IR 05000219/1997002

Insp Rept 50-219/97-02 on 970224-0413.Violations Noted. Major Areas Inspected:Operations,Engineering,Maintenance & Plant Support

ML20141J547
Person / Time
Site:	Oyster Creek
Issue date:	05/19/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20141J502	List: IR 05000219/1997002 ML20141J498 ML20141J522
References
50-219-97-02, 50-219-97-2, NUDOCS 9705280141
Download: ML20141J547 (34)

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U. S. NUCLEAR REGULATORY COMMISSION

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Report N Docket N ; License N DPR-16 Licensee: GPU Nuclear incorporated -

1 Upper Pond Road 1 Parsippany, New Jersey 07054 Facility Name: Oyster Creek Nuclear Generating Station Location: Forked River, New Jersey

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inspection Period: February 24 - April 13,1997 inspectors: Stephen M. Pindale, Senior Resident Inspector (Temporary)

3- Lonny L. Eckert, Radiation Specialist Jason C. Jang, Senior Radiation Specialist Joseph L. Nick, Reactor Engineer (Temporary)

Blake D. Welling, Project Engineer (Temporary)

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Approved By: Peter W. Eselgroth, Chief Projects Branch No. 7 i

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e 1 EXECUTIVE SUMMARY Oyster Creek Nuclear Generating Station Report No. 97-02 This integrated inspection included aspects of licensee operations, engineering, maintenance, and plant support. The report covers about a seven-week period of inspectio This report identifies four issues as apparent violations, which appear in several functional'

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areas. They are all related to configuration control issues whereby system, component or overall plant configuration and status have not been properly maintained. Station personnel, particularly in the operations department, were attentive to self-identify and document several of the issues, resulting in prompt attention to effect further review and evaluation of the specific issue !

l Plant Operations

• Operations personnel failed to identify and address the full effect on the pressure suppression capability of the torus due to the planned on-line maintenance activity, resulting in a configuration in which the torus pressure suppression capability was degraded. Once it was identified, after the specific configuration had existed three separate times, the licensee implemented appropriate interim corrective actions. A I

poor safety review of a related Preventive Maintenance Change Request contributed to this event and represented an additional challenge to tin operators that processed the clearance. (01.2)

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• Nuclear Safety Assessment's efforts in identifying and evalunting a suspected adverse trend as related to switching and tagging problems were very good. Their concerns were communicated to an appropriate level of maragement for resolutio (07.1)

• The " Analysis of Occurrence and Safety Assessment" section of Licensee Event Report 97-03, " Suppression Pool Bypass Flow Created During Preventive Maintenance Due to inadequate Safety Review," was weak.in that it did not discuss in sufficient detail the potential safety consequences as a result of the reported condition. Rather, it focused on the " exposure" to risk, noting that the duration was short (about 10 minutes). (08.1) 1 l

Maintenance

• Routine maintenance and surveillance activities observed by the inspectors were conducted safely and in accordance with station procedures. (M1.3)

• The licensee has been challenged by repeated performance problems with several systems and components (e.g. reactor recirculation control system, station air compressors, and 4 kV cables) that were due to various causes. Substantial efforts by the various station departments have been necessary to address these continued I

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problems, which may divert their efforts from other important activities. The licensee continues to attempt to reach resolution for the individual problems. (M1.4)

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' The failure to meet technical specification requirements related to the trip setpoints for the reactor building ventilation exhaust radiation monitors was'an apparen violation, and the causes for the mis-calibration appear to be symptomatic of -

broader configuration control weaknesses. (M4.1)- ,

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Operational performance problems continued with the reactor recirculation pumps l

} and motor-generator sets. Substantial engineering efforts, including engineering ;

l~ support from an industry expert, provided new insights related to additional causes !

! and corrective actions for these problems. Additional work is expected to occur on '

. the remaining pump and motor-generator set systems. The inspector concluded

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_ that the licensee's efforts in their continued troubleshooting were appropriate.

! Continued attention is warranted to ensure all potential causes for system / pump j transients are identified and corrected. (E1.1)

l l e- An engineer inappropriately manipulated (closed) a heating boiler exhaust damper j while it remained administratively closed, and was an apparent violation of the

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switching and tagging requirements of procedure 108, " Equipment Control." The

. licensee initiated prompt actions to address the specific concerns.~(E4.1)

e Several recent performance problems have occurred that were related to the quality E j'

and completeness of licensee safety determinations and safety evaluations. The licensee's subsequent efforts to evaluate and address these problems have been

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appropriate to date. Followup training and the issuance of a safety review j newsletter properly highlighted the performance weaknesses and provided adequate

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guidance for improved performance. (E7.1)

i j e The licensee failed to conduct a complete 10 CFR 50.59 safety evaluation to

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determine if an unreviewed safety question existed for the removal of the isola * ion i condenser radiation monitors. As a result, they failed to recognize that the two i

radiation monitors satisfied a UFSAR commitment. This is characterized as a j deviation. (E8.1)

!" Plant Sucoort e The licensee effectively implemented the radiation protect lon and security program (R1.1, S1.1)

e The licensee implemented adequate radioactive liquid and gaseous effluent control programs, sufficient to protect public health and safety and the environment. (R1.2)

e The calibration methodology for effluent and process radiation monitoring systems, including calibration data evaluation, was not consistently implemented, and it was unclear which industry standards and NRC Regulatory Guides were used to develop the licensee's calibration procedures. The issue is an unresolved item. (R2.1)

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• Surveillance tests related to air cleaning and ventilation systems were acceptabl However, the licensee's identification that the augmented offgas building pressure was not being maintained as per the UFSAR is an unresolved item. (R2.2)

• The licensee failed to conduct an adequate 10 CFR 50.59 safety evaluation to determine if an unreviewed safety question existed for the removal of the isolation condenser radiation monitors, and is a violation of NRC requirement * The areas toured by the inspectors were well maintained and radiological housekeeping was generally very good. Administrative controls were adequate to caution and inform workers regarding radiological conditions. (R2.4)

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• The licensee provides appropriate monitors to measure and alarm in the event of a criticality accident in areas where special nuclear material is handled, used, or

stored. The emergency procedures were adequate to ensure the safety of personnel in these areas. Personnel attended training and drills to familiarize them ,

with the emergency procedures. (R2.5)

* The Offsite Dose Calculation Manual contained sufficient specification, information, and instruction to implement and maintain the radioactive liquid and gaseous e

effluent control programs. Associated procedures were easy to follow, and training of personnel was good. (R3)

• The licensee met the quality assurance audit requirements. The licensee maintained good quality control programs for the chemistry measurement laboratory. (R7)

• During the inspection period, the licensee identified that there was no charcoal in l the control room as specified in the UFSAR for use in the iodine instrumentation (for accident conditions). The licensee is investigating this issue, which is an -

unresolved item. (R8.1) i

• The licensee demonstrated very good overall response during the annual emergency response exercise. Some minor problems were encountered with communication lines, but overall communications were good. A list of areas for program improvement were developed after the exercise. Overall, technical support center staff performance was assessed by the NRC inspectors as good. (P1.1)

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TABLE OF CONTENTS EAIN.

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, EX EC UTIV E S U M M ARY . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . i TAB LE O F CO NTENTS . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . iv OPERATIONS (40500, 62707, 71707, 93702) . . . . . . . . . . . . . . . . . . . . . . . . . . .1 4 01 Conduct of Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 1

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01.1 General Comments ....,............................ 1-

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O 1.2 Suppression Pool Water Bypass Pathway Created During 3

Maintenance Activity Due to inadequate Review of System 1

Line-up (eel 5 0-219/9 7-0 2-01 ) . . . . . . . . , . . . . . . . . . . . . . . . . 1

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07 Quality Assurance in Operations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 3 07.1 Nuclear Safety Assessment Identification and Review of Multiple Switching & Tagging issues (eel 50-219/97-02-02) .... 3 4 08 Miscellaneous Operations issues ............................ 5 -!

