IR 05000334/1979020

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IE Insp Rept 50-334/79-20 on 790910-14.Noncompliance Noted: Failure to Implement Committed Corrective Action Re Development of Checklist to Aid Tracking Maint Surveillance Test Records
ML19270H757
Person / Time
Site: Beaver Valley
Issue date: 10/12/1979
From: Beckman D, Keimig R
NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML19270H747 List:
References
50-334-79-20, NUDOCS 8001030073
Download: ML19270H757 (13)


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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF INSPECTION AND ENFORCEMENT Region I Report No. 50-334/79-20 Docket No. 50-334 License No. DPR-66 Priority

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Category C

Licensee:

Duquesne Light Company 435 Sixth Avenue Pittsburgh, Pennsylvania 15219 Facility Name:

Beaver Valley Power Station, Unit 1 Inspection at:

Shippingport and Pittsburgh, Pennsylvania Inspection condu te :

September 10-14, 1979 Inspectors:

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D. g Beck jr h, Reactor Inspector date signed f

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Approved by:

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/& /2 - 79 R. R. Keimig, Chief, P tor date signed Projects Se on No.

RO&NS Branch Inspection Summary:

Inspection on September 10-14, 1979 (Recort No. 50-334/79-20)

Areas Inspected:

Routine, unannounced inspection of:

licensee action on previous inspection findings; followup on IE Bulletins; onsite followup of foreign material discovered in ECCS system piping and anomalous behavior of Source Range Nuclear Instruments; review of safeguards DC bus battery monitoring features; review of seismic certification data for relief valves in plant systems; and, a meeting with licensee management concerning Resident Reactor Inspector onsite facilities. The inspection involved 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> onsite and 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> at the corporate offices by one regional based inspector and 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> onsite by three other personnel.

Results: No items of noncompliance were identified. A deviation was identified with respect to a previous licensee commitment to NRC (Failure to implement committed corrective actions, paragraph 2).

Region I Form 12 (Rev. April 77)

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DETAILS 1.

Persons Contacted R. Balcerek, Maintenance Supervisor, BVPS R. Burski, Senior Engineer, BVPS C. Dunn, Vice President, Operations, DLC

    • F. Fiorina, Account Supervisor, American Telephone & Telegraph, Consultant to NRC
  • E. Kurtz, Senior 0A Engineer, DLC
      • F. Lipchick, Station OA, BVPS W. Logan, Structural Design Engineer, DLC
    • W. Marquar dt, Station Office Manager, BVPS
    • M. McCaffrey, Telecommunications Engineer, DLC
    • G. Moore, General Superintendent Power Stations Dept., DLC
    • R. Prince, Account Representative, Bell of Pennsylvania (Part Time Attendance)
    • M. Roble, Account Representative, Bell of Pennsylvania (Part Time Attendance)

L. Schad, Operations Supervisor, BVPS

    • R. Swiderski, Superintendent of Construction, DLC
    • R. Stidham, Supervisor of Engineering Communications, DLC H. Van Wassen, BVPS Project Manager
      • J. Werling, Station Superintendent, BVPS
  • H. Williams, Station Chief Engineer
    • R. Woodling, Senior Engineer, BVPS The inspector also held discussions with and interviewed other members of the operations, maintenance, office, and engineering staffs.

Other Accompanying NRC Personnel The individuals listed below were participants in the meeting discussed in paragraph 9 of this report.

S. Cohen, Communications Specialist, NRC Headquarters R. Keimig, Chief, Reactor Projects Section No.1, Reactor Operations and Nuclear Support Branch, Region I J. Mc0 scar, Chief, Administrative Branch, Region I

  • denotes those present at the exit interview on September 14, 1979
    • denotes those present at the management meeting on September 12, 1979
      • denotes those present at both meetings above.

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2.

Licensee Action on Previous Inspection Findings (0 pen) Unresolved Item (78-30-02):

Licensee to document evaluation of BVPS Unit 2 containment floor liner weld problem applicability to Unit 1 containment.

In response to this item, the station issued Engineering Memorandum (EM) No. 20131 on January 8,1979, requesting DLC Engineering to document the subject evaluation.

