IR 05000219/1988038: Difference between revisions

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{{Adams
{{Adams
| number = ML20245H968
| number = ML20247N416
| issue date = 02/21/1989
| issue date = 07/24/1989
| title = Insp Rept 50-219/88-38 on 881204-890114.No Violations Noted. Unresolved Items Observed.Major Areas Inspected:Maint & Surveillance Activities,Radiation Control & Physical Security,Including Security Events Re Unauthorized Entry
| title = Discusses Safety Insp Rept 50-219/88-38 on 881204-890114 & Forwards Notice of Violation.Violation Classified as Severity Level IV Since Unauthorized Entry Occurred as Result of Isolated Lapse in Vigilance by Security Officer
| author name = Cowgill C
| author name = Kane W
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
| author affiliation = NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
| addressee name =  
| addressee name = Fitzpatrick E
| addressee affiliation =  
| addressee affiliation = GENERAL PUBLIC UTILITIES CORP.
| docket = 05000219
| docket = 05000219
| license number =  
| license number =  
| contact person =  
| contact person =  
| document report number = 50-219-88-38, NUDOCS 8903060109
| document report number = EA-89-108, NUDOCS 8908020393
| package number = ML20245H961
| package number = ML20247N421
| 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 = 27
| page count = 3
}}
}}


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    - 2-    1 JUL .2 4 1989 b -Docket'No'. 50-219    DESIGNATED ORICINAL
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.d U.S. NUCLEAR REGULATORY COMMISSION-


==REGION I==
9 ] 08  Cortified  />ld- Ab TGPU' Nuclear Corporation
Report N /88-38-
.-  Docket N (
License N DPR-16 Priority -- Category C Licensee:  GPU Nuclear Corporation 1_._ Upper pond Road Parsippany, New Jersey 07054
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Facility Name: 0yster Creek Nuclear Generating Station Inspection Conducted: December 4,1988' - January- ~14, 1989 Participating Inspectors: W. Baunack E. Collins D. Lew -
' ATTN: Mr.. Eugene E. Fitzpatrick
J. Wechselberger Approved By:  $  h C. Cowgi41, Q , Reactor Projects Section 1A Date Inspection Summary:
   .Vice President and Director Oyster Creek Nuclear. Generating' Station     -i
Areas Inspected: Routine inspections were conducted by the resident and one region-based inspector (294 hours) of activities in progress during the outage, including
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maintenance and surveillance activities, radiation control, and physical securit In addition, inspectors reviewed two security events involving a potential un-
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authori2ed entry in a vital barrier and loss of security electrical loads, control room loose wires, scram discharge volume foreign material and cleanliness flushing practices, air accumulator testing, torus tiedown bolt material condition, CRD return line piping wall thickness, and refueling operations. The inspectors also conducted a review of the preliminary safety concern process and several particular safety' concerns. Initial review of the operability status of the source range monitors prior to refueling was also conducted. An LER review was conducted to assess any identifiable trends or weaknesse Results: The inspectors identified five unresolved items. The unresolved items included concerns regarding the single failure susceptibility of the standby gas treatment system automatic initiation logics, vital area unauthorized entry, de-letion of SDV flushing requirements and programmatic implementation of cleanliness
  . standards for piping system, completeness of testing of air accumulators associated with safety related valves, and the quality control process used to verify the
,  adequacy of shop spare parts prior to installation. The inspectors developed con-cerns regarding the proper and time'.y processing of preliminary safety concerns L
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identified by employees. Valid safety concerns actually remain open when the tracking system indicates their completed disposition. An error-free refueling of the reactor was completed during the period.
 
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8903060109 890223 PDR ADOCK 05000219
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    . TABLE OF CONTENTS-
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  ' D rywe l l En t ry ( 717 0 7 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ,. . . . . 1 P#~
2.0 Control l Room Loo se Wi re s ( 93702) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 I3.0 I_RM S p i ki n g . P ro b l e'm s . ( 9 3 7 0 2 ) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .
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4.0 - Preliminary Safety Concern Review (71707) . . . . . . . . . . . . . . . . . . . . . . . . . . . " 4  -
  ' 5' 0 Standby. Gas Treatment (SGTS) Logi c (71707) . . . . . . . . . . . . . . . . . . . . . . . . . . .
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5.1 ; Item..................................... ......... -............ 5-5.2' Review.............. ........................................... 6'
,o  5.3 D i s c u s s i o n . . . . . . . . . . . . . . . . . . . . . . .. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6-5.4 Integrated Plant Safety Assessment, Systematic Evaluation     4 Program-(SEP).................  ................................ 7 5.5 Conc 1usicn...................................................... 7-
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6.0 Rag Found in Scram.01schargeLVolume-(93702).......................... 8 6.1- Event.........................  ................................. 8
  ' 6. 2 ' Revi ew of Li cen see Ac ti on s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 8 7.0 . Air Accumulator Testing  (71707)...................................... 10 8.0 Monthly Maintenance Observation  (62/03).............................. 12
 
9.0 Monthly Surveillance Observation (61726).... ........................ 12 9.1 Source Range Monitor Front Panel Test, Procedure 620.4.004...... 12 9.2 Reactor Triple Low Water Level Test and Calibration Procedure 619.3.006.....................................................    . 13 10.0 Backshift Inspection................................................. 13 11. 0 Reacto r Ve s sel Lo st Part s ( 71707) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 12. 0 To ru s T i edown Bol t s ( 71707) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 13. 0 Fuel Suppo rt Ca s ti ng s ( 71707) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 14 14 0 Control Rod Drive Return Line Piping (71707) . . . . . . . . . . . . . . . . . . . . . . . . . 15 15.0' Loss of Security Electrical Loads (81700) . . . . . . . . . . . . . . . . . . . . . . . . . . . 15 16.0 Source Range Monitor Log Integrator Cards (71707) . . . . . . . . . . . . . . . . . . . lo i
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P. 0. Box-388-      '
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Forked River, New Jersey. 08731
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  , 17l.0 Review of Periodic and Special Reports'. (71707). . . . . . . . . . . . . . . . . . . . . . . >17 i:
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  ,18.0 Meetings.............................................................. :18'
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18.1 GeneraliReview Board Meeting....................................  .18:
18. 2 . Qua l i ty Ma intena'nce Team Bri e fi ng. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ' 18 L 19.0 Radiation Protection  (71707).......................................... 18 20.0 Observation'of Physica1' Security (71707).....'.....................l...    '18 eff '-
21.0 Review of ' Licensee Event Reports (LERs)-(90712,L 92700)................    '18 -
22.0 Refueling-Activities................................................. '24
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123.0 Unresolved Items....................................................... 24-24.0 Exit: Meeting..............................  ........................... 24:
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Gentlemen:
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Subject: NOTICE OF VIOLATION (Inspection No. 50-219/88-38)    l This refers to the routine safety inspection conducted between December 4,  !
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1988 and January 14, 1989 at the Oyster Creek Nuclear Generating Section,  '
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Forked River, New Jersey. During the inspection, the NRC reviewed the circumstances ' surrounding an unauthorized entry into a vital area which
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occurred on' January 3,.1989'and which was identified by your staff and reported to'the NRC. The= inspection report, which was sent to you on February 28, 1989,-  '
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indicated that'this-unauthorized entry.was classified as an unresolved item,   i
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The: unauthorized entry occurred when a contract radiological controls  j technician entered a vital area, undetected, after his vital area access  '
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  ' authorization'had been' terminated. Although it was subsequently determined
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_that the technician's access had been erroneously terminated because of an   >
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  - administrative error, this entry' represents a failure to maintain positive access. control to a vital area and constitutes a violation of NRC physical
DETAILS-
' security requirements. The' violation is described in the enclosed Notice of Violatio 'Although this violation could be classified at-Severity Level III in accordance
  :c ~ Drywell Entry
  .with Section C.2 of Supplement III of the " General Statement'of Policy and
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THIS PAGE CONTAINS AND IT NOT FOR PUBLIC DISCLOSUR IT IS INTENTIONALLY LEFT BLAN ,
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Procedure for NRC Enforcement Action," 10 CFR Part 2, Appendix C, 53 Fed. Re (October 13,1988) (Enforcement Policy), the violation has been classified at Severity level IV because: (1) the unauthorized entry occurced as a result of an isolated lapse in vigilance by a security officer, concurrent with a plant worker's inadvertent failure to follow a security procedure; (2) the
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security officer identified the unauthorized entry when the technician exited fromlthe area, and promptly reported it to security management, thereby demonstrating his integrity; and (3) the technician was previously authorized to enter the area, had entered the area on numerous occasions, and his authori-zation would have been continued except for an administrative oversight. The
  ' NRC recognizes that your corrective actions, upon notification of the entry, were prompt and comprehensive. Nonetheless, we are concerned that sufficient measures were not established to preclude such an event from occurrin Therefore, any similar violation in the future may result in escalated enforcement actio OFFICIAL RECORD COPY PROP EA GEN PUBLIC - 0003. PDR  ADOCK 05000219 Q  PDC