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ll. MAINTENANCE (617 26, 62707) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5'

j M1 Conduct of Maintenance .................................. 5 4 M1.1 Maintenance Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 5 M1.2 Surveillance Activities . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 6 M1.3 Maintenance and Surveillance Ac.tivities Conclusions ......... 6

. M4 Maintenance Staff Knowledge and Performance . . . . . . . . . . . . . . . . . . 6 M4.1 Reactor Building Ventilation Exhaust Radiation Monitor Mis-calibration Due to Personnel Error (eel 50-219/97-02-03) ...... 6 111. ENGINEERING (71707,37551,92903) ............................... 8 E1 Conduct of Engineering ............... ................... 8 E1.1 Inadvertent Reactor Recirculation Pump Trip While Operating at Full Power Following Erratic Pump Performance . . . . . . . . . . . . . 8 E4 Engineering Staff Knowledge and Performance . . . . . . . . . . . . . . . . . . 10 E Heating Boiler Tagged-Closed Damper Opened by Unauthorized Individual (eel 50-219/97-02-04) ...................... 10 E7 Quality Assurance in Engineering Activities . . . . . . . . . . . . . . . . . . . . 11 E Licensee Followup for Safety Evaluation / Determination Performance Weaknesses . . . . . . . . . . . . . . . . . . . . . . . . . . . . 11  : ,

E8 Miscellaneous Engineering issues . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 l IV. PLANT SUPPORT (71707, 71750, 82701, 84750, 92904, 93702) . . . . . . . . . . 12 R1 Radiological Protection and Chemistry Controls ................. 12 R General O bservations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 12 R1.2 Implementation of the Radioactive Liquid and Gaseous Effluent Control Program s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 13 R2 Status of RP&C Facilities and Equipment . . . . . . . . . . . . . . . . . . . . . . 14 R Calibration of Effluent / Process Radiation Monitoring System and Flow Measuring Devices (Unresolved item 50-219/97-02-05)............................................ 14 :

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R2.2 Surveillance Tests for Air Cleaning and Ventilation Systems i (Unresolved item 50-219/97-02-06) .................... 17

. R2.3 Removal of the isolation Condenser Radiation Monitors (Violation 5 0-219/97-02-07) . . . . . . . . . . . . . . . . . . . . . . . . . . 18 R2.4 Tours of Plant Are as . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 19 R2.5 Criticality Radiation Monitors ......................... 20 R3 RP&C Procedures and Documentation . . . . . . . . . . . . . . . . . . . . . . . . 21

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R6 RP&C Organization and Administration ....................... 22 R7 Quality Assurance in RP&C Activities ........................ 22 R8 Miscellaneous RP&C issues ............................... 23 !

R8.1 Review of Updated Final Safety Analysis Report (Unresolved item 50 219/97-02-09) ............................. 23

, R8.2 (Closed) IFl 50-219/9 6-09-0 6 . . . . . . . . . . . . . . . . . . . . . . . . . 23 R8.3 (Closed) LER 9 7-002 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 23 R8.4 Periodic Report Review ............................. 24 i P1 Conduct of Emergency Preparedness Activities ................. 24 ;

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P1.1 Emergency Response Drill ........................... 24 i S1 Conduct of Security and Safeguards Activities . . . . . . . . . . . . . . . . . . 25 j S1.1 General Observations . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 l

V. M AN AG EM ENT M EETI NG S . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 i X1 Exit Meeting Summ ary . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 25 ATT A C H M E NT 1 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 26

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ATTA C H M E NT 2 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 27 ATTA C H M ENT 3 . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . - . . . 28 l

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Report Details Summarv of Plant Status The plant was operated at full power during this inspection period (February 24 - April 13, 1997) except for period March 7 through 11,1997. During that time, power operation was reduced to 95% due to the inadvertent trip of the "B" reactor recirculation pump. The licensee investigated the pump trip and adjusted several components tssociated with the pump's motor-generator set before returning to full power operatio OPERATIONS (40500, 62707, 71707, 93702)

01 Conduct of Operations'

01.1 General Comments The inspet,ter:: 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, personnel safety hazards, procedural adherence, and compliance with regulatory requirement Control room activities were found to be well controlled and conducted in a professional mannac 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 independent verification from the control room indications and in the plant that safety system alignment was appropriate for the plant's current operational mod .2 Sucoression Pool Water Bvoass Pathway Created Durina Maintenance Activity Due to Inadeauate Review of System Line-uo (eel 50-219/97-02-01) Inspection S, cope (40500. 62707. 71707)

On March 11,1997, the licensee identified that the pressure suppression feature of the suppression pool (torus) had been inadvertently degraded during planned maintenance activities. The inspector reviewed the controlling job order, system drawings and an associated preventive maintenance program change request. The inspector also interviewed operatic.ns tad maintenance personnel involved with planning and performing the msintenance activit ' Topical headings such as 01, M8, etc., are used in accordance with the NRC standardized reactor inspection report outline. Individual reports are not expected to address all outline topic ,

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b) Observations and Findingg A preventive maintenance (PM) activity was planned and implemented for containment. spray system valve V-21-18 (a 4 inch torus spray valve) on March 11

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1997.' The valve branches from the containment spray system header and - ,

discharges to the air space of the torus. Concurrently, another. containment spray I system header branch, via 6 inch flow test valve V-21-17 remained open during the maintenance. V-21-17. discharges to one of the drywell downcomers. With V-21-17 and V-21-18 open simultaneously, some steam, in the event of a design basis ' ]

accident, would enter the containment spray system piping though V-21-17 I (downcomer) and flow through V-21-18, and ultimately discharge back to the torus

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air volume, effectively bypassing the torus water volume for pressure suppressio The above condition existed for a total time of about 10 minutes. The valve (V-21-18) was cycled (opened and closed) a total of three times for the PM and post- -J maintenance testing activities. The PM activity was a motor operator inspection for !

V-21-18. Prior to the maintenance, the containment spray system had been in it's 1 normal (standby) alignment, in which V-21-17 was open and V-21-18 was close The taggout left V-21-17 open and allowed the maintenance workers to cycle V-21-18 as part of the PM activity. This'PM task had previously pe-formed only while- !

the plant was shutdown and the above syst3m alignment not required for pressure j suppression capabilit The licensee reviewed this event and dMermined thet the root cause was an inadequate safety review (February 1996) of a change that moved the PM (N .

9441M) from.an outage task to an on line task. Previous to this occurrence, this !

activity was performed during refueling outages. As a result of the incomplete 4 l safety review, appropriate precautions weiei not added to the controlling job order and maintenance procedure. The technl cal staff personnel (maintenance  ;

assessment) that reviewed the PM Change riequest for PM 9441M did not have I specific operating knowledge to identify the concern. The r>:ensee also concluded that the control room staff did not identify the potentlO ' " > lypass flow path during development, review and approval of switching urucra to implement the PM activity. The inspector reviewed the associated PM Change Request and confirmed 1 il to be of poor quality for failing to identify the potential to bypass the torus water

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The inspector reviewed the details associated with this event. One of the day shift ,

senior reactor operators identified this event at around 5:00 p.m., after V-21-18 had l already been cycled three times. Upon discovery, V-21-17 was tagged closed for further maintenance and stroking of V-21 18. The licensee reported this condition i to the NRC as per the reporting requirements of 10 CFR 50.72.'

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The inspector found that this potential ficWpath was discussed among some of the onshift SROs during the day that the PM was being conducted. However, their tocus was largely on whether the PM activity represented a pathway between the i primary and secondary containments. Tt;ey correctly concluded that primary l containment integrity was not compromhed by the work, however, they failed to .

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3 recognize the actual torus water bypass concern. The SRO's efforts in returning to the site after his shift had ended to further evaluate (and then report) this event were noteworthy.

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The inspector reviewed procedure 108, " Equipment Control," and identified several areas relevant to this event. Step 9.1.1 (Isolation Switching Order Preparation)

requires the control room operator to determine the appropriate isolation boundaries for the work activities using Attachment 108-6 (Guidelines for isolation Boundaries). j ltem 2.4 of Attachment 108-6 states that the local valve operator does not need to j be tagged for motor-operator valve actuator work (as was the case for this !

maintenance). However, the attachment further states that the effects on the i system if the valve is physically moved during the work shall be considered and additional isolation boundaries shall be added to the Outage as appropriate. In

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addition, step 10.1 of procedure 108 requires the licensed operations supervisor to review the switching order for compatibility with license requirements, station

operating conditions, and temporary modifications to the plant's configuration. The licensee's failure to properly implement the equipment control requirements of procedure 108 is an apparent violation of NRC requirements. (eel 50 219/97-02- !

