IR 05000443/1998004: Difference between revisions

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{{Adams
{{Adams
| number = ML20236W244
| number = ML20196J163
| issue date = 07/30/1998
| issue date = 12/04/1998
| title = Insp Rept 50-443/98-04 on 980517-0704.Violations Noted. Major Areas Inspected:Operations,Engineering,Maintenance & Plant Support
| title = Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-443/98-04 Issued on 980730
| author name =  
| author name = Cowgill C
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
| addressee name =  
| addressee name = Feigenbaum T, Harpster T
| addressee affiliation =  
| addressee affiliation = NORTH ATLANTIC ENERGY SERVICE CORP. (NAESCO)
| docket = 05000443
| docket = 05000443
| license number =  
| license number =  
| contact person =  
| contact person =  
| document report number = 50-443-98-04, 50-443-98-4, NUDOCS 9808050169
| document report number = 50-443-98-04, 50-443-98-4, NUDOCS 9812090273
| package number = ML20236W239
| title reference date = 08-28-1998
| document type = INSPECTION REPORT, NRC-GENERATED, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| document type = CORRESPONDENCE-LETTERS, OUTGOING CORRESPONDENCE
| page count = 20
| page count = 3
}}
}}


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December 4,1998 Mr. Ted '
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Executive Vice President and Chief Nuclear Officer Seabrook Station North Atlantic Energy Service Corporation c/o Mr. Terry L. Harpster P.O. Box 300 Seabrook, NH 03874 SUBJECT: INSPECTION REPORT NO. 50-443/98-04
U. S. NUCLEAR REGULATORY COMMISSION


==REGION I==
==Dear Mr. Feigenbaum:==
Docket No.: 50-443 License No.: NPF-86 l
This letter refers to your August 28,1998 correspondence, in response to our July 30,1998 letter.


Report No.: 50-443/98-04    J Licensee: North Atlantic Energy Service Corporation
Thank you for informing us of the corrective and preventive actions documented in your letter. These actions will be examined during a future inspection of your licensed program.
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Facility: Seabrook Generating Station, Unit 1
Your cooperation with us is appreciated.


Location: Post Office Box 300    l Seabrook, New Hampshire 03874 Dates: May 17,1998 - July 4,1998 inspectors: Ray K. Lorson, Senior Resident inspector Javier Brand, Resident inspector Approved by: Curtis J. Cowgill, Chief, Projects Branch 5 Division of Reactor Projects I
Sincerely,
 
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Original Signed By:
9800050169 900730 PDR ADOCK 05000443 G  PDR .
Curtis J. Cowgill, Chief Projects Branch 5 Division of Reactor Projects Docket No. 50-443 cc w\o ev of Licensee's Resoonse Letter:
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B. D. Kenyon, President - Nuclear Group J. S. Streeter, Recovery Officer - Nuclear Oversight W. A. DiProfio, Station Director - Seabrook Station R. E. Hickok, Nuclear Training Manager - Seabrook Station D. E. Carriere, Director, Production Services L. M. Cuoco, Senior Nuclear Counsel 090034    , of jd
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9812090273 981204 PDR 0 ADOCK 05000443 PM


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l  EXECUTIVE SUMMARY Seabrook Generating Station, Unit 1 NRC Inspection Report 50-443/98-04
Mr. Ted Seabrook Station
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This inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers a 7-week period of resident inspectio Operations:
e Routine operations were performed well and operators were knowledgeable of plant and equipment status. The plant shutdown and cooldown were performed wel l The operators did not control a planned pressurizer level increase well which resulted in exceeding the allowable pressurizer cooldown limits. A subsequent evaluation indicated that the pressurizer integrity was not compromised by this even * Safety-related systems and component material conditions were adequat * The licensee implemented several initiatives to improve the effectiveness of the Nuclear Safety and Audit Review Committee Maintenance:
i e The licensee did not promptly initiate action to confirm the operability of the steam  ,
pressure protection channels on the A and D steam generators (NOV 98-04-01). l The licensee's investigation into lead / lag card methodology issues was thoroug * The licensee promptly identified and investigated an abnormal noise in the emergency feedwater pump room. The engineering evaluation and follow-up of this condition was sound. The decision to repair the leaking valves during the forced outage was appropriate, and the repair activities were effectiv * The licensee performed freeze seal activities well. Minor procedural weaknesses were noted regarding precautions for installing freeze seals near welded joints. The corrective actions for the RC-V-89 pipe leak involving prevention of wetting to insulation had not been completed as scheduled.'
e The forced outage was performed safely. The mode change controls implemented prior to start-up were appropriat Enaineerina:
* The evaluation of an evaporator coil overpressurization event was good. A weakness was noted involving the initial estimate of the maximum coil pressur ii
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cc w/cv of Licensee's Response Letter:
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W. Fogg, Director, New Hampshire Office of Emergency Management D. McElhinney, RAC Chairman, FEMA Rl, Boston, Mass R. Backus, Esquire, Backus, Meyer and Solomon, New Hampshire D. Brown-Couture, Director, Nuclear Safety, Massachusetts Emergency Management Agency F. W. Getman, Jr., Vice President and General Counsel - Great Bay Power Corporation R. Hallisey, Director, Dept. of Public Health, Commonwealth of Massachusetts    j Seacoast Anti-Pollution League D. Tefft, Administrator, Bureau of Radiological Health, State of New Hampshire S. Comley, Executive Director, We the People of the United States      ;
e The licensee responded well to investigate the temporary loss of the normal reactor coolant pump seal cooling flow. The identified causes and corrective actions for this event were adequat e The licensee's actions to improve the reliability of the control building air conditioning system were extensive. The SORC review of an operability determination for a degraded compressor lubricating oil system pressure condition could have been more complete since it did not consider all available technical informatio j Plant Suonort:
W. Meinert, Nuclear Engineer        '
e The radiological control technicians at the radiological controlled area (RCA) l checkpoint and in the field were generally attentive, knowledgeable, and provided I high quality assistance to ensure proper radiological work practices. The RCA  i access turnstile installation was a good initiative to ensure that radiation workers I comply with RCA access control requirements. The inspector identified a poor practice involving a health physics technician who performed an activity on a potentially contaminated system without wearing protective gloves. The licensee's corrective actions for this event were adequat l e The licensee did not implement interim corrective actions to enhance the control of )
cleaning boron from remotely operated valves following an event where a valve was ;
unexpectedly positioned during a cleaning activity,   i i
e Security activities were performed well. The licensee implemented a detailed plan )
to ensure proper vital area access controls were maintained during a planned demonstratio ,
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TABLE OF CONTENTS Paae i
EX EC UTIV E S U M M A RY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii TAB LE O F C O NTENT S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv
!    l. Operations ....................................................1    i 01 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 01.1 G ene ral Comm e nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 02 Operational Status of Facilities and Equipment ................... 1 02.1 Fa cility Tours . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 04 Operator Knowledge and Performance . . . . . . . . . . . . . . . . . . . . . . . . . 2 04.1 Operator Performance Observations . . . . . . . . . . . . . . . . . . . . . . 2 08 Miscellaneous Operations issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 08.1 (Closed) LER 98-004: Minimum Shift Crew Composition . . . . . . . . 3 08.2 Nuclear Safety And Audit Review Committee Meeting . . . . . . . . . 3 11. M ai nt e n a n c e . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 M1 Conduct of Maintenance . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 M1.1 Steam Pressure Lead / Lag Calibrations ....................3 M1.2 Emergency Feedwater Stop Check Valve Leakage . . . . . . . . . . . . 4  )
M1.3 Freeze Seals to Support Repairs of Relief Valves CS.V-250 and SI-V-   l 113 ............................................ 5 M 1.4 Fa n Belt Adjustme nt . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 M 1.5 Surveillance Testing . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 M1.6 Forced Outage Activities .............................7    1 M8 Miscellaneous Maintenance issues . . . . . . . . . . . . . . . . . . . . . . . . . . . . 7 M8.1 (Closed) Service Water Pump Replacement (IFl 96-004-02) ..... 7 M8.2 (Closed) Inspection Follow-up Item (IFI) 50-443/96-10-01....... 8 111. En g i n e e ri n g . . . . . _ . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 E1 Conduct of Engineering . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8  '
E CBA Coil Overpressurization Evaluation ................... 8 E2 Engineering Support Of Facilities And Equipment . . . . . . . . . . . . . . . . . . 9 E2.1 investigation into Loss of RCP Seal injection Flow . . . . . . . . . . . . 9 E2.2 CBA System Design Modifications . . . . . . . . . . . . . . . . . . . . . . 10  i
 