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      ' GPU Nuclear Corporation
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   ,     -2-You are required to respond to this letter and the enclosed Notice, and in preparing your response, you should follow the instructions specified in the enclosed Notice. In your response, you should document the specific actions taken and any additional actions you plan to prevent recurrence. These actions should include strengthening your General Employee Training program by emphasizing the importance of all workers to follow security procedure Further, you should place all Safeguards Information (as defined in 10 CFR 73.21) in an enclosure, so as to allow your letter, without the Safeguards Information, to be placed in the Public Document Room. After reviewing your response to this Notice, including your proposed corrective actions, the NRC will determine whether further NRC enforcement action is necessary to ensure compliance with NRC regulatory requirement Tne enclosed Notice contains details of your security program that have been determined to be exempt from public disclosure in accordance with 10 CFR 73.21 (Safeguards Information). Therefore, those portions will not be placed in the NRC Public Document Room and will receive limited distribution. This letter will be placed in the Public Document Room in accordance with 2.790(a).
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THIS PAGE CONTAINS AND IS NOT FOR PUBLIC DISCLOSUR IT IS INTENTIONALLY LEFT BLAN !
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l-This item will remain an unresolved item pending further inspector review (50-219/88-38-01).


2.0 Control Room Loose Wires l . During the report period the licensee discovered loose wires behind control i  room panels 5F, 6F, and .11R and in panel ER-18 in the 480 volt room. The in-
The response directed by this letter and the enclosed Notice is not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-51
!. spector discussed with the licensee their plans to verify electrical termina-tions prior to plant restart as a result of the extensive maintenance activity conducted in. safety related control panels'. The licensee elected to perform an engineering evaluation to assess the proper method to verify control panel electrical terminations. Inspection Report 50-219/87-04 discusses similar loose connections discovered by the licensee and corrective action taken to resolve the loose wire The loose wires found in the control room are normally terminated in a com-
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pression terminal stri The wires are not lugged and landed, but simply-placed under a spring steel clip with compression force supplied by a set screw. In some cases, two different gauge wires are placed under the same
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compression clip which, with vibration or force exerted on the wire wrap bundle may loosen or pull the wire free from"the terminal strip and break.the electrical connection. The different gauge wire may promote loose termina-tions by allowing the smaller gauge wire to pull free as the larger gauge wire
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absorbs the compression clip force. After the previous loose wires event, the licensee implemented controls and performed extensive inspections to en-sure loose wires would not be a problem during plant operatio The loose wire in panel ER-18 resulted from an HFA relay replacement in the panel when the electrical lead was not relanded following the relay replace-men After discovering the loose electrical connection in panel 11R the licensee elected to reland the wire in the compression terminal strip. Prior to ac-complishing this task the licensee performed an engineering evaluation to determine what the plant response would be to relanding the loose wire. When making the determination the plant response was not as expected. In reviewing this event the licensee determined that the unexpected plant response resulted from personnel error of the plant engineers who reviewed the electrical ter- i mination drawing The inspector examined microfiche electrical termination drawings for panel 11R and found the prints to be of poor quality and diffi-cult to read and were marked "best available". The inspector will review the licensee critique on this event and plans for inspection of electrical ter-minations prior to plant restar .0 IRM Spiking problems During this report period the facility experienced a number of IRM spike On 12/31/88 and again on 1/2/89 full scram signals occurred with the plant in a cold shutdown condition. The licensee could not detect any maintenance or surveillance activities that could have caused the IRM spikes to produce


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Sincerely, William F. Kane, Director Division of Reactor Projects Enclosure:
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Notice of Violation [Contains Safeguards Information (SGI)]*
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cc w/enci (w/o SGI):
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M Laggart, BWR Licensing Manager Public Document Room (PDR)       1 Local Public Document Room (LPDR)       i Nrlear Safety Information Center (NSIC)
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NRC Resident Inspector (w/SGI) *
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State of New Jersey
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       *Contains Safeguards Information l
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l    a full scram. On 1/6/89 the inspectors observed numerous half scrams occur-
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ring from IRM 12 spikes. The spiking problems have been observed on IRM's 11, 12, 13, 15, and 17. The IRM instrument problem has. historically been a problem without resolution. The licensee plans to perform troubleshooting efforts to ensure the IRM's are functional prior to restart from the outag . The inspector will follow licensee activitie .0 Preliminary Safety Concern Review l    During this report period, the inspectors visited the corporate office in Parsippany, NJ to review preliminary safety concerns (PSC) and the PSC proces As a result of this review the inspector expressed concern regarding the length of time to address PSC's, in man,y cases this was greater than one year and involved legitimate safety problems. The licensee committed in a letter (dated 12/3/86) to Region I in response to Inspection Report 50-219/86-24 '
    (dated 11/4/86), to resolve PSC's in a timely fashion. During the inspector's review the new proposed revision to the preliminary safety concerns procedure entitled, " Management of Potential Safety Concerns," 1000-ADM-7330.01 written to incorporate the commitments reflected in the licensee's 12/3/86 letter was examine The proposed revision did not reflect the licensee's commitment to incorporate time constraints in the processing of PSC's to ensure timely resolution of safety concern The second concern the inspectors expressed involved the closure of PSC's with outstanding action items left unresolve The licensee has in the past closed PSC's on the basis of issuing other action items to be resolved such as Tech-nical Function Work Requests (TFWR) or licensing action items (LAI), for ex-ample. The inspectors discovered two examples of this concern in the limited review they conducted of the PSC's. PSC 84-018, reported 10/10/84, addressed a concern regarding a single failure susceptibility of a power supply dis-abling the standby gas treatment system (SBGTS) logics and control power (see paragraph 5.0) and that could render the automatic initiation of SBGTS in-operable. (In addition, operation of the system in a manual mode may be limited and warrants further review.) A TFWR was issued on 1/28/85 to resolve this issue and closed in 10/87 based on the evaluation results of the PSC which concluded that although it was a valid single failure susceptibility, it was in accordance with the plant design basis and the SEP (see paragraph 5.4). It appears that no real evaluation of the safety concern was carried out under the TFW The TFWR simply relied on the evaluation in the PSC which in turn was closed and the action item turned into an apparently untracked TFW Thus, the PSC transmitted the concern to a TFWR which was closed based on th'e PSC evaluation. In addition, this item took three years to close and may have overlooked a valid safety problem (see paragraph 5.0).
 
The second example of a safety concern that remained open and untracked with the PSC considered closed was PSC 86-20 (reported 10/14/86). This PSC ex-presses a safety concern regarding nonfunctioning of the containment spray system automatic initiation logic after a single failure of the 125 VDC power pane LER 86-23 addresses this concer The licensee issued TFWR No. A01692 and Topical Report 038 to justify eliminating the automatic initiation of the
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containment spray system. During this report period the inspector was unable to determine. the status of the TFWR, but was later informed by the. licensee
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that elimination of the automatic initiation of containment spray system was
;  being re-evaluated. In addition PSC-87-003 was written to address another concern with containment ' spray system -logics. This concern is developed in Inspection Report 50-219/87-04. The. concern involves containment spray system-automatic initiation logics in the dynamic test mode.' In;this mode,' coupled with a Lo-Lo reactor water level signal, the drywell- may be-inadvertently sprayed. , Also, another concern involves the dynamic test mode used to
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facilitate torus pool coolin In an accident situati.on, the ability.to. cool the torus may be prevented until the ~ logics may be overridden.(see Inspection
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Report 50-219/87-04).
 