This EM also included a request for evaluation of Unresolved Item No. 78-30-01 concerning the seismic analysis errors reported in LER 78-53 which eventually resulted in the issuance of the NRC Order to Show Cause on March 13, 1979.

The portion of the EM regarding containment liner welds was forwarded to the DLC Structural Engineering Department for disposition.

Durina this inspection, it was determined that apparently no additional action has been taken by the Structural Engineering Department to document the results of the DLC evaluation. Discussion with the licensee indicated that the statements made in IE Inspection Report No. 334/78-30 with regard to the acceptability of the Unit 1 containment were still accurate and no further engineering study was considered necessary. The inspector was unable, however, to determine why the evaluation had not yet been formally documented. The matter of licensee assignment and tracking of such items is further discussed in paragraph 8 of this report.

The matter of documentation for the above licensee evaluation will remain unresolved pending its completion.

(0 pen) Deficiency (79-04-01):

Failure to maintain records of completed Maintenance Surveillance Procedures.

The inspector reviewed the licensee actions described in the DLC response to NRC-Region I dated May 4, 1979, and found the corrective action portion of that correspondence to be implemented as stated.

Review of the licensee's stated " action to prevent recurrence" identified that the prescribed action has not been taken.

This action, specified to be the development and implementation of a checklist for document transmittal and accountability, was to be completed by May 31, 1979. As of Septembar 14, 1979, no action had been taken by the licensee to establish the checklist and thereby fulfill the specific commitment made in the above referenced letter.

Further review by the inspector indicated that some misunderstanding may han occurred on the part of station personnel in that a similar commitment regarding maintenance work request document checklists had been implemented in response to an internal QA audit finding at about the time that the matter of Maintenance Surveillance Procedure record control should have been addressed.

Failure to meet a commitment made in writing to the NRC is considered to be a deviatior. (79-20-01).

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A contributory factor to the licensee's failure to meet the above commitment appears to be the existing methods for licensee assignment and tracking of internal commitment responsibilities. This matter is further discussed in paragraph 8 of this report.

The licensee indicated to the inspector that a reevaluation of the actions specified in the May 4,1979 letter may also be made and could result in modification of the previously stated " action taken to prevent recurrence." The inspector informed the licensee that any such modification should be provided to NRC:RI for review in the licensee's written response to the Deviation identified above.

(Closed) Unresolved Item (78-09-06):

Licensee to obtain ANSI N45.2.ll Design Verification for DPC-0162 for NRC:RI review. As documented in IE Inspection Reports 334/78-09, 334/78-26, 334/79-06, and 334/79-18, modifications performed by the Power Station Engineering Group prior to September 1978 were not conducted in complete conformance with the requirements of ANSI N45.2.11.

Design Change Package 0162, performed during 1977, is one of the modifications involved and the required design verification documentation is not available.

Licensee evaluation of the design change packages which were not completed in accordance with ANSI N.45.2.ll is discussed in IE Inspection Report No. 334/79-18, paragraph 4.c(5) and is considered to be applicable to this item.

Licensee action taken in response to IE Inspection No. 334/79-18 will be reviewed during a future inspection and this unresolved item (78-09-06) is closed for record purposes and will be pursued in conjunction with that report.

(Closed) Infraction (79-09-05):

Failure to establish, maintain and implement procedures for SLCRS filter testing in accordance with Technical Specification 4.7.8.1.c.

The inspector reviewed completed procedure BVT-1.1-2.16.2, Main Filter Bank Cell Airflow Check, Revision 0, performed May 15, 1979; Engineering Memorandum No. 60098, Evaluation of Questionable Test Data, dated June 4,1979; and the BVPS Test Section, Technical Specification Surveillance Procedure Schedule Sheet. These reviews established that the testing was satisfactorily completed prior to plant startup and the actions described in the licensee's letter of September 5,1979 to NRC-Region I had been accomplished.

(Closed) Unresolved Item (79-13-03):

Licensee to improve practices associated with strip chart recorder identification and marking.

Additional direction was provided to the control room operators in the Night Order Book entry of June 13, 1979. A sampling review by the inspector on September 13, 1979, verified that the chart recorders were either operating properly or ad been submitted for maintenance; that the charts reviewed had been appropriately annotated; and, that correctly scaled chart paper was in use.

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3.

