IR 05000219/1998005

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Insp Rept 50-219/98-05 on 980614-0726.No Violations Noted. Major Areas Inspected:Plant Operations,Maint,Engineering & Plant Support
ML20237E509
Person / Time
Site: Oyster Creek
Issue date: 08/26/1998
From:
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20237E506 List:
References
50-219-98-05, 50-219-98-5, NUDOCS 9809010011
Download: ML20237E509 (28)
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U. S. NUCLEAR REGULATORY COMMISSION REGION l

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! Report N Docket N l 72-1004 I t

i License N DPR-16 Licensee: GPU Nuclear incorporated 1 Upper Pond Road Parsippany, New Jersey 07054

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Facility Name: Oyster Creek Nuclear Generating Station l Location: Forked River, New Jersey I l l Inspection Period: June 14,1998- July 26,1998 ,

! l Inspectors: Joseph G. Schoppy, Senior Resident inspector Thomas R. Hipschman, Resident inspector l Joseph T. Furia, Senior Radiation Specialist l Lonny L. Eckert, Radiation Specialist l Stephen M. Pindale, Senior Resident inspector (Hope Creek)

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Approved By: Michele G. Evans, Chief Projects Branch No. 7

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9009010011 980826

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PDR ADOCK 05000219 l G PDR ,

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i l EXECUTIVE SUMMARY l

Oyster Creek Nuclear Generating Station l Report No. 98-05 l This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers about a six-week period of inspectio Plant Operations l

l * Operations conducted safe and high quality fuel handling activities in preparation for the refueling outage. In addition, management provided good oversight. (Section

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  • Operations properly planned and effectively implemented a transition to a new electronic control room narrative log format. (Section 03.1)
  • The General Office Review Board demonstrated a good safety focus and promptly l engaged the station's correction action process to address their safety concern (Section 07.1) * The Independent Onsite Safety Review Group conducted a thorough Technical

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Specification review, identified several value-added improvements, and promptly initiated corrective actions. (Section 07.2)

  • Operators demonstrated good technical specification (TS) awareness, properly l documented applicable limiting conditions for operation, and met all TS i requirements during periods of dieselinoperability. (Section M2.3) l
  • Control room operators responded appropriately to a failure of a containment isolation valve to fully stroke while isolating the reactor water cleanup system on two occasions. The chemistry department provided good support to evaluate chemistry parameter action levels and to provide recommendations concerning continued operation without the reactor water cleanup system. (Section E2.2)

Maintenance l

  • Planning performed a thorough risk analysis and appropriately adjusted the work
week schedule to support an emergent emergency service water (ESW) pump l replacement. Maintenance activities were well-controlled and properly documente (Section M2.1)
  • . Maintenance performed thorough and well documented troubleshooting following two trips of the 'B' reactor recirculation pump. (Section M2.2)

i l * Maintenance properly planned and effectively controlled a diesel generator battery replacement. Maintenance demonstrated a good safety focus in conducting increased monitoring following battery replacement and identified an apparent l

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degraded condition on one of the battery cells. Maintenance took prompt and appropriate corrective action to address the condition. (Section M2.3)

Enoineerina

  • The ESW system engineer demonstrated good system ownership and led the efforts I to promptly address degraded performance on an ESW pump. (Section M2.1)
  • Engineering provided good support and evaluation of the data, following troubleshooting of the 'B' reactor recirculation pump trips. (Section M2.2)
  • Engineering led a well-controlled and effectively coordinated troubleshooting effort to evaluate a main condenser degraded vacuum condition. (Section E2.1)
  • Engineering demonstrated good involvement in evaluating a problem with a reactor water cleanup isolation valve. Engineering root cause evaluation continued at the end of the report period. (Section E2.2)

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  • Engineering demonstrated a good questioning attitude in identifying and documenting a diesel generator switchgear seismic nonconformance. Engineering led the efforts to disposition operability issues, analyze and resolve the .

nonconformance, address deportability, and implement appropriate corrective l actions. (Section E2.3)

Plant Sucoort

  • Radiological controls effectively supported refueling activities. In addition, management provided a good oversight. (Section 01.2)
  • The licensee established an adequate program for the sampling, analysis and assessment of personnel exposure to airborne radionuclides. A programmatic weakness was identified regarding the lack of reverification of the basis documents for this area since 1993. (Section R1.2) 4
  • Appropriate controls for high, locked high and very high radiation areas were established. Control of radiological work for the shipment of a radwaste liner, receipt of radioactive materials (new fuel), and spent fuel pool work were also effective. (Section R1.2)
  • The presentation of continuing training to radiation protection technicians was appropriate with regard to scope and depth of presentation for two classes reviewed during the inspection period. (Section RS.1)
  • The annual review of the radiation protection program was adequate. Weakness in the scope and depth of the summary self-assessment for 1997 was noted; however, the sum total of self-assessments and safety assessments met the requirements of 10 CFR 20.1101. (Section R7.1)

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TABLE OF CONTENTS Paae EX E C UTIV E S U M M ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . ii TA BLE O F CO NT E NTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv OPERATIONS .................................................1

. 01 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 0 General Comments . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 0 Preparation for Ref ueling . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 03 Operations Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . 2 0 New Electronic Operating Log (71707) . . . . . . . . . . . . . . . . . . . . 2 07 Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 -

0 General Office Review Board Concerns (71707) . . . . . . . . . . . . . 2 0 Independent Onsite Safety Review Group Technical Specifications Assessment (71707) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 08 Miscellaneous Operations issues (92901) . . . . . . . . . . . . . . . . . . . . . . . . . 3 0 (Open) Violation 50-219/97-06-05b(EA 97-421; 06014) and Violation 50-219/97-03-01 a . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 11. M AI NT EN A N C E . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 M1 Conduct of Maintenance ....................................4 M Maintenance Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 4 M Surveillance Activities ...............................4 M2 Maintenance and Material Condition of Facilities and Equipment . . . . . . . . . 5 M Emergency Service Water Pump Replacement . . . . . . . . . . . . . . . 5

. M 'B' Reactor Recirculation Pump Trips . . . . . . . . . . . . . . . . . . . . . 6 j M Diesel Generator Battery Replacement . . . . . . . . . . . . . . . . . . . . . 7 ^