The licensee resolved this PSC based on the evaluation performed 'for PSC 86-20 above. Essentially, this item remains open. The licensee has^ initiated a method to status open PSC's which is r good initiative and currently carries only three open'PSC's. This does not include the PSC's mentioned above, though. The status of PSC's receives wide distribution within the corporation including the GORB, but' does not include those PSC's which may have open issues as indicated abov During this review the inspectors noted that some. individuals perceived that PSCs would not be resolved in a timely manner nor were welcomed as a mechanism for resolution of potential safety problems.. Further, some individuals who could be expected to submit a PSC were not knowledgeable of.the proces Based on the above review, the inspectors expressed concern that the process is not working as described, in that concerns are sometimes not resolved in
  'a timely manner, and that some PSCs were closed without appropriate reviews being complete. Additionally, some site staff members appear to have lost confidence in the proces The inspectors plan further review of the PSC process and specific PSC'. .0 Standby Gas Treatment System (SGTS) Logic 5.1 Item As.a result of the Preliminary Safety Concern (PSC) review discussed in paragraph 4.0, inspectors identified one PSC initiated in 1984, #84-018, which addressed, in part, the potential inability of the SGTS initiation logic to meet single failure criteria. Tha specific concern was that upon opening or loss of breaker #20 from Vital AC Panel (VACP)-1, the automatic initiation logic of the SGTS would be disable The PSC evaluation concluded that while the SGTS initiation logic does not strictly meet single failure criteria, this configuration was in accordance with original plant desig It also referred to the Inte-grated Plant' Safety Assessment, Systematic Evaluation Program (SEP) re-view for Oyster Creek, NUREG-0822, evaluation of the loss of VACP- As a result of the PSC evaluation, a Technical Functions Work Request
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4      6 (TFWR) was initiated to review possible design changes. This was sub-J. -
sequently closed evaluating the changes as not required. No modifica-tions were initiated to address this concer ..
  .5.2 . Review
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Inspectors reviewed the electrical configuration of VACP-1. This 120-
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vac panel is supplied from either of two safety related 480 vac uni substations through an automatic transfer switch. Power then.goes through a disconnect switch and a 480/120 v transformer to VACP-1. From review of this configuration, it can be seen that there are several single failures which.could interrupt power to VACP-1, and thus to th ,  SGTS automatic initiation logi Inspectors reviewed the electrical elementary drawing.to determine the ef fects of'a loss of VACP-1-(breaker #20) on the SGTS. Since the SGTS initiation' relays energize on an automatic initiation signal, and the SGTS logic-is energized by VACP-1, the automatic initiation of the SGTS is rendered inoperable without logic control. power. The fans could be started manually, but the system would apparently operate in'a degraded, mode since the preheater: automatic temperature controlswould also not be operatin In addition, VACP-1 breaker #20 supplies power to the reactor building  y (RB) damper controls, turbine building (TB) ventilation . control, main  {
exhaust dampers, drywell and torus exhaust dampers, drywell vent and  '
purge valves and control room ventilation controls for train "A". Th'e exact extent of the impact of a loss of VACP-1 #20 on these components is still under revie .3 Discussion
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The SGTS is a plant engineered safety feature (ESF) reactor building atmosphere cleanup system which functions as a barrier between the radi-ation source and the environs during emergency conditions. Upon initi-  '!
ation~ and secondary containment isolation, the system establishes a  j negative pressure in the RB, thus preventing ground level leakage of  i untreated radioactive material from the RB to the environs; and the sys-  )
tem treats the RB atmosphere prior to exhausting via the plant stac Section 6.5.1.2.1 of the plant updated Final Safety Analysis Report (FSAR) describes the SGTS as consisting of two redundant, full capacity parallel flow. train Section 6.5.1.2.4 of the FSAR describes the sys-tem as automatically starting during the Design Basis Accident (DBA)
upon receipt of an initiation signa The instrumentation and controls section of the FSAR, 7.3, discusses the instrumentation provided to initiate ESF systems, including the SGT This system has both Reactor Protection System (RPS) and Non-RPS initi-
,  ation signal Sections 7.2.2.1 and 7.3.5.2 of the FSAR state that both  ,
the RPS and Non-RPS systems will automatically perform their protective  '
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functions whenever plant conditions exceed preset levels and that no-single failure can prevent the initiating circuits from performing their protective function In addition,10 CFR 50. Appendix A, General Design Criteria 41, Containment Atmosphere. Cleanup, specifies that each ~ system shall have suitable' redundancy to assure that its safety function can m,  be accomplished assuming a single failure.
 
; Integrated plant Assessment Systematic Evaluation' Program (SEP)
Topic VII-3 of the SEP evaluated the effects of a loss of VACP-1 on the ability to place the plant in a safe shutdown condition. A limited
  ' probabilistic risk analysis (PRA), discussed in Appendix D of the SEP, dealt with the contribution to risk of the loss of VACP-1 powered control room indications. The NRC concluded that the increased probability of operator error due to lost indication did not contribute significantly to top events in the fault trees, and thus, loss of VACP-1 was a low importance to risk. 'In addition, SEP Topic VI-7.C.1 evaluated the con-tribution to risk sof automatic bus transfers (ABT), specifically, their contribution to loss of power to redundant unit sub-stations 1A2 and.182.'
Backfitting of redundant power supplies for control room indication was not-recommende PSC 84-018 stated that the current design was acceptable based on the
  'SEP review and acceptance of the design. It is significant to note that the SEP analysis only evaluated VACP-1 from a perspective of providing control room indication and impacting loss of the redundant unit sub-stations'1A2 and 182. The loss of VACP-1 was not reviewed for a loss of automatic ESF protective functions. For these reasons, the inspector questioned the conclusion in PSC 84-01 *
5,5 Conclusions The inspector has concluded, based on a review of the regulations, FSAR, and SEP that the safety function of the SGTS is to automatically initiate during accident conditions to mitigate the consequences of these postu-lated accidents. The licensing basis for this system, as presented in the FSAR, is to automatict.lly initiate when plant conditions exceed pre-
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set levels and that no single failure (electrical) can prevent the in-itiating circuits from performing their protective functio In addition, 10 CFR 50, Appendix A, GDC 41, states that suitable redundancy in com-ponents and features shall be provided so that the system's safety func-tion can be accomplished, assuming a single failur Although still under review, the current Oyster Creek design for the automatic initiation of the SGTS is potentially susceptible to single failures as it has only one power source, VACP-1 and one initiation logic train downstream of VACP-1. The ability of the SGTS to automatically initiate with a susceptibility of single failure will be carried as un-resolved (50-219/88-38-02).
 
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  - 6.0 Rag Found in Scram Discharge Volume 6'.1EEvent-On'12/30/88, after a reactor scram,'the scram discharge volume (SDV) high
,  water level scram signal was bypassed in; order to' resetL the. scram logi Upon reset of the scram logic, the SDV vent'and drain valves opened and the '_'SDV Not Drained" annunciator cleared indicating that the SDV had been drained. Shortly thereafter,.while the SDV vent and drain valves remained open, the "SDV Not Drained" annunciator returned for the. south SDV-. heade Eventually, the SOV instrument' volume "Hi" and "Hi-Hi"~an-nunciators were also received for that- header.~ These indications,showed
  ' that even with the SDV vent and drain valves open, water was entering the SDV. south header and was not being draine The licensee initiated a work request.to inve'stigate t'he inability to-drain the south SD The inboard SDV drain valve ~was disassembled and a rag.was found 'to be blocking-the line. The: inboard drain valve for the north SDV was also disassembled, but no adverse conditions were identified. The source of water was determined to be. leaking scram out-let vr.1ves, which'were subsequently repaire As a result of finding foreign material in the SDV, the licensee initi-ated an effort to fill and drain the SDV five consecutive times in order-to verify the ability' to drain and to remove any other foreign materia This was accomplished successfully. The amount of time to drain the'
 
SDV's was measured with the following results:
NORTH SDV  SOUTH SDV 1  5:06  4:28 :07  4:35 :07  4:35 :11  4:35 :08  4:34 Each filling of the SDV also functionally tested the instrument volume level instrumentatio The licensee conducted a critique of the event in order to determine the source of the foreign material. Based on the deteriorated condition of the rag, the licensee concluded that it was old and had probably been introduced into the SDV in 1984, when the new SDV system was installe I 6.2 Review of Licensee Actions          !
Since the licensee's efforts to identify how and when the rag was intro-       i duced into the SDV were inconclusive, the licensee's approach was to demonstrate the repeated and consistent ability to drain the SDV. The
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licensee al'so desi. red to perform a visual inspection of_'the SDV instru-ment volume while the drain valves were disassembled, using a borescope,
  ~but the physical arrangement of the valves made.this impossibl The-inspector reviewed the' licensee corrective actions and. concluded'that
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the five consecutive SDV fill / drains. demonstrated confidence in the
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ability.to drain..the SDV. They also demonstrated.the operability of the l-#'  instrument volume level instrumentation. The inspector also reviewed-the times required to drain the SDV and concluded that a high degree of-consistency was demonstrated. These' drain down times 'were compared to those experienced during the most recent plant shutdown. No conclusions-could be drawn though, because of the different events measured. The sequence of alarm recorder (SAR) indicates the time of the resetting of .
the scram contacts,'while the five fills / drains were timed from the last SDV. drain valve's opening.
 
