IR 05000333/1998008

Insp Rept 50-333/98-08 on 981123-990110.No Violations Noted. Major Areas Inspected:Operations,Egineering,Maintenance & Plant Support

ML20203E935
Person / Time
Site:	FitzPatrick
Issue date:	02/10/1999
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20203E927	List: IR 05000333/1998008 ML20203E923
References
50-333-98-08, 50-333-98-8, NUDOCS 9902170351
Download: ML20203E935 (30)

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U.S. NUCLEAR REGULATORY COMMISSION

REGION I

Docket No: 50-333  :

License No: DPR-59

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I Report No: 98-08 5

1 Licensee: New York Power Authority i

J Facility: James A. FitzPatrick Nuclear Power Plant

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Location: . Post Office Box 41 -

Scriba, New York 13093 Dates: November 23,1998 - January 10,1999 Ij Inspectors: R. A. Rasmussen, Senior Resident inspector B. S. Norris, Resident inspector J. C. Jang, Senior Radiation Specialist i P. D. Kauffman, Senior Reactor Engineer i

Approved by: J. F. Rogge, Chief Projects Branch 2 Division of Reactor Projects I

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PDR ADOCK 05000333 G pog ,

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EXECUTIVE SUMMARY James A. FitzPatrick Nuclear Power Plant

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NRC IR 50-333/98-08 )

November 22,1998 - January 10,1999 This integrated inspection included aspects of licensee operations, engineering, maintenance, j

' and plant support. The report covered a seven week period of resident inspection and the i

, results of an announced inspection in the area liquid and gaseous radiological effluent controls by a regional specialis .J OPERATIONS Emergency core cooling systems (ECCS) are relied upon during cold shutdown conditions to assure adequate makeup capacity to compensate for a loss of reactor water inventory from an inadvertent draindown of the reactor vessel. The FitzPatrick Technical Specification lists acceptable conditions for ECCS; availability from most conservative (i.e. two ECCS sub-systems .j operable), to least conservative (no ECCS sub-systems operable and containment isolation established to preclude the potential release of tission products). During the restoration from

..the refueling outage, NYPA elected to operate in the least conservative condition with no ECCS

. operable while draining the refueling cavity. Operation in this manner was considered to be an-

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example of non-conservative planning and scheduling. The acceptability of operating without ECCS operable during a draining evolution is being reviewed by NRR. Pending NRR review, this issue' remains unresolved. (URI 50-333/98-08-01) (Section 01.2) I i

Poor control of the configuration for the reactor vessel water level indicating system resulted in -

operators controlling reactor water level with a level indicating system that was partially

disassembled. This resulted in unintentionally draining about 15,000 gallons of water from the reactor vessel. Inadequate procedures were identified by NYPA as the root cause. This event

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also identified weaknesses in operator knowledge of the reactor assembly process, and ,

weaknesses with the plant risk assessment process. Operation with one indicating level i

indication that is undergoing maintenance was considered to be an example of nonconservative planning and scheduling. This non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited Violation.

(NCV 50-333/98-08-02) (Section O1.3)

The reactor restart was conducted in a conservative, well controlled manner. The added supervisory oversight during restart was effective in maintaining prompt resolution to emerging issues. (Section O1.4)

During the recent refe'., ling outage, the licensee identified an improper test method to verify -

containment integrity requirements which resulted from an . inadequate review following a plant modification during the 199F refueling outage. The failure to adequately test prima _ry conte.bment isolation vdvs is a violation of NRC requirements. This licensee-identified and corrected violation .s bdnq w . sated as a non-cited violation. (NCV 50 333/98-08 03) L (Section

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l Executive Summary (cont'd) )

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, MAINTENANCE i The reactor pressure vessel' system leakage test was well controlled and met ASME code requirements. The added oversight of the test coordinator assured retests were conducted and

. signed for as required. (Section M1.2)

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The inspectors considered the procedure for the test of the automatic start and sequencing features for the emergency diesel generators (EDGs), upon a simulated loss of coolant accident coincident with a loss of offsite electrical power, to be acceptable to meet the

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surveillance requirements of the TS. -(Section M2.2)-

ENGINEERING During initial startup from the 1998 refueling outage, NYPA performed a synthesized calibration of the local power range monitors due to one of the three traversing incore probe machines being inoperable. The NRC reviewed the alternative calibration method for technical adequacy hnd for compliance with the TS definition of calibration and found it acceptable for current plant

. operation. (Section E2.2)

PLANT SUPPORT The licensee maintained effective radioactive liquid and gaseous effluent control program !

The Offsite Dose Calculation Manual contained sufficient specification and instruction to !

acceptably implement and maintain the radioactive liquid and gaseous effluent control programs. (Section R1.1)

The licensee had an effective fuel integrity monitoring program. Surveillance tests exceeded TS requirements. (Section R1.2)

The licensee established, implemented, and maintained an effective radiation monitoring -

_ system calibration program, including flow rate measurement systems. As a result of self-assessment initiatives, the licensee implemented efforts to improve radiation monitoring system reliability. . The licensee also established and implemented an effective hydrogen /rqgen

' monitor calibration program. (Section R2.1)

The licensee established, implemented, and maintained an effective air cleaning, surveillance e program with respect to charcoal adsorption surveillance tests, HEPA mechanical efficiency tests, and air flow rate tests. (Section R2.2)

. The licensee established, implemented, and maintained an' effective quality assuran :e program

' for the radioactive effluent control program with respect to audit scope and depth, ai iit team

. experience, and response to audit findinge. The licensee also implemented an effeu ve quality control program to validate measurement results for radioactive effluent samples. (Laction R7.1) .

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l Executive Summary (cont'd)

l NYPA quality assurance auditors identified that the surveillance procedure for calibration of the '

l drywell continuous atmosphere monitor was did not meet the requirements of the FitzPatrick L TSs.- This licensee-identified and corrected violation is being treated as a non-cited violatio (NCV 50-333/98-08-04) (Section R8.1) ,

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, The inspectors noted that both the normal and emergency lighting in one of the emergency switchgear rooms were not working. The inspectors considered it a weakness that the control

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room supervisor was not made aware of a deficient plant condition; specifically, an emergency :

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switchgear room with no normal lighting or emergency lighting. The inspectors were also concerned about a personnel safety hazard, in that the operators may need to respond to an area with no lighting available. In addition, the affected emergency switchgear room was associated with the only operable ECCS pumps. (Section F2.1)

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SUMMARY OF PLANT STATU S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1  :

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' l . OP ERATI ON S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 O1 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 '

01.1 Gene ral Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 .

O1.2 ' ECCS Availability During Reactor Draindown . . . . . . . . . . . . . . . . . . . . . . . . 1

. O1.3 Inadvertent Lowering of Reactor Vessel Water Level During Vessel L .

R eassem bly . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 l . 01.4 Post Outage Startup . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 l 08 ' Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 l 08.1 (Closed) Licensee Event Report 50-333/98-12 . . . . . . . . . . . . . . . . . . . . . . . 5 l 08.2 (Closed) Violation 50-333/97-08-01 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6  ;

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l II . MAI NTENANC E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 i

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M1.1 General Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7

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M1.2 Performance of the Reactor Vessel System Leakage Test . . . . . . . . . . . . . . 8 - {

L M2 Maintenance and Material Condition of Facilities and Equipment . . . . . . . . . . . . . . . 9 (' M2.1 Drywell and Torus Pressure Test . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 i M2.2 EDG and ECCS Load Sequencing Test . . . . . . . . . . . . . . . . . . . . . . . . . . . 10  ;

l l M8 Miscellaneous Maintenance issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 l M8.1 (Closed) Violation 50-333/97-08-02 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 i

l 111. ENG I N EE R I NG . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 1 .