M8 Miscellaneous Maintenance issues (92902) . . . . . . . . . . . . . . . . . . . .. . . . 8 M (Closed) Violation 50-219/97-06-05a(EA 97-421; 05014) and Violation 50-219/9 7-03-01 d . . . . . . . . . . . . . . . . . . . . . . . . . . . 8

lli . EN G I N E ERI N G . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 9 E2 Engineering Support of Facilities and Equipment ....................9 l E Main Condenser Vacuum Degradation . . . . . . . . . . . . . . . . . . . . 9 E Reactor Water Cleanup (RWCU) Isolation Valve Failure to Fully Stroke

...............................................9 E Diesel Generator Switchgear Seismic Nonconformance .......11 E8 Miscellaneous Engineering Issues .............................12 E (Closed) Licensee Event Report 9 7- 1 3 . . . . . . . . . . . . . . . . . . . 1 2 IV. PLA NT S U PP O RT . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 2 R1 Radiological Protection and Chemistry (RP&C) Controls . . . . . . . . . . . . . . 12 R General Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 R Radiological Sampling and Boundary Control . . . . . . . . . . . . . . . 13 R5 Staff Training and Qualifications in RP&C Activities . . . . . . . . . . . . . . . . . 14 R Continuing Training Program Review ....................14 iv

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R7 Quality Assurance in RP&C Activities ..........................15 R Radiation Protection Self-Assessments . . . . . . . . . . . . . . . . . . . 15 R8 Miscellaneous Radiological & Chemistry Issues (84750,92904) . . . . . . . . 16 R (Closed) Violation 5 0-219/9 7-1 1 -03 . . . . . . . . . . . . . . . . . . . . . 16 R (Closed) Unresolved Item 50-219/98-01 -04 . . . . . . . . . . . . . . . 16 R (Closed) Violat'on 5 0-219/9 7-10-09 . . . . . . . . . . . . . . . . . . . . . 16 R (Closed) Violation 5 0-219/9 7-10-10 . . . . . . . . . . . . . . . . . . . . . 16 R (Closed) Violation 5 0-219/9 8-04-01 . . . . . . . . . . . . . . . . . . . . . 17 R (Closed) Violation 5 0-219/9 8-04-0 2 . . . . . . . . . . . . . . . . . . . . . 17 R (Closed) Follow-up Item 50-219/9 8-04-03. . . . . . . . . . . . . . . . . 17 S1 Conduct of Security and Safeguards Activities . . . . . . . . . . . . . . . . . . . . 17 S General Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 17

- M AN AG EM ENT M EETIN G S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 X1 Exit Meeting Summary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 18 ATTA C H M E NT 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1 9 ATT A C H M E N T 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 0 ATTA C H M E N T 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 21 ATTA C H M E N T 4 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 2 3 l

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Report Details Summarv of Plant Statyg The period began with the unit at 70% power for planned maintenance on the 'C'

l feedwater and condensate pumps. Following these maintenance activities, operations 1 increased reactor power to 100% on June 14. On July 3, the 'B' recirculation pump tripped and power decreased to 92%. Later on July 3, operators increased power to 100%. On July 10, the 'B' recirculation pump tripped again and power decreased to 92%.

On July 11, operators increased power to 100%. Operators maintained the unit at or near full' power for the remainder of the period, except for periodic power reductions to

, backwash the main condensers.

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' OPERATIONS 01 Conduct of Operations 01.1 General Comments (71707)

The inspectors conducted frequent reviews of ongoing plant activities and operations using the guidance in NRC inspection procedure 71707. The inspectors observed plant activities and conducted routine plant tours to assess equipment conditions, indications of operator I work-arounds, procedural adherence and compliance with regulatory requirement I Operators conducted control room activities in a professional manner with staffing levels above those required by Technical Specifications. The inspectors verified operator knowledge of ongoing plant activities, the reason for any lit annunciators, safety system alignment status, and existing fire watches. The inspectors also routinely performed q

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independent verification from the control room indications and in the plant that safety

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system alignment was appropriate for the plant's current operational mode.

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01.2 Preparation for Refuelino l Insoection Scoce (60705)

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I in preparation for refueling outage 17R, the inspectors conducted a review of procedures and administrative requirements associated with refueling operations and related activities. The inspectors reviewed portions of the new fuel receipt inspection and observed receiving and processing of the new fue j

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! Observations and Findinos Personnel used three point communications and appropriate procedures. Operators exercised great care to prevent damage to fuel assemblies and channels. Plant staff initiated several corrective actions reports documenting scratches or gouges 10 two fuel channels and a fuel assembly. Reactor engineering placed a hold on use of these components pending a General Electric (GE) review. Radiation control technicians provided good support during fuel handling activities. Radiological controls and operations management provided oversight on various occasions during the activities. The plant staff also addressed minor problems with refueling I

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equipment prior to proceeding with new fuel handling activities and identified a posting deficiency concerning the area around the new fuel, ~ Conclusions -

Operations conducted safe and high quality fuel handling activities in preparation for the refueling outage. _ Radiological controls effectively supported refueling activitie In addition, management provided good oversigh Operations Procedures and Documentation 0 New Electronic Operatina Loa (71707)

l At 12:00 a.m. on July 6,1998, operations transitioned to a new electronic log format. Due to good planning and thorough prior testing, operators experienced a bumpless transfer and effectively implemented the new format. Operations designed the electronic log to replace paper logs kept in the main control room, l

radwaste control room, and chemistry labs. Electronic log entries include Shift

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Technical Advisor (STA) comments and Technical Specification (TS) limiting conditions for operation (LCO) tracking. The inspector noted detailed log entries

! and good TS tracking and awarenes Quality Assurance in Operations 07.1 General Office Review Board Concerns (71707)

On' June 10 and 11,1998, the General Office Review Board (GORB) conducted their bi-monthly meeting at Oyster Creek. In response to Oyster Creek staff's presentations on plant status, significant events, and safety-related activities; the

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GORB demonstrated a good safety focus in requesting that senior plant managers

! respond directly to some of their concerns. The GORB concerns involved tritium l' inventory and production, material condition and potential impact on nuclear safety, and specific administrative guidance for minor maintenance. The GORB technical

. consultant properly documented the GORB concerns and initiated corrective action reports (CAPS 876,877 and 878) to promptly engage the station's corrective action proces .2 Independent Onsite Safety Review Group Technical Specifications Assessment (71707)

