IR 05000219/1989018

Insp Rept 50-219/89-18 on 890724-28.No Violations Noted. Major Areas Inspected:Review of Licensee Actions on Previous NRC Concerns,Atws Rule Implementation,Procurement Control & non-licensed Operators Training

ML20247K461
Person / Time
Site:	Oyster Creek
Issue date:	09/13/1989
From:	Blumberg N, Dev M, Oliveira W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20247K454	List: IR 05000219/1989018 ML20247K449
References
50-219-89-18, GL-85-06, GL-85-6, NUDOCS 8909210175
Download: ML20247K461 (23)

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U.S. NUCLEAR REGULATORY. COMMISSION REGION 1-Report ~N /89-18 Docket N License N DPR-16 Licensee: GPU Nuclear Corporation P. O. Box 388-Forked River, New Jersey 08731 Facility: Oyster Creek Nuclear Generating' Station Inspection at: Forked. River, New Jersey Duration: July 24-28,1989 Inspectors: @ ' Y/OkI M. Dev, Rgheter Engine r Date

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Wh & W. Oliveira, Reactor Engilheer GI/3k1 Date Approved by: N. Blumberg/ Chief, b/ ' //. The inspectors also contacted other licensee administrative and technical personnel during this inspectio .0 Licensee's Actions on Previous NRC Concerns 2.1 (Closed) Inspector Follow-up Item (50-219/86-24-06): Fire Suppress-ion Water System rendered inoperable by improper valve line u In 1986, a valve in the fuel supply line to the diesel drivers fire pump had been found closed by the licensee, rendering the diesel fire pump inoperable. In a follow-up special report (86-06) to the NRC dated July 26, 1986, the licensee identified the cause of this incident to be an improper valve lineup. The NRC performed a follow-up inspection (219/86-24). During this inspection, an addi-tional problem was identified in that the tanks and valves were found mislabelled. During this inspection (89-18), the inspector ____________-__ - _______-__ _ _ _ _ _

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verified that the valve lineup procedure 645.6.004, " Fire Suppression l Water System Valve Lineup" has been corrected and that valves V-9-957, " Fuel Line Isolation for Diesel Fire Pump 1-2", V-9-958,

    " Fuel Line Isolation for Diesel Fire Pump 1-1",. and diesel fuel oil tanks 1-1 and 1-2 were properly labelled.

i 2.2 (Closed) Violation (50-219/86-38-01): Failure to follow a procedure.

l ' Station Procedure 2000-RAP-30325.01, " Alarm Response Procedure," requires a plant shutdown if more than one IRM channel per trip sys-tem becomes inoperable while in the startup mode. Another station procedure 402.2, "IRM Operation During Startup," requires approxi-mately one decade of overlap between the IRM and APRM nuclear instru-mentation to ensure proper respons During a startup event on December 27, 1986, the plant was not shut-down when IRMs 16 and 18 were declared inoperable, nor did the APRM instrumentation achieve approximately one decade of overlap between the IRMs and APRMs. The licensee took prompt corrective action and subsequently provided indoctrination and briefings to the operations staff emphasizing adherence to station operating procedures. The inspector reviewed and discussed these procedures with selected control room operators and found them knowledgeable and capable of implementing these procedures. The inspector also noted that nuclear instrumentation indicating IRMs and APRMs overlap is appropriately monitored as required by the procedures. The licensee's action is considered adequate. This item is close .3 (Closed) Violation (50-219/86-38-02): Failure to follow a procedur Station procedure 312, " Reactor Containment Integrity and Atmosphere Control," requires nitrogen temperature be monitored and maintained at 65"80F during inerting of the drywell and toru During the time period between the December 27, 1986, and January 6, 1987, nitrogen temperature dropped between 55 - 60 F. The licensee took prompt corrective actions and subsequently provided indoctrination and briefing to the operations staff emphasizing adherence to station operating procedures. The inspector reviewed and discussed the procedure with selected control room operators and found them knowledgeable and capable of implementing the procedur The inspector also noted that the nitrogen temperature is adequately maintained and monitored as required by the procedure. The licen-see's action is considered appropriate. This item is close .4 (Closed) Unresolved Item (50-219/87-21-01): Licensee investigate means to assure that the Containment Water Level Monitoring System level impulse signal is not subject to a single failure accident.