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E1 Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11

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E General Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11 l E2 Engineering Support of Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . 12  ;

l- E Calibration of LPRMs with an Inoperable TIP Machine . . . . . . . . . . . . . . . . 12 l

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i- E8 Miscellaneous Engineering Issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 E (Closed) Violation 50-333/97-08-03 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 E8.2 (Closed) Licensee Event Report 50-333/98-09 . . . . . . . . . . . . . . . . . . . . . . 14 -

E8.3 (Closed) Licensee Event Report 50-333/98-07 . . . . . . . . . . . . . . . . . . . . . . 14 E8.4 (Closed) Licensee Event Repr. 50-333/98-10 . . . . . . . . . . . . . . . . . . . . . 15 IV. PLANT SU PPORT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 R1 Radiological Protection ar'd Chemistry (RP&C) Controls . . . . . . . . . . . . . . . . . . . . 16 R Implementation of the Radioactive Liquid and Gaseous Effluent Control Prog rams . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 16

. R1.2 Fuel integrity Monitoring Program . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17

. R2 - Status of RP&C Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 i

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Table of Contents (cont'd)

R2.1 . Calibration of Effluent / Process / Area / Accident Radiation Monitoring Systems and Hydrogen / Oxygen Monitors . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17 R2.2 Air Cleaning Systems . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 R7 Quality Assurance (QA) in RP&C Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 R7.1 Oversight of RP&C by QA . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 R8 Miscellaneous RP&C issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 R8.1 (Closed)I icensee Event Report 50-333/98-11 . . . . . . . . . . . . . . . . . . . . . . 19 F2 Status of Fire Protaction Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . . . . . 20 F Emergency Lighting . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 20 V. MANAG EM ENT MEETING S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 X1 Exit Meeting Sum mary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 ATTACHMENTS ATTACHMENT 1 - Partial List of Persons Contacted-Inspection Procedures Used

- Items Opened, Closed, and Discussed

- List of Acronyms Used

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REPORT DETAILS

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James A. FitzPatrick Nuclear Power Plant NRC IR 50-333/98-08

~ November 22,1998 - January 10,1999 SUMMARY OF PLANT STATUS

' The inspection period began with the unit shutdown and midway through refueling outage number 13 (RFO13). The reactor was restarted on December 16,1998, and RFO13 ended when the main generator was synchronized to the grid on December 21,1998. System restoration, testing, and power escalation continued throughout the period. The period ended '

with the unit at 95 percent powe . OPERATIONS 01 Conduct of Operations  !

l- 01.1 'GeneralComments (71707)

Using NRC Inspection Procedure 71707, the resident inspectors conducted frequent reviews of ongoing 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 the unit was operated in conformance with the Technical Specifications (TS), observations L

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tof infrequently performed surveillance tests and the startup, and verification that logs and records accurately identified equipment status or deficiencies. In general, the conduct of operations was professional and safety-conscious; specific events and noteworthy observations are detailed in the sections below.

L 01.2 ECCS Availability During Reactor Draindown ,

l I a.- Inspection Scooe (60710. 71707)' l L . . .

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The NRC inspectors questioned NYPA's compliance with TS, Section 3.5.F "ECCS- l Cold Shutdown," during the evolution of restoration from refueling operations. .The i L inspectors reviewed the plant conditions in effect, TS, and the Final Safety Analysis  !

i Report (FSAR) in conjunction with this inspectio l Observations and Findinas

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FitzPatrick TS, Section 3.5.F, "ECCS-Cold Shutdown," define the requirements for  ;

I, operational emergency core cooling systems (ECCS) during cold shutdown condition l The specification provides four options depending on activities and conditions existing at  :

. the time. The purpose of Section 3.5.F is to specify the required conditions to assure i'

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adequate inventory makeup capability available to preclude an inadvertent draindown

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and uncovering of the fue ;

i: During the refueling phase of the outage, NYPA satisfied Section 3.5.F.3, which did not i require any operable ECCS. This'section specified that the reactor head must be i i

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removed and spent fuel pool gates opened. This condition provided a large water inventory consisting of the reactor, the reactor cavity, and the spent fuel poo ;

i On November 29,1998, at 9:00 p.m. NYPA entered TS 3.5.F.4 and installed the fuel !

pool gates. The limiting condition for operation (LCO) of Section 3.5.F.4 allowed the requirements of Sections 3.5.F.1-3.5.F.3 to not be met provided other actions were !

taken. This TS section requires that at least one system be restored to operable status within 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> of establish secondary containment integrity within the next 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, :

NYPA promptly established secondary containment. This section also prohibited core !

alterations and certain system operations to preclude an inadvertent draining of the ,

reactor vessel with the ECCS pumps inoperabl On November 29,1998, at 11:05 p.m. NYPA installed the spent fuel pool gate j Several hours later, operators began draining water from the reactor cavity to establish j the conditions required for the next phase of reactor restoration. During the draining, ,

NYPA was operating under Section 3.5.F.4, with no ECCS declared operable. NYPA ;

was planning to drain the water level to below the main steam lines to facilitate the i removal of the main steam line plugs. During this evolution, the NYPA outage risk assessment plan listed inventory control as green, and one train of core spray was considered availabl On November 30,1998, at approximately 9:30 a.m. the NRC questioned NYPA regarding compliance with TS 3.5.F. The NRC inspectors concluded that NYFA should have been operating in either TS section 3.5.F.1 or 3.5.F.2 given the ongoing draindown evolution and the status of the ECCS subsystems at the time. Further, the NRC inspectors noted that TS 3.5.F.4 precluded all operatlons with the potential for draining the reactor vessel; therefore, the ongoing reactor draindown evolution was specifically prohibited. Lastly, the inspectors questioned the establishment of secondary containment in lieu of restoring an ECCS subsystem to service 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> as meeting the intent of the TS versus the literal wordin NYPA evaluated the NRC concerns and concluded that draining of the reactor was a controlled evolution and was not precluded by the requirements of the TS. NYPA stated that the evolution of reactor draindown was essentially an adjustment to the normally established flowrates to maintain reactor water level with the control rod drive system pumps in operation. NYPA noted that maintaining reactor level was a critical function in I

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this condition because CRD flow would result in excessive inventory if the draindown path was isolated. Further, NYPA contended that automatic isolation valves would have stopped the draindown, thus making the evolution of reactor draindown not credible as having the potential to drain the reactor vessel. However, the NRC noted that the automatic holation valves were not required to be operable by T Due to the NRC's concerns, NYPA stopped lowering reactor water level and focused on restoring ECCS to operable. NYPA performed reviews and valve lineups and declared the "B" core spray subsystem operable on November 30,1998 at 2:35 p.m.. The reviews and declaration of operability were the only outstanding items affecting "B" core spray. On December 1,1998 at 6:15 p.m., NYPA completed testing and declared the

"B" residual heat removal (RHR) subsystem operable, satisfying the requirements of TS 3.5.F.1. The RHR system was not able to be made operable earlier because it required

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water in the torus to complete the post-work testing. By the outage schedule, NYPA