L On July 1,1998, the Independent Onsite Safety Review Group (IOSRG) completed Assessment Report 1940-1998-002, Assessment of the OCNGS Technical Specifications. The IOSRG reviewed TS in order to identify any areas where operational practices could allow operators to violate specifications and to identify procedure enhancements to ensure TS compliance. Primarily, the IOSRG -

assessment team conducted a line-by-line TS review and compared TS requirements to operating procedures. IOSRG conducted a thorough assessment and identified many value-added improvements that included proposed TS changes, procedure

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revisions, enhanced TS interpretations, and process changes. The team leader promptly' documented TS-related deficiencies and engaged the Corrective Action Process (CAP) via CAP 943. Although the assessment identified areas for improvements, IOSRG concluded that the overall condition of the TSs and supporting procedures was satisfactory and did not present safety or licensing concern Miscellaneous Operations issues (92901)

08.1 (Ocen) Violation 50-219/97-06-05b(E,, 97-421: 06014) and Violation 50 219/97-03-01 a: Inadvertent shutdown cooling (SDC) system pump trips occurred on April 24,1997, and August 3,1997, due to failure to properly align pressure switches during system startup. Several problems resulted in improper SDC suction pressure switch lineups, which in turn, resulted in inadvertent trips of individual SDC pump These problems included poor communications and shift turnover, and reliance on extensive manipulation of the low suction pressure switches due to an operator work-around (no time delay on the suction pressure switch). Subsequently, a safety evaluation was written to address the addition of a time delay relay for the low suction pressure trip. This technical issue and prior NRC open item were addressed and closed in NRC Inspection 50-219/97-11 (Section E8.5, Unresolved item 50-219/95-09-03). GPUN installed the modification during a SDC system outage in March 1998. However, the modification is not fully operational as it was not tested. Testing is planned to be performed upon SDC system startup during the next refueling outage (Fall 1998), or earlier if the system is required to be placed in service for an unplanned shutdow There were also some significant procedure weaknesses that contributed to the low suction pressure switch mis-alignments. GPUN had made several procedure changes since the above two SDC pump trip events. The inspector reviewed the current revisico of system procedure 305, Shutdown Coo /ing System Operation, and identified i,ome remaining inconsistencies regarding the positioning of the low suction pressure switch isolation valves. The system engineer was knowledgeable of the deficiencies and plans to address them in the near ter The inspector concluded that the licensee made progress in resolving the issues that contributed to the SDC pump trip events. However, additional work remains to be completed on the recent modification and the procedure 305. In particular, the time delay relay for the SDC pump low suction pressure trip needs to be tested to demonstrate operability and procedure 305 alignment inconsistencies need to be resolved to ensure proper as-left and as-found configuration of the low suction pressure switch isolation valves. Pending completion of these activities, these items remain ope _ _ _ -_-. __

4 ll. MAINTENANCE l M1 Conduct of Maintenance l

M1.1 Maintenance Activities (62707)

The inspectors observed selected maintenance activities on both safety-related and

, non safety-related equipment to ascertain that the licensee conducted these

activities in accordance with approved procedures, Technical Specifications, and appropriate industrial codes and standards. The inspectors observed all or portions of the following job orders (JO)
  • JO 00520406 Generalinspection and Cleaning of RPS MG Set 1-1
  • JO 00525938 Remove / Replace 'A' ESW Pump
  • JO 00523738 Unplug Drywell Cavity
  • JO 00523959 Isolation Condenser Level Loop Calibration
  • JO 00525692 Refueling Mast Grapple Repair Maintenance personnel obtained approval for work and conducted activities in accordance with approved job orders and applicable technical manuals and instructions. Personnel appeared knowledgeable of the activities and observed appropriate safety precautions and radiological practices.

l M1.2 Surveillance Activities (61726)

The inspectors performed technical procedure reviews, witnessed in-progress

surveillance testing, and reviewed completed surveillance packages. They verified l that the surveillance tests were performed in accordance with Technical Specifications, approved procedures, and NRC regulations. The inspectors reviewed all or portions of the following surveillance tests

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I * 606.4.002 Cleanup System Valve Operability and in-Service Test

  • 636.4.003 Diesel Generator Load Test l

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e 617.4.002 CRD Exercise and Flow Test and IST Cooling Water Header Check

  • 604.4.016 Torus to Drywell Vacuum Breaker Operability and in-Service Test
  • 680.4.001 Alternate Shutdown Monitoring Instrumentation Channel Check Persennel used the appropriate procedure, obtained prior approval, and completed applicable prerequisites. Personnel used properly calibrated test instrumentation, observed good radiological practices, satisfied technical specification requirements, and properly documented test results. Qualified technicians conducted the tests and appeared knowledgeable about the test procedur M2 Maintenance and Material Condition of Facilities and Equipment M2.1 Emeraency Service Water Pumo Replacement Inspection Scope (62707)

l On July 1,1998, the 'A' emergency service water (ESW) pump exhibited degraded l performance during its monthly operability surveillance. The inspector evaluated the response to the degraded conditio Observations and Findinas On July 1, the ESW system engineer promptly initiated CAP 949 identifying the degraded condition and noted that even though the 'A' ESW pump flow met acceptance criteria, the pump discharge pressure and amps decreased by 20% and flow decreased by 10% since the last performance of the test. The engineer recommended an investigation, as soon as possible, and began preparations for pump replacement. Plant management supported the ESW system engineer and directed planning to expedite the pump replacemen I Planning processed the change to their on-line maintenance (OLM) schedule as a j

' redirect', an emergent activity forced into the schedule, that may require l rescheduling of other activities, but one that can wait for a planner to thoroughly evaluate the deficiency and properly plan the job. Work week managers properly utilized procedure 2000-ADM-3022.01,On-Line Maintenance Risk Management, to evaluate probability risk assessment functions, OLM restrictions, and TS LCO Planning required that maintenance tag both ESW pumps ('A' & 'B') in system I out of service to protect plant personnel during the actual pump replacement; however, they recognized that returning the 'B' ESW pump to service as soon as possible reduced the overall risk. For example, operators removed the 'A' and 'B' ESW pumps from service at 7:30 a.m. on July 7 and returned the 'B' ESW pump to service at 11:01 p.m. on July 8 (while maintenance continued on the 'A' pump).