IV. Plant Support ................................................12 R1 Radiological Protection and Chemistry Controls . . . . . . . . . . . . . . . . . . 12 R1.1 G eneral Comm e nts . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 R8 Miscellaneous Radiological and Chemistry Controls issues . . . . . . . . . . 13 S1 Conduct of Security and Safeguards Activities ..................13 V. Management Meetings ..........................................13 X1 Exit Mee ting Sum m ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 X3 Other N RC Activitie s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 iv i
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PARTI AL LIST OF PERSONS CONTACTED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14
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INSPECTION PROCEDURES USED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . , 14
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ITEMS OPENED, CLOSED, AND DISCUSSED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 i
l l  L1ST O F ACRO NYMS U SED . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 l
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Report Details        l Summary of Plant Status I
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The facility began the period operating at approximately 100% of rated thermal powar. On June 11, the operators declared both trains of the control building air conditioning system (CBA) inoperable, entered Technical Specification 3.0.3, and shutdown the plant. The plant remained shutdown for the remainder of the inspection perio l. Operations _
01 Conduct of Operations 01.'i General Comments (71707)
Using Inspection Procedure 71707, the inspectors conducted frequent reviews of ongoing plant operations, in general, routine operations were performed in accordance with station procedures and plant evolutions were completed in a deliberate menner with clear communications and effective oversight by shift supervision. Control room logs accurately reflected plant activities and observed      ;
shift turnovers were comprehensive and thoroughly addressed questions posed by      )
the oncoming crew. Control room operators displayed good questioning perspectives prior to releasing work activities for field implementation. The inspectors found that operators were knowledgeable of plant and system statu Operational Status of Facilities and Equipment 02.1 Facility Toors (71707,62707)
The inspectors routinely conducted independent plant tours and walkdowns of selected portions of the primary containment, primary auxiliary building, and the emergency diesel generator, emergency feedwater, service water and spent feel pool buildings. These activities consisted of verification that safety-related system configurations, power supplies, process parameters, and operational status were consistent with Technical Specification (TS) requirements, and Updated Final Safety Analysis Report (USFAR) descriptions. The inspector observed that conditions were adequate in these buildings. Some minor material deficiencies were observed that had not been identified by the station staff including:
  * Excessive boron accumulation outboard the 'D' reactor coolant system loo '
  * Minor valve packing and actuator oilleakag * A configuration discrepancy between a main steam valve and its associated drawing.
 
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  * Examples of wetted insulation on safety system pipin )
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  * A non-functioning emergency ligh These issues were identified to the licensee for correction. The inspectors will continue to perform system walkdowns to assess the licensee's performance in      j l
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identification of material deficiencies and also to evaluate the plant material condition.
 
l 04  Operator Knowledge and Performance (71707)
04.1 Operator Performance Observatio_ns Inspection Scope (71702)
The inspectors reviewed ope.ator performance during routine evolutions and in I
respon:.i to plant events. Particular attention was focused on the reactor shutdown
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and cooldown and control of plant shutdown conditions.
 
[ Observations and Findinas The plant shutdown and cooldown activities on June 11, and 12, were performed well and in accordance with operations procedure, OS1000.04," Plant Cooldown From Hot Standby To Cold Shutdown". The inspectors performed several random checks and verified that the TS 3.4.9.1 cooJdown and pressure-temperature limits were satisfie The operators controlled the plant conditions generally well during the shutdown period, however, on June 19,1998, the TS 3.4.9.2 pressurizer cooldown limits were exceeded. The event occurred when the operators raised pressurizer levelin preparation for a maintenance activity that would secure the charging pumps. The charging of the relatively cool reactor coolant into the pressurizer resulted in a cooldown in excess of the 200 F/hourlimi The licensee entered the TS action statement and performed (in consultation with Westinghouse) an engineering evaluation that concluded that this event had minimal impact on the pressurizer integrity. The inspector reviewed this evaluation and noted that the actual plant conditions were bounded by the previously analyzed limiting transient. Additionally, specialist inspectors reviewed the evaluation and found it to be adequat The licensee initiated al adverse condition report (ACR) and implemented an operator standing order to provide guidance for conducting any activity that would change pressurizer level. The inspector reviewed this standing order and found that
;  it provided clear guidance to the operators. The inspector concluded that the
;  operators did not control the pressurizer level increase well, however, the subsequent pressurizer cooldown did not challenge the pressurizer integrity. This
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event is a violation of minor significance and not subject to formal enforcement l  action.


