what actions were needed to preclude recurrence. These did not constitute a violation of NRC requirements. ;
The cause was attributed to a combination of plant conditions in which the reactor building ventilation system was isolated (a contamination control measure) for extended periods to permit removal of highly contaminated, hot, reactor components; and leaks in the Below Refuel Floor Ventilation System. Ventilation isolation resulted in a positive pressure building condition in which loose surface contamination migrated to lower reactor building elevations from the refuel floor.
As immediate corrective actions, temporary repairs were made to the Below Refuel Floor Fan flexible joints and duct work penetrations and minimizing the periods of isolating reactor building ventilation. These actions prevented a recurrence.
Long term actions include minimizing reactor / spent fuel pool temperatures to the extent practical to reduce air convention currents and improving the sequencing of ventilation systems. Additionally, associated duct work will be surveyed and decontaminated. These actions were incorporated into the Action / Commitment Tracking system, c. Conclusions Prompt actions were taken to identify the source and limit the spread of contamination in the reactor building following an anomalous ventilation condition.
Immediate measures appear effective to prevent recurrences. Long term corrective
- actions appropriately addressed the suspected causes.
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R4 Staff Knowledge and Performance in RP&C l
R4.1 Unolanned Exposure incident of Septgnber 17.1998 l
a. Insoection Scoce (83750)
Deviation / Event Report (DER) 98-02288 documented an incident in which a )
maintenance mechanic received an unplanned exposure as a result of not hearing or !
seeing the Electronic Dosimeter (ED) alarms while working in a Locked High l Radiation Area (LHRA). The adequacy of the Root Cause Analysis (RCA) and l corrective actions that resulted from this incident were evaluated. No regulatory i dose limit was exceeded. l b. Observations and Findinas On September 17,1998, a maintenance mechanic received a unplanned dose of 459 mrem while changing a pipe snubber in the Reactor Water Clean Up (RWCU)
Pump Room. The dose was 259 mrem above the ED integrated dose alarm setting of 200 mrem, and the work was conducted in a radiation field of about 7200 mrem per hour, which exceeded the ED dose rate alarm setting of 5000 mrem per hour. The individual was unable to hear the alarm, in part, due to the high noise levelin the area, and was unable to see the alarm because the ED was placed on his upper back (the area expected to receive the highest dose) under his protective clothing. The individual was part of a work group consisting of a second mechanic, a QA technician, and a radiological controls technician. Upon exiting the work area and crossing the step-off pad, a radiological controls (radcon) technician observed the worker's ED alarming. An Electronic Dosimeter Alarm Evaluation form was completed documenting the observed condition, reviewed by supervision, and a DER was generated to initiate management actions to evaluate circumstances resulting in the incident. Critiques were promptly conducted with the cognizant I workers. Details of the incident, with immediate corrective measures and lessons learned, were promptly communicated to the work force through departmental stand-down meetings, shift turnover briefings, and daily management nieetings. i Additionally, plant areas that were high radiation areas and high noise areas were )
evaluated to determine the use of modified dosimetry; (e.g., the use of wireless I remote monitors (WRM)), when working in such areas. The radiation protection staff s;vas briefed on management expectations regarding the use of stay time, WRMs, supervisory oversight regarding dosimeter setpoint changes, and threshold for initiating DERs. I Subsequent to reviewing the DER, the Performance Enhancement Review Committee (PERC) determined that a Team Root Cause Analysis should be performed to identify the underlying programmatic issues and human performance problems, and to develop corrective actions to preclude further occurrences. The i resulting Deviation and Event Analysis described seven contributing causes that I included missed opportunities to learn from past experience, invalid assumptions and inconsistent work practices by the Chief Radiological Controls Technician in establishing the task's radiological controls, and incomplete communications by
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! 29 maintenance supervision / technician regarding the specific job location. Past opportunities to recognize the limitations of EDs in high noise areas were missed as a result of an improper closure of a relevant Quality Assurance audit recommendation, identified in 1996, and failure to initiate a DER for a similar incident that occurred on August 13,1998. Improper radio!ogical work practices included job survey shortcomings, invalid assumptions on worker stay time, poor supervisory oversight for a job in a radiological significant area, and the lack of formal management expectations regarding the use of ALARA reviews, choice / placement of dosimetry, and rigorous communications of changes in job scope. Separate DERs were initiated for the identified procedural non-compliances to better establish their root causes.
To address these shortcomings and human performance deficiencies, broad based corrective measures were developed based on input provided by site departments and third party assessments. Procedures and programs were being revised to strengthen the process for closure of QA recommendations, and DER issuance.
Management expectations were being formalized regarding improving communications of job scope changes to the radcon department, better coordination of tasks in LHRAs with the ALARA group, and closer supervision of the implementation of radiological controls in LHRAs.