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Maintenance technicians effectively used the ESW pump replacement procedure to control and properly document the activity. Maintenance supervisors provided good oversight and used the corrective action process to document a technician's concern regarding a missing phenolic washer (CAP 959). Maintenance conducted an appropriate post-maintenance test. Following a successful operability surveillance,' operators declared the 'A' ESW pump operable at 11:30 p.m. on July I Conclusions 1 The emergency service water (ESW) system engineer demonstrated good system

ownership and led the efforts to promptly address degraded performance on an

ESW pump Planning performed a thorough risk analysis and appropriately adjusted the work week schedule to support the ESW pump replacement. Maintenance activities'were well controlled and properly documented.

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! M 'B' Reactor Recirculation Pumo Trios l Inspection Scope (62707)

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The inspectors reviewed the scope of maintenance on the 'B' reactor recirculation

!' pump after its return to service following two trips on July 3 and July 1 Observations and Findinas

' The maintenance department performed troubleshooting after each trip; however, l these efforts did not identify any failed components or circuits. Following the second trip on July 10, the maintenance department performed more extensive

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c5cks, including disassembly of the exciter, replacing brushes us necessary, and

. replacing the diode wheel.' Additionally, the maintenance department brought in a vendor representative to assist in the troubleshootin The inspectors reviewed two engineering evaluations concerning the 'B' reactor recirculation pump. These evaluations determined that the pump appeared to be operating normally, and would not present any control or transient conditions on the operation of the reactor. Maintenance installed monitoring equipment to capture data needed for further root cause analysis should the pump trip again. Technical specification 3.3.F allows one recirculation pump and loop to be secured and procedures are in place to control the plant in the event of another pump tri Conclusions Maintenance performed thorough and well-documented troubleshooting following two trips of the 'B' reactor recirculation pump. Engineering provided good support l and evaluation of the data.

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M2.3 Diesel Generator Batterv Replacement ' Insoection Scope (62707,717071 l

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! The inspector's observed maintenance's planning and execution of a planned l emergency diesel generator (EDG) battery replacement. In addition, inspectors

! assessed operators' performance relative to TS awareness and LCO tracking with an inoperable EDG.

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. Observations and Findinos

Planning used good risk insights to evaluate and properly schedule an on-line l maintenance activity involving an EDG battery replacement. For example, planning did not schedule conflicting activities during the EDG maintenance window and

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ensured that all necessary replacements parts were available and ready for use.

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At 6:53 a.m. on July 14,1998, operators removed No. 2 EDG from service and entered the appropriate LCO (TS 3.7.C.2). Maintenance technicians used JO N , Diesel Generator Unit #2 Starting Batteries, to effectively control the j group replacement of the EDG battery bank. Maintenance supervisors provided good oversight of the battery replacement. Following installation, maintenance placed the new battery on a float charge, performed procedure 636.2.006, Monthly l

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Diesel Generat. 'atteryInspection, and completed procedure 636.2.012, Diesel Generator Battent; Service Test. Following satisfactory completion of the above, operators conducted an EDG operability surveillance and declared No. 2 EDG

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operable at 8:15 a.m. on July 17.

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l On July 24, maintenance again conducted the monthly battery surveillance on No. 2 EDG to ensure the new battery continued to function as designed. Through this increased monitoring, technicians identified an apparent degraded condition on cell No. 22. Cell No. 22 specific gravity indicated 1.184, whereas the surveillance requirement specified 1.190. Technicians initiated CAP 1024 and operators l promptly declared No. 2 EDG inoperable at 2:00 p.m. Previously on July 24, l operators had declared No.1 EDG inoperable due to a switchgear unit seismic

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concern (see Section E2.3). Operators remained aware of TS requirements and appropriately entered TS LCO 3.7.C.3 for two inoperable EDGs. Operators also l demonstrated a good safety focus and TS awareness concerning core spray components connected to No. 2 EDG as the 'D' core spray pump (powered from No.1 EDG) was concurrently out of service for planned maintenance.

l l Maintenance placed the No. 2 EDG battery on an equalizing charge and monitored cell voltages and specific gravities. Later on July 24, cell No. 22 specific gravity indicated greater than 1.190 and operators declared No. 2 EDG operabl Subsequent engineering evaluation of the condition attributed the apparent low specific gravity reading to electrolyte stratification due to water added to the batteries on July 17, as water level approached the minimum level following the battery service test. Individual cell voltage (including cell No. 22) measurements taken, following discovery of the apparent low specific gravity in cell No. 22, a

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indicated a fully charged battery. Engineering determined that for the measured cell voltages, the specific gravity of the acid in and around the battery plates must be 1.260 or greater for the EDG lead acid batterie Although the license believed that both EDGs remained capable of performing their safety function, management demonstrated a good risk perspective and conservative, safety-conscious decision making in directing the organization's efforts to restore the affected components to operability in a timely manner. At 5:23 p.m. on July 24, operators restored the 'D' core spray pump to an operable status. Operators declared the No.1 and No. 2 EDGs operable at 7:31 p.m. and 7:50 p.m., respectively. During the periods of EDG inoperability, operators maintained good TS awareness, properly tracked all applicable LCOs, and completed LCO action statement requirement Conclusions Maintenance properly planned and effectively controlled a diesel generator battery replacement. Maintenance demonstrated a good safety focus in conducting increased monitoring following battery replacement and identified an apparent degraded condition on one of the battery cells. Maintenance took prompt and appropriate corrective action to address the condition. In addition, operators demonstrated good technical specification awareness, properly documented ,

applicable limiting conditions for operation and met all TS requirements during '

periods of dieselinoperabilit M8 Miscellaneous Maintenance Issues (92902)

M8.1 (Closed) Violation 50-219/97-06-05a(EA 97-421: 05014) and Violation 50-219/97-03-01 d: These two violations involved performance issues during conduct of a routine surveillance and resulted in non-conservative setpoints for radiation monitors that initiate the standby gas treatment system. Specifically, instrument technicians incorrectly interpreted the radiation monitor lo0arithmic scales, and failed to self-check and peer-check their actions during testing activities. In response to the first event (January 22,1997), GPUN's corrective actions primarily focused on human performance. However, the actions were not effective in preventing a second occurrence on July 9,1997. Following the second occurrence, GPUN implemented broader corrective actions, including providing refresher training to allinstrument technicians on reading logarithmic scales, self-checking and peer-checking practices. Engineering also modified the surveillance procedure to change the radiation monitor setpoints such that the new value corresponds to a discrete marking on the logarithmic meter face and to add a meter face drawing to assist interpretation of the logarithmic scale. Finally, instrument technicians and supervisors were given additional training on conducting pre-job briefings. The inspector reviewed the above actions, which included a review of related self-check and pre-job brief training material and surveillance procedures, and determined that the corrective actions were acceptable. Those actions have been effective in preventing ' additional similar occurrences. Based upon satisfactory completion of these actions, these items are closed, t

