IR 05000324/1988024: Difference between revisions

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{{Adams
{{Adams
| number = ML20154B018
| number = ML20206B502
| issue date = 09/01/1988
| issue date = 11/09/1988
| title = Insp Repts 50-324/88-24 & 50-325/88-24 on 880707-0807. Violations Noted.Major Areas Inspected:Onsite LER Review (Unit 1),in Ofc LER Review (Unit 1).unusual Event/Fire Unit 2 & Fire on Diesel Generator Bldg Roof
| title = Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Repts 50-324/88-24 & 50-325/88-24
| author name = Fedrickson P, Levis W, Ruland W
| author name = Verrelli D
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
| addressee name =  
| addressee name = Utley E
| addressee affiliation =  
| addressee affiliation = CAROLINA POWER & LIGHT CO.
| docket = 05000324, 05000325
| docket = 05000324, 05000325
| license number =  
| license number =  
| contact person =  
| contact person =  
| document report number = 50-324-88-24, 50-325-88-24, NUDOCS 8809130173
| document report number = NUDOCS 8811150469
| package number = ML20154A980
| title reference date = 10-13-1988
| document type = INSPECTION REPORT, NRC-GENERATED, INSPECTION REPORT, UTILITY, TEXT-INSPECTION & AUDIT & I&E CIRCULARS
| document type = CORRESPONDENCE-LETTERS, NRC TO UTILITY, OUTGOING CORRESPONDENCE
| page count = 17
| page count = 1
}}
}}


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*  UNITED STATES
[pa es og*'o,
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NUCLEAR REGULATORY COMMISSION 3\ * ^
REGION li h M,  101 M ARIETTA STRE ET, *' s  ATL ANT A, GEORGI A 30323
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Report Nos.: 50-325/88-24 and 50-324/88-24 Licensee: Carolina Power and Light Company P. O. Box 1551 Raleigh, NC 27602 Docket Nos.: 50-325 and 50-324  License Nos.: OPR-71 and DPR-62 Facility Name: Brunswick 1 and 2 Inspection Conducted: July 7 - August 7, 1988 Inspector: _ YY ^ '
W. R. Ruland
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      $lSlYW Da'te Signed f f'
Pe_e -  __ dM//88 W. Levis  ( [''  Da!e Sitned Accompanying Personnel: M. Branch C. Casto L. Garner R. Latta J. Mathis
  < S. Shaeffe Approved by: J'  -
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P. E. Fredrickson, Section Chief  Date Signe f Division of Reactor Projects SUMMARY Scope: This routine safety inspection by the resident inspectors involved the areas of followup on previous enforcement natters, maintenance observation, surveillance observation, operational safety verifica-tion, followup on inspector identified and unresolved items, onsite '
;  Licensee Event Report (LER) review (Unit 1), in office LER review
 
  (Unit 1), Unusual Event / Fire Unit 2, fire on diesel generator
 
building roof, HPCI auxiliary oil pump splice, Residual Heat Removal (RHR) service water (SW) gasket rupture, Automatic Switch Company (ASCO) pressure switch failure, RHR SW temperature limit exceeded, and sustained control room and plant observatio PDR ADOCK 05000324 Q  PDC
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Results: Four violations were identified: failure to properly control work
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approval, resulting in a fire underneath a diesel generator silencer; inadequate corrective action and design control leading to inoperable DC motor operated valves in the High Pressure Coolant Injection
;  (HPCI) system; inadequate corrective action regarding resolution of ,
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silicon bronze bolt cracking; and failure to maintain RHR SW piping temperature (considered licensee identified). No deviations were identifie Continuous onsite NRC coverage (July 18 - August 1,1988) found no significant safety issues regarding operator perfonnanc '
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REPORT DETAILS
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1 Persons Contacted l  Licensee Employees
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)  *K. Altman, Acting Manager - Maintenance W. Biggs, Engineering Supervisor    ,
  *F. Blackmon, Manager - Operations I        '
  *J. Brown, Res. Engineer - Engineering  l
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T. Cantebury, Mechanical Maintenance Supervisor (Unit 1)  !
  *G. Cheatham, Manager - Environmental & Radiation Control  !
R. Creech, !&C/ Electrical Maintenance Supervisor (Unit 2)  i
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  * Dorman, Supervisor - QA    ;
  *K. Enzor, Director - Regulatory Compliance  :
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R. Groover, Manager - Project Construction
  *J. Harness, General Manager - Brunswick Nuclear Project W. Hatcher, Supervisor - Security    i A. Hegler, Superintendent - Operations  '
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R. Helme, Manager - Technical Support J. Holder, Manager - Outages
  *P. Howe, Vice President - Brunswick Nuclear Project  .
L. Jones, Director - Quality Assurance (QA)/ Quality Control (QC)  '
  *M. Jones, Director - On-Site Nuclear Safety - BSEP  I l  R. Kitchen, Mechanical Maintenance Supervisor (Unit 2)  i
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J. Moyer, Manager - Trainin9    r
,  G. Oliver, Manager - Site Planning and Control  l I  *J. O'Sullivan, Project Manager Valves - Projects
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B. Parks, Engineering Supervisor j  *R. Poulk, Senior NRC Regulatory Specialist
<  *J. Smith, Director - Administrative Support  ;
 