. intended to use the inventory from the reactor cavity to fill the toru Overali, the NRC considered NYPA's decision to drain the reactor cavity with no ECCS pumps operable to be an example non-conservative planning and scheduling. The issue will remain an unresolved item pending final NRR review. (URI 50-333/98-08-01) Conclusions Emergency core cooling systems (ECCS) are relied upon during cold shutdown conditions to assure adequate makeup capacity to compensate for a loss of reactor water inventory from an inadvertent draindown of the reactor vessel. The Fitzf-atrick Technical Specification lists acceptable conditions for ECCS availability from most conservative (i.e. two ECCS sub-systems operable), to least conservative (no ECCS sub-systems operable and containment isolation established to preclude the potential release of fission products). During the restoration from the refueling outage, NYPA elected to operate in the least conservative condition with no ECCS operable while draining the refueling cavity. Operation in this manner was considered to be an example of non-conservative planning and scheduling. The acceptability of operating without ECCS operable during a draining evolution is being reviewed by NRR. Pending NRR review, this issue remains unresolved. (URI 50-333/96-08-01)

01.3 Inadvertent Lowering of Reactor Vessel Water Level During Vessel Reassembly Inspection Scope (60710. 71707)

During the process of reassembly of the reactor vessel, the operable reactor vessel water level indicator indicated an erroneously high level. This resulted in operators lowering actual vessel level by 100 inches more than planned. The inspector reviewed the NYPA analysis of this even Observations and Findinos During the reactor reassembly evolution, the NYPA risk assessment required two indications of reactor vessel water level. This was accomplished with (1) the wide range level indicator, which provided indication up to the top of the reactor vessel, and (2) a narrow range indicator which was off scale high. Operators were contrciling reactor water level at approximately 350 inches above the top of active fuel (TAF), by adjusting the water discharge rate to compensate for the constant flow input from the control rod drive cooling water system.

l As part of the reactor disassembly, the wide range water level indicator was modified by disconnecting a portion of the reference leg (specifically, the keep fill condensing pot

and the keep fill system). A temporary standpipe and fill funnel assembly was installed

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. which allowed the level detector to remain operable with the reactor head removed.

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On December 2,1998, as part of the reassembly, the temporary standpipe was being l removed. As the water drained from the temporary standpipe, indicated water level increased. Operators responded to the indicated change and increased the discharge , ,

rate of reactor coolant to maintain indicated water level steady over a 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br /> period. In i addition to the maintenance activity on the reference leg, operators were filling and

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venting the reactor feedwater piping. The fill and vent was an additional feed source -

, causing operators to adjust reactor water discharge rates. The operators maintaining !

level relied on level indication and did not have a means of determining discharge rat :

Once the temporary standpipe was removed, and the normal system piping reinstalled,' 7 j the keep fill flow was initiated. As the reference leg piping filled, the indicated water

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level decreased.' At this point, operators noted the indicated level dropped to 250

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inches; the level was expected to be 350 inches. The second level instrument, which !

comes on scale at 224 inches, was still pegged hig )

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When operators discovered the level discrepancy, a temporary pressure gage was l connected to a reactor vessel low point tap to determine the water level. Once the icw j l level was confirmed to be 250 inches, the level was restored to the inter.ded value. The '

i 100 inch error represented approximately 15,000 gallons of water. Although reactor water level was reduced, the safety significance of this event was low since the water remained well above the TAF during this cold shutdown conditio !

NYPA performed an analysis of the event and developed corrective actions. This event ,

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identified weaknesses in operator knowledge of the reactor essembly process, and

. weaknesses with the plant risk astessment process. The erroneous level indication resulted from maintenance on the only reactor vessel water level indicator that was on-scale at the time as part of reactor reassembly. While NYPA's risk assessment required two level indications, the second indicator was offscale high and was expected to come -

on scale at 224 inches above TAF. NYPA contends that this level transient could not have significantly lowered water level during the draindown evolution because of the automatic isolation feature on the draindown path. This automatic isolation would have

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occurred at 177 above TAF and additional water level indicators would have come on scale for operator use at 224 inches above TAF. Proposed corrective actions included procedural enhancements to address the loss of level indication during the reactor disassembly and reassembly, and providing for an alternate means of level indicatio . NYPA TS, Section 6.8, Procedures, requires written procedures, as recommended in Appendix A of Regulatory Guide (RG) 1.33, November 1972. RG 1.33 requires procedures for removal of the reactor head. Contrary to the above, NYPA procedures for reactor head removal were inadequate in that the effect of disconnecting the level detector piping was not addressed. In addition, the operation with one indicating level

. indication that is undergoing maintenance is an example of nonconsecutive planning ;

. and scheduling. This non-repetitive, licensee-identified and corrected violation is being

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treated as a Non-Cited Violation (NCV), consistent with Section Vll.B.1 of the NRC Enforcorront Policy. (NCV 50-333/98 08-02)

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p l 5  ! Conclusio l I l Poor control of the configuration for the reactor vessel water level indicating system i l- resulted in operators controlling reactor water level with a level indicating system that i

was partially disassembled. This resulted in unintentionally draining about 15,000 -l L gallons of water from the reactor vessel. . Inadequate procedures were identified by l NYPA as the root cause. This event also identified weaknesses in operator knowledge ,

, ~of the reactor assembly process, and weaknesses with the plant risk assessment  :

process. Operation with one indicating level indication _that is undergoing maintenance was considered to be an example of nonconservative planning and scheduling. This  !

non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited  !

Violation. (NCV 50-333/98-08 02) ~

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i 01.4 Post Outage Startup )

1 . Inspection Scope ( 71707) L The inspectors observed activities associated with the reactor startup following the~ i refueling outage. This included operations conduct, resolution of plant problems, and I oversight, b. - Observations and Findinas The reactor startup was conducted in a conservative, well controlled manner. As systems were started and restored, retests were conducted to verify system operabilit Operators were aware of the status of testing and properly addressed identified deficiencies. In one case, the control rod withdraw sequence was stopped and all of the control rods were inserted due to a problem with a control rod failing to mov Throughout the reactor restart evolution, senior NYPA managers provided supervisory oversight of activities. This added supervisory oversight was effective in maintaining prompt resolution to emerging issue Conclusions

. The reactor restart was conducted in a conservative, well controlled manner. The added supervisory oversight during restart was effective in maintaining prompt resolution to emerging issue l 08 Miscellaneous Operations issues

- 08.1'. l (Closed) Licensee Event Report 50-333/98-12: Failure to Meet Primary Containment  !

Leakage Rate Testing Program Requirements (92700)

This Licensee Event Report (LER) identified an improper test method to verify containment integrity requirements. The improper test resulted from an inadequate  ;

review following a change to the plant configuration in accordance with a plant l modification. A modification to replace check valves associated with the reactor water  ;

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L recirculation system mini-purge lines resulted in : r:hange to the test vent point, resulting L

in an improper valve lineup This occ rr . ued d iur n , the 1996 refueling outage (RFO12)

. and was discovered during a review of local leak rate test (LLRT) procedures for the

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current outage (RFO13). s

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The test procedure was revised and performed satisfactorily. Based on the satisfactory  !

test results, NYPA concluded the safety significance of the error was minor. Additional

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i corrective actions included an independent review of all local leak rate test procedure l The inspectors verified selected corrective' actions and considered them acceptable.' ,

This LER is close '

During the current outage, the inspectors observed the performance of several LLRTs  ;

and did not identify any errors in the test methodology. However, the failure to adequately test primary containment isolation valves is a violation of NRC requirement :

This non-repetitive, licensee-identified and corrected violation is being treated as a Non- I

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Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy. (NCV 50-333/98 08-03)

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.08.2 (Closed) Violation 50-333/97-08-01: Failure to Properly implement an Abnormal l Operating Procedure (92901)

. On October 23,1997, t'o identify the cause of a ground on the "B" battery bus, FitzPatrick operators were using AOP-23,"DC Power System B Ground Isolation." The ,

operators became focused on finding the ground and failed to use the breaker  !