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lit. ENGINEERING

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E2 Engineering Support of Facilities and Equipment E Main Condenser Vacuum Degradation 1 j

l Insoection Scoce (37551)

The inspectors conducted a review of engineering actions with respect to finding

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the causes of degraded vacuum in the 'B' main condenser. The inspectors evaluated the impact of the degraded vacuum on the control room staff's operation of the facilit Observations and Findinag Plant staff suspected air in-leakage to the 'B' main condenser since December l 1997. The engineering department coordinated with the operations and maintenance departments to ennduct SF6 testing to identify leakage. Engineering identified several minor leaks in March and April 1998, but did not find the main source of the leakage. Engineering developed an action plan to conduct further testing during the next planned power reduction in June 1998. Maintenance identified and corrected additional leakage, but air leakage into the condenser l continued. Engineering developed a more detailed plan to continue searching for the source of air in-leakage. The engineering department demonstrated effective involvement with other plant organizations. They took the lead to coordinate the activities to locate the source of the leakage, while minimizing the need to manipulate the plant to conduct their action pla Conclusions Engineering led a well-controlled and effectively coordinated troubleshooting effort to evaluate a main condenser degraded vacuum conditio E2.2 Reactor Water Cleanuo (RWCU) Isolation Valve Failure to Fully Stroke Inspection Scooe (71707)

The inspector reviewed the licensee's act'ons, troubleshooting and evaluation for two occurrences of a RWCU isolation valve failing to fully stroke.

j Observations and Findinas i

l On July 6, while closing the RWCU isolation valve (V-16-1) to perform a surveillance test, the valve's indicating lights went from full open indication to intermediate indication. The valve stopped somewhere before the full closed indication. Control room operators responded appropriately to declare the valve inoperable per TS 3.5.A.3 and took the proper actions to isolate the affected i

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I 10 penetration. Troubleshooting efforts determined the t the root cause of the failure of the valve to indicate closed was a high resistance (onnection in the valve's motor control center. Maintenance repaired the connect'on and subsequent valve strokes indicated the valve performed acceptably. Operr.tions declared the valve operable on July On July 19, while isolating the RWCU system in preparation for restoring the 'B'

reactor recirculation pump, the valve had intermediate indication agai Approximately one hour after the operators properly isolated the RWCU system, the l dual indication cleared indicating the valve was fully closed. Troubleshooting of the j valve control circuit did not reveal any evidence of a valve control logic proble However, on this occasion, maintenance tested the valve under conditions that

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better matched the plant's conditions at the time of the anomaly (with the 'B'

recirculation pump out of service and a letdown path established to the condenser).

I Maintenance conducted the previous tests with the valve isolated from the letdown flow pat Results of testing conducted under troubleshooting action plan 98-42 indicated that the valve experienced an increase in load towards the end of the valve strok Engineering evaluated the valve as being fully operable and capable of performing its safety related function. However the valve's existing torque switch setting will not ensure full valve closure under differential pressures that exist when the valve isolates the RWCU svstem under higher flow conditions with the 'B' recirculation pump out of service. Engineering determined that the valve's torque switch setting did not result in unacceptable valve leakage (Appendix J) for containment isolation scenarios as the limit switch setting ensured flow isolation and the valve eventually l fully seated (torque switch relaxed) as the differential pressure across the valve

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decrease Engineering planned to conduct additional testing to further evaluate this

. phenomenon. The inspectors questioned the connection of a motor power monitor (MPM) to an operable valve. At the end of the inspection period, engineering continued to evaluate the effect of test equipment installed on operable safety-related equipment. The inspectors also had questions concerning values

. engineering used in their valve thrust calculations supporting valve operability. This will remain an unresolved item pending NRC review of engineering's evaluation of MPM impact on valve operability and engineering's valve thrust calculations. (URI 50 219/98-05-01).

During the time the RWCU system was isolated, reactor water sulfate, chloride and conductivity parameters exceeded administrative action limits. Chemistry performed a review of an engineering evaluation on the effect of these chemistry parameters exceeding action levels. They determined that it was acceptable to continue to operate at the current levels, but recommended the initiation of a plant shutdown if sulfate levels exceeded 100 pp !

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i 11 Conclusions Control room operators responded appropriately to a failure of a containment isolation valve to fully stroke while isolating the reactor water cleanup system on two occasions. Engineering demonstrated good involvement to evaluate the i problem with the valve and continued to evaluate the root cause at the end of the period. The chemistry department provided good support to evaluate chemistry parameter action levels and to provide recommendations concerning continued operation without the reactor water cleanup syste E2.3 Diesel Generator Switchaear Seismic Nonconformance Inspection Scooe (37551. 71707) I On July 24,1998, engineering identified a condition adverse to quality involving an j emergency diesel generator (EDG) switchgear seismic nonconformance. The

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inspector assessed the actions in response to this condition, Observations and Findmas

.I During an inspection to assess the amount of corrosion on the No.1 EDG base channel support, engineers identified that two interior support channels did not make contact (maximum gap 7/16") with the concrete floor beneath the EDG's switchgear. In addition, inspection of the inside of the switchgear unit revealed three bolts missing from the roof structure connection and one bolt missing at the base wall connection. Engineering determined that these deficiencies invalidated the seismic verification of the EDG switchgear unit and required immediate attention. Engineering promptly initiated CAP 1020 to address the condition and determined that the No. 2 EDG was not similarly affected. Senior Reactor Operators declared the EDG inoperable, made a timely and appropriate 10 CFR 50.72 notification for discovery of a condition that placed the switchgear outside of its seismic design basis (EN 34574), and informed the NRC resident inspecto Engineering provided appropriate guidance to maintenance concerning bolt replacement and torque requirements. Engineering developed and properly evaluated a temporary modification (TMOD 98-48) to install steel shims to close the gap between the base of the channel and the concrete floor. Engineering determined that the shims provided a positive load path for vertical dead and seismic loads to the concrete floor. The shims also restored the stiffness of the base channels and eliminated potential secondary impact and amplification of the switchgear structure during a seismic event. Plant staff planned to re-assess TMOD 98-48 during the Fall 1998 refueling outage when maintenance plans to clean,

. Inspect, and coat the support channel Later on July 24, maintenance completed installation of TMOD 98-48 and replaced the missing bolts. Engineering conducted a seismic qualification walkdown of the switchgear unit and operators declared No.1 EDG operable. Failure to ensure adequate seismic support for the No.1 EDG is a violation of 10 CFP. Part 50,

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Appendix B, Criterion lli (Design Control). This non-repetitive, licensee-identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy. (NCV 50-219/98 05-02)

j Conclusions l

, Engineering demonstrated a good questioning attitude in identifying and l documenting a diesel generator switchgear seismic nonconformance. Engineering led the efforts to disposition operability issues, analyze and resolve the i nonconformance, address deportability, and implement appropriate corrective actions.