o During this outage, a modification was performed on the: scram discharge volume, installing. cleaning connections near the end of each of the six-headers located above the hydraulic control units. The inspector.re--
viewed Short Forms #44841, #44842 and #44840 in order to assess the cleanliness control measures implemented for this modificatio Installation specification, OC MM 328155-001 specified that all' material be maintained in a Class C cleanliness state as defined by ANSI N45. In addition, it specified that the internals' of pipe added by this: modi-fication shall .be flushed. During the modification, a Field Change Re-quest was submitted requesting deletion of the flushing requirement based upon maintaining Class C cleanliness during erection. This request was approved and no flush was performed on the new cleaning connections after they were welded to the SD The inspector reviewed the work packages and quality assurance-inspec-tions and verified that Class C cleanliness was documented until the piping assembly was welded to the SDV header, at which time no more in-ternal visual inspections were possibl l NRC Regulatory Guide 1.37, " Quality Assurance Requirements for Cleaning Fluids Systems and Associated Components of Water Cooled Nuclear Power Plants," essentially incorporates for operating nuclear power plants, the provisions of ANSI N45.2.1-1973, " Cleaning of Fluid _ Systems and As-sociated Components During Construction Phase of Nuclear Power Plants,"
both of which are committed to by the licensee in the Operational Quality Assurance Pla ANSI N45.2.1-1973 states that it is its intent to maintain the specified level of cleanliness during erection so that only water flushing will be required for final cleaning. This indicates that the maintenance of cleanliness during erection does not provide justification for deletion of the requirement to perform a water flush. Based on these requirements, l
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as stated in ANSI N45.'.1-1973, the inspector concluded that the change
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E to the installation specification to waive the flushing requirement was not proper, and that SDV Class C cleanliness has not~been demonstrated by examination of flushing filter _
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0FFICIAL RECORD COPY PROP EA GEN PUBLIC - 0004. /17/89
. The inspector also reviewed licensee Procedures A100-SMM-3900.07, " Main-taining Class C Cleanliness," and A100-GMM-3900.06.." Maintaining Class-C Cleanliness," for implementation of the cleaning requirements of ANSI N45.2.1-1973. These procedures incorrectly reference the 1980 edition of ANSI N45.2.1, which has substantially different flushing requirements from the 1973 edition. Since.the licensee's QA plan delineates Regula-
_ _ _ _ _ - . _ _ . . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ __ . - - - - - _ _  _ _ _ _ _ _ _ _ _ _ _ _ - . - . - - -- - - --
  : tory Guide 1.37, which endorses the 1973 edition of ANSI N45.2.1, the i
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licensee is committed to the 1973 requirements (as inodified by the OQA).


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Discussions with QA personnel indicated that it is not unusual for the flushing. requirements to be waived based on maintenance of cleanliness
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during erection. .The inspector concluced that the licensee's procedures were not adequate'in_that they do not implement the cleaning requirements of ANSI N45.2.1-197 The licensee't deletion of the system flushing. requirement from'the SDV-modification and the licensee's programmatic implementation of the cleaning requirements of ANSI N45.2.1-1973 will be an unresolved item (50-219/88-38-03).
 