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i Mr. Ted Seabrook Station Distribution w/cv of Licensee Response Letter Region i Docket Room (with concurrences)
3 Conclusions The plant shutdown and cooldown were performed well. Operators did not control a planned pressurizer level increase well which resulted in exceeding the pressurizer cooldown limit Miscellaneous Operations issues (92901)
Nuclear Safety information Center (NSIC)
0 (Closed) LER 98-004: Minimum Shift Crew Composition: Failure to maintain a senior reactor opermor (SRO) in the control room as required by TS 6.2.2. On May 4, the shift manager momentarily (for about one minute)left the control room after assuming the control room SRO position. The inspector performed an in-office review of this event and concluded that the licensee's root cause of personal oversight was appropriate. This f ailure to maintain the required control room staffing levels constitutes a violation of minor significance and is not subject to formal regulatory actio .2 Nuclear Safety And Audit Review Committee Meetina The inspectors observed portions of the routine Nuclear Safety And Audit Review Committee (NSARC) meeting on June 9, and 10, The inspectors noted that the topics discussed were important to plant safety and that the NSARC members participated in the discussions. The two day meeting format was one, of several licensee initiatives, developed to improve the effectiveness of the NSARC. The inspector considered it positive that the licensee was developing initiatives to improve the effectiveness of the NSAR II. Maintenance    l M1  Conduct of Maintenance M 1.1 Steam Pressure Lead / Lao Calibrations Inspection Scop _e The inspectors reviewed the licensee's investigation into technical and performance issues involving the calibration of lead / lag control cards, Observations And Findinas The inspector discussed in inspection Report 98-02 several deficiencies involving the licensee's initial response to out of calibration steam generator pressure lead / lag cards. These deficiencies included a f ailure to promptly confirm, on May 4,1998, the operability of the A and D steam generator pressure lead / lag cards after three of six B and C steam generator channels were known to be inoperable. The licensee, after subsequent NRC questioning, tested the A and D steam generator channels, on May 5, and found that one channel on each steam generator was inoperabl I t              ;
PUBLIC NRC Resident inspector        l C. Cowgill, DRP        l R. Summers, DRP C. O'Daniell, DRP B. McCabe, OEDO C. Thomas, PD l-3, NRR l
 
J. Harrison, PD l-3, NRR        -l
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'R. Correia, NRR        I D. Screnci, PAO, ORA DOCDESK Inspection Program Branch, NRR (IPAS)        j DOCUMENT NAME: G:\ BRANCH 5\RPLY-LTR\S89804.RPY To receive a copy of this document, indicate in the bos: *C" - Copy without ettschrnent/ enclosure "E" = Copy with attachment / enclosure *N" = No copy OFFICE Rl/DRP QQ l  Rl/DRP pf f l  /  l  l  l NAME RSummers 1  CCowgill %
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DATE (0/y/98  10/y/98 i OFFICIAL RECORD COPY l
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The failure to promptly test tne A and D steam generator channels resulted in operation in excess of six hours without placing the inoperable channels in the trip condition as required by TS Table 3.3 3, Action 18. The inspector considered the safety significance of this issue minimal since the inoperable channels remained functional, and each steam generator had two operable pressure protection channels, as require l Appendix B Criterion XVI, Corrective Action, states, in part, that measures shall be established to promptly 'dentify conditions adverse to quality. Contrary to the above, on May 4,1998, the licensee did not promptly act to determine the operability status of the A and D steam generator pressure lead / lag card channel This is a violation of 10 CFR 50, Appendix B, Criterion XVI. (NOV 98-04-01). l The licensee performed a thorough review of the calibration methodologies used for the lead / lag control cards. This review, included forming an event team, consulting with vendor experts, and bench top testing. The review concluded that the control    '
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cards, that had been calibrated using the initial licensee method described in Inspection Report 98-02, were within the TS operability limits. Thus, unresolved item 98-02-01 is closed. Additionally, the licensee submitted Licensee Event Report 98-005-00,on May 29, which adequately described this event and corrective    j actions for the calibration methodology concerns. Licensee Event Report 98-005-00 is close Conclusions The licensee did not promptly establish that two of the steam pressure protection lead / lag control card channels were inoperable on May 4. The licensee's subsequent investigation into the calibration methodology for these cards was thoroug M1.2 Emeraency Feedwater Stop Check Valve Leakaoe 1 Inspection Scoce On June 6,1998, a nuc: ear systems operator (NSO) identified a " banging" noise in the emergency feedwater (EFW) pump room. The licensee had previously evaluated this type of noise and attributed it to main feedwater system backleakage through EFW stop check valves FW-V82 and V88. These valves separate the pressurized l main feedwater system from the normally depressurized EFW discharge heade Previously the system engineer installed thermocouple to monitor and trend the
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! EFW discharge pipe surface temperatur The backleakage, if severe enough, could result in steam void formation in the EFW    j discharge piping or steam binding of the EFW pumps. The inspectors performed several system walkdowns, interviewed plant personnel, ana evaluated the licensee's response to this issue. Additionally, the inspectors reviewed the licensee's actions to repair these valves during the forced outag I
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5 Observations and Findinas The NSOs promptly checked the EFW discharge pipe temperature indications and noted that the temperature had increased by about 15 F (from 130 F to 145 F).
 