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E8 Miscellaneous Engineering issues (90712)

l E (Closed) Licensee Event Reoort 97-13: Reactor Building Ventilation Ductwork May i Not Meet its Seismic Design Basis Since Original Construction. Details for this issue are described in NRC Integrated inspection Report 50-219/97-09 Section E2.4. The issue related to the licensee's identification that the associated supply ventilation ductwork, which comprises a portion of the secondary containment boundary, may be in conflict with seismic design bases. The LER remained open while the licensee conducted a detailed review to identify all relevant design and licensing basis information to make their final determination, and to determine whether hardware modifications would be required. On June 30,1998, the licensee irJormed the NRC via a docketed letter that LER 97-13 had been canceled as their subsequent review determined that the event was not ' reportable. Based on calculations and ventilation walkdowns, engineering concluded that the existing installation of the reactor building ventilation supply ducts ensures that the secondary containment boundary would be maintained if a design basis seismic event occurred. Licensing determined that the seismic adequacy of the reactor building ventilation supply ductwork is consistent with the design basis as stated in Facility Description and Safety Analysis Report (FDSAR), Volume I, Section V-2. Inspector in-office review of the licensee's letter dated June 30,1998, and FDSAR Sections V-2.3.1 and V-2. concluded that the condition was not outside of the Oyster Creek design basi This LER is closed and cancele IV. PLANT SUPPORT R1 Radiological Protection and Chemistry (RP&C) Controls R1.1 General Observations (71750)

During radiologically controlled area (RCA) tours the inspectors observed that technicians posted proper warning signs, personnel wore appropriate dosimetry, personnel conducted adequate radiological monitoring of personnel and materials leaving the RCA, and technicians maintained monitoring instrumentation functional and in calibration. Technicians maintained radiation work permits (RWPs) and survey status boards up-to-date and accurate, monitoring instruments were l

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l i 1 functional and in calibration. They observed activities in the RCA and verified that personnel complied with the requirements of applicable RWPs, and that workers remained aware of the radiological conditions in the area.

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l R1.2 Radiological Samolina and Boundary Control

Insoection Scoce (83750)

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The inspector reviewed the programs for: (1) air sampling and analysis for

, radioisotopes within the plant; (2) control of radiological boundaries; and, (3) review l of certain radiological work.

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l In the area of radiological sampling at the work place, the review included:

l procedures for sample collection and analysis (Procedure Nos. 6630-ADM-4212.01 l and 6630-QAP-4224.21); basis for determining average alpha, beta and gamma i derived air concentrations; calibration and control of counting instruments utilized in air sampling and analysis; and, inclusion of sampling results into individual dose-of-records. In addition to reviewing procedures and records, interviews with cognizant l l licensee personnel were also conductod.

L in the area of the control of radiological boundaries, direct observation of various plant areas within the radiologically controlled areas (RCAs) was conducted to evaluate compliance with NRC requirements contained in 10 CFR 20 Subparts G, I and J. Special attention was focused on the control of access to posted high, locked high and very high radiation area Review of radiological work included direct observations of work in progress within the RCA, including the shipment of spent resin to a waste disposal site and the receipt and initial inspection of fresh (non-irradiated) fuel. Specific attention was focused on the control of work by the radiation protection staff, including surveys, postings and As Low As Reasonably Achievable (ALARA) instructions.

! Observations and Findinas

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Air Samolina and Analysis l

l The review of the program, including procedures, basis documents, sample results and calibration records 1. dicated that the licensee had established an appropriate program for air sampling and analysis, it was noted that the basis calculations performed in 1993, for determining isotopic mix within the plant were conducted based on samples collected in 1991, and that the licensee did not establish by procedure or practice the periodic re-verification of the plant isotopic mix. The inspector performed an independent assessment of the current isotopic mix and its impact on the average alpha, beta and gamma derived air concentrations (DACs),

and determined that no significant changes had occurred, on average, but that certain specific isotopes within the average had changed by upwards of an order of magnitude since the licensee had performed its basis calculations. Since the average values had not significantly changed, the licensee was in compliance with _ . _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - - - _ _ _ _ _ _ _ _

the NRC requirements found in 10 CFR 20.1501. The licensee's Radiological l Engineering Manager indicated that periodic verification of the isotopic mix _would be performed in the future. (IFl 50-219/98-05-03)  ;

Radiological Boundaries / Conduct of Radiological Work l

Tours of various portions of the RCA indicated that the licensee has properly '

established boundaries for contaminated areas, and high, locked high and very high '

radiation areas throughout the facility. Appropriate postings as required under c 10 CFR 20, Subpart J, were in plac !

l On June 16,1998, the licensee shipped a 14-215 liner from its facility to a licensed {

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land disposal facility. Direct observation of the liner shipment, contained in an NRC-approved shipping cask for which a certificate of compliance has been issued, including the performance of radiological surveys for the release of the vehicle from j the RCA, was made. Licensee controls for this work evolution were determined to i be appropriate. In addition, an assessment of the shipping paperwork to verify i compliance with 10 CFR 71.5 was also performed. All documents reviewed were  !

determined to be in accordance with the applicable portions of 10 CFR Parts 20,61 and 71 and 49 CFR. Also during this inspection period, the licensee received new j (un-irradiated) fuel from a vendor in preparation for the 17R refueling outag l Appropriate receipt surveys were conducted in accordanco with 10 CFR 20.190 l Additionally, new fuel inspections and work in the spent fuel pool in preparation for i the 17R refueling outage were observed to have appropriate radiological control Conclusions

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The licensee established an adequate program for the sampling, analysis and )

assessment of personnel exposure to airborne radionuclides. A programmatic  !