01) J Conclusions Operations personnel failed to identify and address the full effect on the pressure suppression capability of the torus due to the planned on-line maintenance activity, and this is an apparent violation of NRC requirements. A poor safety review of the PM Change Request by maintenance assessment personnel contributed to this event and represented an additional challenge to operators processing the clearanc Once it was identified, after the specific configuration had existed three separate

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times., the licensee implemented appiopriate interim corrective actions.

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07 Quality Assurance in Operations )

l 07.1 Nucluar Safety Assessment identification and Review of Multiole Switchina &

, Taaaina isspes (eel 50-219/97-02-02) Insoection Scoce (40500. 62707. 71707)

The licensee's Nuc! ear Safety Assesement (NSA) identified that several deviation 4 reports related to switching and tagging had occurred since July 1995. NSA subsequently initiated a deviation report (DR), which was assigned to the Operations department for review. The inspector reviewed the associated NSA monthly report and related documentation, interviewed NSA personnel and independently assessed the available informatio . - . - - - -

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~ Observations and Findinas in response to several DRs initiated since the beginning of 1997, NSA commenced a detailed review of the DR database to identify whether broader weaknesses or trends related to switching and tagging existed at Oyster Creek. As a result of their review, NSA initiated a separate DR (97-107) on February 14,1997, which identified a total of 23 DRs since July 1995 that involved either process or i implementation problems during switching and tagging activities. The DR was {

assigned to Operations to respond to the issues, and it reported that the large !

number of DRs was indicative of an adverse trend which requires elevated I management attention.

NSA assigned an initial root cause classification as managerial methods; corrective actions from prior deviations not effective at preventing recurrence. Although operations management did not agree with NSA's determination, operations

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acknowledged the development of an adverse trend in 1997. Operations management stated that they have initiated several corrective actions to improve performance in this area. Some of the actions include 1) implementing independent verification of allisolation and restoration switching orders,2) implementing changes to the computerized tagging program (TRIS) to provide enhanced i inforrnation to operators regarding the status of components, 3) and providmg refresher training to operators on switching and tagging activities and processes, including management's expectations regarding performance.

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The inspector reviewed the listing of the 23 DRs that were listed by NSA. NSA -

placed the DRs in three general categories; development, execution, and proces The majority of them were in the development and execution categories. Some of the DR examples include the followin * Three control rods were valved out of service for maintenance on January 25-26,1997, however, the technical specification surveillance requirement (4.2.D) was not considered, which requires a test of remaining control rods within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. This iequirement was avoided by three hours (control rods returned to service), and the DR was characterized as a "near miss "

• An ambiguous system outage alignment resulted in establishing an improper flowrate for cooling water to a turbine building closed cooling water system heat exchanger (January 22,1997).

• 125 Vdc panel "D" was inadvertently de-energized while tagging an associated normal power source to panel "D" due to de-energizing an incorrect DC supply breaker (October 9,1996).

At the end of the inspection, DR 97-107 remained open. The licensee's efforts in identifying and documenting the individual deficiencies were good, as was NSA's identification of the apparent adverse trend. The inspector further considered these problems to be indicative of possibly larger configuration control issues (e.g. overall system / component status) on a site-wide basis. As such, the numerous switching

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and tagging deficiencies for which corrective actions have not been fully effective is - ;

un apparent violation. (eel 50-219/97-02-02)'

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.NSA's efforts in identifying and evaluating a suspected adverse trend as related to i switching and tagging problems were very good. Their concerns wer ;

communicated to an appropriate level of management for resolution. -The collective l

. issue of configuration control is being considered separately on a broad basi Miscellaneous Operations issues 08,1- (Ooen) Licensee Event Report (LER) 97-03: Suppression Pool Bypass Flow Created During Preventive Maintenance Due to inadequate Safety Review. This event is discussed in detailin Section 01.2 of this report. The proposed corrective actions for this_ event appear appropriate and will be reviewed during a subsequent  :

inspection.- The inspector concluded that the " Analysis of Occurrence and Safety Assessment". section of this LER was weak in that it did not discuss in sufficient

' detail the potential safety consequences as a result of this align nent. Rather, it focused on the " exposure" to risk, noting that the duration was short (about 10 minutes). The inspector discussed this LER weakness with the licensee, who state that they would further review and evaluate the issue. Pending completion of the licensee's review and subsequent LER revision, and verification of the licensee's corrective actions, this LER remains ope . MAINTENANCE (61726, 62707)

M1 Conduct of Maintenance M1.1 Maintenance Activitiet Insoection Scope (62707)  !

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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 portions of the following job orders (JO): 1 o JO 512916 Move Control Rod Blade Guides in the Fuel Pool; o JO (Various) No. 2 Emergency Diesel Generator Preventive / Corrective !

Maintenance Activities (System Window); and l i

e JO (Various) Troubleshoot and Repair "B" Reactor Recirculation Pump and l Motor-Generator Set Tn l

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i- . Observations and Findinas The inspectors concluded that the above activities had been approved for performance and were conducted in accordance with approved job orders and <

applicable technical manuals and instructions. Personnel performing the activities f were knowledgeable of the activities being performed and were observing appropriate safety precautions and radiological practice M1.2 Surveillance Activities  ; Insoection Scope (61726) l

The inspectors performed technical procedure reviews, witnessed in-progress surveillance testing, and reviewed completed surveillance packages. They verified

that the surveillance tests were performed in accordance with Technical Specifications, approved procedures, and NRC regulations. The inspectors reviewed

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j all or portions of the following surveillance tests-l

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• 612.4.001 " Standby Liquid Control Pump and Valve Operability and Inservice Test;"  !

• 617.4.001 " Control Rod Drive Pump (B) Operability Test;" and ,

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• 636.4.003 " Diesel Generator (2) Load Test." Observations and Findinas  ;

A properly approved procedure was in use, approval was obtained and prerequisites were satisfied prior to beginning the test. Surveillance test instrumentation was properly calibrated and used, radiological practices were adequate, technical specifications were satisfied, and personnel performing the tests were qualified and knowledgeable about the surveillance test procedur M1.3 Maintenance and Surveillance Activities Conclusions The maintenance and surveillance activities observed by the inspectors were conducted safely and in accordance with station procedure M4 Maintenance Staff Knowledge and Performance M4,1 Reactor Buildina Ventilation Exhaust Radiation Monitor Mis-calibration Due to Personnel Error (eel 50-219/97-02-03) Insoection Scope (61726. 71707)

On February 28,1997, the licensee found that a January 1997 calibration left both reactor building (RB) ventilation exhaust radiation monitors set non-conservatively high, which would have resulted in a delayed actuation of the standby gas

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treatment system (SGTS). The inspector reviewed the licensee's followup i activities, including the associated formal critique sessio I l Observations and Findinas I Control room operators were manually starting the SGTS using the trip test function )

of the RB ventilation exhaust radiation monitors, when they noticed the system did )

not start at the expected setpoint of about 13 mR/hr. They subsequently l determined that the A-1 and A-2 monitor setpoints were at 30 mR/hr and 40 mR/hr, i respectively. The operators promptly declared the two radiation monitors inoperable, and followed the applicable technical specification (Table 3.1.1, item J.2) action statement (isolate the RB and run the SGTS).

On March 3,1997, the licensee conducted a formal critique to identify the root

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causes for the common mis-calibration of the two radiation monitors. The '

licensee's initial review identified that the two instruments were mis-calibrated on January 22,1997, as part of a quarterly surveillance (calibration). Their critique determined the root cause to be personnel error in that the action taken was not in accordance with accepted craft practices, expected expertise, and level of qualit l Specifically, the technician stationed at the meter indications (in a control room i back panel) misinterpreted the logarithmic scale on both radiation monitor module i He interpreted the as-found readings to be about 101/2 compared to the actual'and j expected 13 (mR/hr), and then adjusted each of them about three marks past the l

"10" mark. However, on the logarithmic scale, three marks past '10" is "40." j l

The inspector attended the March 3,1997, critique and reviewed the associated critique report. The critique meeting was effectively conducted. The report noted several observations as well as the root cause. The most significant observations were 1) the same technician correctly read logarithmic scales earlier that same day during other instrument calibrations, 2) the location / position of the lead technician (also in the control room but mostly located at a different panel) precluded his visual observation of the meter indications, and 3) neither technician located in the control l room questioned the common out of specification (as-found) readings, especially l considering that these instruments typically have not demonstrated drift problems in the pas The licensee's corrective actions included conducting refresher training to all instrument technicians as related to logarithmic scales, including the critique report in " Events" training for instrument technicians, and conducting a peer review training session (by the technicians involved with this event), stressing self-checking and maintaining a questioning attitud The safety impact of this event was minimal. The licensee analyzed potential offsite radiation dose rates with the higher (incorrect) setpoints, and determined that ;

the results were still well below the 10 CFR Part 20 limit of 2.0 mR/hr. The inspector reviewed the associated control room alarm response procedure and found that it appropriately lists the expected system and operator action . . . .

