IR 05000029/1990016

Insp Rept 50-029/90-16 on 900821-1001.Violation Noted.Major Areas Inspected:Daytime & Backshifts in Areas of Plant Operations,Radiological Controls,Maint & Surveillance, Emergency Preparedness,Security

ML20197H720
Person / Time
Site:	Yankee Rowe
Issue date:	11/06/1990
From:	Rogge J NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20197H702	List: IR 05000029/1990016 ML20197H697 ML20197H715
References
50-029-90-16, 50-29-90-16, NUDOCS 9011200091
Download: ML20197H720 (20)

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U.S. NUCLEAR REGULATORY COMMISSION

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REGION.I

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l Report'No:

50-29/90-16 w.

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Docket No:

50-29

' Licensee No: DPR-3 J

Licensee:

Yankee Atomic Electric Company 580 Main Street -

'l Bolton, Massachusetts 01740-1398

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Facility Name: Yankee Nuclear Power Station

' Inspection at:

Rowe, Massachusetts

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' Inspection Conducted: August 21 - October 1,1990

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Inspec; ors

T. Koshy[ Senior Resident Inspector

. M. Markley,. Resident Inspector

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- Approved By:

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L: 6kogge, Chief, Reactor PfdjectTSection 3A'

' Dale.

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-Insocction Summary: Insocction on August 21 - October 1.1990 Reoort No. 50-20/90-16

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' Ateas Inspected: Routine inspection on daytime and backshifts in the areas of: plant operations; 3.<,

radiological controls; mainten2nce and surveillance; emergency preparedness; security; Lengineering and technical support; and sa.'ety assessment and quality verification, a.

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Results:

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General Conclusions on Adequacy. Strengths or Weaknesses in Licensee Programs

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The licensee'is addressing age-related equipment needs as evidenced by the installation J

of a new excore nuclear instrumentation system. However, plant operation with transient material in the vapor container is of concern. -- A weakness was noted in operator w

knowledge and training for upgrades to the Safety Parameter Display System for Critical Safety Functions.

' A ' strength ' was noted in radiation protection : and physical security program

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implementation.

However, the use of antiquated maintenance tools contributed to personnel exposure during main coolant pump mainto :ance.

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IJnresolved Items -

Three unresolved items were identified:

Seismic qualification of the-safe shutdown. system (SSS) boric acid mix tank

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mixing motor support fixture (Section 3.2);

j Fitness-for-Duty. (FFD) program implementation 'for two licensee identified-I

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incidents (Section 7.2); and Licensee action to assess 'the size, rating and quality of electrical fuses 'Section

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8.1).

3:

Violations

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' One violation was ideritified during this inspection period regarding inadequate corrective -

action to preclude repeated failure of station emergency lighting (Section 5.3),

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EXECUTIVE SUMMARY Plant Occrations

~ Yankee Nuclear Power Station (YNPS, Yankee or the plant) continued the core XXI refueling outage which began June 23,1990. The projected outage was extended to facilitate replacement of the station emergency diesel generators (EDGs). On August 2 - 27, 1990, the resident

..nspectors conducted a special inspection of the FDG (50-29/90-14). The findings of the special inspection were the subject of an enforcement conference held in the NRC Region I office on September 21,1990, On September 28,' 1990, an Unusual Event was declared due to pressure boundary leakage at an-incore detector conoscal connection. At the time, the plant was in Mode 3 (Hot Standby). The heatup was terminated and the plant was returned to Mode 5 (Cold Shutdown).

At the conclusion of the inspection, testing to identify the location of leakage was ongoing.

During the initial shutdown hot leak inspection on June 24, 1990, the NRC identified a temporary scaffold installed in the pressurizer cubicle which existed in-place throughout the operating cycle. Additional examples of transient equipment storage practices in the vapor containment were noted (Section 3.5). A weakness was noted in operator training for upgrades

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to the Safety Parameter Display System for Critical Safety Function status trees (Section 3.6).

Radiological Controls Radiological planning and program implementation was generally good. However, maintenance department use of an antiquated manual lapping device contributed to the 18 puson-rem being expended to correct motor-to-pump leakage on the No. 4 main coolant pump (Section 4.1).