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weakness was identified regarding the lack of reverification of the basis documents for this area since 190 !

Appropriate controls for high, locked high and very high radiation areas were established. Control of radiological work for the shipment of a radwaste liner, receipt of radioactive materials (new fuel), and spent fuel pool work were also i effectiv R5 Statf Training and Qualifications in RP&C Activities R 5.1 Continuina Trainina Prooram Review Insoection Scope (83750)

Review of the licensee's continuing training program included selected training documents, lesson plans and training objectives, interviews with cognizant health physics and training personnel, and direct observation of two sessions of radiation protection technician continuing training.

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___ _ _ _ _ _ _ _ _ --__ O a 15 Q_ observations and Findinas Continuing training for health physics technicians is provided several times each calendar year. For 1998, the licensee was conducting a portion of its spring training cycle during the time of this inspection, which included three days of training. The training sessions audited included presentations on the 17R refueling j outage and on the detection and decontamination of tritium (reactor coolant tritium activity has increased during the current operating cycle). Attendees included one shift of radiation protection technicians and their supervisor, while the presentations were made by the ALARA coordinator and one member of the technical training staf Conclusions The presentation of continuing training to radiation protection technicians was l<

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appropriate with regard to scope and depth of presentation for two classes reviewed during the inspection perio R7 Quality Assurance in RP&C Activities l l

R Radiation Protection Self Assessments Insoection Scope (83750)

Records of audits and appraisals in support of the annual requirement to review the radiation protection program content and implementation in accordance with 10 CFR 20.1101 and self-assessments performed by various members of the Radiation Protection and Nuclear Safety staff during 1997 and 1998 were reviewed.

I f Observations and Findinas For 1997, the licensee conducted a series of radiation protection self-assessments in addition to safety assessments performed by the Nuclear Safety Assessment (NSA) Department. The licensee also performed a summary self-assessment (Audit No. 6610-PA-97-005)to ensure that all of the programmatic areas within the l_ radiation protection program were properly evaluated in support of meeting the requirements contained in 10 CFR 20.1101 for an annual review of the radiation

! protection program. While the various self assessments and safety assessments l~ were determined to be of appropriate scope and technical depth, this summary self-l assessment was determined to be very limited in scope and depth of review. For example, only portions of the ALARA and radiological postings programs were included in the audit. Additionally the assessment indicated that a review of previously preformed Radiological Controls Department self assessments was also cart of the audit; however none of the Radiological Engineering assessments were reviewed during this audi Each assessment reviewed was performed by knowledgeable technical specialists or qualified auditors. The NSA Department performs a biennial audit of the radiation protection program, with the next audit scheduled to begin later this year, n

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l l 16 l Conclusions The annual review of the radiation protection program was adequate. Weakness in the scope and depth of the summary self-assessment for 1997 was noted; however j the sum total of self assessments and safety assessments met the requirements of j 10 CFR 20.110 R8 Miscellaneous Radiological & Chemistry issues (84750,92904)

A Region i inspector reviewed the licensee's implementation of corrective actions for several previously identified violations and concerns as discussed belo l

R8.1 (Closed) Violation 50-219/97-11-03: Failure to follow procedures. The licensee submitted its corrective actions in response to this violation by letter dated April 10, l 1998. As part of this inspection, independent verification of the corrective actions I submitted was conducted. All corrective actions proposed were appropriately implemented. This item is close R8.2 (Closed) Unresolved item 50-219/98-01-04: Contaminated clothing outside the RCA. The licensee initiated a corporate-level investigation which was unable to determine the owner of the contaminated clothing. Licensee tests and records confirmed that the portal monitors throughout the plant had the ability to detect the level of contamination found on the clothing and that this level was sufficient to cause an alarm actuation. The clothing was removed to a secure storage location to await disposal as radwaste. Based on the location of the clothing and the level of contamination present, this event had little or no safety consequence to the public health and safety. While this event represents a failure by the license to properly controllicensed material, this failure constitutes a violation of minor significance and is not subject to formal enforcement actio H8.3 (Closed) Violation 50-219/97-10-09: Failure to establish and implement adequate Radiation Monitoring System (RMS) calibration guidance. The inspector reviewed ;

the' calibration techniques used in a recent calibration of the overboard discharge l RMS. The proceduralinadequacies pertinent to this RMS had been corrected and no further proceduralinadequacies were noted. The licensee planned to continue to address RMS calibration procedural inadequacies pertaining to other RMS and complete all actions by December 199 R8.4 (Closed) Violation 50-219/97-10-10: Failure to establish and implement procedures to verify that the design basis relative to air balance affecting the augmented off gas building, new radioactive waste building, and turbine building area ventilation systems, as described by the UFSAR, were maintained at a negative pressure. The licensee has established calibrated differential pressure gauges in each of these three buildings and the equipment operators monitored the differential pressure during their rounds. The licensee self-identified that this issue also pertained to the Old Radioactive Waste (ORW) building and had completed similar corrective actions for the ORW building by the time of the inspection.

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R8.5 (Closed) Violation 50-219/98-04-01: Failure to perform and document a safety evaluation in accordance with 10 CFR 50.59 to support the use of radiologically contaminated water (tritium) from the condensate transfer system for makeup to the I shell side of the isolation condenser system during standby conditions, instead of the demineralized water transfer system (a normally non radioactively contaminated system), as described in the design bases. The licensee took action to administratively preclude the use of the condensate storage tank as the makeup source of water for isolation condenser evaporative losses. Special training j pertaining to tritium was provided to operations staff to help increase both their 1 awareness on tritium and their ability to maintain radioactive effluents releases ALARA. As described in NRC Inspection Report No. 50-219/98-04,the projected l' doses to the public constituted a small fraction of the regulatory limit. Therefore,

although no safety evaluation had been performed prior to using condensate transfer water for makeup, recent evaluations show no safety consequence for the chang R8.6 (Closed) Violation 50-219/98-04-02: Failure to provide for the monitoring, sampling, and analysis of tritium vapor released from the isolation condenser system L to unrestricted and controlled areas in order to demonstrate compliance with the dose limits for individual members of the public. The licensee calculated past doses from the isolation condenser pathway (see NRC Inspection Report 50-219/98-04 section R1.1) and established a program which provided for the monitoring,