R. Starkey, Manager - Nuclear Safety and Environmental Services  l
,  V. Wagoner, Director - IPBS/Long Ran:le Planning  !
{  R. Warden, !&C/ Electrical Maintenance %pervisor (Unit 1)  !'
l  B. Wilson, Engineering Supervisor 1  *A. Worth, Engineering Supervisor    ;
T. Wyllie, Manager - Engineering and Construction  l
 
l  Other licensee employees contacted included construction craftsmen, '
,  engineers, technicians, cperators, office personnel, and security force )
J  members, j  * Attended the exit interview    I l  Note: Acronyms and abbreviations used in the report are listed in
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'l Followup on Previous Enforcement Matters (92702)
Violation 324/86-16-01 Failure to Declare a Support inoperable
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  (CLOSED)
When Required by Procedur The inspector reviewed the licensee's response to the violation dated August 14, 1986. This incident revealed a l,  programatic deficiency in that the previous test data sheets for visual examinations were used only for recording examination results. The decision on whether a deficiency requires an LCO was made solely by the ,
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ISI support coordinator with no independent review. Corrective actions 1  included imediate review of outstanding WR/Jos issued by the ISI group, i l
No other improper determinations of LCO requirements were found. The
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inspector verified, through record review, that the corrective actions implemented by the licensee were completed in accordance with the response,    ,
>      L (CLOSED) Violation 325/87-02-02 and 324/87-02-02, Failure to Adequately Establish Chlorine Monitor Annunciator Procedure. The inspector reviewed l the corrective actions taken in the licensee's response dated April 2 ,
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1987, and found them implemented adequately. The inspector also reviewed 4  OER 2-87-02, Chlorine Detector Failure Requiring Isolation of Control l  Building Ventilation. The OER gave an in-depth analysis of the correction ,
of the chlorine detector drip rate problem, revisions to annunciator '
]  procedures, and other procedure related problem The inspector concurs with the evaluation results and corrective actions take No significant safety matters, violations or deviations were identifie . MaintenanceObservation(62703)    l
 
The inspectors observed maintenance activities, interviewed personnel, and reviewed records to verify that work was conducted in accordance with ,
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approved procedures. Technical Specifications, and applicable industry c
;  codes and standards. The inspectors also verified that: redundant i j  components were operable; administrative controls were followed; tagouts i were adequate; personnel were qualified; correct replacement parts were
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i used; radiological controls were proper; fire protection was adequate; quality control hold points were adequate and observed; adequate i  post-maintenance testing was performed; and independent verification
:  requirements were implemented. The inspectors independently verified that selected equipment was properly returned to servic l  Outstanding work requests were reviewed to ensure that the licensee gave !
I  priority to safety-related maintenanc Numerous maintenance items were l l  reviewed throughout the reporting perio The inspectors '
-  observed / reviewed, in detail, those portions of the following maintenance
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activities:    '
;  88-AUDP1 Unit 1 HPCI Auxiliary 011 Pump Motor Lead Repair i
'      l 88-AULU1 HPCI Auxiliary Oil Pump 011 Changeout Due to High Moisture l
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i 88-AUNH1 Replace Directional Control Valve for CR0 30-15  !
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88-ATW3 Replaced the 2B Heater Drain Pump Motor
 