7.0 Air Accumulator Testing The licensee conducted :esting on the air accumulators to various safety re-lated air operated valves during this report period. This testing was con-ducted in response to concerns identified by the Oyster Creek Emergency Operating Procedure (EOP) Inspection (IR 50-219/88-200) and Generic Letter 88-14, Instrument Air Supply Problems Affecting Safety Related Equipmen The E0P Inspection had identified deficiencies in the method by which the containment ventilation exhaust valves, V-27-1 and V-27-2, and intake valves, V-27-3 and V-27-4, were tested. -The testing method as identified by the E0P-inspection does not verify that the accumulator check valve would seat and be leak tight upon loss of instrument air nor verify that the accumulator had a ' sufficient free air volume available to stroke the valve under design con-ditions. The Oyster Creek instrument air system is not a safety related syste In the original scope of the accumulator testing, the licensee identified eighteen accumulators which they intend to test during this outage. These accumulators affect containment isolation, the Standby Gas Treatment System (SGTS) and condensate makeup to the isolation condensers. The inspectors have closely followed licensee testing activities. An additional 22 accumulators which are associated with SGTS valves were planned to be accomplished by the end of the next refueling cycl >
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Presently, the licensee has tested sixteen of the eighteen accumulators. Of the sixteen accumulators tested, eight accumulators required corrective main-tenance. Four of the accumulators which required work provide an air supply to the main. steam isolation _ valves'(MSIV), and the other four accumulators which required work provide an air supply to the SGTS trains' inlet and outlet
. valve The. accumulators and associated piping to the MSIVs provide an air supply to assist in seating the valves upon a loss of instrument air. Previous local leak rate tests have shown that if no air is supplied to the MSIVs in the shut" direction, .the valves will not pass the local leak rate test acceptance cri-teria. This fact' was identified in Preliminary Safety Concern 88-07 which was written in January 1988 but is not yet resolved. The accumulator piping to'the MSIVs consists of brass pipes and fittings. While attempting to per-form the test on these accumulators, the licensee discovered that there was excessive leakage from these fittings. The procedure had specified that all fittings were required to be snooped and the leakage corrected; however, due to the large number of fittings which were leaking air, the procedure could '
not be completed.-  The' licensee subsequently performed an air pressure drop test on' the accumulator and associated piping to quantify the leakage. The air pressure in the accumulator had dropped from 100 psig to 40 psig in a half hour period. The licensee evaluated the brass piping design to be inadequate and is presently modifying the piping by replacing the brass piping with steel piping and welding the steel piping together instead of utilizing fitting The licensee has also concluded that this piping must meet seismic require-ments and is presently installing seismic supports to the pipin The accumulators to the SGTS train inlet and outlet valves provide the motive force to-shut the valves upon a loss of instrument ai The absence of ac-cumulators upon a loss of instrument air would result in the standby SGTS train's inlet and outlet valves failing open. Since the SGTS trains share a common header downstream of the outlet valves of each train and upstream of the inlet valves of each train, the operating SGTS flow will recirculate backwards though the standby SGTS train, thereby rendering the operating SGTS ineffective. During the testing of these accumulators, the licensee deter-mined the pressure drop in the accumulators and associated piping to be ex-cessive. In an attempt to reduce the air leakage rate, the licensee replaced the isolation check valves to the the accumulators. The replacement of the check valves reduced the air leakage to a 1.5 psig to 3.0 psig pressure drop from 100 psig over a 30 minute period. The licensee, however, has evaluated this reduced leakage still to be unacceptable to ensure that the accumulators can perform their design function. As a result, the licensee is presently replacing the valve operators on the valves to further reduce the air leakage rat The licensee is continuing their evaluation of the test result and considering changes to their SGTS procedures to ensure that the SGTS will l  perform its design functio l
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The licensee has not completed their evaluation of the test results. Plans,    i however, are underway to review all forty accumulators for seismic require-ments. The licensee has not decided whether further accumulator testing on l-
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  : Region ~11 Docket Room (with concurrences)'
  . Management Assistant,-DRMA
  :Section' Chief,,DRP; n  r" LA', Drorcerick,1 Project: Manager, NRR'
/, .  -: Robert J;- Bores,L DRSS ~    l 1 ' ,. . ,  J. ' Lieberma'n, CE (w/SGI)*
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other accumulators are necessary. The inspector asked based upon the high'
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percentage of maintenance work required as a result of accumulator testing to date and the questions concerning the ability of the valves to perform I,
their design function, if the remaining accumulators should be. tested prior to restart from the outage. The inspectors will continue to follow the lic-
  . ensee's actions. This item is unresolved (50-219/88-38-04).
l l  8.0 Monthly Maintenance Observation u,s Portions of the fellowing maintenance activities were observed.
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Accumulator testing on Containment Ventilation Exhaust Valves, V-27-1 and V-27- s
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Troubleshooting of Main Steam Isolation Valve, NS03A, to determine cause for local leak rate testing failur Repair of Feedwater Check Valve, V-2-7 Items. inspected included a review of pertinent work packages, observation of valve tagouts, verification that technical specification requirements for secondary containment were satisfied, appropriate radiological precautions were taken in accordance with the Radiation Work Permit, required administra-tive approval was obtained prior to the start of work and quality assurance hold points were observe The accumulator test on valves V-27-1 and V-27-2 involved the disassembly of the accumulator from its associated piping and seismic supports to verify that there was no water accumulation. It was noted during the reinstallation of the accumulator that the work package did not specify any-torque specifica-tions for the strap bolts which held the accumulator to its seismic suppor The inspector reviewed the original installation specification for the modi-fication and noted that the installation specification listed torque values for these bolts. The inspector identified this fact to the licensee. The  .
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licensee subsequently changed the work package to torque these bolts in ac-cordance with the installation specificatio Although further discussion with the Technical Functions personnel raised
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questions on whether or not torque values were in fact required for these bolts, it was evident that the requirement for torque values for the rein-
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stallation of these bolts was not considered. The inspector had no further  :
questions on this matte i I
9.0 Monthly Surveillance Observations    '
9.1 Source Range Monitor Front Panel Test, Procedure 620.4.00 As a result of the frequent short periods which were observed on the source range monitors (SRM), the inspector observed the performance on the SRM Front Panel Test. The short period alarm was jumpered out of f
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Lm'*~  . service.during this period of-time. A reitew'of the licensee's eval'--        u L    ation to refuel without this alarm function was performe It" was al so'. _
m    note'd on the' surveillance that this alarm was* tagged out:of serviceLand that the Rod Block alarm was lockedDin as 'a result. off the Refueling -
m    .' Bridge being: tagged out of service. l The test- dataiwere observed'to be c:. -
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within thel Technical Specification ~ acceptance criteria. Proper approval from the Group Shift: Supervisor was obtained prior to the start of.the'
  ,  surveillance. _No' unacceptable conditions were identifie .2 Reactor Triple L'ow Water Level Tes't and Calibration, procedure 619.-31006; V  -
Portions of.the ReactoF Triple Low Water Level Test and'Calib' ration. sur- ,
    .veillance'were-observed from the control room and at RK02. Review of        .
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this' surveillance included theLverification of tes't instrument calibra-tion', test' values 'were within the acceptance _ criteria, appropriate ad-
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ministrative approval recei.ved and proper. documentation. performed. No unacceptable conditions were' identifie .
  ' 10.0 Backshift Inspection-
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e NRC inspection of-' licenses activities during deep backshift hours were con '        >w ducted on the'following dates:
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Saturday, December 31, 1988-
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    . Thursday, Janua'ry 5, 1989-
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    ' Friday, January 6, 1989
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Saturday,. January 14,.1989        '
Areas:of inspection included the observation of fuel loading in the reactor vessel,- fuel. pool gate gasket replacement, Source Range Monitor Front Panel Testing, control room activities and performance of plant tour .0 Reactor Vessel = Lost Parts Dur.ing this outage, a large number of items had inadvertently fallen 1n' to the reactor vessel. These items include two bullet lights', a Hansen air fitting,      .
a magic marker cap,'a metallic band, a metallic ring, a " screen-like" material, an acorn. nut, a lens cap, a lens to a pair of s'afety glasses, three pieces of tie wrap:and a few short pieces of duct tape. Although a majority of these          i items have been removed from the reactor vessel, the licensee intends to leave some of these items in the reactor vessel during the upcoming operating cycle. These items include a few short pieces of duct. tape, a lens to a pair of safety glasses and a lens cap. A General Electric analysis was presented to show that the items would disintegrate upon heatup and that chemistry changes were acceptabl I The inspector reviewed portions of the licensee's Reactor Vessel Lost Parts" Analysis. This analysis includes a reference to the General Electric Lost Part Analysis which was performed for the licensee during the 10R Outag No concerns were identified by the inspector in the review of General Elec-tric's evaluatio In light of the large number of items which have fallen l
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y  4 in the reactor. vessel during this outage and the fact that many 1tems in.the
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reactor vessel were identified by.use of an underwater camera device the per-m 1 sons who may have inadvertently and unknowingly dropped the ' items, improvement
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in the observation of cleanliness during maintenance and operational activi-
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  - ties is needed. Licensee staff members acknowledged the need- for improvement
,  and agreed to review their ' practice .0 Torus Tiedown Bolts Independent visual examinations were conducted by the licensee and the in-
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spector on the torus anchor bolts. This examination was conducted in response to anchor bol_t anomalies identified in the Hatch Nuclear Plant.'s Mark I con-tainment. The majority- of the anchor bolts at. the Hatch , plant"were deformed with a maximum deflections between 1/4 inch and onelinc The visual examination.of the anchor bolts at Oyster Creek included checking for deformation in the saddles, bearing plates, base plates, anchor bolts and grouting. Positions of ,the bolts in?the slotted holes were also noted. The licensee examined 90% of the anchor bolts and identified no deformation or anomalie The inspector examined 20% of the anchor bolts with similar find-ings. .The inspector had no further questions regarding this matte .0 Fuel Support Castings During the replacement of fuel support castings '(pieces) in the contrel rod drive tubes, the licensee experienced difficulty in properly setting the castings. The licensee had removed 26 castings earlier 4 the outage to sup-port the exchange of control rod blades. Most of these 26 castings were not properly' seated. Additionally, the licensee believed that three peripheral castings were not f ully seate After unsuccessful ati.empts to seat the castings, the licensee contracted General Electric to' assist in the seating of the castings. With General Electric assistance, the licensee was able to seat all castings. ' One of the castings, however, required minor filing to allow the guide pin on the lower core support plate to fit into the casting. Further evaluation of i.he three peripheral castings indicated that two of the castings were fully seated.
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The third casting was easily seated after it was slightly touched. The lic-ensee determined that this casting may have lifted off its seat when its fuel bundle was removed from the cor The licensee believes that a potential contributor to the difficulty in seat-ing the castings was the fact that some castings were not returned to their original position Although these castings are supposedly identical, the licensee hypothesized that they may have become form fitted in their positions in the core plat In previous outages, the entings were returned to their original positien as a result of the method of exchanging control rod blades, and thus no fit problems existed. The licensee was able to obtain an original plant drawing showing the uniquely numbered castings in their original core position. Utilizing the serial numbers enscribed in the castings, the cast-ings were returned to their original position A - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - - -
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It was observed, however, that some of the castings which the licensee had-difficulty ~ seating were already.in the original' position. -It, appeared to the inspector that operator _ technique.was also a potential contributor to the difficulties in seating the castings. - General Electric personnel _ had experi-enced a easier effort in seating the castings. .Another potential contributor
.. to the' seating difficulty was the weight of the tool utilized to seat the castings. The licensee. believed that use 'of a heavier tool would have as-sisted.in properly seating the casting After seating the. castings, the licenser verified that all castings which were moved during the outage were properly seated. Additionally, more.than sixty-other castings which were not moved were verified:to be fully seated. The inspector had no further questions regarding this matte .0 Control-~ Rod Drive Return Line Piping On December 28, 1988, while performing weld inspections on the control rod
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drive return-line, the licensee identified that a section of piping was sig-nificantly' thinner than called for in Revision 6 of the General Physics In-service Inspection Program Schedule (15IPS). The ISIPS had called for a schedule 80 piping which has a nominal thickness of 300 mils. The portion of piping,.however. had readings between 200 mils and 220 mil This portion of piping is located between the CRD return line manual isolation valve, V-15-29, and weld NC-4-20 and forms part of the reactor pressure boundar The piping is a 3 inch pipe six inches in-legth. The licensee's t evaluation of this piping' determined that .it was not thinned from erosio This evaluation was based opon the uniform thickness of the piping and the abrupt change in wall thickness at tna location of weld NC-4-20. The evalu-ation further._ postulated that the piping was probably installed schedule 40 piping during original plant construction, althwgh no installation documen-tation can ce locatec: The licensee performed a stress. analysis on the'
schedule 40 piping m htermined thn it was acceptable in accordance with ANSI B31.1. The inspet. tor reviewec m licensee's evaluation of pipe stresses and identified no unacceptable cono m on .0 Loss of Security Electrical Loads On December 10, the licensee was preparing to perform maintenance to replace a breaker which supplies safeguards loads. Upon deenergizing the safeguards loads, security personnel realized that compensatory actions could not be maintained because of the unanticipated loss of certain safeguards function As a result, the security loads were reenergized almost immediately and the maintenance activities deferre The inspector interviewed operations and security personnel concerning the j circumstances surrounding this event. Operations personnel had believed that security personnel maintained a set of safeguards electrical diagrams and that ]
they were cognizant of the loads which would be lost upon performance of this l maintenance. On the contrary, security personnel did not have such diagrams j
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in-their-possession and stated that even if they did have.them they.did not
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l  -have adequate knowledge to utilize them. Security personnel had relied 'on
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memorytto determine what safeguards equipment they wouldLlos It wasl evident that the miscommunication resulted in an' inadequate review of g , . the impact of the maintenance activity. The safeguards engineer at the plant w'as not informed nor consulted in the performance this work. It was also i  noted from interviews with personnel that this practice had occurred many-L  times in the past, however, appropriate compensatory measures were taken.