The increased temperatures were attributed to either an increased ambient room temperature or an increased backleakage flow through the check valves. Since the noise was potentially caused by steam bubble formation and collapse within the piping, the licensee performed ultrasonic testing, and determined that detectable steam voids had not formed inside the piping. The inspector noted that the ultrasonic testing was performed at the most susceptible pipe location (i.e adjacent to the check valves where the highest temperatures had been noted). The licensee concluded that the EFW system was operabl The inspector questioned the impact of the " banging" on the pipe integrity. The engineering department presented an evaluation for an earlier similar condition (i.e.


l back leakage via the "D" steam generator feedwater recirculation valve FW-V153)
that had been performed in February 1997. This review indicated that the short term pressure wave impact could be accommodated by the piping system and its supports, but cautioned that continuous wave impact could lead to localized component deformation. The licensee visually inspected the associated piping and components and concluded that no degradation existe During the forced outage, the licensee, with assistance of the valve manufacturer,
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disassembled and inspected both the FW-V82, and 88 valves. Minor steam cutting i
  '"%.s    North Atlantic Energy Service Corporation ;
! was observed on the seat and disc of each valve. The licensee repaired the FW-V- l l 82 valve and replaced the FW-V-88 valve. The inspectors observed the valve component inspections, foreign material exclusion (FME) controls, and portions of I
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the valve repair activities and did not identify any deficiencies.
l      Scabrook, Nil 03874 Atlaintic
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! Conclusions The licensee promptly identified and investigated an abnormal noise in the emergency feedwater pump room. The licensee's engineering evaluation and follow-up of this condition was sound. The decision to repair the leaking valves during the forced outage was appropriate, and the repair activities were effectiv M1.3 Freeze Seals to Suonort Reoairs of Relief Valves CS-V-250 and SI-V-113 Insoection Scone l
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mechanical maintenance technicians to support repairs on the reactor coolant pumps seal return relief valve (CS-V-250), and the cold leg safety injection relief valve (SI-V 113). The inspector performed field walkdowns of the proposed freeze seals in the mechanical penetration area and in the demineralized alley prior to implementation, reviewed the work packr'ges, applicable procedure, interviewed the work supervisor, and observed portions of the work. The inspector also I
Ref: ACR 98-1417  !
AR#98009418 i      ACR 98-2038  ,
l      AR#98016104  )
[      AR#98014612  i
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United States Nuclear Regulatory Commission l  Attention: Document Control Desk      i Washington, DC 20555-0001      ;
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Seabrook Station    !
Reply to Notice of Violation  -  !
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  ' North Atlantic Energy Service Corporation (North Atlantic) provides in the enclosure its  !
response to the Notice of Violation described in Inspection Report 98-04.


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I Shodld you have any questions concerning this response, please contact Terry L. Harpster, ,
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Director of Licensing Services, at (603) 773-7765.


observed removal of relief valve SI-V-113, to evaluate the licensee's corrective actions to prevent the wetting of insulation while removing the valv b. Observations and Findinas:
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The briefings conducted by the mechanical supervisor, prior to performing the freeze seal, were excellent. The work packages were thorough and included an adequate on-line maintenance assessment. Additionally, required precautions, and system lineup contingencies were included to prevent or mitigate the consequences of a freeze seal failure. The inspector observed proper field coverage by fire protection, health physics technicians and management. The oversight group performed the required liquid penetrant test of the affected pipes before and after the freeze, which confirmed adequate pipe condition The mechanical supervisor demonstrated a good questioning attitude by performing field walkdowns and verifying actual as built conditions of the associated piping, prior to performing the freeze seals. As a result, for valve CS-V-250, he identified that the distance between two welds in the area of the freeze seal was much closer than specified in the work package, and that the freeze jacket was located within one inch (at both ends) of a pipe weld. The inspector immediately questioned whether or not the procedure requirement, "that pipe sections to be frozen shall not contain welded joints could be met", when taking in consideration the potential growth of the freeze sea Design engineering reviewed the inspector's questions and concluded that the growth of the freeze seal (ice plug) is limited to the size of the freeze jacket, and therefore no freezing of the weld joints would occur. In addition, engineering stated that this procedural requirement was primarily a concern when applying the freeze seal to carbon steel piping. The licensee revised the procedure to clarify the requirement The freeze seals and subsequent work to repair the subject relief valves were completed successfully. The inspector verified that the licensee took periodic temperature readings to confirm that the freeze seal plug did not extend beyond the freeze jacket dimension The inspector observed that water leaked onto the insulation jacket below SI-V-133 during its removal. One of the corrective actions for the RC-V-89 pipe leak was to prevent the wetting of insulation. This action had not been implemented as
Very truly yours,
; scheduled and reflected a minor weakness in the corrective action program l implementatio c. Conclusion:
The licensee performed the freeze seal activities well. The inspec.or found the work package and associated on-line maintenance and freeze seal evaluations adequat Management and oversight support was observed. Procedural weaknesses were noted regarding the precautions for installing the freeze seal near welded joint _ _ - _ _ _ _ - - _ _ _ _ _ _ _ _ _ _ _  _ _ _ _ _ _ _ ______ _ ___ _ _ _-________
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The corrective actions for the RC-V-89 pipe leak involving the prevention of wetting of insulation had not been completed as schedule M1. 4 Fan Belt Adiustment The inspector observed mechanical maintenance technicians adjusting the drive belt tension on an emergency switchgear room ventilation supply fan. The inspector noted three issues that challenged the technician's ability to adjust the belt tension to within the limits specified by maintenance procedure, MS0523.48," Sheave Alignment and Belt Tensioning". These issues included the measurement tolerances that could be achieved (force and deflection), using the deflection gage, and the method used to identify the belt measurement location. The licensee indicated that the inspector's concerns would be reviewed and addressed. The inspector concluded that this response was appropriat M1.5 Surveillance Testina The inspector observed surveillance terting and confirmed that the tests were properly controlled, test instrumentatica was within its required calibration periodicity and accuracy, test acceptance limits were met, and that the test scope was adequate. The following test activities were observed and no significant deficiencb were noted:
NORTH ATLANTIC ENERGY SERVICE CORP.
    * Emergency Diesel Generator Load Testing
    * Turbine Driven Emergency Feedwater System Testing
    * Emergency Battery Testing M1.6 Forced Outaae Activities The inspector noted that the licensee controlled the forced outage activities safely and repaired a number of key safety and operational components prior to the plant start-up. These activities included: repair of packing and control air leaks on main steam, and letdown system valves, replacement of three main steam safety valves, repair of emergency feedwater system recirculation, and isolation valves, extensive modifications to the control building air conditioning (CBA system), and completion of a majority of the 18 month surveillance tests. The inspectors verified that activities were properly scheduled, and that plant conditions were adequate to perform the intended activity. Additionally, the inspector attended a Mode Four change meeting and noted that the appropriate issues were identified for resolution I    prior to the mode chang M8 Miscellaneous Maintenance issues M8.1 (Closed) Service Water Pumo Replacement (IFl 96-004-02) Inspection Report 96-10 discussed the use of a non-qualified lifting device during removal of a service water pump. The device in question had been previously load tested and, inspected immediately prior to performing the lift, however, the licensee had not properly evaluated this device prior to use. The licensee subsequently performed an