;  No significant safety matters, violations, or deviations were identifie f SurveillanceObservation(61726)    l
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I  The inspectors observed surveillance testing required by Technical Specifications. Through observation, interviews, and record review, the ;
inspectors verified that: tests conformed to Technical Specification !
requirements; administrative controls were followed; personnel were !
qualified; instrumentation was calibrated; and data was accurate and complete. The inspectors independently verified selected test results and proper return to service of equipmen .
I The inspectors witnessed / reviewed portions of the following test I
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activities-
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1-MST-RBM11M Rod Block Monitor Chai.nel Functional Test  l 2-MST-APRM11M APRM Rod Block Functional Test f
i 2-MS'i-APRM21W APRM 15% Trip and Inoperable Channel Functional Test !
Calibration    >
i 2-MST-APRM27Q APRM 12% Rod Block Channel Functional Test  j 2-MST-IRM11W IRM Channels A, C E & G Functional Test 2-MST-SRM11W SRM Cnannel Functional Test (Setpoint Calibration) l PT-01. RSCS Operability Functional Test  !
I No significant safety matters, violations, or deviations were identifie !
      : Operational Safety Verification (71707)  i I
The inspectors verified that Unit 1 and Unit 2 were operated in compliance ;
with Technical Specifications and other regulatory requirements by direct !
observations of activities, facility tours, discussions with personnel, ;
reviewing of records and independent verification of safety system statu !
The inspectors verified that control room manning requirements of 10 CFR t 50.54 and the Technical Specifications were met. Control operator, shift !
supervisor, clearance, STA, daily and standing instructions, and  !
jumper / bypass logs were reviewed to obtain information concerning j operating trends and out of service safety systems to ensure that there 1 were no conflicts with Technical Spacifications Limiting Conditions for l Operations. Direct observations were conducted of control room panels, !
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instrumentation and recorder traces important to safety to verify
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operability and that operating parameters were within Technical ,
Specification limits. The inspectors observed shift turnovers to verify that continuity of system status was maintained. The inspectors verified the status of selected control e an annunciator t Operability of a selected Engineered Safety Feature division was verified weekly by ensuring that: each accessible valve in the flow path was in its correct position; each power supply and breaker was closed for j components that must activate upon initiation signal; the RHR subsystem
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cross-tie valve for each unit was closed with the power removed from the valve operator; there was no leakage of major components; there was proper ;
lubrication and cooling water available; and a condition did not exist which might prevent fulfillment of the system's functional requirement I instrumentation essential to system actuation or performance was verified operable by observing on scale indication and proper instrument valve lineup, if accessible,    '
The inspectors verified that the licensee's health physics policies /pocedures were followed. This included observation of HP ,
practices and a review of area surveys, radiatie:: work permits, postin l and instrument calibratio ;
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The inspectors verified that: the security organization was properly '
ranned and security personnel were capable of performing their assigned I functions; persons and packages were checked prior to entry into the protected area; vehicles were properly authorized, searched and escorted L within the PA; persons within the PA displayed photo identification i badges; personnel in vital areas were authorized; and effective !
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compensatory measures were employed when require The inspectors also observed plant housekeeping controls, verified l position of certain containment isolation valves, checked various i clearances, and verified the operability of onsite and offsite emergency !
power source No significant safety matters, violations, or deviations were identifie . Followup on Inspector Followup and Unresolved Items (92701)
(OPEN) Inspector Followup Item 325/86-11-04 and 324/86-12-04, Poor Quality RRll Procedures. The inspector reviewed the licensee's current l program for upgrading maintenance instructions and procedures which includes the RRll procedure project. The priority for upgrading has been those procedures utilized in current plant modification Due to the ongoing decrease in plant modification work and consequently the number of i maintenance procedures which are being reviewed and upgraded, the licensee is planning on continuing upgrades on a more scheduled and prioritized ,
basis. This changeover is due to occur in the mid 1989 time frame pending I budget limitations. A considerable amount of maintenance procedures still l
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        }.f l La NOV o 91303  <
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Carolina Power and Light Company ATIN: Mr. E. E. Utley    '
Seni..r Executive Vice President Power Supply and Engineering and Construction P. O. Box 1551 Raleigh, NC 27602 Gentlemen:
SUBJECT: INSPECTION REPORT NOS. 50-325/88-24 AND 50-324/88-24 Thank you for your response of October 13, 1988, to our Notice of Violation  :
issued on September 2,1988, concerning activities conducted at your Brunswick facilit We have evaluated your response and found that it meets the require-ments of 10 CFR 2.20 We will examine the implementation of yoJr corrective actions during future inspection We appreciate your cooperation in this matte [


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Sincerely, L
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David M. Verrelli Chief I keactorProjects$ ranch 1 DivisionofReactorProjects
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cc: R. B. Starkey, Jr. , Manager BrunswickNuclearProject
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! L. Harness, State of North CarolinaPlant General Manager l
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bec: NRC Resident Inspector DRS Technical Assistant
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i  require review and upgrad The current staffing levels involved appear to be adequate to properly review procedures which are utilized in plant modificatio ;
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RII y RII f    l RCartoll:ser  PFr ri son   I 11/$/88  11/$/88   (
 