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  'The inspector brought this problem to the attention of the licensee. Although a formal critique' was not performed the licensee held ~a meeting with the dif-
  'feront departments involved. The licensee intends to' generate a list of saf eguards loads and their associated power supplies for use after this~ outag It the interim, an increased awareness of the problem and the utilization of the safeguards engineer should prevent similar occurrences and not jeopardize plant securit The inspector had no further questions regarding this matte .0 Source Range Monitor Log Integrator Cards During pre-refueling checks the licensee found defective log integrator cards in the source range monitors (SRM). These log integrator cards were replaced prior to refueling, but af ter the card replacement the SRM!s exhibited fre-quent large spikes coupled with short period alarms. Prior to refueling, the 4  inspectors discussed a concern regarding SRM operability with the Plant Engi-neering director.. The licensee explained that they had delayed refueling for approximately five hours to trouble shoot the SRM's and found no problem The inspector reiterated his concern regarding SRM operability and the large number of short periods. The licensee stated that with regard to refueling, the SRM's were operable to accurately measure neutron level. The inspector agreed to review SRM behavior after some fuel assemblies had been loaded in the core around the SRM's. After this had occurred, the SRM still exhibited the same behavior. In addition, when one source was moved adjacent to SRM 21, a normal source range response was observed and did not display the ab-normal spiking prevalent with SRM's 22, 23 and 24. Again the licensee deter-mined that the source range instrumentation was adequate to measure source level neutrons during. refuelin Subsequent to refueling the licensee performed troubleshooting of the SRM's to include response time testing of the log integrator cards. The licensee conducted bench testing of the log integrator cards removed from the SRM's after completing refueling operations. The licensee determined that the 390 microfarad capacitors used in the log integrator dampening circuit were de-fectiv These log integrator replacement cards with the defective 390 microfarad capacitors were obtained from shop spares and bench tested prior to plant installation. Inspection Report 50-219/88-33 identified a violation regarding
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raised a concern regarding reliability of these spare-parts due .to the inade-quate storage. During the 88-33 inspection the licensee assured the inspector that before any shop spares would be used in the plant sufficient bench test-ing would be conducted to . verify the acceptability of the spare part. Thisi
,  event raises the question of the acceptability of the QA process used to verify shop spare part adequacy for plant installation and also raises.the question of previous shop spare part installation'in the plan Another concern results from the guidance provided by the technical manual with regard to the, correct replacement module number. Apparently the tech- I nical manual provided different module numbers on the functional block diagram, the SRM applicability table and the manual secti.on on operation and trouble-shooting of the log integrator. In addition, the one module number indicated'
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is for a process radiation monitor with a difference in the card capacitance
  .which explains the large spikes and short period alarms exhibited by the SRM' The inspector questioned whether the technical manual could provide sufficient guidance to the technician to repair safety instrumentation and control cir-cuitry. The licensee critiqued this event. This item is unresolved 'pending further NRC review (50-219/88-38-05).
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17.0 Review of Periodic and Special Reports
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Upon receipt, periodic and special reports-submitted by the licensee pursuant to Technical Specification requirements were examined by the inspecters. This
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review included the following considerations: the report includes the infor-mation required to be reported to the NRC; planned corrective actions are adequate for resolution of identified problems; and the reported information is vali The following reports were reviewed:
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Monthly Operating Report for November 1988
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Special Report 88-01, Fire Diesel Pump 1-2
This report identified nonfunctional fire diesel . pump 1-2 due to debris in the suction bell, which degraded pump performance. A rag was found in the suction bell, but the licensee was unable to determine how it was i
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introduced into the system. The licensee offers that a potential intro-duction point is through inspection ports on the relief valve. The in-spector was unable to understand how a rag introduced in the system at j  a relief point could happen to partially block the pump suctio System L  material cleanliness requirements during maintenance should be reviewed j  by the licensee. The inspector had no concern i
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  .18.0 Meetings-
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,  18.1 Ge'neral' Review Board Meeting f
The inspectors attended portions of the General Office Review Board
,    meeting / conducted at Oyster Creek on January 10 and 11,1989. The meet-
_ ing appeared to be a very worthwhile effort.to improve plant and' company
    ' performance.'
18.2 Quality Maintenance Team Briefing-On 12/21/88, resident inspectors met with a maintenance manager and were , , ,
briefed on the licensee'r, plans to enhance maintenance performance. The plans include the initiation of a pilot program using ~ the Quality Main-tenance Team (QMT) concept with eventual full implementation of the QMT '
    . concept; s 19.0 Radiation Protection During entry to and exit from the RCA, the inspectors verified that proper-
    . warning signs were posted, personnel entering were wearing proper dosimetry, personnel and materials -leaving were properly monitored for radioactive con-tamination and monitoring instruments were functional and in calibratio ,
Posted extended Radiation Work permits (RWPs) and survey status boards were reviewed to verify that they were current and accurat The inspector ob-served activities'in the RCA to verify that personnel. complied with the re -
quirements of applicable RWPs and that workers were aware of the radiological-conditions in the are .0 Observation of Physical Security During daily tours, the inspectors verified that access controls were in accordance with the Security Plan, security posts were properly manned, pro-tected area gates were locked or guarded and that isolation zones were free of obstructions. The inspectors examined vital area access points to verify that they were properly locked or guarded and that access control'was in accordance with the security pla .0 Review of Licensee Event Reports (LERs)
21.1 In-Office Review of LERs A review was conducted of LERs submitted to the NRC to verify the details were clearly reported, the cause appears accurate and is supported by report details, corrective actions appear' to be appropriate, and that no further information is required. The following LERs were reviewed:
87-34: Inadequate emergency lighting was identified to exist in several areas. Additional lighting to correct the condition has been installe _ _ _ _ - _ _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ .
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87-35: Reactor scram with the reactor shut down during. the performance of a surveillance test. The cause was attributed to a degraded cable
  -
    ' connection due to normal wear. The cable will.be repaired and similar conditions inspecte .
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^ *'  '87-36: Improper control of a' high radiation area due to personnel error
-    and procedural' noncomplianc Several workers were terminated and-
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stricter guidance was , issue .
?    87-37: Failure to perform two monthly ' gaseous' effluent dose calculation Subsequent calculations showed releases--to be~
      ~
due to personnel erro well below the limit. Corrective action consisted of training additional
    ~
as    personnel to perform the calculation ''
87-39: Discovery of a failure to meet 10 CFR 50, Appendix R criteria in an emergency diesel generator lockout relay circuit. Corrective action was taken to modify the circuitry to comply with 10 CFR 50, Appendix R cruari . 87-42. ~orus oxygen sample line isolation valves were identified as not meeting single failure criteri Modifications will be made to isolation
  . valve ci,rcuitry to meet single failure criteri ': '(Voluntary Report) Fuel pool cooling system piping . radiation shielding' inadvertently removed creating a highl radiation area. . Properly labelled permanent shielding was installed and personnel training was conducte m 87-42: This event was discussed in Inspection ReportL50-219/87-3 : (Voluntary Report) This event was discussed in Inspection Report 50-219/87-3 ..
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88-01: This event was discussed in Special Inspection. Report 50-219/88-0 : Containment high range radiation monitors failed to meet accept-ance criteria during calibration due to incorrect post installation testing. The monitor was repaired, tested and returned to service, a  ;
vendor document control program established, and personnel training con-ducte : Containment particulate monitor sample line isolation valves con-trol circuitry found not to meet single failure criteria since installa-tion in 1976. Standing orders have been issued until the deficiency is corrected in accordance with the Integrated Living Schedul : Isolation condenser automatic actuation pressure sensors tripped at valves greater than those specified in Technical Specifications. The !
sensors were adjusted to trip within setpoint limits. Long term correc- ;
tive action is to replace the instruments with an analog trip syste I i
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88-05: A tagging error caused components of both standby gas' treatment-systems to be inoperable. Corrective action consisted.of additional
!:  m personnel. training. This event was reviewed.in Inspection. Report 50-
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219/88-09.~
y  '88-06: 'During surveillance testing,. four 'of eight ' isolation condenser y
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pipe break sensors tripped at a differential pressure greater than specified in the Technical-Specifications. The sensors were adjusted
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to trip within unit Four sensors will be. replaced during theccurrent-outage with newer design switches and their performance evaluate : During surveillance testing, two isolation condenser-automatic actuation pressure sensors tripped at' valves greater than specified in . .
the. Technical Specifications. Corrective action is the same as described in LER 88-0 ~
88-11: This event was reviewed in detail in Inspection Report 50-219/
88-2 : Testing showed Main Steam Isolation Valve NSO4A leaked excessivel The valve was repaired, a new valve stem was installed, and a' successful r  leak rate test' performed. The valve inspection and repair procedure will'
also be revise .
88-15: A main steam isolation valve closure occurred during surveillance testing, while the plant was in cold shutdown. The valve closure oc-curred with a required jumper fell offland shorted to ground. A project has been initiated to eliminate the need for. installing jumpers during surveillance testin : During surveillance testing, four isolation condenser automatic actuation pressure sensors tripped at values greater than those specified by the Technical Specifications. The corrective action is the same as that specified in LERs 88-04 and 88-05'.
88-17: (Voluntary Report) Three control rod accumulators low pressure alarms were present for greater than one hour in violation of plant pro-cedure The event was discussed with all operating shift personne : Inadvertent actuation of the "B" isolation condenser due to opera-tor erro The operator immediately realized his error and took correc-tive actio The actuation lasted for less than seven seconds. This report was made required reading for operator : This report described the circumstances associated with both isolation condensers being inoperable. An enforcement conference was held relative to this matte The enforcement conference is dist.ussed in Inspection Report 50-219/88-3 l
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y 88-21: ,This report described the circumstances associated with both
  ' isolation condensers being in an unanalyzed' condition. A_special review-of.this' event was conducted by an Augmented Inspection Team. Resultsl of the team review are documented in Inspection Report 50-219/88-8 .  .88-22: This . report described the loss of an emergency bus due to a' ground :
fault. -This-event was reviewed in Inspection Report 50-219/88-8 : Drywell airlock not . leak rate tested in accordance with Appendix J. The review of this event is discussed in Inspection Report 50-219/-
    '
i-88-2 : Main steam isolation valve closure signal occurred during sur -
veillance testing due 'to procedural deficiency. The surveillance proce-dure will be revise : A reactor scram signal was received with the plant shutdown-due to an error in a newly revised procedure. The procedure will be revised and this LER made required reading for qualified reviewer G 88-27: Several fire watch tours were missed due to inadequate guidance
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and direction being provided to temporary contractor employees. Correc .
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tive action consisted of establishing a log which requir'es hourly signa-tures. Also, 'a procedure will be developed which will define fire watch responsibilities and the log requirement : Six containment isolation valves were identified which do not meet NUREG-0737 criteria which requires deliberate operator' action to open an isolation valve after the isolation signal is reset. A modification will be performed on these valves so they will meet requirements in ac-cordance with GPUN's Integrated Living Schedul .2 Onsite Review of Licensee Event Reports (LERs)
The following LERs were reviewed to determine the report was adequate in assessing the event, the cause appeared accurate, corrective actions appeared appropriate, generic applicability was considered, and that the licensee review and evaluation were complete and accurat : Safety limit violation caused by personnel error while removing reactor recirculation pumps from service. This event was discussed in detail in Inspection Report 50-219/87-29. During the event, it was noted the fuel zone level monitor (F2LM) did not activate when all five recir-culation pumps were trfpped. Corrective actions specified in this LER
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and also discussed in a September 20, 1987 letter, Clark to Murley, were:
  "GPUN will evaluate the fuel zone water level instrumentation s tem for the appropriateness of the recirculation pump trip signal and the pos-i  sible addition of an alarm." During this inspection, the licensee's i l  committed-to evaluation of this matter was reviewed. A Licensing Action !
L  Item (LAI) 87178.06 was written to initiate this evaluation. This LAI .
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was closed out on December 7, 1987, based on the performance of an evaluation. The inspectors review of this evaluation showed that it dealt'with a February 1987 event in which a recirculation pump M-G set field relay failed and did not at all address the corrective actions specified in the LER. The performance of an 1    propriate evaluation
,  and its apparent acceptance was discussed with sne licensee. Subse-quently, a supplemental response to the LAI, dated January 16, 1989, was prepared to address the F2LM, and concluded the installed instrumentation is adequate. Also, the evaluation of an alarm installation was stated to have been completed and determined no alarm was necessar The writ-ten evaluation associated with the alarm determination was not available on site. Additional management attention to the review of committed-to evaluations appears to have been warranted in this instance to assure that a correct evaluation was performe , 88-08, and 88-10: These three LERs all reported standby gas treatment system automatic initiations. The basic cause of all events was water in the off gas inlet line and its associated drain to a sump located at the base of the stack. In addition to procedure changes, personnel counseling on the importance of strict performance compliance, and the required reading of two of the reports by Radwaste Operations personnel, a modification to the off gas drain was made and a new seal pot was installed to increase the water level over the drain lin This modification should eliminate this problem which had been experienced for a long tim : This LER reports a reactor scram signal which occurred while moving the reactor mode switch from shutdown to refuel during testin This apparent cause was attributed to mechanical wear of the mode switc Corrective action included an evaluation of a mode switch replacement during a future refueling outage. This evaluation has been completed and it was recommended not to install a new mode switch at this tim The license'e has, however, committed to purchase a new improved mode switch. One of the reasons for not installing a new mode switch is that there is no clear evidence a problem exists with the installed switc The inspector had no further questions relative to this matte : This report describes the discovery of a reactor coolant sample line containment isolation valve being inadequately supported. It was reported the valve was supported by rope and a pipe support upstream of the valve was broken. In addition to correcting the physical deficien-cies, corrective action consisted of making this LER required reading to all plant en;.neering personnel. During this inspection, it was de-termined additional corrective action was taken in that a critique of the event was performed which identified root causes and specified addi-tional corrective actions. The inspector had no further questions rela-tive to this matte _ _ _ _ _ _ _ _ _ - - _ - _ _ _ _ _ - _ - _ _ _ - - _ _ - _ _ _ _ _ _ _ _ . . _ - _ _ - _- _ __-__
 