IE Bulletin Followup The inspector reviewed the IE Bulletins (IEB) listed below in order to determine whether the following actions were taken by the licensee and IEB requirements and licensee commitments were met:

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The written response was submitted within the time period stated in the IEB; The written response included the information required to be

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reported; The written response includes adequate corrective action commitments

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based on the information presented in the IEB and the licensee's response;

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Licensee management had forwarded copies of the written response to appropriate onsite management;

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Information discussed in the licensee's response was accurate; and,

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Corrective action taken by the licensee was as described in the written response.

IEB 79-04 - Incorrect Weights for Swing Check Valves Manufactured by Velan Engineering Company The licensee's response to NRC-Region I, dated April 30, 1979, identified the existence of six inch and three inch check valves of the subject types installed in seismically designed station systems.

This response indicated that the seismic analyses associated with these valves would be completed prior to the station's return to power operation from the NRC's Order to Show Cause dated March 13, 1979.

The licensee's submittal of July 11, 1979, in response to the Show Cause Order, stated that all lines having six inch check valves have been reanalyzed and have computed stresses which are within allowable limits for all lines.

The submittal further stated that all three inch check valves are located in lines which were analyzed via hand calculations.

These calculations utilized assumed valve weights which were, in each case, greater than the actual, correct valve weight.

The licensee has concluded that, for all cases, no additional evaluation is necessary.

The inspector reviewed the documents discussed and had no further questions on this matter.

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IEB 79-07 - Seismic Stress Analysis of Safety Related Piping This matter was the subject of NRC Orders dated March 13 and August 8, 1979.

The licensee's response to the IEB notes that all information requested by the IEB will be submitted to the Director of Nuclear Reactor Regulation, USNRC in accordance with the requirements of the NRC Order, dated March 13, 1979.

This response is considered to be acceptable based upon the resolution of associated issues by NRC:NRR.

Any further IE inspection activity in this regard will be conducted pursuant to the NRC Orders above.

IEB 79-09 - Failures of GE Type AK-2 Circuit Breakers in Safety Related Systems The licensee's response to NRC-Region I, dated May 17, 1979, identified seven breakers of the type addressed by the IEB. The inspector's review of the BVPS OM Sections, discussed as part of the inspection of IEB 79-11 below, confirmed that the licensee's submittal is accurate in this respect. The licensee's letter further states that a preventive maintenance program which complies with the guidance in IEB will be developed and implemented by September 30, 1979.

This aspect of the licensee's response will be reviewed by NRC:RI following its implementation (79-20-02).

IEB 79-11 - Facility Overcurrent Trip Device in Circuit Breakers for Engineered Safety Feature Systems The licensee's response to the subject IEB stated that no Westinghouse type DB-50 and DB-75 breakers equipped with overcurrent devices, as discussed in the IEB, are used in the plant's ESF systems.

The inspector reviewed the following references to confirm the licensee's conclusion.

BVPS OM Section 1.36.1, 4KV Station Service System - Listing of

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Major Components, Revision 2.

All breakers noted are ITE Type SNK 350 or SHK 250.

BVPS OM Section 1.37.1, 480V Station Source System - Major Component,

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Revision 4.

All breakers noted are GE type AK-3A-50, Bus supply or the breaker, or AK-3A-25, load supply breakers.

BVPS OM Section 1.39.1, 125VDC Control System - Major Components,

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Revision 3.

The breakers were all noted to be of the following types:

Battery Breakers - GE Type AK-2A-25 or AK-2A-50.

Distribution Switchboard Breakers - ITE, 2 pole breakers.

Distribution Panels -

Heineman, 3 x 1 pole breaker.

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A sampling review of applicable electrical drawings confirmed the licensee's findings.

IEB 79-13 Cracking in Feedwater System Piping The portions of the licensee's response and actions associated with the identification, repair, and testing of cracked piping and implementation of a feedwater piping inservice inspection program have been reviewed as discussed in IE Inspection Reports Nos. 334/79-08 and 79-15.

During this inspection, the licensee's response to IEB Sections 5.b and c were reviewed with respect to adequacy of emergency procedures and detection methods for feedwater line break accidents. The inspector reviewed licensee procedures to determine that they addressed the identification of and response to feedwater line breaks.