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The inspector reviewed MP-52, Standards for Pre)aring and Maintaining . l
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Maintenance Procedures, Revision 003, dated Apri' 14, 1988, for accuracy l 1  and scope as compared to recommended criteria expressed in NUREG/CR-1369, !
)  Revision 1. Procedures Evaluation Checklist for Maintenance, Test and  l
<  Calibration Procedures Used in Nuclear Power Plants. No discrepencies  !
!  were note MP-52 now incorporates Procedures Administration Manual :
j'  philosophy for format and content and has been regularly upgraded. The
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i  inspector also interviewed both licensee and contract personnel in regards f i  to the current maintenance procedure upgrade process and future  l
!  expectation This item should remain open pending the development of a j
!  more discernible upgrade schedule and prioritization program for the  ,
I  remaining maintenance procedure upgrade !
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I  (CLOSED) Inspector Followup Item 325/86-17-03 and 324/86-18-03, Review of :
1  Unauthorized Control Room Pressure Boundary Extension. The inspector  ;
reviewed OER 2-86-025, approved on August 14, 1986, which adequately  [
detailed the breakdown in coninunications and work controls which lead to i l
the event. The licensee performed a smoke test on July 10, 1986, which !
l  demonstrated that a positive pressure still existed with both fire doors !
!  lef t open. The inspector concurs that, in terms of control room  t
{'  habitability, the presence of a security guard to close the wired open I door in case there had been any in-leakage, further mitigates the  t I  potential for loss of habitabilit However, the guard was not stationed !
j  at the door for control room habitability concerns. The licensee's  j
,  actions to prevent recurrence included operations shift personnel training, a new plant security open door policy, and the posting of all !
doors and hatches into the control raom with information regarding the !
consequences of leaving doors open and who can authorize this action. The t inspector has no further questions,    t i
j  (OPEN) Inspector Followup Item, 325/86-24-03 and 324/86 25-03, Review of I l Lonegren Relief Valve Test Program. The inspector reviewed the licensee's I j  assessment of the Lonegren Model LCT-11 Emergency Core Cooling System relief valve failures dated October 7,198 The functional testing i  included bench testing and disassembly examinatio The results found j  that nine out of the ten valves tested exhibited unsatisfactory result i The problems, which were generic to all the valves, included severe
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;  corrosion of springs, stems, guides, and steps. In addition, cutting and i pitting were noted on discs and seating surfaces. The licensee also
!  reviewed the consequence of these relief valve failures on plant operating ;
J  procedures and concluded that frequent cycling of the valves via routine l j  bypassing of ECCS keepfill station PCVs may be contributing to many 4  premature failure Certain ECCS keepfill PCV setpoints have been reset I i  in order to avoid needless cycling of both primary systen and keepfill '
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Currently, planned corrective actions include either rebuilding of the i LCT-11 model or, if necessary, r9 placement of the entire valve with Model .
LCT-1 This item will remain open pending completion of the valve (
upgrade which is scheduled for late 198 j i
  (CLOSED) Inspector Followup Item 325/87-31-02 and 324/87-35-02, Work ;
Controls for Fasteners Loosened on Operable Equipment. The inspector ;
reviewed various licensee inter-office memorandums concerning the  :
resolution of the continued instances of non-conformances involving work I controls for loosening fasteners . The subject has received appropriate l management attention and involvement in the issue resolution. The  !
inspector also reviewed Q-List Evaluation No. 88-10, approved on March 22, i 1988, along with the associated work package documentation. The analysis L provided appears to be adequate in determining that cable tray covers are i not seismic or fire protection related, and are not required from a :
structural integrity standpoin Based on these conclusions, the !
licensee's intention is to downgrade the cable tray cover, as a component ,
part, to non Q-List classification and replace the damaged or missing j covers in the course of normal iaodification activity as they are  j discovered. The inspector hd no further question !
t (OPEN) Inspector Followup Item 325/88-21 05, Review of hPCI Door OER. On I July 11, 1988, an additional event occurred regarding a HPCI door. At l approximately 5:00 p.m., an inspector accompanied the Unit 2 shift foreman on tour of the Unit 2 reactor building. The inspector and shift foreman found the north door of the HpCl room wide open, with a large 4* square ;
fan in the doorway. No maintentt.ce personnel were in the area. The !
Maintenance staff had been prforming MAC testing in the area. At the i time the door was found open, the CO system had been rendered inoperable for personnel safety. The shift foreman stated that maintenance personnel l
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were given pennission to open the door for ventilation, provided an !
individual is present to close the door in case of fir Both events !
will be reviewed when the original OER is issue l I
  (CLOSED) Unresolved item 325/88-18 07 and 324/88-18-07, Adequacy of l Actions to Identify and Correct Silicon Bronze Bolt Problem. Upon further :
NRC review of LER 1-88-006 and its supplement, this matter is considered a i violation. Violation: Inadequate Corrective Actions Taken to Identif ;
and Correct Silicon Bronze Bolt Failures (325/88-24-03 and 324/88-24-03)y .
I I  (CLOSED) Unresolved Item 325/88-21-07 and 324/88-21-07 Valve Operability i With Respect to Starting Resisters in DC Notor Control Centers. This issue and related issues regarding DC motors were identified as a  ,
violation in AIT report 325,324/88-2 The violation will be i administratively tracked using this report. Violation: Inadequate Corrective Action for problems Identified in DC Motor Operated Yalves (325/88-24-04 and 324/88-24-04).
 
Two violations and no deviations were identifie _- .  .- .
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! Onsite Review of Licensee Event Reports (92700)
;  The below listed LERs were reviewed to verify that the information
!  provided met NRC reporting requirements. The verification included
)  adequacy of event description and corrective action taken or planned, ,
1  existence of potential generic problems and the relative safety
;  significance of the even Onsite inspections were performed and concluded that necessary corrective actions have been taken in accordance ,
with existing requirements, license conditions and comitment t (CLOSED) LER 1-85-59. Reactor Scram Due to Primary Containment Group 1 ,
Isolation Along with Trip of Reactor Core Isolation Cooling System and i
,  Trip / Lockout of Diesel Generator No. 4. This event was previously inspected in inspection report 50-325/87 31. The licensee has submitted a
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supplemental response, dated June 27, 1988, to the subject LER. The '
. revision exhibited a more clearly defined scope of event investigation, !
I  corrective actions, and reportability concerns. Resolutions to various :
1  technical concerns which my or may not have been related to the root i l cause of the event were analyzed and improvements were made as part of the j  corrective actions taken. The inspector reviewed the completed work i 1  package and verified the associated procedural change !
i No significant safety matters, violations, or deviations were identified.
 