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l    Additional corrective actions included: an inspection and audit of other lifting l    devices that had the potential for use in safety-related applications. The inspector
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found the licensee's corrective actions to be reasonable and complete, and did not identify any violations. This item is close M8.2 (Closed) Inspection Follow-uo item (IFI) 50-443/96-10-01: Transmitter l   Configuration Control. The inspectors identified in Inspection Report 96-10 that (    plastic FME caps had been installed on safety-related service water pressuro transmitters. The plastic caps could potentially challenge the environmental qualification of the affected transmitters in a harsh environmental. The licensee performed an investigation which included a field walkdown of all related equipment to identify and correct any non-conforming condition The licensee inspected approximately 500 transmitters in both safety and non-safety related applications, and identified that a total of 82 transmitters had the improper cap installed. Three of the transmitters were located in a harsh environment but were not required for operability. The inspector noted that the licensee's initial evaluation of these transmitters did not address whether they were required for operability. This was considered a minor weakness; otherwise the corrective actions for this condition were reasonable and complete. No violations l    were identified, and this item is close !!!. Enaineerina E1 Conduct of Engineering E CBA Coil Overoressurization Evaluatio_q Insoection Scope (37551)
Executive Vice ' resident and ChiefNuclear Officer
The inspector reviewed Engineering Evaluation EE-98-021, that was performed following a control building air conditioning (CBA) system evaporation coil overpressurization even Findinas and Conclusions The evaluation concluded that the maximum allowable coil pressure had not been exceeded. This conclusion was based on a statement by the individuals involved with the event that the observed pressure was within the first increment on the high pressure source regulator gage. The inspector noted that this was equivalent to a pressure of approximately 160 psig, which was not consistent with the observed plastic deformation of the test gage pressure sensing tube which had a maximum range of 350 psi The inspector questioned the design engineer regarding the pressure difference between the individual's observation, and the observed gage deformation that should not have occurred below the maximum indicated pressure on the gage. The
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  .cc: H. J. Miller, NRC Region 1 Administrator
;  J. T. Harrison, NRC Project Manager, Project Directorate 1-3 R. K, Lorson, NRC Senior Resident Inspector
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l NYN-98105 REPLY TO NOTICE OF VIOIfATION Carriere, D. COMELEC Cook,N. e-mail Cuoco, L. M. e-mail Garfield, G. e-mail ;
Gillespie, J. Millstone 475/5 Kacich, R. M. e-mail Makowicz, M. J. e-mail Millstone Nuclear Licensing e-mail 1 NSARC Distribution e-mail Quinlan, W. J. e-mail File 0001  01-48 i File 0020  01-48 l RMD  02-06 l
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licensee did not have any data regarding the pressure at which the gage would elastically deform, and subsequently performed an analysis and test of a similar gage which indicated that the plastic deformation could have occurred at a pressure range of 400-600psig. The licensee also noted that this pressure range would not challenge the evaporator coil integrity. Additionally, the licensee successfully performed a visual, and pressure test of the evaporator coi l Conclusions      l The licensee performed a good evaluation of an control building air conditioning l system evaporator coil overpressurization event. A minor weakness was noted j involving the initial evaluation of the maximum coil pressure. The licensee subsequently revised the analysis and developed an adequate basis for the coil pressure.
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l E2 Engineering Support Of Facilities And Equipment (37551,40500)
E2.1 investigation into Loss of RCP Sealiniegtion Flow inspection Scoce:
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On June 11,1998, while cooling the plant down from Mode 3 to Mode 5, reactor l
coolant pump (RCP) seal injection flow was lost to the all four running RCPs for l approximately 40 seconds. The low RCP flow alarm alerted the control room  l l operator (RO) to this condition. The RO promptly restored the sealinjection flow
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and verified that the other RCP parameters were normal.
 
l The licensee initiated a root cause analysis (ACR 98-1735)for system engineering .
to evaluate this event, and determine the necessary corrective action l Additionally, the system engineer evaluated other recently reported instances of seal l injection flow problems. The inspector reviewed RCP pump performance data,  !
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analysis.
 
, Observations and Findinas:
The sealinjection flow is designed to cool the RCP seals. The RCP seals prevent j reactor coolant system (RCS) leakage along the pump shaft from reaching the containment atmosphere. Flow to the RCP seals is normally provided by the  i charging system. Alternatively, RCS system backleakage, cooled by the thermal  I barrier heat exchanger, will cool the RCP seats if the primary method is unavailable.
 
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The backup method is not preferred, however, since it supplies unfiltered water to I the RCP seals that could potentially lead to seal degradatio The licensee consulted with the pump vendor, and determined that the RCPs seals would not be damaged by this temporary interruption in the normal RCP seal  !
injection flow. The evaluation concluded that the most like cause for this event I was an inadvertent blockage of flow to the operating online filter element in the seal i i
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water system combined with an inadvertent closing of the inlet valve during the e filter swap or a momentary blockage at the inlet valve. The operators stroked the inlet isolation valve to ensure proper operation, and the operations manager briefed personnel on the event.
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i A separate evaluation, conducted by the system engineer, identified another l potential cause involving operation of the charging flow control air-operated valve (CS-V-121) during shutdown conditions (i.e, Mode 3). In this condition, the charging valve is required to control at a low flow band with high differential pressure, causing the valve to operate virtually closed. The system engineer initiated a request to have design engineering evaluate this conditio Conclusion:      i I
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E2.2 CBA System Desian Modifications l Inspection Scope The inspectors reviewed the modifications implemented to improve the reliability of the CBA system after both trains were declared inoperable requiring a plant shutdown per TS 3.0.3 on June 11,1998. The inspector also reviewed operability determination (OD) 98-010 that had been performed per operating procedure, OE 4.5, " Operability Determination" to evaluate a reduced compressor lubricating oil pressure, a Observations and Findinos
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The CBA system has experienced previous operating problems as documented in Inspection Report 97-08, and LER 97-018. On June 11,1998, the 2A CBA compressor became inoperable before the 2B CBA compressor, which had been removed from service to install a modification, could be returned to service. The licensee formed a project team to identify the required actions to restore the
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reliability and operability of the CBA system.