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            .i The dual channel of containment water level monitoring system receives water level impulse signal from a single upturned water collecting tube in the bottom of the toru Scrap and debris collected in the upturned sensing signal tube in the event of an       ;

accident had potential to impair the dynamic response of this i instrument. To resolve this concern the licensee added a backflush l capability to the dual channel water level instrument detector lines L and revised procedure 604.3.017, " Wide Torus Level Calibration" to facilitate backflush of the level instrumentation lines every time a l l surveillance or testing was performed. The inspector reviewed ' several completed surveillance and calibration tests and observed that they were performed satisfactorily. The licensee's action.is ! considered complete. This item is close .5 (Closed) Deviation (50-219/88-14-03): Failure to provide training for ignition source fire watche Two jobs involving ignition sources were observed during'+.he orevious inspection. In both cases, the fire watch individuals assigned to the job were not performing their duties in accordance with plant ) procedures and were not fully aware of their duties as a fire watc The site training program did not include " hands on" training for the i fire watch personnel to meet their commitment to the guidelines of the " Nuclear Plant Fire Protection Functional Responsibilities, Administrative Controls, and Quality Assurance", dated June 1977, and  ; NFPA 51B. The guidelines were provided to the licensee in a NRC l letter of August 8, 197 ;

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The inspector reviewed the revised Station Procedure 120.1 " Welding, l Burning and Grinding Administrative Procedure". The Training Depart- 1 ment had also developed and implemented their " hands on" training ] commitment to the NRC guidelines and NFPA 51B. This training is ' similar to the approved training provided to the Fire Brigade Team The inspector also reviewed the training records of the mechanics . and was satisfied with the adequacy of the training. This item is l close .0 Anticipated Transient Without Scram (ATWS) Rule (10 CFR 50.62) (IP 25020) Bpckground On July 26, 1984, the Code of Federal Regulation (CFR) was amended to include Section 10 CFR 50.62, Requirements for reduction of Risk from ATWS Event for Light-Water-Cooled Nuclear Power Plants. An ATWS is an expected operational transient (such as loss of feedwater, loss of condenser vacuum, or loss of offsite power) which is accompanied by a failure of the reactor trip system (RTS) to shutdown the reacto !

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i ;. l 6 For each boiling water reactor, three systems are required to mitigate the consequences of an ATWS event, namely:

 (A) Alternate Rod Injection (ARI) System (B) Standby Liquid Control (SLC) System (C) Recirculation Pump Trip (RPT)

3.1 Review Criteria The systems and equipment required by 10 CFR 50.62 are not designated as safety-related. However, the equipment is part of the broader class of structures, systems, and components important to safety defined in 10 CFR 50, Appendix A, General Design Criteria (GDC).

Accordingly, they are designed, fabricated, erected and tested to quality standards commensurate with the importance of the safety functions they perform. In addition, NRC Generic Letter (GL) 85-06, Quality Assurance Guidance for ATWS Equipment that is not Safety- , related, details quality assurance requirements applicable to such equipmen In general, equipment installed in accordance with the ATWS Rule is required to be diverse from the existing RTS, and be testable at power. The diversity consideration reduces the potential for common mode failures that could result in an ATWS leading to unacceptable plant condition .2 ATWS Rule Implementation Review and Findings Alternate Rod Injection (ARI) The licensee installed the ARI system during the 1988 refueling outage. The system utilizes a two-out-of-two logic which causes , immediate energization of the ARI valves when either the reactor vessel high pressure trip setpoint, or the reactor vessel low-low water level trip setpoint is reached, or the manual ARI trip actuated by an operator from the control room. The inspec-tor reviewed licensee's documentation pertaining to the imple-mentation of the ATWS Rule. The modification design descrip-tion, MDD-0C-643A, Alternate Rod Injection System (listed in Attachment-1) provides QA scope for the ARI related plant modifications. The inspector discussed the ARI modifications with cognizant licensee personnel, and interviewed operations staff to verify licensee compliance to the ATWS Rule. Following observation and examinations were made by the inspector:

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Preoperational ARI system function time test demonstrated that rod injection of all control rods started within 15 seconds and completed within 25 seconds of ARI initiatio The test satisfied the ARI system function time require-ments approved by the NRC safety evaluation report (SER) to _ _ _ . . ___-____-__-__---___-___________ _ _ ___ _ _ -

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the Topical Report NEDE-31096-P (listed in Attachment-1).