The inspector confirmed that the Emergency Operating Procedures of the BVPS OM Section 1.5.3.4, Revision 18, provide sufficient guidance for operator recognition and response, consistent with the licensee's response.

No items of noncompliance were identified.

4.

Foreign Material in Low Head Safety Injection (LHSI) System Piping On September 10, 1979, during performance of Operating Surveillance Test No.1.11.1, Safety Injection Pump Test (SI-P-1A), Revision 18, the 1A LHSI Pump was observed to have a recirculation flow of approximately 140 gpm at 186 psig. The acceptance criterion stated in the above procedure (pursuant to ASME Code,Section XI, IWP 3000) is 235-255 gpm.

Check valve 1-SI-29, located in the pump's recirculation line was suspected to be the cause of the flow reduction due to its history of disc binding.

(This valve is also discussed in IE Inspection Report No. 334/79-17.)

Valve I-SI-29 was removed from the system and replaced on September 11, 1979.

During that maintenance, two fragments of a red plastic material (approximately 1 cubic inch each) were found in the valve and adjacent piping. These pieces of plastic initially were thought to be part of an electrical conduit bushing and were considered, on that basis, to be the only pieces of significant size within the system.

Upon completion of valve repairs on September 13, 1979, the licensee again performed the pump surveillance test and found that the pump again would not produce the recirculation flow necessary to meet the acceptance criterion stated in the procedure.

The observed flow during this performance was approximately 210 gpm.

This value was also below the manufacturer's specified minimum pump flow for continuous pump operation.

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The I-SI-29 check valve was reopened and a third piece of red plastic material was found in the inlet port and was extracted.

Further, during this time period, the pieces were identified to be fragments of a WILC0 No. HN-4-L fire hose nozzle and appeared to constitute only a small percentage of the total nozzle.

The nozzle normally consists of six pieces:

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Two cylindrical body segments approximately 2 inches in diameter and 3 inches long made of the red plastic material; Two rubber ring gaskets;

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A brass diffuser, approximately 3 inches long and 3/8 inch in diameter, having a flanged end about 1 inch in diameter; and,

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A nut which secures the diffuser within the nozzle.

With the exception of the three fragments discussed above, all other parts of the nozzle Ere unaccounted for. The recovered fragments were identified to be from both body segments.

Upon removal of the third fragment, valve 1-SI-29 was reassembled and OST 1.11.1 was again performed on September 13, 1979.

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resulted in flow data within the acceptable range and higher than the actual flows observed during all testing back through 1978. All pump parameters were normal. The data for both tests conducted on September 13, 1979, were reviewed by the inspector and the data for the last test were confirmed to be acceptable.

Based on the potential for the remaining nozzle fragments still to be within reactor plant systems, the inspector addressed the following cmcerns to the licensee and requested they be included and documented in the licensee's evaluation:

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The possible locations of fragments not found and the potential consequences, including possible entry into the core, damage to ECCS pumps / components, and further potential for line blockage;

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The occasions on which the nozzle could have entered the system and the adequacy of the administrative controls which should have prevented such entry;

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The effects of the nozzle material on reactor plant system materials and fluids including chemical content, behavior of nozzle materials in various system operating environments, and any potential with adverse effects noted; and, 1678 121

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Documentation of plans for the search for and removal of any remaining nozzle fragments from reactor plant systems, including the methods to be employed.

During the period of September 10-13,1979, the 1A LHSI pump was considered to be inoperable with respect to Technical Specifications.

The inspector reviewed the licensee's activities in this regard and determined that the applicable requirements and action statements of the Techncial Specifications had been observed properly by the licensee.

In order to assure the continued operability of the suspect pump, the licensee intends to perform daily surveillance of the pump in accordance with OST 1.11.1 until such time as the Onsite Safety Committee has reasonable assurance that line blockage or pump degradation is not an immediate problem. At that time, the licensee intends to progressively lengthen the surveillance interval to weekly, then monthly.

The licensee's actions in regard to the above will be reviewed during future inspections.

The acceptability of those actions and operability of the potentially affected systems will remain unresolved pending further review by NRC:RI (79-20-03).

During review of documentation associated with this matter, the licensee was unable to provide the inspector with the records of performance of OST 1.11.1 on September 10, 1979.