. sequence detailed in the procedure. -The power supply breakers for the high pressure coolant injection (HPCI) pump suction valves from the torus (23MOV-57/58) were to be ,

opened prior to opening the circuit breaker for the HPCI logic power supply (71DCB2-6). -

This was to prevent the valves from inadvertently opening due to a false signal that the i

level in the condensate storage' tank was low. The operators immediately recognized  ;

the error and restored the HPCI system to the normal lineu !

NYPA determined the root cause to be personnel error; I,e., the operator did not  :

-adequately use self-checking and proper place-keeping during the evolution. _ Corrective

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actions included formalizing the management expectations related to place-keeping, l and revising the procedure to improve the human factors aspects.

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The inspectors reviewed the completed Deviation and Event Report (DER 97-1475), the L revised procedure, and the violation response and considered them acceptable. The l inspectors also observed the procedure being implemented during a recent ground l

isolation and identified no procedural errors. This violation is closed.

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11. MAINTENANCE M1 Conduct of Maintenance

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M1.1 GeneralComments (61726,62707).

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Using NRC inspection Procedures 61726 and 62707, the resident inspectors periodically. observed various maintenance activities and 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. In general, maintenance and

' surveillance activities were conducted acceptably, with the work requests (WRs) and ,

necessary procedures in use at the work site, and with the appropriate focus on safet Specific activities and noteworthy observations are detailed in the inspection report. The inspectors reviewed procedures and observed all or portions of the following ,

maintenance / surveillance activities: '

TST-87 Primary Containment Pressurization Test ST-39P Containment integrity Component Verification TOP-288 HPCI and RCIC Turbine Slow Roll and Overspeed Testing Using Auxiliary Package Boiler AP-19.08 Infrequently Performed Tests or Evolutions ST-9C . Emergency AC Power Load Sequencing Test and 4KV Emergency *

- Power System Voltage Relays Instrument Functional Test ,

ST-9E LOCA Bypass of EDG Shutdown Logic Functional Test -l MP-4.02 Reassembly of Reactor Vessel After Refueling i OP-1 - Main Steam System RAP-7. Control Rod Scram Timing Evek:,h ISP-24 - Rod Block Mor.itor instrument Functional Test /Calibradon l- MP-31.14 Reactor Feed Pumr,7urbine Mechanical Control Line-Up ,

~ ST-5D APRM Calibration l ST-3B Core Spray Simulated Automatic Actuation Test  !

, ST-3M Core Spray Testable Check Valve Testing l ST-30 Core Spray Loop Keepfull Check Valve Functional Test l ST-3S ' Core Spray Outboard injection Bypass Switch Functional Test i

. ST-3U Core Spray Hold Pump Min Flow Check Valve Reverse Flow Test ,

ST-3P NB Core Spray Loop NB Ouarterly Operability Test I

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ST-2P ' RHR Shutdown Cooling Simulated Automatic isolation Test !

ST-2T

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LPCI Cross Connect Valve Position Instrument Functional Test i

ST-2Y . RHR Heat Exchanger Performance Test

[ ST-2A E/AF RHR Loop NB Auto Control Bypass Switch Functional Test l ST-2A G/AH RHR A LPCI Testable Check Valve 10AOV-68NB Test

! ST-39J . Leak Testing of RHR and Core Spray Testable Check Valvcs l l ST-39H RPV System Leakage Test and CRD Class 2 Piping System Functional .

L Test i l ST-2AUAM . . RHR Loop NB Quarterly Operability Test i ST-2AP RHR Loop A Containment Spray Line Level Switch Functional Test ,

j- . ST-2RNRB RHR Service Water Loop NB Monthly Operability Test i

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l 8 ST-2XA/XB RHR Service Water Loop A/B Quarterly Operability Test ST-4K HPCI Turbine Slow Roll and Overspeed Test ST-4G HPCI System Inoperable Test ST-4H HPCl/RCIC Testable Check Valve Testing ST-4T . HPCl/RCIC Exhaust Vacuum Breaker Check Valve Functional Test ST-4W HPCI Drain Pot Drain to Torus Stop Check Valve Functional Test ST-4Y HPCI Steam Supply Valves Auto Control Bypass Switch Functional Test ST-4N - HPCI Quick-Start, Flow Rate and Inservice Test l ST-4M HPCI Torus Suction Operability Test i M1.2 Performance of the Reactor Vessel System Leakage Test 1 Insoection Scope (61726)

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Foiiowing the restoration from the refueling outage, NYPA performed a system leakage

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test for the reactor pressure vessel. The inspectors obse-;od the test briefing and portions of the test, and reviewed the test procedure r id several of the retests to cerure

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the testing requirements were appropriately imp!sr..ented.

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! Observations and Findinas i The reactor pressure vessel system leakage surveillance test, ST-39H, was conducted ( to demonstrate the integrity of the reactor pressure vessel and associated piping

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systems. The procedure also tested piping and components worked on %g the

! outage. The test met the requirements of the American Society of Mechanical L Engineers (ASt.E), S'ction X In addition to the general pressure boun:lary area inspections, specific observations were required to satisfy specific maintenance activity retests. Ma'ntenance which required retest was tracked by the NYPA work request system. The inspecters reviewed a sample of these retests to assure the specific joints being tested were within

. the pressurized test boundary, and that the test pressure was appropriate. No .i deficiencies were identifie j l A test coordinator was appointed to manage the overall performance of the test, manage the accomplishmwnt of the specific retests, and to assure leakage and l deficiencies identified during the test were appropriately dispositioned. The test l coordinator was effective in this role and added an additional level of oversight to this

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The reactor pressure vessel system leakage test was well controlled and met ASME L code requirements. The added oversight of the test coordinator assured retests were

conducted and signed for as required.

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L M2 Maintenance and Material Condition of Facilities and Equipment

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M2.1 . Drywell and Torus Pressure Test Inspection Scooe (61726)

l l The inspectors observed portions of the pressure test that NYPA performed to verify the l Integrity of the primary containment after reinstallation of the temporary access hatch l cut in the torus wall. The inspectors reviewed the completed test procedure, the TS, and the NYPA Request Relief and associated NRC Safety Evaluation. In addition, the inspectors toured the drywell and torus prior to the pressure tes Observations and Findinos

' During the refueling outage, NYPA removed a section of the torus wall to facilitate the installation of new suction strainers for the FitzPatrick ECCS pumps. In an August 1998 l

letter to the NRC, NYPA requested relief from the requirements of the ASME Boiler and -

Pressure Vessel Code,Section XI, with respect to testing after the reinstallation of the l hatch.~ On September 21,1998, the NRC approved the request to perform a pressure

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test and visual examination vice the normal pneumatic pressure test required by

- 10CFR50, Appendix J, " Primary Reactor Containment Leakage Testing for Water-

' Cooled Power Reactors."