This operating time was found within the calculated ARI function time of 27 seconds per licensee calculation (C-1302-641-5450-003).

- Inputs coming from nuclear safety-related systems to the non-Class IE ARI system are isolated through existing Class IE isolation devices internal to the Foxboro cabine Five new ARI valves have been added to the scram air header located in the reactor building. The two new ARI block valves added between the filter / regulator and backup scram valves assure that incoming control air to the scram air header will be blocked during ARI system actuation. The three new ARI vent valves added to scram air header down-stream of the backup scram valves provide separate path for venting the scram air header during ARI system actua-tio This three vent path arrangement assures scram air header depressurization within the required time even if one of the three vent valves were to fail during an ARI system actuatio ARI system equipment is qualified to anticipated opera-tional occurrences and located in a non-harsh environment which is outside the drywel I

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A normal / bypass svitch enables the ARI logic testing during plant operation. The status and availability condition are annunciated in the control roo ARI instrument loops are powered from 120V AC vital power system and are backed by 125V DC secondary sources, and as l such loss of offsite power would not affect ARI system function. The 125V DC power to the ARI control logic is supplied by Battery A with Battery B as a backup. Battery charger for both batteries A and B are powered from vital AC power source which is backed up by a diesel generato Review of the licensee's failure mode effect analysis j (FMEA) of Reactor Protection System (RPS) and ARI Common

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Power Supply in Control Panel 18R ( listed in Attachment-1) indicated that power interruption in either ARI or RPS will i not affect the performance of other syste l

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The ARI system instrument components are diverse from i existing RPS components with the exception of the reactor l low-low level trip channels. The reactor low-low level l trip function utilizes Rosemount transmitter and Foxboro l Spec 200 modules for instrument loop signal processin ! Since the RPS also uses the same type of signal processing, l the ARI modifications at OCNGS do not meet the diversity l l _____-__________-__--__a _

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requirements stated in the ATWS Rule. In'accordance with the ATWS Rule, the ARI system components are required to be diverse from the RPS from sensor output to the final actuation device. Because of safety risk associated with common mode failure, diversity in ATWS mitigating equipment was regarded an important factor for the implementation of the ATWS Rule. In order to comply with the ATWS Rule, the licensee has been required to change out the existing trip units before the end of the next refueling outag Discussion with the licensee management indicated that the licensee, in conjunction with the BWR Owners Group, is evaluating the adequacy of the existing system for satisfy-ing the ATWS Rule requirements. Pending review of-the adequacy of the licensee's actions to comply with the ATWS Rule, the diversity issue pertaining to ARI equipment remains an Unresolved Item (50-219/89-18-01).

B. Standby Liquid Control System The standby liquid ccstrol system is designed to bring the reactor to a cold shutdown cot.dition from the full power steady state operating condition at any time in core life independent of control rod system capabilities. In compliance with the ATWS requirements the facility Technical Specifications were amended and approved by the NRC (Amendment No.124). During the 1988 refueling outage the licensee implemented the intent of this amendment by a combination of concentration, Baron-10 enrichment, and flow rate of sodium pentaborate solution. A minimum of 15 wt% of solution and 35 atord Boron-10 enrichment at a 30 gpm pump flow rate satisfy the ATWS Rule equivalency requirement and assure.that the reactor is shutdown before unacceptable containment conditions-develo The inspector reviewed Station Procedure (SP) 828.9, Secondary System Analysis- Liquid Poison; result of Liquid Poison Pump Operability Test conducted in accordance with SP 612.4.001; and typical liquid poison analysis pertaining to sodium pentaborate concentration and tank volume. The modification met the require-ments of the ATWS Rule for the Liquid Poison Syste C. Recirculation Pump Trip (RPT) The function of the RPT is to reduce the severity of thermal transients on fuel elements by tripping the recirculation pumps * early in the transient events, such as turbine trip or load rejection. The rapid core flow reduction increases void content and thereby introduces negative reactivity in the reactor to reduce the thermal power. During the 1978 refueling outage the licensee conducted the RPT modifications to satisfy a NRC letter of September 1, 1976, indicating the NRC's position that a RPT _ _ _ _ _ - ____-