! InOfficeLicenseeEventReportReview(90712)  i f The below listed LER was reviewed to verify that the information provided i  met NRC reporting requirement The verification included adequacy of i  event description and corrective action taken or planned, existence of -
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  (CLOSED) LER 1-88-15 Automatic Isolation of Units 1 and 2 Comon Control l  Building Heating, Ventilating, Air Conditioning System and Emergency Air
 
Fil vation Syste l No significant safety matters, violations, or deviations were identified, i Unusual Event / Fire Unit 2 (93702)  !
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i  The licensee declared an Unusual Event at 11:30 p.m., due to a fire in i
 
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Unit 2 from the 2B Heater Orain Pump motor. The fire existed for greater j than 30 ninutes from initial detection. At 11:15 p.m., the Unit 2 breez j j  way 20' south fire alarm was received in the control room followed shortly l i by the bus 2C motor overload alann. Personnel were dispatched to the !
I  breezeway and the 2C switchboard to inYestigate. A fire was confirmed at i  11:20 p.m. due to the presence of smoke in the breezeway area and the fire l
j  alarm sounding. Motor amperage for HDP 2B was reading 200 amps (140 J  normal value). HDP 2C was started at 11:20 p.m. and HDP 2B was secured.
 
]  The fire brigade assembled and entered the 2B HDP room at 11:28 p.m. and i
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j extinguished the fire in the moto The fire was reported out at  ,
11:48 p.m. and the Unusual Event terminated at 11:49 An NRC inspector was in the control room at the time of the event and reported that operator actions were prudent and in accordance with procedures. The
;  fire was confined to the HDP motor and no release of radioactive materials
;  from the plant occurre The inspectors will review the post fire  ;
investigation report during future routine inspection No significant safety matters, violations, or deviations were identifie !
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10. Fire on Diesel Generator Building Roof (93702)    ,
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The licensee declared an unusual event at 7:54 a.m. on July 24, 1988, due  !
 
to a fire on the DG building roof near the No.1 DG exhaust silencer. At  r 7:42 a.m. the fire was reported to the control room. At 7:44 a.m. the
'1 fire alarm was sounded. The fire brigade mustered at the assembly point at 7:50 a.m. Fire extinguishers were raised to the roof and the firc was  i reported out at 7:59 a.m., at which time the unusual ever.L was terminate !
hRC inspectors were present in the control room and at the assembly poin (
and noted that the licensee's response to the event was controlled, well  I executed, and in accordance with established procedure .
The fire occurred in soee 8 X 8 timbers used to support the west end of  !
the DG No. 1 exhaust silencer The timbers ignited due to the performance  i of PT-12.2A, the monthly DG load test, which had been performed for the Nu. 1 DG earlier that norning. The timbers were placed to support removal  l and replacement of the exhaust silencers, which is scheduled for the week  l of September 17. Installation of the wood support structure, which was  [
installed on all four diesel generators, was completed on July 22, 198 '
The work was authorized by a Plant Services Authorization form, which is  I Appendix H to MP-14A, Corrective Maintenance, included with this form was  l a Fire Protection Engineering Review fonn Attachment 2 to FPP-014 This form, completed on July 8,1988, stipulated that fire retardant sheeting  ;
be placed over the cribbing and that a fire extinguisher be present at the  j work site. This stipulation was made since non-fire retardant wood was  !
used in this application due to the unavailability of treated wood in the  i 8 X 8 size. Attachment 2 of FPP-014 has provisions for a fire inspector's  t signature to ensure that the appropriate stipulations were satisfie i This block was not signe The licensee's initial review of the event determined the following:
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The timbers should not have been in place unless the diesels were  i under clearanc In fact, it was the understanding of the individual  {
completing Attachment 2 that the structure would not be put in place  (
until the diesels were tagged out. The operating state of the  :
diesel, however, was not specified as a prerequisite to putting the  I structure in plac l
 
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   * The lack of independent review / controls to ensure that personnel complied with the stipulated conditions for use of the wood ,
 
contributed to the even * The magnitude and impact of the fire were within the liinits
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l  previously analyzed by fire protection engineerin I l  This event in itself does not appear to be safety significant. However, i  the lack of controls in perfoming maintenani:e on safety related
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structures and components is a weakness that must be addresse The use y
of a Plant Services Authorization form as the controlling document for this work is not appropriate. Appendix ! to MP-14A defines plant services
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work for that which control by WR/JO is not required. in addition, Appendix ! provides examples of work items which do not require review by :
the shif t foreman. The installation of non-fire retardant wood support
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I  structures in direct contact with Q structures is not exempted from WR/JO !
:  control or SF approval as specified in Appendix ! and J. Therefore, a !
WR/JO should have been used and SF approval obtained prior to performing !
j  this wor Criterian V to Appendix B of 10 CFR 50 requires that I j  activities affecting quality shall be prescribed by documentad !
instructions or procedures of a type appropriate to the circumstance !
The lack of controls in place during the installation of the wood support structure is in violation of this requirement. Violation: Fire on Diesel !
Generator Building Roof (325/88-24-01 and 324/88-24-01). ;
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No significant safety matters, one violation and no deviations were [
identified.
 