' In addition to the project team, the licensee, formed a team comprised of external individuals with significant industrial air conditioning system experience to independently review the system performance. The Team identified several l-problems that appeared to contribute to the system degradation. These factors  )
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REPLY TO NOTICE OF VIOLATION NRC Inspection Report 98-04 describes a violation where Nonh Atlantic did not take action to promptly determine the status of the A and D steam generator steam pressure protection channels. North Atlantic's response to this violation is presented below.
 
The licensee developed extensive system modifications to address these factors including: re-sizing of the compressor supply line, relocation of the thermal expansion valves, elimination of low flow areas and loop seals in the compressor suction line, de-rating of the compressor, and installation of a surge tank at the inlet to the compressor suction. The system modifications and testing were on-going at the completion of the period, however, the inspector noted that the licensee's corrective actions appeared comprehensive, and involved the technical support of outside experts. The Station Director also indicated that the long-term plan is to procure a replacement safety-related system that will cool the control room by chilled water instead of the present freon expansion syste Operability determination (OD) 98-010, was initiated to evaluate an unexpected decrease in the 5B CBA compressor lubricating oil pump differential pressure on May 17,1998. The evaluation was completed and accepted by the Station Operation Review Committee (SORC) on June 17,1998. The evaluation concluded that the 5B CBA compressor was operable following the unexpected drop in oil pressure based on: the differential pressure was within the manufacturer's limits, all other compressor operating parameters were normal, and a compressor crankcase oil sample was normal. The final evaluation did not discuss, however, any potential cause(s) for the lubricating oil pressure decrease, nor provide an assessment regarding the length of time that the oil pump could have been expected to functio The compressor was disassembled and inspected by an off-site vendor on July 13, 1998. The inspection results identified a number of non-conforming conditions within the compressor, and its internal lubricating system. The inspector noted that the SORC review was not complete in that the inspector did not consider this information during its review of OD 98-010. The licensee indicated that the i information would not have changed the final operability determination for this compresso j Conclusions The licensee's actions to improve the reliability of the control building air conditioning system were extensive. The SORC review of an operability determination for a degraded compressor lubricating oil system pressure condition could have been more complete since it did not consider all available technical information, i


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1. Description of the Violation:
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The following is a restatement of the violation:
10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, states, in part, that measures shall be implemented to ensure that conditions adverse to quality are promptly identified.


IV. Plant Support R1 Radiological Protection and Chemistry Controls R1.1 General Comments Inspection Scoce During the inspection period the inspector toured the radiological controlled area (RCA) on several occasions (including inside containment during the forced shutdown) to observe radiological control practice Observations and Findinas The radiological controls technicians at the RCA checkpoint and in the field were attentive, knowledgeable, and provided high quality assistance to ensure proper radiological work practices. The installation of an access control turnstile was a good initiative to address licensee identified problems with radiation worker adherence to RCA entrance and access requirement On June 26,1998, the inspector identified a health physics (HP) technician not wearing protective gloves while taking a smear of a breached and potentially contaminated component. The operations and maintenance technicians involved with this activity were wearing protective gloves. The HP department initiated ACR 98-1905 to investigate this inciden The licensee concluded that Seabrook's program does not formally require HP technicians to wear protective clothing (i.e; gloves) whenever a potentially ,
contaminated boundary is breached, but instead relies on the technicians judgement l to determine the required level of protection. The HP manager acknowledged that !
wearing gloves for this type of activity was a good industry practice and that the technician should have been wearing protective gloves'. The licensee coached and counseled the individual, and briefed the other HP technicians on the issue. The inspector considered the licensee's actions to be adequat l Conclusion The radiological control technicians at the radiological controlled area (RCA)
checkpoint and in the field were generally attentive, knowledgeable, and provided high quality assistance to ensure proper radiological work practices. The RCA access turnstile was a good initiative to ensure that radiation workers comply with RCA access control requirements. The inspector identified a poor practice involving a health physics technician who performed an activity on a potentially contaminated system without wearing protective gloves. The licensee's corrective actions for this event were adequat I l
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l Contrary to the above, on May 4,1998, NAESCO did not take action to promptly identify l  whether steam pressure protection channels for the A and D steam generators were operable, !
 
after three of the six steam pressure protection channels for the B and C steam generators were found to be inoperable. Subsequent testing determined that two of the six A and D steam generator pressure protection channels v'ere inoperable.
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13 R8  Miscellaneous Radiological and Chemistry Controls issues The inspector reviewed an event involving a motor-operated reactor coolant system valve (RC-V-87) that had been remotely operated while an individual was cleaning boron residue from the valve. The licensee initiated an ACR to review this event, and develop corrective actions. The inspector discussed this issue with the Radiological Protection Manager who indicated that existing program guidance for cleaning motor and air operated valves did not require notification of control room or work control personnel and that program guidance would be strengthened in this are The inspector was concerned that the cleaning activity could result in a personnel or equipment hazard and questioned the licensee regarding the interim actions taken to ensure that on-going cleaning activities were properly controlled. The licensee indicated that the personnel assigned to perform the cleaning activity had been informed of the inadvertent valve stroking event, but no other controls had been implemented. The licensee subsequently suspended the boron removal activities pending development of the revised standards. The inspector concluded that not implementing the interim controls represented a corrective action program weaknes S1  Conduct of Security and Safeguards Activities The inspectors observed security force performance during inspection activitie Protected area access controls were found to be properly implemented during random observations. Proper escort control of visitors was observed. Security officers were alert and attentive to their duties. The inspectors noted that the     ,
licensee developed and implemented a detailed plan, that included participation by    !
I    local law enforcement personnel, to ensure that proper vital area access controls
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were maintained during an off-site demonstration.


l V. Manaaement Meetinas X1  Exit Meeting Summary The inspectors presented the inspection results to members of licensee l    management, following the conclusion of the inspection period, on July 8,1997.
This is a Severity Level IV Violation (Supplement 1).


l   The licensee acknowledged the findings presented.
2. Reply to the Notice of Violation:
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Reason for the Violation North Atlantic agrees with the violation. On April 27, 1998, North Atlantic commenced calib' ration activities for six (three each on the B and C steam generators) steam generator l  pressure protection channels. On May 1,1998, the lead / lag value for three channels were found l outside of the acceptability limits specified in the calibration procedures. The three channels were calibrated and returned to service. On May 4,1998, North Atlantic determined that the lead / lag values addressed in the calibration procedures were Technical Specification allowable l  values and, as such, the as-found data was outside the Technical Specification limits.