It was this performance which initially identified the unacceptably low pump recirculation flow.

BVPS Operating Manual, Chapter 1.55A requires that such deficiencies and their disposition be documented on the cover sheet of the respective OST.

Discussion with the plant operators concerning this requirement indicated to the inspector that a clear understanding of this requirement did not exist and, that in other similar cases, unsatisfactory test data may have been inadvertently discarded.

The inspector advised licensee management of these conversations and of the apparently missing data from the September 10 performance of OST 1.11.1.

The licensee stated that the missing OST had been reviewed and circulated among the station staff during the initial evaluation of the pump problem and that effort would be made to locate the missing records.

Additionally, action will be taken to ensure that all personnel who perform surveillance testing are advised of the need to preserve incomplete or unacceptable test data as quality assurance records and to properly document the deficiencies and resolutions on the appropriate documents.

The licensee committed to complete these actions by October 1, 1979. This matter will remain unresolved pending the availability of OST 1.11.1 of September 30, 1979, for inspector review and review of the actions taken to ensure compliance with the documentation requirements of OM Chapter 1.55A (79-20-04).

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5.

Source Range Nuclear Instrument Channel N31 Abnormal Behavior During a routine review of Shift Supervisor logs, the inspector noted that on September 6,1979, while at 100% power, the NI Source Range Channel N31 began displaying a positive count rate even though the detector high voltage power supply was supposedly disconnected by Permissives P-6 (1/2 intermediate range channels above setpoint) and P-10 (2/4 power range channels above setpoint).

A similar count rate of a lower magnitude was also observed on Channel N32. The licensee has removed temporarily the high voltage power supply fuses from Channel N31 to ensure that high voltage is removed from the instrument detectors and is investigating the cause and permanent corrective action. The licensee has caution tagged the N31 power supply fuses and made the control room operators aware of the abnormal system arrangement.

The inspector confirmed that the procedural requirements of the BVPS Operating Manual, Section 1.2.4.6, Abnormal Procedure - Malfunction of Nuclear Instrumentation, are being applied and that the applicable Technical Specifications are being properly observed.

The inspector noted that Source Range channels are not required to be operable for the existing Mode 1 (Power Operation) conditions.

The licensee's procedures referenctd above provide the actions to be taken for operation Modes 2-5 with one or both source range channels inoperable.

The inspector had no further questions at this time but informed the licensee that the results of troubleshooting will be reviewed during a subsequent inspection (79-20-05).

No items of noncompliance were identified.

6.

Safeguards DC Bus Battery Status Monitoring Recent NRC inspections at several other operating facilities identified a problem with battery status monitoring. A review of the respective facilities'125 volt DC systems designs showed that the batteries are connected to their busses via either a fuse or circuit breaker. All electrical monitoring of the DC system and condition of the battery chargers, is sensed on the bus side of the battery fuse or circuit breaker.

If the fuse or circuit breaker would open, the circuit between the battery and its bus, bus voltage would be likely maintained by the battery charger and, because all monitoring comes off the bus, there would be no direct, conclusive indication that the battery has been disconnected.

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A similar configuration exists at BVPS with respect to the 125 VDC battery breakers and bus instrumentation. The facility's Technical Specifications 4.8.2.3 and 4.8.2.4 require surveillance of breaker position as part of the system's operability tests.

Additionally, direct indication of the status of each battery breaker is provided on the control room vertical control board.

The breaker indication is secured by breaker auxiliary contact position and is conspicuous to the plant operator. Based on the inspector's evaluation and the availability of direct breaker position indication, the potential problem discussed above does not appear applicable to BVPS Unit 1.

No items of noncompliance were identified.

7.

Review of Seismic Certification Data for the Reactor Vessel Over-pressurization System Protection The licensee installed the Reactor Overpressurization Protection System during 1977-1978 and, in correspondence to NRC:NRR, dated November 23, 1977, categorized the installation as an interim system.

This categorization was due to relief valves installed in the system's nitrogen supply lines, not possessing the required seismic certification at the time of installation.

In a letter to NRC:NRR, dated December 14, 1978, the licensee stated that the existing valves had been qualified for seismic service by testing to the seismic design requirements of the Beaver Valley Power Station. On that basis, the letter stated that the system status would be upgraded to a permanent modification.