For the test, NYPA developed temporary surveillance procedure TST-87, " Primary Containment Pressurization Test." The procedure was a one-time test and was

!- conducted in accordance with administrative procedure AP-19.08," Infrequently Performed Tests or Evolutions." The inspectors reviewed the administrative procedure

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. and verified that the requirements foi a special evolution were implemented, including l . the assignment of a senior manager for oversight of the test. The procedure required a - ,

I minimum test pressure of 45 psig, to be maintained for at least 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The.45 psig limit l is the maximum calculated containment pressure following a design basis loss of

[ coolant acciden ,

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< Prior to the test, the inspectors verified that the test procedure was consistent with the NRC Safety Evaluation, which accompanied the September 1998 letter, and the TS

. Bases, Section 3.7.A. The inspectors also toured the torus, prior to filling, and the l drywell and determined that post-outage housekeeping was acceptable and there were no s;gnificant discrepancies.. In addition, the inspectors observed some of the l

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preparations and a portion of the actual pressurization. The test was completed

satisfactorily, with no leakage detected during NYPA's visual (VT-2) inspection. ; The L ,

" inspectors reviewed the completed surveillance test package and identified no L - discrepancie c.- Conclusion

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! NYPA performed a static pressure test of the primary containment after the torus

[ - temporary access hatch was reinstalled. The test procedure was consistent with the

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NRC Safety Evaluation and the bases of the TSs. The test was performed with no problems, and was completed satisfactorily, with no leakage detecte M2.2 EDG and ECCS Load Sequencing Test Inspection Scooe (61726)

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' The inspectors reviewed the NYPA surveillance procedure for demonstrating the automatic start feature of the emergency diesel generators (EDGs) and the associated sequencing on to the emergency buses of the ECCS pumps. The FSAR and TSs were

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reviewed to verify that the applicable testing requirements were incorporated into the -

procedure. The inspectors observed portions of the surveillance test as it was 'i performe ' Observations and Findinas

' The preferred electrical lineup at FitzPatrick during power operation is to supply all in-

- house loads from the main generator via the normal station service transformer (T-4).

An alternate lineup is to supply the emergency switchgear (Buses 10500 and 10600) j

from offsite sources via the reserve transformers (T-3 and T-2, respectively). In the !

event of a loss of the normal and reserve power supplies, the emergency power system ;

can energize the emergency buses. The emergency power system consists of two independent divisions, each division having two EDG In accordance with TS Surveillance Requirement (TSSR) 4.9E.4, FitzPatrick is required

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to perform a functional test of the emergency power system or,ce during each operating ,

cycle, typically during a refueling outage. The test was a simulation of a loss _of coolant i accident (LOCA) coincident with a loss of the normal and reserve power supplies. The I

. purpose of the test was to demonstrate that the ED_Gs automatically start, and that the

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ECCS pumps load onto the emergency buses in the proper sequence. The surveillance requirement was completed satisfactorily using procedure ST-90, " Emergency AC L Power Load Sequencing Test and 4KV Emergency Power System Voltage Relays .

Instrument Functional Test." l The inspectors observed one of the pre-evolution briefs, completion of the prerequisites for testing the 10600 Bus, and portions of the actual tests. The inspectors also reviewed the surveillance test procedure, the TSs, and the FSA ] Conclusion The inspectors considered the ' procedure for the test of the automatic start and sequencing features for the emergency diesel generators (EDGs), upon a simulated l

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-. loss of coolant accident coincident with a loss of offsite electrical power, to be

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. acceptable to meet the surveillance requirements of the TS.

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M8 Miscellaneous Maintenance issues ,

i M8.1 - (Closed) Violation 50-333/97-08-02: Failure to Enter the LCO for PCIS During  ;

Troubleshooting (92902)

J l In October 1997, during maintenance activities to identify and repair an electrical

l ground, the primary containment isolation system (PCIS) function of the outboard HPCI '

, steam supply valve was unknowingly rendered inoperable.. Specifically, the fuses were l removed from the valve logic circuit. Neither the operators nor the maintenance 1-planners recognized that removal of the fuses disabled the PCIS function, which existed undetected for about sixteen hours. During the maintenance, a technician accidently

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shorted the logic circuitry and caused an invalid ESF actuation. The ESF actuation l alerted the operators to the inoperable PCIS function of the valve, and the operators ,

immediately took the appropriate actions, as required by TS 3.7. !

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NYPA determined the root cause to be personnel error on the part of the operators and maintenance planners for failing to recognize the total plant impact of removing the fus !

A contributing cause was a weak surveillance test procedure; i.e., it did not contain I l guidance associated with the TS LCO. The corrective actions included a review of the ;

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event with all licensed operators and work planners, revising the affected surveillance i procedure, and reviewing other surveillance and operating procedures that direct fuse i removal to assure TS requirements are include j The inspectors conducted an in-office review of the NYPA violation response, the associated DERs, the procedure review results and the revised procedures. In addition, the inspectors discussed the issue with several planners and shift managers as to their understanding of their respective responsibilities related to work planning and approva l The inspectors considered the actions taken to be acceptable and had no further questions. This violation is close Ill. ENGINEERING l l'

E1 Conduct of Engineering i

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E1.1 ' General Comments (37551)

Using NRC Inspection Procedure 37551, the resident inspectors frequently reviewed design and system engineering activities and the support provided by the engineering organizations to plant activities. Specialist inspectors in this area used other procedures during their reviews of engineering activities; these inspection procedures are listed, as l applicable, for the respective sections of the inspection report.

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i E2 Engineering Support of Facilities and Equipment

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' I E Calibration of LPRMs with an Inoperable TlP Machine -

l u l Inspection Scope (37551. 61726)

1-During initial startup from the refueling outage, NYPA identified that one of the three traversing incore probe (TIP) machines was inoperable. NYPA informed the inspectors

- that a synthee'r.ed calibration of the local power range monitors (LPRMs) was possible by means of G,e core monitoring computer.

l- The inspectors questioned whether the synthesized calibration was an acceptable altemative to the normal calibration method of using the TIP system. The inspectors reviewed the TSs, the FSAR, the calibration procedure, NYPA's analysis for the

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computer calibration method and vendor supporting information. In addition, the inspectors discussed the issue with the NRC's Region I office and the Office of Nuclear Reactor Regulation (NRR) to determine if NYPA's proposed method was acceptable.

. Observations and Findings System Description: During reactor operation, the reactor power is measured using the LPRMs. There are 31 LPRM strings distributed symmetrically within the reactor core; each LPRM string has four detectors, spaced evenly from the bottom to the top of the core. The LPRM detectors are calibrated using the TIP system. The TIP detectors

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traverse a tube located adjacent to the tube that houses each LPRM string. FitzPatrick has three independent TIP machines, each one able to traverse eleven LPRM tube i The center LPRM string is common to allow the three TIP machines to be normalize l On December 28,1998, the resident inspectors noted that the "C" TIP machine had -

tripped several times on high torque while trying to traverse the C-9 position. NYPA declared the "C" TIP machine inoperable for trouble-shooting activities. Due to the "C" TIP machine being inoperable,10 of the 31 LPRM strings could not be calibrated using the normal method. During subsequent discussio is, the inspectors learned that NYPA used a " synthesized calibration" method for the LPiRMs associated with the TIP tubes that could not be traversed. The synthesized method used the TIP information that was available and calculated the power in the remaining locations by means of the 3D MONICORE computer program. FitzPatrick TS, Table 4.1-2, states that the LPRMs J must be calibrated every 1000 effective full power hours (EFPH) by "TIP System l l Traverse." TS Section 1.0.F.2 defines instrument channel calibration as "... the j L adjustment of an instrument signal output so that it corresponds ... to a known value of j

! the parameter which the instrument monitors. ..." The inspectors questioned whether

the synthesized method met the TS definition of a calibratio I i

The inspectors discussed the concem with personnel in the NRC Regional office, the