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Technical Specification 3.1, Table 3.1.1 (Item J.2) requires that the RB ventilation exhaust radiation monitors initiate the SGTS at a trip setting of less than or equal to - ;

4 17 mR/hr. This event relates to overall configuration control of plant systems and .;

, components to the extent that a common error occurred but was not detected by j existing mechanisms (self-check, questioning unexpected as-found results,' and  !

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failure of supervisory procedure review to identify the common apparent

performance degradation for the monitors). ~This is a configuration control issue to i

uthe extent that technicians and surveillance procedure reviewers failed to recognize .

, 'that equipment status and function was compromised, and is an apparent violation i of Technical Specification 3.1. (eel 50-219/97-02-03)'  ! Conclusions The operators were alert to' identify that the trip setpoint for the RB ventilation ,

exhaust radiation monitors were non-conservatively high while manually starting j SGTS. The failure to meet technical specification requirements is an apparent

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' violation, and the causes for the mis-calibration appear to be symptomatic of j broader configuration control weaknesse '

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lit. ENGINEERING (71707,37551,92903)

E1 Conduct of Engineering E1d Inadvertent Reactor Recirculation Pumo Trio While Operatina at Full Power -

Followino Erratic Pumo Performance

. Inspection Scope (37551. 71707) 1 On March 7,1997, while operating at full power, the "B" reactor recirculation (RR)

pump tripped. The inspector monitored plant parameters following ths unexpected >

RR pump and motor-generator (M-G) trip. The system engineer was interviewed to assess the licensee's planned and implemented actions to determine the cause for the "B" M-G set trip, as well as additional recent performance problems with the RR control syste , Observations and Findinas During routine operations, control room annunciators awociated with the "B" loop of the RR system alarmed (pump differential pressure low, and drive motor breaker _

lockout). The operators immediately noticed that the "B" RR pump and M-G set had ,

tripped, and they entered abnormal operating procedure ABN-3200.02,

" Recirculation Pump Trip." As per the procedure, the operators closed the "B" RR loop discharge valve. in addition, core engineering personnel were notified of the pump _ trip. _The overall plant response to the RR pump trip was normal, with reactor power stabilizing at about 95% of full powe . _ , . _ _ ._

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[ tThe "B" RR pump tripped at 10:27 a.m. on March 7. Earlier that day, between  !

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g 2:00 a.m.' and 4:30 a.m., the contiof room operators noted several operational i- anomalies within the "B" RR flow control system. They observed about a 10 MW L E thermal power decrease and a corresponding minor flow reduction in the "B" RR i p flowrate. They also noticed that the amperage indication was oscillating on the "B"  !

p ' RR M-G set, and subsequently placed the "B"_ RR flow controller in manua :

Subsequently, th6 3perators identified that the "B" RR pump flowrate was

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oscillating about 200 gpm (between about 30,800 gpm and 30,600 gpm) in a

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sinusoidal fashion. There was no obvious operations or maintenance activity in .

progress that would have caused the oscillations. A single cycle (full wave) took i

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about five minutes. In response to the above indications, engineering and- ,

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operations personnel discussed RR system performance and possible monitoring and  :

troubleshooting activities. No conclusions were reached regarding the course of  !

l action to take before the RR pump tripped at 10:27 l An additional recent similar problem was identified regarding the "B" RR pump on February 20,1997,'as documented in NRC Inspection 50-219/97-01. At that time, a flowrate reduction of about 1300 gpm was observed over about a 21 hour2.430556e-4 days <br />0.00583 hours <br />3.472222e-5 weeks <br />7.9905e-6 months <br /> perio ]

The' licensee had suspected that the tachometer was the cause for the proble I Also,- at the time, the licensee had noted that the pump speed signal became i

" noisy" during the time of reduced flow. They were continuing to monitor and i troubleshoot the problems at the end of that inspectio Following the "B". RR pump and M-G set trip on March 7,1997, the licensee developed a formal troubleshooting plan (Action Plan 97-06). A recorder was installed to monitor voltage regulator parameters. During the troubleshooting

. activities, the M-G set field breaker open fuses were tagged out to prevent energizing the RR pump with the M-G set drive motor operating.' Actual tachometer generator speed was measured and compared to the indicated spee The licensee brought.'an industry expert onsite, who was experienced with RR pump and M-G set problems, to assist them in troubleshooting the "B" RR system. The individual reviewed prior maintenance and performance history, reviewed voltage regulator adjustment procedures and techniques, and retrieved data from the licensee's troubleshooting effort The industry expert assisted the licensee in adjusting the voltage regulator, which

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provided enhanced stability of the control system. He also identified that the RR control system components appeared less stable when ambient temperature is cold j in combination with the RR pumps operating at high speeds. The industry expert ,

assisted by providing' guidance for " tuning" additional critical components in the  !

control system. While the "B" RR pump remained shutdown, the licensee also  !

replaced the tachometer as an additional precautio The "B" RR pump and M-G operated normally for the remainder of the inspection  ;

period. The licensee was continuing their evaluation of system performance, j

including environmental conditions on the electrical components. They were in the process of evaluating which pumps' controls (voltage regulator) would be adjusted, l

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as well as other system window planned maintenance, during the upcoming power reduction (April 18,1997). Conclusions Operational performance problems continued with the RR pumps and M-G set Substantial engineering efforts, including engineering support from the industry expert, provided new insights related to additional causes and corrective actions for these problems. Additional work is expected to occur on the remaining RR pump systems. The inspector concluded that the licensee's efforts in their continued troubleshooting were appropriate. Continued attention is warranted to ensure all potential causes for system / pump transients are identified and correcte E4 Engineering Staff Knowledge and Performance ,

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E4.1 Heatina Boiler Taaaed-Closed Damoer Opened by Unauthorized Individual (eel 50-219/97-02-04) Inspection Scope (62707,71707)

On March 19,1997, while restoring an existing taggout for the No. 2 heating boiler, operations personnel identified that an individual had inappropriately mispositioned a damper that was part of the taggout boundary. The inspector reviewed the licensee's followup of this event by reviewing relevant documentation and interviewing station personne Observations and Findinas Operations personnel reported that, while removing tags and repositioning components to normal position, the No. 2 boiler discharge damper was partially open. The associated switching and tagging order to support the system outage

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had tagged the discharge damper closed to support boiler stack cleanin Operations personnel subsequently learned that personnel who were supporting a vendor for nozzle tip diffuser replacement had opened the red-tagged and closed damper to aid in natural circulation cooling in the heating boiler firebox. The operations department then submitted a deviation report for further review and followu The project engineer for ongoing heating boiler work was aware of maintenance activity to clean the heating boiler stack. He was also aware that the cleaning activity had been completed and that component restoration was imminen Nevertheless, he had checked the position of the damper and had apparently partially opened it. There were no personnel safety implications of this particular event. However, there was regulatory significance in that the project engineer was not authorized to operate that dampe _ _ _ . _ _ ___ . .____ _ _._ _ _ ___ _ _ _._ _ _ .._ -.