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..         g would significantly limit the consequences of an AWTS event for Oyster Creek. These modifications consisted of installing new electrical hardwares, including an independent second trip coil in each of the five recirculation pump 4160 V AC switchgear breakers, as well as rewiring several existing circuits. The RPT .is a Modified Hatch Logic using an one-out-of-two taken twice logic design to trip the five recirculation pumps in the

, ' event of high reactor pressure or low-low water level. Each trip coil receives an input from two of the four sensor channels containing a relay and normally closed high pressure, and low-low level contacts. Input from Both of the channels supply-ing a pump trip coil are required to actuate the recirculation pump tri The inspector reviewed modification documentation, including the operators training lesson plan, and interviewed selected plant operators and a training instructor to determine the licensee compliance to the ATWS/RPT. The licensee actions appeared to be appropriate. The inspector did not have any further questions at this tim .3 C_onclusions The ATWS related design, procurement, installation and testing were  ; conducted in accordance with approved procedures by qualified

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personnel to assure systems performance as required, except the ARI diversity issue discussed in paragraph 3. Operating procedures, associated plant drawings and FSAR have been updated to reflect the as-built system configuration, and operating staffs were tr ained and indoctrinated on these changes accordingly. Procedures for routine testing of these systems have also'been established and implemente , Permanently installed means for bypassing the system during mainten-ance and testing with continuous indication of the bypass status in the control room have been provided. The OCNGS ATWS mitigation sys-tem design and plant modifications assure that once the mitigation action is initiated, the action goes to completion and that the sub-sequent return to normal operation requires delicate operator actio The plant emergency operating procedure has been established for manual initiation of the ATWS Mitigation system from the control room. The personnel responsible for supervising and implementing the ATWS Rule are knowledgeable and capable of implementing the pla The ATWS related plant modifications were conducted under the purview of the licensee's QA program, and QA control was exercised on these activitie In general, the licensee has appropriately complied with the ATWS Rule except for the ARI diversity issue discussed in paragraph 3.2.A. In addition, the licensee is awaiting NRC/NRR guidance and recommendations as stated in the NRC safety evaluation report dated November 4, 1988 (listed in Attachment-1) to incorporate operability

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l and surveillance requirements for ARI and RPT in the facility Technical Specifications. Pending amendment of the Technical Specifications and review of the adequacy of the licensee's actions, the ARI and RPT operability and surveillance requirements are considered an Unresolved Item (50-219/89-18-02).

I 4.0 Procurement Control (IP 38701) j 4.1 Scope The scope of the inspection was to verify the implementation of the licensee QA program for procurement of safety-related and important-to-safety items in conformance with regulatory requirements, licen-see commitments, and industries guides and standard .2 Program Implementation Review and Findings The licensee has established an automated material management system (AMMS) procedure 7200-ADM-6231.07 (listed in Attachment-1) which is a computerized data base for initiation and procurement of equipment and materials, and for warehouse operations at DCNGS. In addition, station procedure 125.2, Conduct of Spare Parts Engineering (listed in Attachment-1) delineates methods used by the plant engineering personnel for identification and establishment of quality and tech-nica'. requirements to support spare parts procuremen The inspector reviewed selected procurement documentation, including items procured for the station ATWS related modifications discussed in paragraph 3.0. Documentary evidence was available on site, which confirmed that these materials and equipment met the procurement requirements and purchase specifications. Materials were procured from qualified vendors whose quality assurance program was consistent with the licensee procurement requirements. The vendor qualification was based on vendor audits and surveillance. Procurement documents for safety-related items had specified special instructions, includ-ing provision of regulatory requirements of 10 CFR 23, Reporting of Defects and Nonconformances. The licensee's receiving inspections verified vendor compilance to the licensee's purchase specifications and instructions, including validation of test reports and certifi-cate of compliance. Any nonconformances generated based on receiving inspections were properly dispositioned by the licensee through material nonconformance reports (MNCRs). During this inspection the licensee's Site Audit QA group was conducting an audit (SOC-89-02) of the station material program and warehouse activities. The audit had identified two noncompliance in the areas of record maintenance, and management control of contract service The inspector noted that the OCNGS material management group has an array of redundant and obsolete material management procedures which appeared to be ineffective in controlling the station procurement I i I