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11. HPCIAuxiliaryOilPumpSplice(25576)  j
,  As the result of a previous EQ violation (325,324/88-21-02) cealing witn [
SBGT SCR controllers, the licensee began to re-evaluate the qualification !
status of their skid mounted equipment; HPCI and SBGT in particular. This l
 
review, which was performed by BESU and completed on June 6,1988,
!  identified several possible EQ concerns. One item in particular was that
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the HfCI auxiliary oil pump motor splice connection may not be qualifie ,
As a result of this deficiency, NCR A-88-016 was issued and the splices :
!  were examined on July 15, 1988, and July 24, 1988, for Unit 1 and Unit 2, }
J  respectively. The inspection results are documented in EER 88-0349 and '
!  EER 88-0371. The licensee concluded that the splices in Unit 1, s,hich i consisted of one in line tape splice, five parallel tape splices, and one
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i l  three wire tape splice were acceptable based on tM quality of splice construction, the observible characteristics of the splico in relation to .
4  those qualified in existing qualification data packages, and the l
{  environment in which it must function (ie., harsh for radiation only), i j  The licensee, however, replaced these splices en July 30, 1988, with a !
j  type covered by their present Qualification riata package '
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i I  The results of the inspection of the Unit 2 splice revealed that the a
splices were unacceptable. These splices also consisted of one in line, i
!  five parallel, and one three wire tape splice. However, due to the poor l
!  workmanship noted, lack of outer jacketing tape and three cases where i holes were observed through the insulating tape, the licensee concluded i l  that qualification was indeterminate for these splice Operations was i i  infonned of the situation and HPC1 was declared inoperable although it was l l  not required to be operabie at that time since the reactor was in mode 3 j
!  with pressure less than 113 psig. The splices were replaced with  ,
qualified splicos prior to Unit 2 resuming power operation l l  The lack of a qualified splice on the HPCI Unit 2 auxiliary oil pump is a j  violation of 10 CFR 50.49 requirements. However. since the violation was I
;  ider.tified as a result of licensee corrective actions to a previous (
:  violation (325,324/88-21-07), no Notice of Violation will be issued.
 
I
!  No significant safety matters, one violation and no deviations were j  identified.
 
l 12. RHR Service Water Gasket Rupture (63702)
l  On July 17, 1988, with Unit 1 in mode 3 and shutdown cooling established i  on the A RHR loop, the licensee experienced a gasket failure on the RHR service water line. The failure occurred at a 4" cleanout connection
{
upstream of flow control valve E11-F068A. The failure occurred at a copper nickel raised face blind flange. The gasket used was an unreinforced EPDM gasket. When the leak was reported, shutdown cooling was secured and damage from spraying water assessed. Water accumulated in l  the north core spray room and had wetted down several components, including a lighting distribution panel, MCC IXJ. and valve 1-E21-F015A, Affected components were examined, necessary repairs made, and items
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restored to operable statu The failed gasket was replaced with a ,
reinforced type. To verify that other gaskets in the RHR service water l  system were operable, the licensee perfortned a pressure test on the (,
)
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syste Flow was reduced in accordance with OP 17 such that system '
!  discharge pressure was approximately 400 lbs. Other gaskets were walked ,
1  down and no other leaks were foun l 1      !
i  The licensee has experienced other gasket failures on the Unit 2 service (
)  water piping on April 30, 1986, and December 24, 1987. The failures were l
]  evaluated nnd determined to be attributable to three factors. These  e i  facters were unreinforced EPOM gaskets, raised face flange joint, and l carbon steel piping materials. As a result of these failures, the  l 1l  licensee compiled a libt of such high risk joints in both units. During 1  the last refueling outage of Unit 2, 22 of the 55 joints on the discharge side of the RHR SW purps were replaced. Prio.' to the latest failure on
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Unit 1, 20 of the 51 high risk joints were scheduled for replacement during the next refueling outage. $1nce the latest failure was a joint j  not previously considered to be high risk, the licensee is re evaluating
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their position to determine if other gaskets should be replaced.
 
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,  This item will undergo further review along with the previous unresolved t
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item (324/87-43-06) which was written after the December 24, 1987 failur l One potentially significant issue involving ruptured gaskets was j
identified; no violations, or deviations were identifie l i
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13. ASCO Pressure Switch Failure (62703)
)  On July 25, 1988, while attempting to establish shutdown cooling on Unit
      *
I  2 the control room operator was unable to open valve 2-E11 F009 from the I  control room. This valve is the inboard isolation valve on the common RHR
!  shutdown cooling suction line. It is off recirculation loop A and it is !
1  not paralleled by a redundant line. Therefore, the opening of this valve t i  is essential in the removal of core decay heat from the reactor during the
!  shutdown cooling trode of operation. However, at the time of this event, '
i  the condenser was available and the reactor had sufficient steam pressure j  to remove decay heat through that path. Additione.lly, the drywell was  1 accessible and the F009 valve could have been manually opene l
      !
j  Trouble pressureshooting (ender WR/J0-88-AUFA1 switch 0 832-PS-N018A-1) contacts detemined were open, that a relatedthe not allowing interlock control switch to open the F009 valve from the control roo This l  particular switch is designed as part of the control logic to allow the
;  valve to open only when reactor pressure is less than 140 psig to protect
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the low pressure RHR system. The particular switch was equipped with a i  dual set of microswitch controls. The licensee wrote an engineering  .
;  evaluation (EER 88-0369) to allow a temporary repair of the problem by  !
j  using the spare switch.
 