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Procedure.108, " Equipment ' Control," is a general procedure for the control o maintenance and includes the method for obtaining permission and clearance for- !

personnel to work. The stated purpose of the procedure is ensure that switching' !

operations are performed' consistent with equipment protection, nuclear safety

~ l concerns, and regulatory requirements. Step 4.2 (General Practices) of the  ;

procedure requires that the act of positioning components shall be controlled and l coordinated by the operations department. Step 5.2.2 of the procedure requires i that components bearing a red tag shall ngt be operated or activated except as J permitted by Attachment 108 3 (Testing Electrical Components). ' Attachment 108- )

, 3 was not applicable for this occurrence. Failure to implement the above  ;

requiremehts of procedure 108 in an apparent violation. (eel 50-219/97-02-04) {

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The licensee responded promptly to this occurrence. The licensee implemented J

disciplinary actions for the inappropriate actions. In addition, a plan was developed 1 to achieve corrective actions and to prevent recurrenc ) Conclusions The failure to implement the switching and tagging requirements of procedure 108 in an apparent violation of NRC requirements and is a further example of deviating from established system configuration control requirements. Operations personnel were attentive and responsive to the unexpected condition of the heating boiler exhaust damper, and initiated prompt actions to effect further review and actio E7- Quality Assurance in Engineering Activities E2d t_icensee Followuo for Safety Evaluation / Determination Performance Weaknesses , Insoection Scoos (37551. 40500)

The inspector reviewed the licensee's actions to address recent problems concetriind the adequacy of 10 CFR 50.59 safety evaluations and safety determinations.' The inspector interviewed personnel and reviewed relevant training material and other documentatio Observations and Findinas in response to severalissues related to performing 10 CFR 50.59 safety evaluations and safety determinations, as documentec' in several recent NRC inspection reports, the licensee has initiated steps to improve performance. Several of the problems were related to engineering personnel not completing a thorough review of the UFSAR or other design basis information, resulting in incomplete safety review The licensee enhanced the safety review training program to emphasize the need to completely investigate and accurately answer the safety determination screening questions. The training focused on the tools available to the engineers to assist in their' reviews (e.g. computerized UFSAR database). The inspector reviewed the training handout and found it to be useful for the training need . .- . -

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The licensee has established an Oyster Creek safety review process assessment .!

team to assess the overall effectiveness of implementation of the safety

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determination element of the safety review process. The team's charter included 1)

reviewing the specific NRC-cited examples (violations), 2) selecting and reviewing a ,

j sample population of safety determination reviews performed prior to, during, and j 4 since the 16R refueling outage (including those for procedure changes, temporary :

, modifications, troubleshooting, and engineering configuration changes), and 3) l

reviewing' safety evaluation implementation associated with a sample of the above -!

l- safety determinations (from item 2). ,

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" The licensee prepared a " Safety Review Newsletter," issued in April 1997. The ;

j purpose of the newsletter is to inform safety review process users and independent i l'

reviewers of important safety review issues and developments. It was sent to j

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about 700 GPUN safety review process users. The licensee plans to continue to i issue periodic newsletters as a communication tool on the subject matter. The i F inspector reviewed the document and found that it provided useful insights to GPUN

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safety reviewer J j .. Conclusions

1 j' The inspector concluded that the licensee's efforts to evaluate and address several I

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recently identified performance problems associated with safety determinations and

. safety evaluations were appropriate to date. The results of their continuing reviews

[ will be assessed upon completion. Followup training and the issuance of a safety.

I review newsletter properly highlighted the performance weaknesses and provided i' adequate guidance for improved performance.

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E8- Miscellaneous Engineering issues

E (Closed) Unresolved item 50-219/97-01-03: Isolation condenser vent radiation

!- monitors removed from service prior to addressing all relevant UFSAR commitments. This item is administratively closed, and has been re-characterized

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as noted in Section R2.3 of this report.

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I IV. PLANT SUPPORT (71707, 71750, 82701, 84750, 92904, 93702)

I i-l R1- ' Radiological Piotection and Chemistry Controis RL1 General Observations  !

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y During entry to and exit from the radiologically controlled area (RCA), the inspectors J l verified that proper warning signs were posted, personnel entering were wearing

. proper dosimetry, personnel and materials leaving were properly monitored for j radioactive contamination, and monitoring instruments were functional and in  ;

, calibration. During periodic plant tours, the inspectors verified that posted extended I i

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Radiation Work Permits (RWP) and survey status boards were ' current and accurat They observed activities in the RCA and verified that personnel were complying with the requirements of applicable RWPs, and that workers were aware of the

radiological conditions'in the area.

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R12 a Imolementation of the Radioactive Liould and Gaseous Effluent Control Proarams Inspection Scope (84750-01)

i The inspection consisted of: (1) plant tours, including the control room; (2) review of liquid and gaseous effluent release permits; (3) review of airborne tritium quantification techniques; and (4) review of unplanned /unmonitored release  !

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. Observations and Findinas .

i 'The inspectors toured the control room and selected radioactive liquid and gas i processing facilities and equipment including effluent radiation monitors and air l cleaning systems. No inoperable equipment was noted during the tour.

. During review of selected radioactive gaseous effluent discharge permits, the j

, inspectors determined that the discharge permits were complete and met the i

, . Technical Specification /Offsite Dose Calculation Manual (TS/ODCM) requirements )

' for sampling and analyses at the frequencies and lower limits of detection '

established in the TS/ODCM.

i l At the inspectors' request, the licensee demonstrated their capability for monitoring ;

and quantifying the airborne tritium. The licensee calculated the total amount of '

l water loss using the plant makeup water inventory. The licensee assumed a certain

. percent of water loss due to an evaporation from the spent fuel pool and other

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!. components (e.g., air ejector) to the environment via the main stack. The water j vapor released through the main stack. The licensee calculated the airborne tritium j release using waterborne tritium measurement results (e.g., reactor coolant and

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spent fuel pool). Calculated airborne tritium released through the main stack for the

. - fourth quarter of 1996 was to be about 4.1 curies. The inspectors compared this

. value to what was reported in the 1996 Annual Report; during the same quarter, about 5.5 curies of airborne tritium was reported to have been released. The

inspectors stated that the licensee's assumptions and calculation methodologies

were goo '

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At the time of the inspection, the inspectors noted that there were no unplanned /unmonitored releases since the previous inspection conducted in

February 1996, with the exception of slightly contaminated water release on ,

t September 17-18,1996, (See Inspection Report No. 50-219/96-09). l

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c.- Conclusions  :

I Based on'the above review, the inspectors concluded that the licensee generally I maintained and implemented the effluent control programs effectively. It was noted I that the inadvertent release on September 17,1996, was caused by inidequate- )

work process controls and was not'a programmatic deficiency in implementation of

'the effluent controls progra R2 Status of RP&C Facilities and Equipment B2d Calibration of Effluent / Process Radiation Monitorina System and Flow Measurin Devices (Unresolved item 50-219/97-02-05) Insoection Scone (84750-01)

The inspectors reviewed the most recent calibration results and calibration  ;

procedures for the effluent / process radiation monitoring system (RMS) and' effluent !

flow measuring device The inspectors noted that the licensee's technical specifications and UFSAR were not specific in regard to RMS calibration methodology. Therefore, the inspector j applied the following general industry practice attributes to determine calibration - l adequacy:  !

(1) . Electronic calibration results; (2) Plateau checks; 'I (3) Number of calibration sources (and linearity determination); and 1 (4) Evaluation of the secondary calibration data to validate the primary j calibration result i

The inspectors also utilized the following documents as a basis to determine l whether the calibration procedures contained sufficient detail and guidance to verify conversion factors (calibration factors):

e NRC Regulatory Guide 1.21, " Measuring, Evaluating, and Reporting i Radioactivity in Solid Wastes and Releases of Radioactive Materials in Liquid -

and Gaseous Effluents from Light Water Cooled Nuclear Power Plants, i February 1979"; j e NRC Regulatory Guide 4.15, " Quality Assurance for Radiological Monitoring Programs (Normal Operations)- Effluent Streams and the Environment,

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February 1979";

e . ANSI N42,18,1980, " Specification and Performance of On-Site i instrumentation for Continuously Monitoring Radioactivity in Effluents"; and c e EPRI TR-102644, " Calibration of Radiation Monitors at Nuclear Power Plants, March 1994".

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L Observations and Findinos ' I

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The Instrumentation and Controls Department had the responsibility for performing t electronic and radiological calibrations for the radiation monitors and associated ,

l ' flow measurement devices. -The inspectors reviewed the most recent calibration results for effluent and sampler flow measuring devices. All calibration results of 4 the flow measurement devices met the licensee's acceptance criteri !