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An Adverse Condition Report (ACR) was initiated and the event was determined to require a
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l  License Event Report (LER) pursuant to 10CFR50.73(a)(2)(i) due to the operation of the plant l
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from May 1,1998 to May 4,1998 in a condition prohibited by the Technical Specifications.
and licensee managers and technical staff leads was conducted on July 2,1998, to discuss the corrective actions to restore the control building air conditioning (CBA)
system.


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North Atlantic submitted LER 98-005-00 on May 29,1998.


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On May 1,1998, Station management requested Maintenance personnel to investigate the applicability of this condition on the A and D steam generators. This was not acted upon
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PARTIAL LIST OF PERSONS CONTACTED Licensee W. Diprofio, Unit Director J. Grillo, Technical Support Manager R. White, Design Engineering Manager J. Peterson, Maintenance Manager G. StPierre, Operations Manager B. Seymour, Security Manager J. Linville, Chemistry and Health Physics Manager INSPECTION PROCEDURES USED IP 37551:  Onsite Engineering IP 61726:  Surveillance Observation IP 62707:  Maintenance Observation IP 71707:  Plant Operations IP 71750:  Plant Support Activities IP 92700:  Onsite Followup of Written Reports of Nonroutine Events at Power Reactor Facilities ITEMS OPENED, CLOSED, AND DISCUSSED Ooened NOV 98-04-01, Failure To Promptly identify inoperable Steam Pressure Channels  j Closed:
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I U RI, 98-02-01, Lead / Lag Control Card Methodology    ]
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LER 98-004-00, Minimum Shift Crew Composition    i LER 98-005-00,Inoperability Of Steam Pressure Protection Channels    l IFl 96-004-02, Service Water Pump Replacement IFl 96-010-01, Transmitter Configuration Control l
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when the ACR was discussed during the daily Station meeting on May 5,1998, insufficient  l
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information was presented regarding the potential for the condition to be common to the other steam generators and as a result the operability of the A and D steam generators was not pursued' >
promptly.      !
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This event was included in the scope of a comprehensivbreview of lead / lag issues. North !
Atlantic performed a thorough review of the calibration methodologies used to determine NLL !
card time constants. This review was proactive and involved forming an event team, consulting l with vendors and bench top testing.


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The cause of this event is that the Corrective Action System was not effectively implemented ,
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after the initiation of the ACR. There was not adequate communication between the departments regarding the specifics of the channel calibration failures and the potential for common failures - l was not adequately questioned / pursued. As a result, action was not promptly initiated to validate ,
the status of the A and D steam generators in a timely manner. l
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Corrective Actions
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1. The individuals involved in initiating and reviewing this Adverse Condition Report have l been coached and counseled.    !
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Date When Comoliance Will Be Achieved    i
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North Atlantic is currently in compliance with 10 CFR 50 Appendix B, Criterion XVI
  " Corrective Action."


LIST OF ACRONYMS USED ACR Adverse Condition Report ASME American Society of Mechanical Engineers CAS Central Alarm Station CBS Containment Building Spray EDG Emergency Diesel Generator EFW Emergency Feedwater FME Foreign Material Exclusion LCO Limiting Condition for Operation NSARC Nuclear Safety and Audit Review Committee RHR Residual Heat Removal SORC Station Operations Review Committee SUFP Startup Feedwater Pump SW Service Water TDEFW Terbine Driven Emergency Feedwater Pump TS Technical Specifications UFSAR Updated Final Safety Analysis Report WR work request
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Revision as of 02:41, 9 December 2021

Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Rept 50-443/98-04 Issued on 980730
ML20196J163
Person / Time
Site: Seabrook NextEra Energy icon.png
Issue date: 12/04/1998
From: Cowgill C
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To: Feigenbaum T, Harpster T
NORTH ATLANTIC ENERGY SERVICE CORP. (NAESCO)
References
50-443-98-04, 50-443-98-4, NUDOCS 9812090273
Download: ML20196J163 (3)
v • d • e


Text

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December 4,1998 Mr. Ted '

Executive Vice President and Chief Nuclear Officer Seabrook Station North Atlantic Energy Service Corporation c/o Mr. Terry L. Harpster P.O. Box 300 Seabrook, NH 03874 SUBJECT: INSPECTION REPORT NO. 50-443/98-04

Dear Mr. Feigenbaum:

This letter refers to your August 28,1998 correspondence, in response to our July 30,1998 letter.

Thank you for informing us of the corrective and preventive actions documented in your letter. These actions will be examined during a future inspection of your licensed program.

Your cooperation with us is appreciated.

Sincerely,

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Original Signed By:

Curtis J. Cowgill, Chief Projects Branch 5 Division of Reactor Projects Docket No. 50-443 cc w\o ev of Licensee's Resoonse Letter:

B. D. Kenyon, President - Nuclear Group J. S. Streeter, Recovery Officer - Nuclear Oversight W. A. DiProfio, Station Director - Seabrook Station R. E. Hickok, Nuclear Training Manager - Seabrook Station D. E. Carriere, Director, Production Services L. M. Cuoco, Senior Nuclear Counsel 090034 , of jd

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9812090273 981204 PDR 0 ADOCK 05000443 PM

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Mr. Ted Seabrook Station

cc w/cv of Licensee's Response Letter:

W. Fogg, Director, New Hampshire Office of Emergency Management D. McElhinney, RAC Chairman, FEMA Rl, Boston, Mass R. Backus, Esquire, Backus, Meyer and Solomon, New Hampshire D. Brown-Couture, Director, Nuclear Safety, Massachusetts Emergency Management Agency F. W. Getman, Jr., Vice President and General Counsel - Great Bay Power Corporation R. Hallisey, Director, Dept. of Public Health, Commonwealth of Massachusetts j Seacoast Anti-Pollution League D. Tefft, Administrator, Bureau of Radiological Health, State of New Hampshire S. Comley, Executive Director, We the People of the United States  ;