.
l  As a result of this failure, combined with other recent setpoint drif t .
,  problems associated with ASCO tripoint pressure switches being used in  !
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high pressure applications where a low setpoint ($ required, the hRc  ;
requested the licensee to review all other applications of this type of '
!  pressure ; witch. To address these concerns, EER 88-0376 was developed by i l  engineering and reviewed by the PNSC. This evaluation provided the  l j  following information:    '
  *
)  Review of procurement specification BSEP 252-091 indicated that  !
l correct pressure and setpoint range were specifie ;
3        l
)
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Review of purchase order B26127 indicated that ASCO certified  '
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compliance to the above specificatio l
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Field verification of all installed ASCO tripcint pressure switches J  indicated that label plate data showed that rated over-range pressure was within system design pressur I  *
1  The failure of pressure switch 2-B32-PS-h018A-1 was a random failure j  and not indicative of a generic proble i i
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A failure analysis of pressure switch 2-832 PS-N018A-1 would be i
!  perfomed as soon as the switch can be removed (presently scheduled for early September).
 
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Based on the above, the pressure switches app +ar corru t for this applica- ,
tion. However, to improve reliability the licensee is workiy with ASCO *
!
on the instrurent drif t proble The licensee indicates that their
!  current calibration frequency of once per nonth is sufficient to ensure I operabilit During a plant safety meeting where this issue was j discussed, a member indicated that Yankee Rowe had recently istved a 10 CFR Part 21 report on problems similar to those being experienced at BS5P.
 
]  A review of +.he Part 21 report and followup of the licensee failure ;
1  analysis is identified as an inspector Folicwup Item: ASCO Pressure ;
j  Switch railure (325, 324/88-24-02),   !
No significant safety matters, violations oc deviations were identifie ,
14. RHR SW Temperature Limit Exeteded (71707)  j t
!  During review of the data associated wita the performance of the 1A RHR ;
i heat exchanger performance test conducted on July 17, 1988, and pressure j l testing of gaskets subsequent to the repair of a failed gasket, the
{  licensee detemined that they had exceeded a design limit of 120 degrees F j  on the RHR SW piping dcwnstream of the RHR heat exchange The chart i i  recorder which monitors this parameter showed that before stabilizing, the l  RHR service water temperature downstream Of the heat en hanger pealed at
]  215 degrees F when initially putting the heat exchanger in servicat. The !
j  120 degrees F limit was based on the current stress analysis for the RHR }
i  SW piping downstream of the RHR heat exchange To detennite if this i
 
piping was still operable, the licensee prepared EER 88 0365 and  ;
j  re-evaluated the stress model for this piping tssu ,ing a temperature of i
:  215 degrees The analysis considered thenral and dead weight loads and j  shewed ; hat system design stresses bad nM been exceede Further (
j analysis will be done to detemine if the piping would have remained
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operable at 215 degrees F during design ba$1s earthquake conditions to i  detemine reportability of the even Thi$ item 15 due by October 1 ;
J  198 i I
, Te allow for continued operation, the licensee justified short term  t stismic qualification of the RHR SW piping downs' ream of the RHR heat l
: exchanger with a limit of 186 degrees F. The operating procedure for RHR, I i OP-17, was revised for tJnit 2 in Revision 27 (cated July 2.?,1988) to '
i reflect the 186 degrees F limit and inposed &n operating linit of 170 !
) degrees F. The temperature limit for long tera qualWication is due for I complet!cn by November 1,1988.
 
]      [
Failure to maintain terperature below 420 degroes F on RHR $W piping is a
! violation of 10 CFR 50, Appendix B, Criterion V. The operatir.g procedure l j  was inadequate to keep the temperature from exceedirg 120 tegrces !
j  However, since all the requirements of 10 CFR 2. Appendix C, were  i l      '
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i l  satisfied the violation is considered licensee identified and no notice i
,  of violation is being issue LIV: Failure to maintain RHR SW piping >
temperature below 120 degrees The inspectors will follow the
', licensee's corrective actions during future routine inspections (325/
4  88-2405and324/88-24-05).
 
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No significant safety matters, one licensee identified violation, and no l 4  deviations were identified.
 