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The inspectors' review of the most recent calibration results for the effluent / process RMS indicated the following: i e Containment High Range Radiation Monitors (2 channels)

(1) Electronic calibration was performe (2) Plateau check data was not available,

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j (3) One radiation level was used (8.154 R/hr). -

(4) No evaluation of the secondary calibration data to validate the primary calibration results was availabl ')

i e Main Stack High Range Noble Gas Effluent Monitor 'j (1) Electronic calibration was performe :

(2) Plateau check data was not availabl (3) Performed the primary Calibration using two Xe-133 sources (5.521 pCi/cc and 0.5697 Ci/cc) i (4) No evaluation of the secondary calibration data to validate the primary calibration results was availabl e Turbine Building Vent High Range Noble Gas Effluent Monitor (1) Electronic calibration.was performe (2) Plateau check data was not availabl (3) One radiation level was used (8.170 R/hr).

(4) No evaluation of the secondary calibration data to validate the pnmary i calibration results was availabl ) i e Main Stack Normal Range Noble Gas Effluent Monitor (1) . Electronic calibration was performe (2) Plateau check data was not available, but high voltage setpoints were verifie (3) Four solid sources (Ba-133) were use (4) No evaluation of the secondary calibration data to validate the primary calibration results was available, o Turbine Building Vent Normal Range Noble Gas EffluentM' onitor j (1) No electronic calibration data was availabl l (2) Plateau check data was not available, but high voltage setpoints were verifie (3) Three solid sources (Sr-90) were use (4) No evaluation of the secondary calibration data to validate the primary calibration results was availabl l

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e Offgas Building Exhaust Vent Noble Gas Monitor l

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(1). No electronic calibration data was availabl (2) Plateau check data was not available, but high voltage setpoints'were I verifie '(3) Two solid sources (Sr-90 and Ba-133) were use (4) No evalu'ation of the secondary calibration data to' validate the primary calibration results was available.

! Steam Jet Air-Ejector Monitor (2 Channels)

(1) No' electronic calibration data was availabl :(2) Plateau check data was not availabl (3) : Two radiation levels were used (30.7 mR/hr and 123.6 mR/hr).

(4) No evaluation data was availabl e Main Steam Line Radiation Monitors (4 Channels)

(1) No electronic calibration data was availabl (2) Plateau check data was not availabl (3) One Radiation level was use (4) No calibration data evaluation was availabl e Liquid Radwaste Effluent Line Monitor (1) No electronic calibration data was availabl (2) Plateau check data was not available, but high voltage setpoints were verifie (3) ' Two solid sources (Ba-133 and Cs-137) were use (4) No evaluation of the secondary calibration data to validate the primary calibration results was availabl o Reactor Building Service Water Effluent Line Monitor (1) Electronic calibration was performe (2) Plateau check data was not avei!::bie, but high voltage setpoints were ,

verifie .

l (3) Four solid sources (2 Ba-133, Cs-137, and Co-60) were use '

(4) No evaluatinn of the secondary calibration data to validate the primary calibration results was availabl e Turbine Building Sump No.1-5 Radiation Monitor-(1) No electronic calibration data was availabl .!

(2) Plateau check data was not available, but high voltage setpoints were l verifie l (3) Two solid sources (Co-60) were use (4) No evaluation of the secondary calibration data to validate the primary calibration results was availabl l l

With regard to the high range RMS, it should be noted that the NRC has accepted j one point calibration methodology to minimize personnel exposures in order to  ;

effect ALARA. Accordingly, electronic calibration is relied on for RMS calibratio ~

For example, the licensee used one radiation level (8.154 R/hr) for Containment )

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, High Range Radiation Monitors (2 channels), which was acceptable. However, the e licensee's method did not demonstrate and document that the detector response )

was linear in the cross over region between the low and high ranges of RM *

The inspectors noted that licensee calibration practices were inconsistent. It can be

noted from the above that high voltage setpoints were not evaluated in all cases, and various numbers and types of calibration sources were used which could effect

, response linearity. The inspectors questioned the licensee as to how they were

complying with NRC Regulatory Guides 1.21 and 4.15 or how they had established acceptable alternative calibration practices. . The licensee was not able to locate any analysis of these NRC Regulatory Guides prior to the end of the inspection. The

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. inspectors questioned the licensee as to: (1) which standards and N3C Regulatory

Guides were used in the preparation of their calibration procedures; and (2) how

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calibration' data was evaluated. These questions were being' researched by the licensee at the end of the inspection. The licensee informed the inspectors that -!

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other calibration documentation existed, but was not immediately available for

!- review.

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l The inspectors also noted that the licensee did not perform trending or tracking of 1 i- the secondary calibration factors to the primary calibration factors. Consequently, j the inspectors were not able to validate the primary conversion factors which were used for alarm setpoint calculations of the effluent monitors.

i-s- Conclusions

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Based on the above reviews, the inspectors determined the following conclusions:

• The licensee's RMS calibration practices were not consistent, and i

e- It was also not clear which industry standards and NRC Regulatory Guides

. were used in the development of the licensee's calibration procedures.

Based on these conclusions, the inspectors stated that this matter was considered to be unresolved pending further more comprehensive reviews. This unresolved l item consisted of: (1) review of the licensee's positions on NRC Regulatory Guides

! 1.21, 4.15 and other relevant guidance, and (2) review of the above RMS l

electronic and radiological calibration records back to 1990. (URI 50-219/97-02-06)

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BL2 Surveillance Tests for Air Cleanina and Ventilation Systems (Unresolved item -

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50-219/97-02-06)

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- Insoection Scooe (84750-01)

The inspectors reviewed the licensee's

(1) most recent surveillance test results, !

and (2) performance summaries to determine the implementation of technical specification (TS) requirements and UFSAR commitments for the following systems:

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• Standby Gas Treatment System (TS requirement),

• TurLine Building Ventilation System (UFSAR),

• Old and New Radwaste Building Ventilation Systems (UFSAR), and I

• Offgas Building Ventilation System (UFSAR). Qbservations and Findinos i Test methodologies for the SGTS were good and surveillance test results were within the licensee's TS acceptance criteri Turbine building air flow tests were performed and test results were within the established criteria. Offgas building and the new radwaste building HEPA tests were performed and the test results were within the established criteri I With regard to plant air balance, the licensee had not retrieved all of the requested information prior to the end of the inspection. After the inspection, the inspectors were informed that the licensee f.ad initiated a deviation report on April 7,1997, pertaining to the discovery that the differential pressure being maintained in the AOG building was not being maintained at -0.25"W.G. as per UFSAR Section 9.4.4.2.3. These matters will'oe the subject of further review (URI 50-219/97-02-06). Conclusions Filtration test results were acceptable. However, this area requires further review to determine the licensee's conformance with regulatory requirement R2.3 ftemoval of the Isolation Condenser Radiation Monitors (Violation 50-219/97-02-07) Inspection Scoce (84750-01. 92904)

The inspection consisted of: (1) plant tours; (2) review of the licensee's 10 CFR 50.59 analysis supporting removal of the isolation condenser system radiation monitors; (3) review of the licensee's NUREG-0737 Action item II.K.3.14 commitment; (4) review of the NRC safety evaluation (a December 12,1981, letter to the licensee) pertaining to NUREG-0737 action item II.K.3.14 commitment; (5)

Operator Training Ler, son Plan No. 69, " Process Radiation Monitors", (December 8, 1980); (6) Section 3.2.4.5 of the UFSAR, " Operating Philosophy" (Isolation Condenser), Revision 0; and (7) other relevant documents such as engineering memorand Current licerisee isolation condenser tube leak detection capabilities are assessed in Section R3 of this repor .. -. -. - - ..~ - - . - - - - - - - . - - -- - . - . . _ .

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19 [ Observations and Fin &eas

b A permanent change (December 1995) was made to the facility, as described in the l safety analysis report (SAR) involving the removal of the isolation condenser radiation monitor without considering all relevant portions of the Updated Final Safety Analysis Report (UFSAR) to determine if an unreviewed safety question

) existed. Specifically, the inspectors noted that Section 1.9.31 of the UFSAR, a j section containing commitments regarding the isolation condenser system radiation monitor, was not reviewed during the conduct of the subject 10 CFR 50.59 i' assessment and documented by a' written safety evaluation which provided the basis that the change did not involve an unreviewed safety question.