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n ,'- / 11 activities. The licensee representative indicated that the material management department procedures are currently being reviewed for developing a comprehensive station procedure for procurement control consistent with licensee QA program and corporate policy. The inspector did not have any further questions at this tim Following procurement documents were reviewed: Material Requisitions Items A124-89-0227 Repair of DL11-W Board 7240-89-5350 Nut Glands 7240-89-5517 0-Rings 7240-89-5581 Repair Kit-CRD Gear Reducer 7240-89-0401 Valve Assembly, GE-846D87-4P014 7240-89-5435 Bolt Assembly- Fire Protection 7240-88-0163 Boric Acid Granular 5370-87-0089 Sodium Pentaborate with enriched B-10 5513-87-0039 Steel Tubulars and Plates 5513-87-0038 Unitrust Angle Fittings 4.3 Conclusions Licensee's administrative controls of the station materials program and warehouse-activities appear adequate. The material management personnel have been trained and indoctrinated in implementing quality assurance requirements in the procurement process at OCNGS The computer based AMMS is used to create corporate data acquisition to facilitate issuance of procurement documents, update status of acquisition activities, and record receipt of materials. Quality Classification of equipment and materials, and engineering evalua-tion, including dedication and approval process of spare parts and commercial grade items are appropriately defined and controlle One weakness is the array of redundant and obsolete material manage-ment procedures that are currently being reviewed for consistency with QA program and corporate polic .0 Plant Procedures (IP 42700) 5.1 Scope , The licensee's plant procedures were reviewed to ascertain whether the procedures are implemented in accordance with regulatory require-ments, technically adequate, and changed in accordance with the Technical Specifications (TS) requirement .___ _ ________________- _-_- __-_-_-_ - _-______ _ -

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5.2 Plant Procedures Program 5.2.1 Program Administration The principal procedures for administering the plant procedures program are Station Procedure 107, " Procedure Control," and Work Management System (WMS) Procedure A000-WMS-1220.14, "Prepara-tion, Review, and Approval of Work Procedures." The Safety Review Manager, through the Document Center (DC)', is primarily responsible for administering the program. DC responsibilitie include: initiating periodic reviews of controlled procedures; issuing a quarterly status of the procedures and a monthly status of procedures overdue for their biennial review; and providing the plant with copies'of the changes to the latest procedures. All completed surveillance, test, and maintenance procedure records are also maintained by the D .2.2 Program Review The inspector selectively sampled operational, abnormal alarm condition, maintenance, and administrative procedures. The procedures reviewed were.in compliance with the Technical Specifications (TS), and were current, controlled, properly filed, and readily retrievabl Some of the plant procedures selected are addressed in paragraph 2.0. The documents. reviewed or referenced in this inspection are lis:ed in Attachment- The inspector verified program implementation by reviewing the following types of procedures with the responsible personnel:

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Eight operations procedures such as welding, burning, grinding and fire watch instructions performed in accord-ance with Station procedures 120.1 and 12 NSSS response alarm procedure 2000-RAP-3024.0 Five maintenance procedures such as A-100-GME-2918.51 and 52 for maintaining limit torque valves and MOVATS testing, respectivel Five Instrument and Control (I&C) procedures such as procedure 604.3.017 for performing wide range torus level calibration Test procedures such as Technical Specification surveil-lance procedure 665.5.005 for performing the drywell airlock leak rate.

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Eleven administrative procedures such as procedure 1000-ADM-1291.01 for the nuclear safety and environmental L impact review and approval of document The technical content of procedures were reviewed by the inspector. Operations procedure 617.4.014 was reviewed with l- emphasis on the concerns identified in the 10 CFR 50.62 ATWS Rule regarding the time required for. initiating and completi.ig ' an alternate rod injection evolution. Changes to the procedure adequately addressed the requirements of 10 CFR 50.6 The licensee's biennial review of the maintenance, construction and facility (MCF) procedures appears protracted. Out of the 60 overdue procedures, 42 procedures have been rewritten, revised or updated. The remaining 18 procedures are to be completed by April 199 The inspector witnessed the following activities for procedural adherence:

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A continuous fire watch round beir.g performed by an equip-ment operator (EO) in accordance with station procedure 12 Two E0s closing and tagging out Fire Protective Post Indicating Valves V-)-13 ared V-9-18 in accordance with station procedure 105, " Control of Maintenance." A senior reactor operator verified their action A craftsman returning a controlled key to in the Control Room and performing proper documentation per station procedure 108, " Equipment Control."