The root cause analysis and supporting corrective action records were discussed with the RES Department Manager and General Supervisor - Health Physics. Based upon information provided in these interviews and documents, the inspector concluded that this incident was a violation of TS 6.11(A)1.b. "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 which continuously integrates the radiation dose rate in the area and alarms when a preset integrated dose is received." Implicit in this specification is that the use of stub a device is sufficient to inform the worker that established dose and dose rate limits are being exceeded.
The root cause analysis was thorough and identified various program controls that were ineffective in preventing the incident. The immediate and long term corrective actions were comprehensive and appropriate to prevent a recurrence. Additionally, at the request of the inspector, the licensee re-evaluated the incident of September 17,1998, to determine whether the affected individual could have potentially received an exposure in excess of the limits stated in 10 CFR 20. This evaluation, provided to the NRC on November 13,1998, addressed various scenarios with changes in task duration, dosimetry location, work area dose rates, shielding, source distance, and worker assignment. From this review, it was determined that no substantial potential existed for the individual to receive a dose in excess of the regulatory limits.
Accordingly, this licensee identified and corrected violation is being treated as a l
Non-Cited Violation (NCV), consistent with Section Vll.B.1 of the NRC Enforcement
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Policy. (NCV 50 333/98-07-06)
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c. Conclusions The licensee identified an unplanned exposure event that was the result of deficiencies in the application of radiological controls, including failure to recognize and properly assess a precursor event of a similar nature, in accordance with the
- established corrective action process, the licensee conducted a thorough and comprehensive root cause assessment and planned and completed corrective actions designed to prevent recurrence appeared acceptable. This finding was identified as a Non-Cited violation. (NCV 50-333/98-07-06)
R7 Quality Assurance in RP&C Activities R7.1 Review of Outaae Related QA Reoorts a. Inspection Scope (83750)
Quality assurance (QA) audits and surveillances for pre-outage preparations and outage activities were reviewed to determine the effectiveness of the oversight in assessing radiological controls program performance.
b. Observation and Findinos Quality assurance audits A98-19J," Site Contractor Activity Controls," and A98-18J, " Radiation Protection Plan," effectively evaluated the contractor radiation protection technician queiication program and implementation of the ALARA program, respectively. In particular, audit A98-18J provided an evaluation that'
identified issues such as drywell constant air monitor calibrations, inconsistent worker practices, and the disposition of past findings. DERs and corrective actions items were appropriately developed.
- Weekly QA surveillance reports (SR Nos. 2035,2056,2057,2058, and 2061) of pre-outage preparations and plant support activities conducted during the outage noted proper worker practices and evaluated overall ALARA program effectiveness.
Evaluations of the ALARA program were based on comparison of actual dose with departmental aose goals based on monitored data.
Surveillances were conducted weekly and directed at significant activities in progress. Third party evaluations were routinely provided by technical specialists from other utilities and the corporate staff. Findings relevant to improving radiological controls and industrial safety were promptly comraunicated to site departments at shift turnover briefings.
c. Conclusions Quality assurance audits /surveillances and third party evaluations effectively identified factors that could degrade radiological control program performance. a Findings were promptly communicated to the workforce at shift turnovers and planning meetings to expeditiously improve performance.
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V. Manaaement Meetinas X1 Exit Meeting Summary The inspectors presented the inspections results to members of the licensee management at the conclusion of the inspection on December 4,1998. The licensee acknowledged the findings presented.
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i ATTACHMENT 1 PARTIAL LIST OF PERSONS CONTACTED Licensee
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M. Anderson Radiation Protection Supervisor, Torus Modification D. Bel! Senior Quality Engineer '
J. Bracey Administrative Coordinator, RES R. Brown Radiation Protection Supervisor G. Brownell Licensing Engineer P. Brozenich Operations Manager M. Colomb Site Executive Officer W. Comstock Quality Assurance Audit Coordinator R. Converse General Manager Maintenance D. Cristafulli Radiation Protection Supervisor S. Dull Chief Radiological Controls Technician, Torus i J. Fitzgerald Construction Services Manager- !
N. Hoy Project Manager, Torus Modification D. Lindsey General Manager, Operations I J. Maurer General Manager, Support Services A. McKeen Manager Radiological and Environmental Services !
F. Mitchell Radiation Protection Supervisor l C. Moreau Quality Assurance Auditor l K. Neal Senior Chemical / Nuclear Engineer {
R. Patch Director Quality Assurance K. Peper Health Physics, General Supervisor !
R. Plasse (Acting) Licensing Manager l S. Pointon Chief Radiological Controls Technician i M. Redding Public Affairs
, D. Ruddy Director, Design Engineering J. Sloyka Journeyman Radiological Controls Technician G. Tasick Manager, Design & Analysis D. Vandermark Quality Assurance Manager !