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Title 10 CFR 50.59, " Changes, Tests and Experiments", section (a)(1), states that the holder of a license (i) may make changes to the facility as described in th.e

safety analysis report, and (ii) make changes in procedures as described in the j safety analysis report without prior Commission approval, unless the proposed i change involves a change in the Technical Specifications or involves an unreviewed I

safety question. Section (b)(1) states that the licensee shall maintain records of j changes ... made pursuant to this section. Those records must include a written -

j safety evaluation which provides the basis for the determination that the change

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does not involve an unreviewed safety question. Conclusio_nl i-

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The inspectors concluded that failure to perform an adequate safety evaluation l required by 10 CFR 50.59 constituted a violation of NRC requirements. (VIO 50 ..

219/97-02-07)  !

H2A Tours of Plant Areas  ;

' Insoection Scope (71750) ,

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. The inspectors toured various areas of the site including areas within the

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radiologically controlled area (RCA). .I Observations and Findinas The inspectors observed the general conditions and radiological housekeeping in the various areas of the facility. Equipment and buildings were kept in good conditio .

Radiological housekeeping was very good in most areas. However, one minor radiological concern was identified by the inspectors to licensee representatives and immediate corrective actions were taken to resolve the concern. Administrative controls were adequate, including information signs warning workers regarding radiological conditions, access controls to radiological areas, and barriers to prevent inadvertent entry into high radiation area , .

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L 20 Conclusions The areas toured by the inspectors were well maintained and radiological 1 housekeeping.was generally very good. A minor concern was brought to the licensee's attention for resolution or correction. Administrative controls were adequate to caution and inform workers regarding radiological condition .

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B2d - Criticality Radiation Monitors l Insoection Scope (71750)

i The inspectors reviewed the licensee's compliance with NRC regulations (10 CFR <

l 70.24) which require a monitoring system in each area where special nuclear material (quantity exceeding 700 grams) is handled, used, or stored. The j monitoring system must be capable of detecting a criticality accident and may not i be further than 120 feet from the material, in addition, emergency procedures are l required for each are .i Observations and Findinas The inspectors noted that the licensee has installed radiation / criticality detection j monitors on the refueling floor. These monitors would detect a criticality from fuel ,;

being loaded into the reactor. The inspectors asked the licensee's staff about a l separate monitor for the new fuel vault. The licensee stated that the detector on-the west end of the refueling floor would also detect a criticality accident in the new fuel vault. The new fuel vault was within 120 feet of the criticality monito R But the inspectors were concerned that the licensee did not have documentation regarding the effect of concrete shielding between the fuel and the monito j Specifically, it was unclear whether the concrete would shield the radiation from a j criticality accident to a lower level than the detection level on the existing radiation i criticality monitor. The licensee's staff responded that the concrete would not shield the radiation enough to create a detection problem. The licensee's staff had -i responded to this concern in the past, but no documentation was found to verify this calculation. The licensee performed and documented a calculation (dated April 8,1997) that concluded that the radiation monitor (C-5) was adequate to detect a criticality accident in the new fuel vault. The inspectors reviewed the licensee's !

calculation and noted that the calculation appeared accurate with appropriate assumptions and methodolog The inspectors also reviewed the licensee's emergency procedures and training-pertaining to a criticality alarm. The licensee includes the response training in the-annual site training. Site evacuations are periodically rehearsed during emergenc preparedness drills. The inspectors concluded that this training was adequate to

! ensure that all personnel in a special nuclear materials area withdraw to an area of

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safety. The training and drills are conducted on a frequency that allows personnel to become familiar with the evacation pla i i

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The licensee provides appropriate monitors to measure and alarm in the event of a  !

criticality accident in areas where special nuclear materialis handled, ~used, or f stored. The' emergency procedures were adequate to ensure the safety of j personnel.in these areas. Personnel attended training and drills to familiarize them- l with the'ensergency procedures, j

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R3 RP&C Pror,edures and Documentation j i jn199& don Scope (84750,01. ti2701) i

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The inspectors reviewed the ODCM implemented at the Oyster Creek site, including: ]

-(1) dose factors, (2) setpoint calculation methodology, and (3) bio-accumulation  ;

- g factors for aquatic sample media. The inspectors alto reviewed the.1995 and 1996 }

annual radioactive effluent report l l

The inspectors also reviewed the following selected chemistry and radiological ' I controls procedures to determine whether the licensee could implement the i radioactive liquid and gaseous effluent control programs effectively, j

! Gaseous and Uquid Cinumn Control Procedures, {

e Airborne Tritium Measuring Procedures, i e Irnplementation of Section 6.16 of TS, lodine Monitoring, and I Implementation of Section 6.15.2.(4) of TS, System Leak Tes i Observations and Findinas l

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All necessary parameters, such as effluent radiation monitor setpoint calculation j methodologies, site-specific dilution factorsi and dose factors were listed in the i ODCM. The licensee adopted other necessary parameters from NRC Regulatory l Guide 1.10 !

The inspectors reviewed the 6995 and 1996 annual radioactive effluent release l reports. The annual reports also summarized the assessment of the projected j maximum individual and population doses resulting from routine radioactive airborne  ;

and liquid effluents. Projected doses to the public were well below the technical j specification (TS) limits. The inspectors determined that.there were no anomalous  ;

measurements, omissions or adverse trends in the report !

I Section 6.16 of TS, lodine Moniuring, stated that "the licensee shall implerne,1t a  !

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program which will ensure the capability to accurately determine the' airborne iodine  !

concentration in vital arees under accident conditions. This program shall include i the following: . (1) training of personnel; (2) procedures for monitoring; and  ;

(3) provisions for maintenance of sampling and analysis equipment." The l Radiological Control / Safety Department (RC/S) had responsibilities to implement i these requirements. The inspectors toured RC/S Department laboratory and discussed with the representative. The inspectors noted that the licensee had  ;

provisions to satisfy the above requirements by establishing 25 iodine sampling i stations throughout the plant, i I

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Section 6.15.2.(4) of the TS requires that the isolation condenser system leak test j j must be performed once every 24 months. In practice, the chemistry department :

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takes a grab water sample from the isolation condenser tank and analyzes it every ;

, month. The grab samples (500 ml of water) were analyzed using a gamma - j p spectrometer (1000 seconds counting). The laspectors reviewed selective . j

' analytical data for 1996 and 1997. The analytical results indicated that there were :

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no tube (or system) leaks to the isolation condenser tank, i

2 Conclusions

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i Based on the above review, the inspectors determined that the licensee's'ODCM I 3 contained sufficient specification, information, and instruction to' implement and l

, maintain the radioactive liquid and gaseous effluent control programs. The

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inspectors also determined procedures were detailed and easy to follow.- Training of j personnel, procedures for iodine monitoring, and provisions for maintenance of ]

sampling and analysis equipment were goo '

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i R6 RP&C Organizathm and Administration i

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The inspectors reviewed the organization and administration of the radioactive liquid ,

! and gaseous effluent control programs, and discussed with the licensee changes j made since the last inspection, conducted in February 1996. No changes since the l last inspection of this' program area were noted. Staffing levels appeared to be' '

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. appropriate for the conduct of routine dutie I L \

. R7 Quality Assurance in RP&C Activities I i insoection Scope (84750-O'll i The inspection consisted of reviews of the: (1) Quasty Assurance (QA) Audits

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required by Section 6.5.3 of the TS, and (2) Quality av urance program for

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Chemistry measurement laboratory required by Section 6.8.1.i of the TS.

, Observations and Findinas

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o The inspectors noted that the audit frequency was changed from 12 months to l 24 months for ODCM/REMP. Howevert the licensee audited certain functions of the ]

, effluent control programs annually. The 1996 chemistry audit covered a portion of l j the effluent contro! program. The inspectors reviewed the audit scope and plan for I j_ licensee Audit S-OC-97-03, "ODCM/REMP," which had been initiated at the time of j the inspection. Scope and content of the audit were good. The inspectors noted that subj1ct matter experts were used as members of audit team The QA/ Quality Control (GO) program for analyses of effluent samples was  ;

conducted by the chemistry department. The chemistry laboratory participated in i the interlaboratory QC and the intralaboratory comparison programs. The inspectors i reviewed the QC data for intra /interlaboratory comparisons and noted that the ]

majority of OC data were within the licensee's acceptance criteria. Discrepancies !

were investigated and resolve ;

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l 23 1 Conclusions Based on the above review and discussions with the licensee, the inspectors l determined that the licensee met the audit requirements and QA and QC programs '

for the radioactive liquid and effluent control program l R8 Miscellaneous RP&C lssues l R8.1 Review of Uodated Final Safety Analysis Report (Unresolved item 50-219/97-02-09)

l A discovery of a licensee operating their, facility in a manner contrary to the Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a special focused review that compares [ plant practices, procedures and/or parameters to the UFSAR descriptio While performing the inspections discussed in this report, the inspectors reviewed the applicable portions of the UFSAR that related to the areas inspected. Several i inconsistencies were noted between the wording of the UFSAR and the plant practices, procedures and/or parameters observed by the inspectors (see R2.2 and below).