- A plant engineer preparing the outline for a new I&C procedure 621.3.038, " Stack and Turbine Building Radiation Gaseous Effluent Monitoring (RAGEMS) System." Also observed a plant engineering supervisor reviewing the revision to station procedure 416.1, " Operation of Ester-line Angus Type MRL (Multipoint Recorder / Logger) Record-ers." The activities were performed in accordance with station procedure 10 Electricians replacing a limit switch 0-ring in accordance with maintenance procedures A100-GME-2918.51, "Limitorque Valve Maintenance," and A100-GME-2918.52, "MOVATS Testing."

- Mechanics performing a snubber test activity in accordance with a temporary change (7-24-89-2) to maintenance proce-dure 675.1.507, " Functional Testing of Gergen-Patterson I Hydraulic Snubber."

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, 5.3 Conclusions l-The plant procedure program is adequately documented, administered, and implemented. The procedures and procedure changes are techni-cally adequate, clear and retrievable. The program is also audited and monitored by QA to ensure compliance with the regulatory and s Technical Specification requirements. A weakness in the licensee's procedures pertaining to maintenance, construction and facility (MCF) was noted as not being completed in a timely manner. MCF management-is, however, committed to complete their biennial reviews of their overdue procedures by the end of the yea No violations or deviations were observe .0 Non-Licensed Staff Training (IP 41400) 6.1 Scope The scope of this inspcction was to review the implementation of the licensee's non-licensed staff training progra .2 R_oview of Program Implementation and Findings The inspector selected activities discussed in paragraphs 2, 3, and 5 to review their impact (lessons learned) on the licensee's implemen-tation of their non-licensed staff training program. The review and examination included:

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Observing activities performed by mechanical and electrical maintenance personnel, plant engineers, equipment operators, and QA/QC personne Interviewing personnel performing the activities and their supervisors, as well as the Training Departmen Examining training records and chenges to lesson plan Examining licensee's Non-Licensed Operators training accredita-tion status with the Institute of Nuclear Power Operations (INPO).

In early 1989 the Direct,r of Oyster Creek Nuclear Generating Station (OCNGS) introduced a " Plan of Excellence." Each department was required to subscribe to the OCNGS Plan of Excellence and formally develop their objectives for achieving excellence. The Training Department's objectives included: ensuring that the training program is re-accredited by INP0 in 1990; improving the Maintenance Training Program and staff; and improving the Fire Brigade Training Progra To achieve these objectives, the Training Department is developing a _ ________ - _____ _ ____ - _ _ _ a

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>     15 detailed plan with' milestones for achieving INPO re-accreditation in accordance with INPO criteria. The addition of the " Mechanical Maintenance Technician Training Program,"'and provision for a Mechan-ical Maintenance Laboratory as well as increasing the Maintenance Training' staff with qualified instructors are apparent improvements in licensee's awareness to achieving quality. These improvements and the increase in Fire Brigade Training at the Fire Academy from two day sessions bientally to two day sessions annually are examples of   j active management involvement in the non-licensed staff training effor The inspector observed the following activities being performed in accordance with approved procedures by trained and qualified personnel:
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Fire watch round performed by an equipment operator (EO) who was also a qualified fire brigade member. Station procedure 12 regarding fire watches for hot work activities was discussed with the E0. The E0 was not familiar with the contents of the procedure because EOs do not conduct these fire watches. Thc Training Department is evaluating the need to include station procedure 120.1 in the fire brigade trainin On-the-job-training (0JT) performed during the functional test of a snubber. The mechanics received the fundamentals of testing snubbers at the Training Center and subsequently performed ;he OJi in the. Maintenance Sho A new plant engineer preparing an outline for a new procedur The engineer and the inspector walked down the procedural steps and discussed his training to perform the task. The inspector was satisfied that the engineer was properly traine An Operations supervisor reviewing a memorandum from Operations Control pertaining to latest procedure changes that affected Operations for circulating it to the operations staff for informatio The inspector followed up on the training affected by the 10 CFR 50.62 "ATWS Rule" requirements. Documents reviewed included an Operator Training Document Action Item Tracking System (OTDAITS) work sheet which listed the Standby Liquid Control System (SLCS) lesson plans updated because of the ATWS Rule. Also listed in the OTDAITS were courses affected by the Alternate Rod Injection (ARI) System and the Recirculation Pump Trip (RPT) AWTS Rule requirements. The inspector reviewed Licensed Operator Requalification Cycle 88-7 which addressed the ATWS Rule, and found appropriat In March 1989, the Training Department conducted a self-assessment in preparation for a similar assessment to be conducted by INPO in _ _ . _ _ _ _ _ _ _ _ _ _ _ _ _ _ __-_ -__ _ _ _ _ _ _ _ _ _ _ _ -