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NRC Project Manager, and representativ3s from the NRR Technical Specification, instrumentation and Control, and Reactor Systems Branches. The concems were: 1)

i L Did the existing FitzPatrick TS allow NYPA's proposed method of synthesized y calibration, using the 3D MONICORE, for the LPRMd and 2) Was the proposed

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synthesized method an acceptable alternative as a means to calibrate those LPRMs that were not accessed with a TIP7 NYPA indicated that their actions to calibrate the LPRMs in this manner was consistent with standard industry practice and that the actions comply with TS. NYPA committed to provide written clarification of the 3D MONICORE signal synthesis methodology for additional staff revie The NRC steff concluded that there was no clear restriction that prevented the use of this LPRM calibration method. Additionally, any affect on plant safety due to the decreased accuracy of this calibration method would be very small. Subsequently, the NRC determined that a restrictive, verbatim interpretation of the TS would result in licensee action that was not justified by the immediate safety implications of the issu Thus, the NRC found NYPA's interpretation of TS Table 4.1-2, to perform a synthesized j calibration of the LPRMs to be acceptable for current plant operation and will further i review the licensee submitta I l

On January 18,1999 NYPA submitted a letter to the NRC explaining, in detail, their I methodology for the synthesized calibration of the LPRMs. Also, NYPA stated that they I would be submitting a TS amendment to clarify the TS requirements in this area to !

prevent further confusio ;

As part of their trouble-shooting effort, NYPA visually inspected the "C" TIP machine, the probe and cabling and found no indications that would have inhibited inserting the !

probe into the tubes. NYPA then placed a nitrogen (N 2) purge on the "C" TIP tubing to l remove any residual moisture from the system; the purge was removed and they again i attempted to run the "C" TIP machine. The TIP was unable to run through the C-9 and C-1 positions; the C-1 position is the common TIP tube used for normalization of all the TIP probes. If the C-1 pos$on is unavailable, then the "C" TIP machine is inoperabl After additional Na purging, the C-1 position became usable and a TIP run was obtained for all positions except C-9. Subsequently, on January 18,1999, during a TIP run, the

"C" TIP was unable to traverse the C-1 position. NYPA is continuing to trouble-shoot the "C"TIP machine to identify the root cause for the repeat problem Conclusion During initial startup from the 1998 refueling outage, NYPA performed a synthesized calibration of the local power range monitors due to one of the three traversing incore probe machines being inoperable. The NRC reviewed the alternative calibration method for technical rWequacy and for compliance with the TS definition of calibration and found it acceptable for current piant operatio E8 Miscellaneous Engineering issues E (Closed) Violation 50-333/97-08-03: Erroneous Removal of Components from l

Environmental Oualification Program (92903)

NRC Inspection Report 50-333/97-08, identified a violation of NRC requirements related to the erroneous removal of four HPCI system pressure switches from the environmental qualification (EO) program. The licensee attributed the error to poor

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personnel performance. The responsible personnel made a non-conservative i assumption and overlooked a safety function of the switche '

! Corrective actions included: briefing of engineering personnel, revision of the procedure ,

p for performing EQ evaluations, performance of an e aent-of condition review, and a plant modification to correct the specific condition with the pressure switches. . The

extent-of-condition review identified several adGonal components that were not i appropriately classified in the EQ program. These issues were addressed by the ,

licensee's corrective action program.' The inspectors reviewed the procedure change !

and verified the pressure switches were added to tim EQ component list. This violation' l l Is close i

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l l E8.2 (Closed) Licensee Event Report 50-333/98-09: Errorin Exclusion Region of Power- l 1-Flow Map. (92700) ,

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While reviewing calculations for fuel cycle 14, NYPA reactor sngineering questioned the !

size of the exclusion region of the power-flow map. On August 7,1998, NYPA was informed by the vendor, General Electric, that an error in the cycle 13 exclusion region i calculation had been discovered. The effect of the error was that the Exclusion Region in use since cycle 13 began was incorrect and should have been larger.

! The inspectors conducted an on-site follow-up and completed an in-office review of the l LER. The inspectors reviewed the technical issues associated with the LER, including ;

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review of cycle 13 operatings history, and performed reviews of DER 98-1880, Saferj - i Evaluation JAF SE-96-052, Revision 2, and conducted control room inspections of rev%ed operating procedures. The inspectors reviewed the operating history and determined that the plant had not operated in the corrected Exclusion Region and therefore there was no violation of NRC requirements. In addition, the inspectors verified selected corrective actions were properly completed prior to unit restart from the refueling outage. The

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inspectors concluded that the root cause and corrective actions were reasonable. This

! LER is close E8.3 (Closed) Licensee Event Report 50-333/98-07:. High Pressure Coolant injection System Declared Inoperable Due to DC Ground on "B" Station Battery Bus (92700)

On July 31,1998, the HPCI system was declared inoperable following the identification of an electrical direct current (DC) ground on the "B" station battery motor control cente The HPCI auxiliary oil pump (23P-150) breaker opened due to water intrusion through a degraded cable penetration conduit seal on top of motor control center (MCC) 71BMCC-

! 4. The water was from a service water line being drained for repair. The water was l' being drained into a 55 gallon barre! vs hich overflowed onto the floor, seeped through a l floor plug, and onto MCC 71BMCC-4. The draining evolution was not monitore Operator response to this event was documented in NRC inspection Report 50-333/98-04 and engineering review was documented in NRC inspection Report 50-333/98-0 During this inspection, the inspectors conducted an onsite review of the LER, the associated DER, and supporting procedural revision documentatio ' NYPA determined the cause for the DC ground condition to be inadequate managerial methods. Specifically, clear guidance was not provided for the conduct of evolution L , - - _ _ -

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activities continuing through the shift change period and the need to provide continuous j monitoring of dyaamic activities with potential risks (e.g. draining into a limited volume).

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Based on an in-office review of DER 98-01746 and reviseo plant procedures OSD-4, i

" Shift Tumover and Log Keeping," OSD-18, " Equipment Status Control," and AP-12.03, t " Administration of Operations," no new issues were revealed and no violations were l

identified. The inspectors verified that selected corrective actions were properly completad or progressing according to NYPA's documerited schedule. The root cause

! cad corrective actions appear appropriate to prevent recurrence. This LER is c'osed.

l l E8.4 (Closed) Licensee Event Report 50-333/98-10:- Primary Containment Electrical

Penetration Inoperable Unier Postulated Conditions (92700)

l On August 13,1998, three containment electrical penetrations were determined to be inoperable under certain postulated fault conditions. NYPA did not correctly translate the applicable design basis for safety-related systems and components into the plant design. Specifically, the effects of a short circuit on the protection scheme for some containment penetrations were not analyzed for the condition of a loss of coolant l accident coincident with a foss of offsite power. Further NYPA engineering evaluation determined that primery coniainment electrical penetrations X109 and X100D were also subject to the same condition. This issue was identified by the NRC during the j engineering team inspection and resulted in a violation. The violation and initial

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corrective actions were documented in NRC Inspection Report 50-333/98-0 During this inapar+ian, the inspectors conducted an onsite review of the LER, DER 98-L 01910, Safety Evaluations98-037 and 98-040, and modification documentation used to l install fuses in the affected electrical circuits prior to unit restart from the refueling l outage. The LER accurately documented the condition and resulting corrective action ]

The inspectors verified selected corrective actions and concluded that the corrective actions were acceptable. This LER is closed.

l IV. PLANT SUPPORT l

Using NRC Inspection Procedure 71750, the resident inspectors routinely monitored the performance of activities related to the areas of radiological controls, chemistry, emergency preparedness, security, and fire protection. Minor deficiencies were discussed with the appropriais management, significant observations are detailed below. Specialist inspectors in the same areas used other procedures during their reviews of plant support activities; these inspection procedures are listed, as applicable, for the respective sections of the inspection report.