UFSAR ltem 1.9.50 response described that a " dedicated analyzer and charcoal cartridge are provided in the control room." UFSAR ltem 1.9.50 reflects the requirement of NUREG-0737 ltem lil.D.3.3, " Improved Inplant lodine instrumentation Under Accident Conditions." On April 3,1997, the licensee issued a deviation report (Deviation Report No.97-222). The deviation report described j that the licensee identified that there was no charcoal in the control room to satisfy the UFSAR ltem 1.9.50 requirement. The licensee was investigating this issue during the current NRC inspection. Pending completion of their review and subsequent NRC followup, this item is unresolved. (URI 50-219/97-02-09)

R8.2 (Closed) IFl 50-219/96-09 06: Adequacy of the licensee's review of the event (unplanned radioactive liquid release) and its corrective actions. Dose assessment was completed and included in the 1996 annual report. Quantification techniques of radioactive releases were appropriat l R8,3 LClosed) LER 97-002: Reactor building vent radiation monitor setpoints exceeded I Technical Specification limit (17 mR/hr) due tc personnel error. The licensee identified that the i.mproper setpoints (30 mR/hr and 40 mR/hr) of the reactor 1 building ventilation exhaust radiation monitors on February 28,1997. The improper l setting occurred on January 22,199 The licensee promptly corrected the improper setpoints and investigated other setpoints. Appropriate action was taken to address the inadequate performance of the individuals involved. The safety significance of this particular event is minimal because the offsite dose rate would be about 0.1 mR/hr as a result of a 40 mR/hr setpoint, well below the 10CFR2O limit of 2.0 raR/hr See Section M4.1 for additional discussion of this event. This LER is close l

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R8.4 Periodic Report Review The 1996 Effluent Release Report was reviewed and found to be acceptabl l

P1 Conduct of Emergency Preparedness Activities P Ememency Response Drill l Insoection Scope (71750) ,

I The inspectors observed a portion of the annual emergency response drill and staff l performance during the drill at the licensee's Technical Support Center (TSC). I

Observations and Findinas  ;

Overall, accident assessment, plant response monitoring, and event classification activities in the TSC were good. Good communications were evident among the j various emergency response facilities (i.e., simulator control room, technical support center, operations support center). The technical support center (TSC) provided properly-supporteJ bases for technical conclusions. The Emergency Director in the TSC conducted good briefings with his staff that were sufficiently frequent, were of good detail, and were consistent with changing plant conditions. The inspectors l noted that the ecmputer used to simulate the control room indications was i

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unavailable for approximately 90 minutes during the middle of the day. The licensee's exercise participants suspended the exercise and resumes when the computer system became available. The NRC resident office was iniormed that a drill was in progre ss and the simulated emergency classificatio The inspectors at: ended the licensee's exercise critique the day following the exercise. The licensee's critique focused on major observations and comments, and a performance asuessment was discussed. During the critique, the licensee noted all minor deficiencies. The problem with the computer used to simulate the control room indicators was discussed. Communications were a principle problem due to a failure of several phone lines. The inspectors assessed the licensee's critique a being good. The licensee was very self critical and identified many areas fcr future improvemen Conclusions The licensee demonstrated very good overall response during the annual emergency response exercise. Some minor problems were encountered with communication lines, but overall communications were good. A list of areas for program improvement were developed after the exercise. Overall, TSC staff perfo.mence was assessed by the NRC inspectors as good. The inspectors noted no particular exercise strengths or exercise weaknesses in the TS .

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S1 Conduct of Security and Safeguards Activities  !

S1.1 General Observations l

During routino tours, access controls were verified in accordance with the Security ;

Plan, security posts were properly manned, protected area gates were locked or '

auarded, and isolation zones were free of obstructions. Vital area access points were examined and verified that they were properly locked or guarded, and that ;

access control was in accordance with the Security Pla !

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! MANAGEMENT MEETINGS X1 Exit Meeting Summary 4

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A verbal summary of preliminary findings was provided to the senior licensee i management on May 7,1997. During the inspection, licensee management was l periodically notified verbally of the preliminary findings by the resident inspector No written inspection material was provided to the licensee during the inspectio j No proprietary information is included in this repor l l

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ATTACHMENT 1

. PARTIAL LIST OF PERSONS CONTACTED

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Licensee (in alohabetical order)

d T. Blount, Emergency Preparedness Manager G. Busch, Manager, Regulatory Affairs i D. Croneberger, Director, Equipment Reliability 1'

J. Hildebrand, Plant Maintenance Director S. Levin, Director, Operations and Maintenance

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K. Mulligan, Manager, Plant Operations M. Roche,' Director, Oyster Creek '

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R. Shaw, Radiological Controls / Safety Direr .-

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D. Slear, Director, Configuration Control R. Tilton, Manager, Nuclear Safety Assessi.nent

- NRC (in alohabetical order)

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L. Eckert, Radiation Specialist

. J. Jang, Senior Radiation Specialist J. Nick, Reactor Engineer (Temporary)

S. Pindale, Senior Resident inspector (Temporary)

8. Welling, Project Engineer (Temporary)

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i 27 i ATTACHMENT 2 i INSPECTION PROCEDURES USED )

I Procedure N Title l

40500- Effectiveness of Licensee Controls in Identifying, Resolving,- l and Preventing Problems  !

37551 Onsite Engineering 61726 Surveillance Observation i I

62707 Maintenance Observation

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71707 Plant Operations

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-71750 Plant Support 82701 Operational Status of the Emergency Preparedness Program 84750-01 Radioactive Waste Treatment, and Effluent and Environmental Monitoring l 92700 Onsite Followup of Written Reports of Nonroutine Events at J Power Reactor Facilities 92901 Followup - Operations i

92902 Followup - Maintenance I

92903 Followup - Engineering ,

92904 Followup - Plant Support 93702 Onsite Event Response

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ATTACHMENT 3 ITEMS OPENED AND CLOSED Opened Number Tvoe Description 97-02-01 eel Apparent violation - Suppression pool pressure suppression capability degraded due to inadequate review of system line-up. (01.2)

97-02-02 eel Apparent violation - Multiple switching and tagging issues (trend identified by Nuclear Safety Assessment). (07.1)

97-02-03 eel Apparent violation - Reactor building ventilation exhaust radiation monitor mis-calibration due to personnel error and ineffective post-work review. (M4.1)

97-02-04 eel Apparent violation - Heating boiler tagged-closed exhaust damper mis-positioned by unauthorized individual. (E4.1)

97-02-05 URI Inconstatent implementation of calibration methodology for effluent and process radiation monitoring systems. (R2.1)

97-02-06 URI Augmented offgas building pressure not being maintained as per the UFSAR specified value. (R2.2)

97-02-07 VIO The licensee failed to conduct an adequate 10 CFR SD.59 safety evaluation to determine if an unreviewed safety question existed for the removal of the isolation condenser

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radiation monitors. (R2.3)

97-02-09 URI No charcoal in the control room as specified in the UFSAR for use in iodine instrumentation during accident conditions. (R8.1)

97-03 LER Suppression Pool Bypa.,s Flow Created During Preventive Maintenance Due to inadequate Safety Review. (08.1)

s Closed Number Description 97-01-03 UPI isolation condenser radiation monitors inappropriately removed from service. (E8.1)

} 96-09-06 IFl Licensee followup of unplanned radioactive liquiJ releas (R8.2)

97-02 LER Reactor building ventilation exhaust radiation monitor setpoints exceeded '.echnical Specificatior. "mit. (R8.3)