A. Zaremba Licensing Manager !
NRC:
R. Barkley Project Engineer, Region i T. Moslak Radiation Specialist, Region i R. Nimitz Senior Radiation Specialist, Region i R. Rasmussen Senior Resident inspector J. White Chief, Radiation Safety Branch, Region I l
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' Attachment 1 2 INSPECTION PROCEDURES USED IP 37551: Onsite Engineering IP 60710: Refueling Activities IP 61726: Surveillance Observations IP O a it Opera ions I (P 71750: Plant Support IP 83750: Occupational Radiation Exposure l IP 90712: In-Office Review of Written Reports i IP 92700: Onsite Follow-up of Written Reports of Nonroutine Events at Power Reactor )
Facilitit t j IP 92901: Operations Follow-up ;
IP 93803: Safety Systems Outage Modification Inspection i
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l, Attcchment 1 3 ITEMS OPENED, CLOSED, AND DISCUSSED Opened
, 50-333/98-07-01 NCV Separated Hose During Torus Dewatering 50-333/98-07-02 IFl Safety Relief Valve Setpoint Drift 50 333/98-07-03 NCV Missed Pressure Testing of Pressure Isolation Valves l 50-333/98-07-04 VIO Implementation of Main Steam Safety Relief Valve Modifications 50-333/98-07-05 NCV Surveillance Testing of the Pressure Suppression Chamber 50-333/98-07-06 NCV Unplanned Exposure Incident of September 17,1998 i
Closed 50-333/98-02-01 VIO Failure to Carry Out the Actions of the Correct Procedure During the Use of Emergency Operating Procedures 50-333/98-02-03 VIO An Operator Aide Was Not Adequate Concerning Plant Operations During Degraded Flow Conditions 50-333/98-02-02 VIO A Plant Procedure Was Not Adequate Concerning Assigned Duties of the On-Shift Licensed Operator 50-333/98-02-06 VIO Exclusion of Low Power or shutdown Conditions from the Annual Operating Test 50-333/98-02-07 VIO Exclusion of the Emergency Plan From the Annual Operating Test 50-333/98002 LER Safety Relief Valve Setpoint Drift 50-333/98006 LER Missed Pressure Testing of Pressure Isolation Valves 50-333/97010 LER Surveillance Testing of the Pressure Suppression Chamber 50/333/97-02-01 VIO Failure to Take Adequate Corrective Actions Following Previous Shutdown Cooling Isolation Events 50-333/97-05-02 URI Control Rod Drive Closure Bolt Crack Indications 50-333/98-07-01 NCV Separated Hose During Torus Dewatering 50-333/98-07-03 NCV Missed Pressure Testing of Pressure Isolation Valves 50-333/98-07-04 VIO Implementation of Main steam Safety Relief Valve Modifications 50-333/98-07-05 NCV Surveillance Testing of the Pressure Suppression Chamber 50-333/98-07-06 NCV Unplanned Exposure incident of September 17,1998 Discussed None
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AttachmInt 1 4
LIST OF ACRONYMS USED
'ACT Action / Commitment Tracking
. ADS Automatic Depressurization System ALARA As Low As Reasonably Achievable e AOP ' Abnormal Operating Procedure
.AP. Administrative Procedure-ASME American Society of Mechanical Engineers COLR Core Operating Limits Report CRD Control Rod Drive DER- Deficiency and Event Report ECCS Emergency Core Cooling Systems ED - Electronic Dosimeter EOP Emergency Operating Procedures ESF- Engineered Safety Features fcaw Flux Cored FHSRO Fuel Handling Senior Reactor Operator FME Foreign Material Exclusion FSAR Final Safety Analysis Report GENE General Electric Nuclear Energy HPCl High Pressure Coolant injection INPO Institute of Nuclear Power Operations ISI Inservice Inspection IST Inservice Testing IVV1 In-Vessel Visual inspection LER' Licensee Event Report LHRA Locked High Radiation Areas'
.LPCI ' Low Pressure Coolant injection -
' MSSRV Main Steam Safety Relief Valve MT Magnetic Particle NCV Non-Cited Violation NDE Non-Destructive Examinations NRC- Nuclear Regulatory Commission NSE Nuclear Safety Evaluation NYPA New York Power Authority ORG Operational Review Group PERC Performance Enhancement review committee PORC- _
Plant Operating Review Committee
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- PLCO : Potential Limiting Conditions for Operation pqr Procedure Qualification Record QA Quality Assurance Radeon Radiological Controls RCA Root Cause Analysis RCIC Reactor Core isolation Cooling RHR ' Residual Heal R. mcval RWCU Reactor water Clean Up RWP Radiation Work Permit-SRV Safety Relief Valve
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- o. Attrchmsnt 1 5 in TOP Temporary Operating Procedure
- ' TP Training Procedure
TS Technical Specification TS LCO Technical Specification Limiting Conditions for Operation UT Ultrasonic Examination wps Weld Procedure Specification
- WRM. Wireless Remote Monitors l
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