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j 15. Sustained Control Room and Plant Observatton (71715)  ;
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      ;
I  NRC began 24 hour coverage of the licensee's activities on July 18, 198 l Region !! initiated the additional oversight because the liebnsta's ,
1  continued problems with failure of plant equipment and questions ;
j concerning the supporting management system Continuous coverage ended !
j on August 1,1988, when Unit 2 reached 1001 power, NRC performed ;
j  additional inspections during this tine as part of the AIT, report h :
j  325,324/88-2 l 1 Major events and procedures reviewed by the inspectors are included !
l  throughout this report. No major findings resulted from the continuous ;
j  coverag The coverage included extended control room observation, !
i a*.tendance at licensee management turnover reetings, and observation of f j  surveillance tests and maintenance activitie j
)  No significant safety matters, violations or deviations were identifie l i
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16. Exit Interview (30703)    !
I l  The inspection scope and fitidings were sumari7ed on August 5,1988, with l l  (hose persons indicated in paragraph 1. The inspectors dr. scribed the !
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areas inspected and discussed in detail the inspection finJings listed l
,  belo Dissenting coments were not received from the licente, t Proprietary information is not contained in this report,  i l
)
 
Item hmeer  Description / Reference Paragraph
 
325, 324/88-24-01 VIOLATION - Fire on Diesel Generator Building
!  Roof (paragraph 10).
 
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)  325, 324/68 24-03 VIOLATION - Inadequate Corrective Actions Taken !
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to Identify and Correct Silicon Bronn Bolt t l  Failures (paragraph 6).  !
      !
1  325, 324/E8-24<04 Inadequate Corrective Action for Problers !
j  Identified in DC Motor Operated Valves (paragraph
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j  6). l
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325, 324/88-24 02 IFl - ASCO Pressure Switch Failure (paragraph (  13).    }
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325, 324/38-24-05 LIV - Fallere to Maintain RHR SW Piping Tempera- l ture Below 120 Degrees F (paragraph 14).
 
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17. List of Abbreviations for Un:t ad 2 ,
AIT Augmented Inspection T A0 Auxiliary Operator APRM Average Power Range Monitor ASCO Automatic Switch Company BESU Brunswick Engineering Sub 'Jnit BSEP Brunswick Steam Electric Plant CO Carbon Dioxide i.,P$L Carolina Power & Light Company CRD Control Rod Drive CWIP Circulating Water Intake Pump DC Direct Current DG Diesel Generator EER Engineering Evaluation Report EPDM Ethylene Propylene Dipolymer EQ Environmental Qualitnation ESF Engineered Safety Feature F Degrees Fahrenheit HDP Heater Drain Pump HP Health Physics HPCI High Pressure Coolant Injection HSD Hot Shutdown I&C Instrumentation and Control IE NRC Office of Inspection and Enforcement IFI Inspector Followup Item IPBS Integrated Planning Budget System IRM Intermediate Range Monitor ISI Inservice Inspection JC0 Justification for Continued Operation LC0 Limiting Condition for Operation LER Licensee Event Report LIV Licensee Identified Violation MAC Motor Actuator Characterizer MCC Motor Control Center MP Maintenance Procedure NCR Non-Conformance Repott
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NOUE Notice of Unusual Event NRC Nuclear Regulatory Comission NUREG Nucl% r Regulation OER Operating Experience Report OP Operating Procedure PA Protected Area PAM Procedures Administration Manual PCV Pressure Control Valve PNSC Plant Nuclear Safety Committee PT Periodic Test Q Quality QA Quality Assurance QC Quality Control
>
 
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RCIC Reactor Core Isolation Cooling RHR Residual Heat Removal RRIL Regulatory Related Instrument List RSCS Rod Sequence Control System RWM Rod Worth Minimizer SDGT Standby Gas Treatment SCR Silicon Controlled Rectifier S/D Shutdown SF Shift Foreman SRM Source Range Monitor STA Shift Technical Advisor SW Sere:ce Water TS Technical Specification URI Unresolved Item WR Work Request WR/JO Work Request / Job Order l
l l
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Latest revision as of 14:28, 6 December 2021

Ack Receipt of Informing NRC of Steps Taken to Correct Violations Noted in Insp Repts 50-324/88-24 & 50-325/88-24
ML20206B502
Person / Time
Site: Brunswick  Duke Energy icon.png
Issue date: 11/09/1988
From: Verrelli D
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To: Utley E
CAROLINA POWER & LIGHT CO.
References
NUDOCS 8811150469
Download: ML20206B502 (1)
v • d • e


Text

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}.f l La NOV o 91303 <

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Carolina Power and Light Company ATIN: Mr. E. E. Utley '

Seni..r Executive Vice President Power Supply and Engineering and Construction P. O. Box 1551 Raleigh, NC 27602 Gentlemen:

SUBJECT: INSPECTION REPORT NOS. 50-325/88-24 AND 50-324/88-24 Thank you for your response of October 13, 1988, to our Notice of Violation  :

issued on September 2,1988, concerning activities conducted at your Brunswick facilit We have evaluated your response and found that it meets the require-ments of 10 CFR 2.20 We will examine the implementation of yoJr corrective actions during future inspection We appreciate your cooperation in this matte [

Sincerely, L

David M. Verrelli Chief I keactorProjects$ ranch 1 DivisionofReactorProjects

!-

cc: R. B. Starkey, Jr. , Manager BrunswickNuclearProject

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! L. Harness, State of North CarolinaPlant General Manager l

bec: NRC Resident Inspector DRS Technical Assistant

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Document Control Desk l

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RII y RII f l RCartoll:ser PFr ri son I 11/$/88 11/$/88 (

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