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    !88-12: This, report describes.certain Appendix'R deficiencies identified
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    :in;elgetromatic relief valves and cleanup' system valves circuitry. , Part
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   ,   ,of'the: analysis and nfety assessment of the occurrence described the use of-~ torus cooling'using the containment spray system. dynamic test mode 1''  .and the emergency. service water system.- The use~of this method of torus
,
;,   cooling is affected by a Preliminary Safety Concern.(87-003) which de-s-scribes certain elements of the containment spray systemLlogics during; the dynamic test mode. The PSC. explains'that the capability'of the dynamic test mode to cool the torus.from the control. room may not be avai.lable due to logic circuit. considerations. Ther inspectors felt thise consideration was:not addressed in the LER. The Preliminary Safety Con-cern.(PSC) had been closed out but because of recent interest in:PSCs "
5/ :. :. /B9    ]/p /89  5/ /89 0FFICIAL RECORD COPY PROP EA GEN PUBLIC - 0005. /12/89
was stated by the licensee as be.ing reevaluated. The resolution;of this matter is being tracked.by, Unresolved Item 50-219/87-04-0 : :
     '
88-13i This-report describes a Main Steam, Isolation Valve (MSIV) stem failure.: As part of the description of the event,'.the licensee described a Technical Specification violation in that a half scram was not inserted
    :for approximately 50 minutes while there was no steam flow in the "A"'
steam.line with.both.MSIVs indicating open. The decision-not to insert
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the' half scram was made for safety concerns to permit a maintenance fore-man to = inspect the' outboard MSI The LER states'when it was determined, the foreman was clear of the outboard MSIV, the half scram.wasm inserte This statement is not entirely true for, based on discussions with per-sonnel and an entry in the log book, the half scram was in fact inserted,.
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based on information developed during the evaluation before the outboar MSIV.was inspecte Since many options were available to the crew such as believing the out-board valve position indication and not inspecting the valve to removing
    'the' half. scram for only the two or three minutes the outboard valve was being inspected. The failure to insert the required half scram was dis-cussed in detail with members of the onshift crew during the event and with licensee managemen From these discussions, it was determined-management above the 'shif t level made the ~ decision not to insert-the half scram for personnel safety considerations. Certain levels of managerrent and members of the shift crew, during the event immediately recognized and expressed the view that a half scram must be inserted. The Group Shif t Supervisor (GSS) was aware that someone had been dispatched to inspect the outboard MSIV but it could not be determined who actually required-the inspection be performed. After being dispatched, the fore-man could not be contacted by radio or page .
        '
The inspectors discussed the failure to insert the half scram, and the ;
details associated with the decision with licensee management. During I this discussion the need to minimize disagreement during shift operation and to effectively control the dispatching of personnel which could have an effect on decisions which have a potential to affect Technical Speci-fication requirements was identified. It is realized that during an event the time to make decisions is limited. However, events of this
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nature 'should be evaluated in order'to more effectively address poss'ibl '
future. events.- The licensee. indicated the event had been reviewed and that~ the' GPUN letter. describing command responsibilities ha'd.been re-emphasized with the GSSs. No enforcement action relative to the delayed insertion of a half. scram when'.its insertion was indicated is. going to
,  be taken'since'it was identified ~ by the licensee and met -the criteri of 10 CFR 2, Appendix C'(50-219/88-38-06).
 