The inspector reviewed the licensee's documentation in this regard and determined that, rather than being qualified by testing, the valves had been qualified for seismic service by analysis.

The Crosby Valve and Gage Company's Seismic Oualification Report, No. EC606, Revision 0, dated May 9, 1979, presented the calculations and conclusions which establish the seismic certification of the valves.

This report stipulates that the analyses were conducted in accordance with DLC Contract No.

BV-EC-62594, which provided the vendor the amplified seismic response spectra and acceptance criteria for the analyses.

This report concluded that the calculated seismic and operating stresses are within the allowable values of Section VIII of the ASME Code. The inspector considered this information as confirmation of the seismic capability.

With respect to the licensee's letter of December 14, 1978, the inspector requested that a corrected letter be submitted which states that the valves were certified by seismic analysis rather than testing.

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8.

Licensee Administration of Comitments During this inspection, two examples of problems associated with licensee adainistration of commitments were identified as discussed in paragraph 2 cf this report. Similar problems are also discussed in IE Inspection Report No. 334/79-18.

Inspector review of the licensee's activities in regard to assignment, followup, and closeout of internal and external commitments indicated that most such commitments are tracked informally with no mechanism for escalation to management attention should problems be encountered during their disposition.

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The station had recently initiated development of a more formalized

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commitment control system in order to avoid further problems with comitment administration. Discussion with licensee management indicated that the system would be applied to commitments made to NRC as well as internal licensee commitments between DLC departments and contractors.

The development and implementation of the licensee's comitment control program will be reviewed during future inspection (79-20-06).

9.

Management Planning Meeting for Resident Reactor Inspector Facilities During this inspection, on September 12, 1979, a meeting among NRC, licensee, and telephone company respresentatives was conducted to present and discuss the specific requirements and needs for establishment and support of the Resident Reactor Inspector's office for Beaver Valley Power Station, Unit 1.

Attendees are denoted in paragraph 1 of this report.

Implementation of the Resident Reactor Inspection Program for Unit 1 is scheduled to occur concurrent with the end of fiscal year 1979.

The licensee has provided an office, which meets the requirements of 10 CFR 50.70, within a newly constructed administration building for that use.

Discussion with licensee representatives indicated that, although minor construction work was still in progress, the building and office would be available for occupancy by the inspector on or about October 1,1979.

The meeting discussions included the arrangements necessary for installation of telecommunications equipment including the identification of responsibility for various aspects of the installation, maintenance, and billing.

Agreement was reached on the details of installation for commercial voice, telecopy, and emergency voice telephone equipment. Arrangements were made for installation of the required cabling, availability of emergency AC power sources for emergency telephone circuits, and the schedule for installation. Availability and location of normal AC power outlets for the support of office equipment were also established and the licensee was provided with a tentative floor plan of equipment arrangement to assist in the installation.

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Arrangements for office security, access to members of the public and media, and office janitorial service were also discussed. Additional discussion will be required with the licensee to establish a suitable location for inspector contact with the public and media representatives.

The licensee presently considers the existing office to be unsuitable for such contacts and will assist in making alternative arrangements.

Arrangements end schedule for the delivery and erection of NRC office furniture were also discussed and were left pending the final implementation by the Region I office.

NRC representatives presented the licensee with an overview of the Resident Reactor Inspection Program with regard to its initial implementation, current planning for expansion of the program, and tentative plans for additional manning during fiscal year 1980. This discussion included plans for providing a Resident Reactor Inspector for Beaver Valley Power Station, Unit 2 and the assignment of a Unit Reactor Inspector for Beaver Valley Power Station, Unit 1.

At the conclusion of the formal meeting, the attendees inspected the Resident Reactor Inspector's office and held additional discussions regarding the detailed installation of office services.

10. Unresolved Items Unresolved items are matters about which more information is required to determine whether the items are acceptable items, items of noncompliance, or deviations. Unresolved items are discussed in paragraphs 2 and 4 of this report.

11.

Exit Interview A management meeting was held with licensee personnel (denoted in paragraph 1) at the conclusion of this inspection.

The purpose, scope and findings of the inspection were disucssed as they appear in the details of this report.

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