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L L E 199 INPO also visited the Training Department in June 1989 and was , satisfied that the Ninetiertcisd Staff Training Program was on  ! schedule for their re otAreditation assessment v, sit in 199 ! 6.3 Conclusions Management's involvement in the quality of the Non-Licensed Staff Training Program is adequate as evidenced by the improvements in.the maintenance training effort, addition of a new maintenance building, and responsiveness of the Training Department to the ATWS Rule, as well as increase in the training staf .0 Document Control Program (IP 39702) 7.1 Scope The scope of the inspection was to ascertain whether the implementa-tion of the quality assurance (QA) program related to document centrol was in conformance with the Technical Specifications and regulatory requirement .2 Program Implementation The program is administered by the Safety Review Manager in accord-ance with station procedure 103, " Station Document Control." The inspector witnessed and examined the following activities for the ATWS Rule related plant modifications, procurement activities, non-licensed training activities and plant procedures discussed in previous paragraphs:

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Processing of Form 103-1 for biennial review of procedure 200-0PS-3024.2 Storing and filing documents in accordance with Regulatory Guide 1.8 Retrieving records such as surveillance's performed in accor-dance with I&C procedure 604.3.017 regarding torus level calibration Preparing, maintaining and distributing documents such as the quarterly " Procedure Review Cycle."

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8.0:. Audit Program (IP 40704) 8.1 Scope The scope'of the inspection was to ascertain whether qualified

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personnel were conducting audits in conformance with the Technical Specifications and regulatory requirements.

I' 8.2 Program Implementation The program is administered by the DC QA MOD /0Ps Audit Supervisor in-accordance with procedure 6130-QA-7206_.01, " Quality Assurance Audits." Review by the inspector of the following indicate that audit program was satisfactorily implemented:

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Training and qualification of QA auditors. Training records of selected QA auditors were reviewed. Their qualification and training appeared consistent with the requirements of QA procedu e 6130-QA-7202.01, " Indoctrination and Certification of QA MOD /0PS Personnel."

-- Planning, scheduling, preparation and conduct of audits and follow-up. The station audit group has appropriately imple-mented these QA audit activities as evidenced by the review of i QA audits S-0C-88-09 and S-0C-88-20. The inspector also

reviewed additional nine 1988-89 QA audits, including a joint utilities management audit, and thirty QA monitoring report The audits were comprehensive and had thoroughly evaluated the l licensee quality controlled activities. Ma'nagement response to L the audit findings and corrective actions implementation were in general appropriate and timel .3 Conclusions The licensee's QA program implementation of audit and monitoring the station quality control activities appear adequate and effectiv .0 Unresolved Items Unresolved items are matters about which more information is needed to ascertain whether they are acceptable items or violation. Two unresolved items are discussed respectively in paragraphs 3.2.A and 3. .0 Management Meetings l: Licensee management was informed of the scope and purpose of this inspection at the entrance interview on July 24, 1989. The findings of the inspection were discussed with licensee representatives during the cour:a of this inspection and presented to the licensee management at the July 28, 1989 exit interview (see paragraph 1.0 for attendees).

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At no time during this inspection was written material provided to the licensee by the inspectors. The licensee did not indicate that any proprietary information was involved within the scope of this inspection.