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R1 Madiological Protection and Chemistry (RP&C) Controls >

R Implementation of the Radioactive Liquid and Gaseous Effluent Control t'rograms Inspection Scope (84750)

The inspectors reviewed: radioactive liquid and gaseous effluent release permits; effluent control procedures; unplanned or unmonitored release pathways; the 1997 and the first palt of the 1998 semi-annual effluent reports; and the Offsite Dose Calculation Manual (ODCM).

The inspectors toured the control room and other areas, and reviewed selected l radioactive gas processing facilities and equipment, effluent and process radiation monitoring systems, and verified the reactor building plant air balance and air cleaning system : Observations and Findinas All effluent radiation monitors and air cleaning systems were operable at the time of the plant tour. The various buildings were routinely maintained at a slight negative pressure, as described in the FSA The ODCM provided descriptions of the sampling and analysis programs, which were established for quantifying radioactive liquid and gaseous effluent concentrations, and for calculating projected doses to the public. All necessary parameters, such as effluent radiation monitor setpoint calculation methodologies, and site-specific dilution factors, were listed. Radioactive gaseous effluent release permits and monthly dose projections

, were complete. There were no unplanned or unmonitored radioactive gas releases in l 1997 or 1998. The effluent control procedures reviewed were detaile The Semi-Annual Radioactive Effluent Reports for 1997 and the first part of the 1998 provided data indicating total released radioactivity for liquid and gaseous effluents. The

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assessment of the projected maximum individual doses resulting from routine

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radioactive airt.)rne and liquid effluents were included, as required. Projected doses to l the public were well below the TS limits. There were no anomalous measurements, omissions or adverse trends in this repor Conclusions The licensee maintained effective radioactive liquid and gaseous effluent control l programs. The Offsite Dose Calculation Manual contained sufficient specification and l instruction to acceptably implement and maintain the radioactive liquid and gaseous effluent control programs.

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R1.2 Fuel Integrity Monitoring Program Insoection Scooe (84750)

The inspection consisted of a review of reactor coolant system (RCS) data for fuel integrity parameters, as listed in TS, Section 3/4.6.C," Coolant Chemistry," and associated procedures; and interviews with chemistry staff regarding the fuel integrity .

monitoring progra ; Observations and Findinas i

Surveillance results for RCS samples (e.g., iodine dose equivalent, conductivity, and chloride lon) were well below the TS limits. Procedures were adequate and easy to follow. The responsible individuals in the Chemistry Department had good knowledge of testing methodologie ; Conclusions I l

The licensee had an effective fuel integrity monitoring program. Surveillance tests !

exceeded TS requirement R2 Status of RP&C Facilities and Equipment R Calibration of Effluent / Process / Area / Accident Radiation Monitoring Systems and i Hydrogen / Oxygen Monitors l l Inspection Scope (84750)

The inspectors reviewed the most recent electronic and radiological calibration results for the following selected effluent / process / area radiation monitoring systems (RMS),

system flow rates, and hydrogen / oxygen monitors:

Radiation Monitorina Systems: liquid radwaste discharge monitor, service water discharge monitor, reactor building closed loop cooling water radiation monitor, main stack noble gas monitors (normal and high range), refuel floor exhaust radiation monitor, reactor building exhaust radiation monitor, turbine building exhaust monitors !

- (normal and high range), radwaste building exhaust monitors (normal and high range), ,

and offgas radiation monitor ]

. Flow Rate and Exolosive Gas Monitors: liquid radwaste flow rate, main stack flow rate, turbine building vent flow rate, radwaste building vent flow rate, and oxygen / hydrogen monitors Observations and Findinas All RMS calibration results reviewed were within the acceptance criteria defined by the licensee's procedures. The calibration data indicated that the RMS was responding in a _ , - - _

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linear manner. Chemistry staff and a system engineer performed for tracking and ,

trending efforts to obtain sufficient information to assess RMS performanc j l

Calibration results of flow rate measurement systems and hydrogen and oxygen  ;

monitors were within the licensee's acceptance criteri j I Conclusions

The licensee established, implemented, and maintained an effective radiation l monitoring system calibration program, including flow rate measurement systems. As a result of self-assessment initiatives, the licensee implemented efforts to improve radiation monitoring system reliability. The licensee also established and implemented an effective hydrogen / oxygen monitor calibration progra R2.2 Air Cleaning Systems Insoection Scooe (84750)

The most recent surveillance test results (in-place HEPA and charcoal leak tests, air capacity tests, pressure drop tests, and laboratory tests for the lodine collection efficiencies) were reviewed with respect to: the standby gas treatment system (SBGT),

the control room ventilation system, and the technical support center syste Observations and Findinas All surveillance results were within the TS acceptance criteria. The responsible individual had appropriate knowledge of testing methodologies and acceptance criteri Conclusions The licensee established, implemented, and maintained an effective air cleaning surveillance program with respect to charcoal adsorption surveillance tests, HEPA mechanical efficiency tests, and air flow rate test R7 Quality Assurance (QA)in RP&C Activities R Oversight of RP&C by QA Inspection Scope (84750)  ;

i The inspection consisted of a review of: the 1998 audit, OA policy of the measurement laboratory, and implementation of the measurement laboratory quality control program for radioactive liquid and gaseous effluent sample Observations and Findinag Audit findings from 1998 did not identify any significant regulatory or safety issues. The scope and technical depth of the audit were sufficient to assess the quality of the

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i radioactive liquid and gaseous effluent control programs. Individuals with experience in l

. radioactive affluents control and chemistry participated as audit team member ;

The QA Support Program consisted of the quarterly distribution of test or quality control ,

samples and issuance of a performance evaluation report. Quality control charts for the !

gamma spectrometry and proportional counters were frequently reviewed by licensee .

. staff and used as a oechanism to assess laboratory performance, Conclusions .

The licensee establit hed, implemented, and maintained an effective quality assurance ,

program for the rajic active effluent control program with respect to audit scope and -

. depth, audit tesm exparience, and response to audit findings. The licensee also impkmented an effective quality control program to validate measurement results for radioactiw effluent sample ;

R8 Miscellaneous RP&C losues R (Closed) Licensee Event Report 50-333/98-11: Failure to Meet Drywell Continuous  !