. sampling, and analysis of tritium vapor released from the isolation condenser system. The licensee also planned to take additional actions such as repairing the leaking valve seat in the upcoming refueling outage to reduce the amount of tritium released from this pathway.

l l R8.7 (Closed) Follow-uo item 50-219/98-04-03: Regulatory Guide 1.21 compliance pertaining to 1995 and 1996 tritium releases from the isolation condenser system and licensee projected dose calculations to the public. The inspector discussed the ;

implications of current and past tritium releases from the isolation condenser release pathway as it relates to the annual effluent release report (required by Regulatory Guide 1.21). The inspector described means by which past annual effluent release ,

reports could be amended by the licensee's staff. However, projected dose to the l public for 1995 and 1996 were found to be minimal, and therefore the updating of I the annual report was unnecessar S1 . Conduct of Security and Safeguards Activities

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l S1.1 General Observations (71750)

During routine tours, the inspectors noted that security controlled vital and protected area access in accordance with the security plan, properly manned security posts, locked or guarded protected area gates, and maintained isolation zones free of obstruction . _ _ _ _ _ _ _ - _ - - _ _

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18 MANAGEMENT MEETINGS X1 Exit Meeting Summary The inspectors provided a verbal summary of preliminary findings to senior licensee ,

management on June 19, July 22, and July 30,1998. During the inspection i period, inspectors periodically discussed preliminary findings with licensee management. Inspectors did not provide any written inspection material to the licensee. No proprietary information is included in this repor l l

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ATTACHMENT 1 PARTIAL LIST OF PERSONS CONTACTED Licensee (in alphabetical order)

G. Busch, Manager, Nuclear Safety & Licensing S. Levin, Director, Operations and Maintenance D. McMillan, Director, Equipment Reliability K. Mulligan, Plant Operations Director J. Perry, Plant Maintenance Director M. Roche, Director, Oyster Creek D. Slear, Director, Configuration Control l R. Tilton, Manager, Assessment NRC (in alphabetical order)

L. Eckert, Radiation Specialist J. Furia, Senior Radiation Specialist T. Hipschman, Resident inspector S. Pindale, Senior Resident inspector (Hope Creek)

J. Schoppy, Senior Resident inspector

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ATTACHMENT 2 INSPECTION PROCEDURES USED Procedure NA Title 37551 Onsite Engineering 60705 Preparation For Refueling 61726 Surveillance Observation 62707 Maintenance Observation  ;

71707 Plant Operations 71750 F e at Support 83750 Occupational Radiation Exposure 84750 Radioactive Waste Treatment, and Effluent and Environmental l

Monitoring 92700 Onsite Followup of Written Reports of Nonroutine Events at Power 1 Reactor Facilities 90712 Inoffice Review of Written Reports of Power Reactor Facilities 92901 Followup Operations 92902 Followup - Maintenance 92903 Followup - Engineering 92904 Followup - Plant Support 93702 Onsite Event Response

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l ATTACHMENT 3 l ITEMS OPENED AND CLOSED l

Opened l

Number Tvoe Description 98-05-01 URI Engineering's evaluation of MPM impact on valve operability and engineering's valve thrust calculations. (E2.2)

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98-05-03 IFl Periodic verification of the isotopic mix l for air sampling and analysis. (R1.2) ]

Opened \ Closed I i l Number Tvoe Description j

98-05-02 NCV Emergency diesel generator switchgear seismic nonconformance. (E2.3)

Closed Number Tvoe Description 97-03-01 d VIO Non-conservative setpoints for radiation monitors that initiate the standby gas I treatment system. (M8.1)

97-06-05a VIO Inadequate corrective action in response (EA 97-421; 05014) to non-conservative setpoints for radiation monitors that initiate the standby gas treatment system. (M8.1)

97-10-09 VIO Failure to establish and implement adequate RMS calibration guidanc (R8.3) l

'97-10-10 VIO Failure to establish adequate oversight i over the New Radioactive Waste Building, 3 Turbine Building, and Augmented Off Gas Building Ventilation Systems. (R8.4)

97-11-03 VIO Failure to properly maintain and review ;

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required records for the free release of '

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materials from the radiologically

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controlled area. (R8.1)

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t 22 98-01-04- URI Radiologically contaminated clothing found outside the RCA. (R8.2)

98-04-01 VIO Failure to perform and document a safety evaluation in accordance with 10 CFR 50.59. (R8.5)

I 98-04-02 VIO Failure to survey for releases from the isolation condenser system. (R8.6)

-98-04-03 IFl Regulatory Guide 1.21 compliance pertaining to 1995 and 1996 tritium releases from the isolation condenser system and licensee projected dose

! calculations to the public. (R8.7)

Closed / Canceled I

97-13 LER Reactor Building Ventilation Ductwork l

May Not Meet its Seismic Design Basis l

Since Original Construction. (E8.1) ,

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Discussed Number Tvoe Description

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97-03-01 a VIO Inadvertent shutdown cooling system

pump trips due to failure to properly align l pressure switch during system startu (08.1)

97-06-05b VIO Inadequate corrective action in response I (EA 97-421; 06014) to shutdown cooling system pump trip (08.1) l l

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ATTACHMENT 4 LIST OF ACRONYMS USED l

l ALARA As Low As Reasonably Achievable CAP Corrective Action Process CFR Code of Federal Regulations DACs Derived Air Concentrations DRP Division of Reactor Projects DRS Division of Reactor Safety , EDG Emergency Diesel Generator l

ESW Emergency Service Water l FDSAR Facility Description and Safety Analysis Report GORB General Office Review Board l GPUN General Public Utilities (GPU) Nuclear IOSRG Independent Onsite Safety Review Group IST In-Service Test JO Job Order

LCO Limiting Conditions for Operation LER Licensee Event Report l MPM Motor Power Monitor NRC Nuclear Regulatory Commission NRR Office of Nuclear Reactor Regulation NSA Nuclear Safety Assessment OCNGS Oyster Creek Nuclear Generating Station OLM On-Line Maintenance ORW Old Radioactive Waste PDR Public Document Room QA Quality Assurance RCA Radiologically Controlled Area RMS Radiation Monitoring System RP&C Radiological Protection and Chemistry RWCU Reactor Water Cleanup RWP Radiation Work Permit

'SDC Shutdown Cooling STA Shift Technical Advisor TMOD Temporary Modification TS Technical Specification UFSAR Updated Final Safety Analysis Report ,

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