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88-20: This report describes .the discovery that the_. isolation condenser steam vents and' the emergency service water discharge lines,from the containment spray heat' exchangers radiation monitors would not ' fulfill
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their design function of radiation' leak detection due to .high radiation background following an accident. During this inspection the inspector reviewed the four alarm response. procedures associa'ted with, these moni-y  tors, of the four procedures involved only one had a note included which indicated the monitor was subject to high background radiation and may alarm.without a ~ leak being present. The necessity of upgrading prece-
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dures to reflect discoveries made which could have an affect- on operator .
responses was discussed with licensee management. . Management immediately took steps to insure that ' reviews are conducted of LERs to assure that appropriate procedures are changed if necessar .0 Refueling Activities Ref ueling began on January-6,1989 and was completed on January 13. This
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refueling'was conducted without any error. The only problems observed were
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  .those associated with Source Range Monitors (Sections 9.1 and.16.0). This
  . smooth and error-free refueling is in contrast to the' earlier defueling. de-scribed'in Inspection Report 50-219/88-33. The significant improvement is the result of the application of lessons learned from refueling and.the. con-siderable management attention paid to the refueling.
 
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23.0 Unresolved Items
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Unresolved items are matters for which more information is required in order to ascertain whether they are acceptable, violations, or deviations. Unre-solved items are discussed in paragraphs 1.0, 5.0, 6.0, 7.0 and 16.0 of this repor .0 Exit Interview A summary of the results of the inspection activities performed during this report period were made at meetings with senior licensee management at the end of this inspection. The licensee stated that, of the subjects discussed at the exit interview, no proprietary information was include t L-___-_--_ ._
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Latest revision as of 21:08, 30 January 2022

Discusses Safety Insp Rept 50-219/88-38 on 881204-890114 & Forwards Notice of Violation.Violation Classified as Severity Level IV Since Unauthorized Entry Occurred as Result of Isolated Lapse in Vigilance by Security Officer
ML20247N416
Person / Time
Site: Oyster Creek
Issue date: 07/24/1989
From: Kane W
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To: Fitzpatrick E
GENERAL PUBLIC UTILITIES CORP.
Shared Package
ML20247N421 List:
References
EA-89-108, NUDOCS 8908020393
Download: ML20247N416 (3)
v • d • e


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- 2- 1 JUL .2 4 1989 b -Docket'No'. 50-219 DESIGNATED ORICINAL

9 ] 08 Cortified />ld- Ab TGPU' Nuclear Corporation

,

' ATTN: Mr.. Eugene E. Fitzpatrick

.Vice President and Director Oyster Creek Nuclear. Generating' Station -i

,

P. 0. Box-388- '

Forked River, New Jersey. 08731

Gentlemen:

Subject: NOTICE OF VIOLATION (Inspection No. 50-219/88-38) l This refers to the routine safety inspection conducted between December 4,  !

1988 and January 14, 1989 at the Oyster Creek Nuclear Generating Section, '

Forked River, New Jersey. During the inspection, the NRC reviewed the circumstances ' surrounding an unauthorized entry into a vital area which

.

occurred on' January 3,.1989'and which was identified by your staff and reported to'the NRC. The= inspection report, which was sent to you on February 28, 1989,- '

indicated that'this-unauthorized entry.was classified as an unresolved item, i

!

The: unauthorized entry occurred when a contract radiological controls j technician entered a vital area, undetected, after his vital area access '

,

' authorization'had been' terminated. Although it was subsequently determined

_that the technician's access had been erroneously terminated because of an >

- administrative error, this entry' represents a failure to maintain positive access. control to a vital area and constitutes a violation of NRC physical

' security requirements. The' violation is described in the enclosed Notice of Violatio 'Although this violation could be classified at-Severity Level III in accordance

.with Section C.2 of Supplement III of the " General Statement'of Policy and

,

'

Procedure for NRC Enforcement Action," 10 CFR Part 2, Appendix C, 53 Fed. Re (October 13,1988) (Enforcement Policy), the violation has been classified at Severity level IV because: (1) the unauthorized entry occurced as a result of an isolated lapse in vigilance by a security officer, concurrent with a plant worker's inadvertent failure to follow a security procedure; (2) the

,

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security officer identified the unauthorized entry when the technician exited fromlthe area, and promptly reported it to security management, thereby demonstrating his integrity; and (3) the technician was previously authorized to enter the area, had entered the area on numerous occasions, and his authori-zation would have been continued except for an administrative oversight. The

' NRC recognizes that your corrective actions, upon notification of the entry, were prompt and comprehensive. Nonetheless, we are concerned that sufficient measures were not established to preclude such an event from occurrin Therefore, any similar violation in the future may result in escalated enforcement actio OFFICIAL RECORD COPY PROP EA GEN PUBLIC - 0003. PDR ADOCK 05000219 Q PDC

_ _ _ .

. . , .

' GPU Nuclear Corporation

"

, -2-You are required to respond to this letter and the enclosed Notice, and in preparing your response, you should follow the instructions specified in the enclosed Notice. In your response, you should document the specific actions taken and any additional actions you plan to prevent recurrence. These actions should include strengthening your General Employee Training program by emphasizing the importance of all workers to follow security procedure Further, you should place all Safeguards Information (as defined in 10 CFR 73.21) in an enclosure, so as to allow your letter, without the Safeguards Information, to be placed in the Public Document Room. After reviewing your response to this Notice, including your proposed corrective actions, the NRC will determine whether further NRC enforcement action is necessary to ensure compliance with NRC regulatory requirement Tne enclosed Notice contains details of your security program that have been determined to be exempt from public disclosure in accordance with 10 CFR 73.21 (Safeguards Information). Therefore, those portions will not be placed in the NRC Public Document Room and will receive limited distribution. This letter will be placed in the Public Document Room in accordance with 2.790(a).

The response directed by this letter and the enclosed Notice is not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, PL 96-51

Sincerely, William F. Kane, Director Division of Reactor Projects Enclosure:

Notice of Violation [Contains Safeguards Information (SGI)]*

cc w/enci (w/o SGI):

M Laggart, BWR Licensing Manager Public Document Room (PDR) 1 Local Public Document Room (LPDR) i Nrlear Safety Information Center (NSIC)

NRC Resident Inspector (w/SGI) *

State of New Jersey

  • Contains Safeguards Information l

l

1

)

0FFICIAL RECORD COPY PROP EA GEN PUBLIC - 0004. /17/89

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Region ~11 Docket Room (with concurrences)'

. Management Assistant,-DRMA

Section' Chief,,DRP; n r" LA', Drorcerick,1 Project: Manager, NRR'

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