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Atmosphere Monitoring System Technical Specification Surveillance Requirements

. (TSSR)

- Inspection Scope (90712. 92700)

During a recent QA audit of the RP program, NYPA identified that the TSSR for the ,

drywell continuous atmosphere monitoring (CAM) system was not being me !,

The inspectors reviewed the LER, FitzPatrick TSs, and the applicable surveillance test ;

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procedure. In addition, the inspectors discussed the issue with QA and RP personnel, confirmed that the required TS LCO actions were implemented, and determined that the ;

corrective actions were on schedule, j i ' Observations and Findinas  ;

On September 29,1998, FitzPatrick QA auditors identified that the procedure used to

' satisfy a TSSR for the'drywell CAM system was not adequate. Specifically, TSSR 4.6.D.4 (which references Table 4.6-2) required that the CAM particulate and gaseous instrumentation be calibrated once every 3 months. The NYPA procedure for testing the l '

CAM (RP-RESP-03.01, "Drywell Constant Air Monitor") only tested the efficiency of the instrument and did not include the entire instrument range nor a verification of the alarm response, as required by the TS definition of a channel calibration, in accordance with

.TS 3.6.D.6, if the CAM was not operable (i.e., not calibrated), then operation could continue for up to 30 days provided grab samples of the containment atmosphere were analyzed daily. The failure to properly calibrate the drywell CAM system is a violation of

NRC requirements. 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 50-333/98-08-04) ,

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The inspectors verified that the compensatory samples were taken once the condition was ' identified.' in addition, the inspectors confirmed that both channels of the CAM ,

'were successfully calibrated, and that the procedure was appropriately revised. _ The inspectors performed on-site and in-office review of the LER and considered the root 'l i

cause and corrective actions to be reasonable. The description and analysis of the event, !

as described in the LER, are consistent with the inspectors' understanding of the even l l The LER is close I Conclusions  !

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NYPA quality assurance auditors identified that the surveillance procedure for calibration - !

g of the drywell continuous atmosphere monitor did not meet the requirements of the : {

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FitzPatrick TSs. This is being treated as a Non-Cited Violation. (NCV 50-333/98 08-04) j

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F2 . Status of Fire Protection Facilities and Equipment j I

F2.1 - Emergency Lighting i

Insoection Scooe (64704)  ;

l l During a plant tour, with the plant in a refueling outage, the inspectors noted that both i l

the normal and emergency lighting in the north (Division ll) emergency switchgear room ,

were not working. The inspectors reviewed the occurrence and the NYPA resolutio i Observations and Findinas -

! i l The inspectors questioned the control room supervisor (CRS) as to the reason for the i l room being dark. The CRS informed the inspectors that the normal lighting had been ;

de-energized for maintenance. He was unaware that the battery pack for the '

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emergency lighting was depleted and that the room was dark. Division ll ECCS includes ,

the "B" core spray (CS) pump and the "B" and "D" RHR pumps. At the time that the 6 inspectors identified the issue, the only operable ECCS pumps were the "B" CS and "B" t RHR pumps; all of the Division i pumps were inoperable. Because the unit was in a cold shutdown condition, there was no violation of NRC requirement t l- 1 As a result of the inspectors concems, NYPA installed temporary hand held lights in the area, promptly restored the normal lights, initiated a DER to document the event, and subsequently confirmed the recharging of the emergency light.

L Conclusions I

. The inspectors noted that both the normal and emergency lighting in one of the

- emergency switchgear rooms were not working. The inspectors considered it a

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weakness that the control room supervisor was not made aware of a deficient plant b " condition; specifically, an emergency switchgear room with no normal lighting or , emergency lighting. The inspectors were also concerned about a personnel safety

hazard, in that the operators may need to respond to an area with no lighting available.

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In addition, the affected emergency switchgear room was associated with the only operable ECCS pumps.

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V. MANAGEMENT MEETINGS l X1' Exit Meeting Summary ,

l The inspectors presented the inspection results to members of the licensee management at the conclusion of the inspection on January 26,1999. The licensee l- acknowledged the findings presented.

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ATTACHMENT 1 PARTIAL LIST OF PERSONS CONTACTED -

NYPA J. Alexander, Administrative Coordinator, Radiological & Environmental Services (RES)

N. Avrakotos, Emergency Preparedness Coordinator M. Colomb, Site Executive Officer D. Cristafulli, Radiation Protection Training Program Administrator D. Lindsey, General Manager, Operations J. Maurer, General Manager, Support Services A. McKeen, RES Manager K. Peper, Health Physics General Supervisor K. Pushee, Respiratory Protection Supervisor-J. Ratigan, Radiological Engineer D. Ruddy, Director, Design Engineering K. Szeluga, Dosimetry Supervisor D. Vandermark, Quality Assurance Manager A. Zaremba, Licensing Manager INSPECTION PROCEDURES USED IP 37551 Onsite Engineering IP 60710 Refueling Activities IP 61726 Surveillance Observations IP 62707 Maintenance Observations IP 64704 Fire Protection Program IP 71707 Plant Operations IP 71750 Plant Support Activities IP 84750 Radioactive Waste Treatment, and Effluent and Environmental Monitoring !

IP 90712 In-Office Review of Written Reports of Nonroutine Events at Power Reactor Facilities IP 92700 On-site Follow-up of Written Reports of Nonroutine Events at Power Reactor Facilities IP 92702 Corrective Actions IP 92901 Followup - Operations IP 92902 Followup - Maintenance IP 92903 Followup - Engineering l

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l Attachment 1 ITEMS OPENED, CLOSED, AND UPDATED l OPENED l

URI 98-08-01 Acceptability of No ECCS Pumps Available During Reactor Cavity Draindown - {

1 l NCV 98-08-02 Inadequate Procedure to Control Reactor Vessel Level During I Reassembly i NCV 98-08-03 Failure to Adequately Test Primary Containment isolation Valves i

NCV 98-08-04 Failure to Take Compensatory Actions for Inoperable CAM Closed NCV 98-08-02 Inadequate Procedure to Control Reactor Vessel Level During Reassembly NCV 98-08-03 Failure to Adequately Test Primary Containment isolation Valves NCV 98-08-04 Failure to Take Compensatory Actions for Inoperable CAM VIO 97-08-01 Failure to Properly implement an Abnormal Operating Procedure -

VIO' 97-08-02 Failure to Enter the LCO for PCIS During Troubleshooting VIO 97-08-03 Erroneous Removal of Components from EQ Program LER 98-07 HPCI System Declared Inoperable Due to Ground on "B" Station Battery Bus LER 98-09 Error in Exclusion Region of Power-Flow Map LER 98-10 Primary Containment Electrical Penetration inoperable Under Postulated Conditions LER 98-11 Failure to Meet Drywell CAM System TS Surveillance Test Requirements LER 98-12 Failure to Meet Primary Containment Leakage Rate Testing Program l Requirements j UPDATED j

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Attachment 1 LIST OF ACRONYMS USED l l

l ASM American Society Of Mechanical Engineers i CAM Continuous Atmosphere Monitor i DER Deficiency and Event Report  !

ECCS . Emergency Core Cooling System EDG Emergency Diesel Generator EFPH Effective Full Power Hour EO Equipment Qualification ESF Engineered Safety Feature

.FSAR Final Safety Analysis Report l HEPA High Efficiency Particulate Activity HPCI High Pressure Coolant injection l LCO Limiting Conditions for Operation l LER Licensee Event Repod LLRT Local Leak Rate Test LPRM Local Power Range Monitor i LPCI Low Pressure Coolant injection  !

MCC Motor Control Center  ;

NCV Non-Cited Violation  !

NRC- Nuclear Regulatory Commission- 1 NRR NRC Office of Nuclear Reactor Regulation l

NYPA New York Power Authority '

ODCM Offsite Dose Calculation Manual PCIS Primary Containment isolation System l OA Quality Assurance I OC Quality Control RCIC Reactor Core Isolation Cooling

.RCS Reactor Coolant System i RFO Refueling Outage RG Regulatory Guide j RHR Residual Heat Removal RMS Radiation Monitoring System RP&C Radiological Protection and Chemistry RTID Radiological Technical Information Document RWP Radiation Work Permit SBG Standby Gas Treatment System TAF Top of Active Fuel TIP . Traversing incore Probe

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TS Technical Specification L TSSR Technical Specification Surveillance Requirement l URI Unresolved item WR Work Request l

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