W. Meinert, Nuclear Engineer '

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i Mr. Ted Seabrook Station Distribution w/cv of Licensee Response Letter Region i Docket Room (with concurrences)

Nuclear Safety information Center (NSIC)

PUBLIC NRC Resident inspector l C. Cowgill, DRP l R. Summers, DRP C. O'Daniell, DRP B. McCabe, OEDO C. Thomas, PD l-3, NRR l

J. Harrison, PD l-3, NRR -l

'R. Correia, NRR I D. Screnci, PAO, ORA DOCDESK Inspection Program Branch, NRR (IPAS) j DOCUMENT NAME: G:\ BRANCH 5\RPLY-LTR\S89804.RPY To receive a copy of this document, indicate in the bos: *C" - Copy without ettschrnent/ enclosure "E" = Copy with attachment / enclosure *N" = No copy OFFICE Rl/DRP QQ l Rl/DRP pf f l / l l l NAME RSummers 1 CCowgill %

DATE (0/y/98 10/y/98 i OFFICIAL RECORD COPY l

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'"%.s North Atlantic Energy Service Corporation ;

l- g g.gjg P.O. Box 300  !

l Scabrook, Nil 03874 Atlaintic

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(603>4:4932i  ;

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The Northea6t Utilities Sptem f

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Ref: ACR 98-1417  !

AR#98009418 i ACR 98-2038 ,

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United States Nuclear Regulatory Commission l Attention: Document Control Desk i Washington, DC 20555-0001  ;

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Seabrook Station  !

Reply to Notice of Violation -  !

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' North Atlantic Energy Service Corporation (North Atlantic) provides in the enclosure its  !

response to the Notice of Violation described in Inspection Report 98-04.

I Shodld you have any questions concerning this response, please contact Terry L. Harpster, ,

Director of Licensing Services, at (603) 773-7765.

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Very truly yours,

NORTH ATLANTIC ENERGY SERVICE CORP.

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'Ted CTFeigenba O f/hn - '

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Executive Vice ' resident and ChiefNuclear Officer

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.cc: H. J. Miller, NRC Region 1 Administrator

J. T. Harrison, NRC Project Manager, Project Directorate 1-3 R. K, Lorson, NRC Senior Resident Inspector

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l NYN-98105 REPLY TO NOTICE OF VIOIfATION Carriere, D. COMELEC Cook,N. e-mail Cuoco, L. M. e-mail Garfield, G. e-mail ;

Gillespie, J. Millstone 475/5 Kacich, R. M. e-mail Makowicz, M. J. e-mail Millstone Nuclear Licensing e-mail 1 NSARC Distribution e-mail Quinlan, W. J. e-mail File 0001 01-48 i File 0020 01-48 l RMD 02-06 l

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REPLY TO NOTICE OF VIOLATION NRC Inspection Report 98-04 describes a violation where Nonh Atlantic did not take action to promptly determine the status of the A and D steam generator steam pressure protection channels. North Atlantic's response to this violation is presented below.

s.

1. Description of the Violation:

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The following is a restatement of the violation:

10 CFR 50, Appendix B, Criterion XVI, Corrective Actions, states, in part, that measures shall be implemented to ensure that conditions adverse to quality are promptly identified.

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l Contrary to the above, on May 4,1998, NAESCO did not take action to promptly identify l whether steam pressure protection channels for the A and D steam generators were operable, !

after three of the six steam pressure protection channels for the B and C steam generators were found to be inoperable. Subsequent testing determined that two of the six A and D steam generator pressure protection channels v'ere inoperable.

This is a Severity Level IV Violation (Supplement 1).

2. Reply to the Notice of Violation:

Reason for the Violation North Atlantic agrees with the violation. On April 27, 1998, North Atlantic commenced calib' ration activities for six (three each on the B and C steam generators) steam generator l pressure protection channels. On May 1,1998, the lead / lag value for three channels were found l outside of the acceptability limits specified in the calibration procedures. The three channels were calibrated and returned to service. On May 4,1998, North Atlantic determined that the lead / lag values addressed in the calibration procedures were Technical Specification allowable l values and, as such, the as-found data was outside the Technical Specification limits.

An Adverse Condition Report (ACR) was initiated and the event was determined to require a

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l License Event Report (LER) pursuant to 10CFR50.73(a)(2)(i) due to the operation of the plant l

from May 1,1998 to May 4,1998 in a condition prohibited by the Technical Specifications.

North Atlantic submitted LER 98-005-00 on May 29,1998.

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On May 1,1998, Station management requested Maintenance personnel to investigate the applicability of this condition on the A and D steam generators. This was not acted upon

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l promptly. Additionally, the individuals involved in the preparation and review of the ACR did not recognize that the out of tolerance condition identified on the B and C steam generators should have called into question the adequacy of similar channel calibrations performed on the A and D steam generators. By May 4,1998, North Atlantic had not pursued the validation of the i status of similar channels on the A and D steam generators in a timely manner. In addition, Page 1 of 2 1 . , -

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when the ACR was discussed during the daily Station meeting on May 5,1998, insufficient l

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information was presented regarding the potential for the condition to be common to the other steam generators and as a result the operability of the A and D steam generators was not pursued' >

promptly.  !

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This event was included in the scope of a comprehensivbreview of lead / lag issues. North !

Atlantic performed a thorough review of the calibration methodologies used to determine NLL !

card time constants. This review was proactive and involved forming an event team, consulting l with vendors and bench top testing.

The cause of this event is that the Corrective Action System was not effectively implemented ,

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after the initiation of the ACR. There was not adequate communication between the departments regarding the specifics of the channel calibration failures and the potential for common failures - l was not adequately questioned / pursued. As a result, action was not promptly initiated to validate ,

the status of the A and D steam generators in a timely manner. l

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Corrective Actions

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1. The individuals involved in initiating and reviewing this Adverse Condition Report have l been coached and counseled.  !

Date When Comoliance Will Be Achieved i

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North Atlantic is currently in compliance with 10 CFR 50 Appendix B, Criterion XVI

" Corrective Action."

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