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ATTACHMENT-1 Document Reviewed: 1.0 Procedures / Correspondence MDD-0C-643A, QA Plan of Scope for Alternate Rod Injection System, Rev 3 Station Procedure 103, Station Document Control, Rev. 21 Station Procedure 105, Control of Maintenance, Rev. 29 Station Procedure 107, Procedure Control, Rev. 16 Station Procedure 108, Equipment Control, Rev. 40 Station Procedure 2000-RAP-3024.01, Alarm Response Procedure, Rev. 27 Station Procedure 120.1, Welding, Burning, and Grinding Procedure, Rev. 2 Station Procedure 102.2, Continuous Fire Watch Instructions, Rev. 2 Station Procedure 125.2, Conduct of Spare Parts, Rev 4 Station Procedure 130, Conduct of Technical Review and Safety Review By Plant Review Group, Rev. 4 Station Procedure 416.1, Operation of Easterline Angus Type MRL (Multi-point Recorder / Logger) Recorder, Rev. 2 Station Procedure 604.3.17, Wide Range Torus Level Calibrations, Rev. 8 Station Procedure 617.4.014, Alternate Rod Injection Logic Test, Rev. 2 Station Procedure 621.3.038, Stack and Turbine Building Radiation Gaseous Efficient Monitoring System (RACEMS) (Draft) Station Procedure 828.9, Liquid Poison, Rev 4 Station Procedure 612.4.001, Liquid Poison Pump Operability Test, Rev 14 Station Procedure 2000-0PS-3024.28, Standby Liquid Control System-Diagnostic and Restoration Actions, Rev 2 Station Procedure 823.7, Chemical Analysis of Boron, Rev 2 Safety Evaluation, SE-328232-001, Use of Enriched Sodium Pentaborate Solution in the SLCS, Rev 0 Modification Proposal # 202-76-1, Recirculation Pump Trip Modification, Rev 4 Completion Report 202-76-11, RPT Modification Administrative Procedure 2000-ADM-1291.0), Procedure for Nuclear Safety and Impact Review and Approval of Documents, Rev. 1 Policy and Procedure 7200-ADM-6231.01, Automated Material Management System (AMMS), Rev 1 Policy and Procedure 7210-ADM-6231.01, Contract Purchase Requisitions, Rev 2 Report No. 3731-045, Failure Mode Effect Analysis of RPS and ARI Common Power Supplies in Control Panel 18R, Rev 0 NRC Letter dated August 3,1988, Issuance of Amendment-124 (Oyster Creek Nuclear Generating Station Technical Specifications) Test Procedure 665.5.005, Drywell Airlock Leak Rate Test, Rev.13 Test Procedure 675.1.507, Functional Testing of Gergen-Patterson Hydraulic Snubber, Rev. 5 Oyster Creek Plan of Exce11ance Quality Assurance Procedure 6130-QA-7202.01, Indoctrination Training and Certification QA/ MOD /0PS, Rev. 7 L- .- - - - - _ - - . - - -. _ _ _ _ _ _ _ _ _ _ _ - - - - - - - _ _ _ _ _ _ _ _ _ - - - - - - - _ _ __ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _____

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ , _ _ 6ttachment-1 2

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QA Audit Report Nos. S-0C-88-01, 09, 14, 15, 17, 19, 20, 89-01; v-0C-87-04 } j Cooperative Management Audit Program Report 0-COM-89-05, June 7,1989 ' Status of Completion and Schedule 1989 OC Site Audits Open Audit Finding Status QA Monitoring Reports l Regulatory Guide (RG) 1.88, Collection, Storage, and Maintenance of l Nuclear Power Plant Quality Assurance Records Procedure Review Cycle (May 3, 1989) Licensed Operator Requalification Cycle 88-7 Operator Training Document Action Iten; Tracking System (OTDAITS) USNRC Safety Evaluation Report on Oyster Creek Nuclear Generating Station Compliance with ATWS Rule 10 CFR 50.62, dated November 4, 1988 2.0 Plant Drawings ARI System - ATWS Elementary Diagram, 3431-E0578, Rev 2 Recirculation Pump Trip Elementary Diagram,19529, Rev 1 l l I

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