Emergency Preparedness (EP)

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- The licensee conducted an EP drill on September 11,1990, to test facility communications. On September 25,1990, the licensee conducted the annual EP exercise. Licensee performance for both of these activities was generally good (Section 6.0)

Maintenance and Surveillance Following overhaul, two of the station emergency diesel generators were returned to service i

without demonstrating they would perform satisfactorily in service. This was the subir da special NRC inspection. The licensee determined the EDGs did not have adequate ca; o

support emergency loads and decided to replace them. New diesel generator installation und

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l preoperational testing was the subject of another special NRC inspection. The licensee installed l

a new excore nuclear instrumentation system during the core XXI refueling outage (Section 5.2).

l Inadequate corrective action for station eraergency lighting failures resulted in an apparent violation of 10 CFR 50, Appendix B quali y assurance requirements (Section 5.3).

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Executive Summary

becurity No deficiencies were observed in routine physical security performance and program implementation. The licensee identified two incidents regarding fitness-for-duty (FFD) One involved three individuals reporting for duty under the influence of alcohol. The other involved five individuals inadvertently omitted from the random drug testing pool. These items are unresolved pending specialist inspector review (Section 7.2).

Encineering and Technical Sucoort The licensee adequately resolved NRC identified items regarding ECCS TS guidance and the i

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surveillance practice on the use of low span gas for calibrating the high range post-accident hydrogen analyzer. Licensee corrective measures for commercial grade fuse verification are ongoing (Section 8.1).

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Safety Assessment and Ouality Verification The licensee Plant Information Reporting (PIR) program is adequately addressing items less significant than LERs for appropriate corrective action (Section 9).

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' TABLE OF CONTENTS

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Persons Contacted............................... -...... -..

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Summary of Facility' Activities................................

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1lant Operations (IP 71707,'71710,93702)

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Plant Operations Review...............................

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Engineered Safety Feature System Walkdown

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. Review of Temporary Changes, Switching and Tagging 3'

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Unusual Event During Plant.Heatup.......................

3.5 Vapor Container Hot Leak Inspection.......................

3.6 Safety Parameter Display System Upgrades...................

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Radiological Controls (IP-71707)

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Main Coolant Pump Maintenance...... -....................

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Mairtenance and Surveillance (IP 61726, 62703).....................

5.1:. Maintenance of Emergency Diesel Generators 7~

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' Excore Nuclear Instrumentation Upgrades....................

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Station Emergency Lighting....... -.....................

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Emergency Preparedness...................................

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g 6.1 Emergency Preparedness Drill

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. Emergency Preparedness Exercise........................

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Secu rity (I P 71707)......................................

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. Observations of Physical Security........................

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7.2 Fitness for Duty Program Implementation....................

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Engineering and Technical Support (IP 71707, 37828, 92700, 92702)

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. (Closed) Notice of Violation (88-02-02 Commercial Grade Fuses Installed

<in Vapor Containment Alarm and Configuration......-_..,,.....

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,(Closed) Inspector Followup Item (87-11-09) Calibration.of Hydrogen Analyzer Using Low Span Gas

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8.3 (Closed Inspector Follow Item (87-11-01) Clarification Needed for TS Sections 3.5.5 and 4.5.2 (ECCS)........................

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Safety Assessment and Quality Verification (IP 40500,90712)...........

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9.1-PIR 90-001, Primary Drain Collecting Tank (PDCT) Overflow.......

9.2 PIR 90-002, Omission of TS Required Item from Auxiliary Operator Log shee t.......................................

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Review of Periodic Reports (IP 90713).........................

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Management Meetings (IP 30703)............................

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DETAILS i

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Persons Contacted

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. Yankee Nuclear Power Station

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I N. St. Laurent, Plant Superintendent

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T. Henderson; Assistant Plant Superintendent

e Yankee Atomic Electric Comoany (YAEC)

a J. Thayer, Vice President and Manager of Operations R. Mellor, Projects Manager

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The inspector also interviewed other licensee employees during the inspection, including

members of the operations, radiation protectica, chemistry, instrument and control,

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maintenance, reactor engineering, security, emerg ncy preparedness, training, technical services and general office staffs, l

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' Summary of Pacility' Activities

- Yankee Nuclear Powr r Station

Yank e the plant) continued the core XXI

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refueling outage which began Jua

..,1990. 'Inc projected outage was extended to i

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facilitate replacement of the station emergency diesel generators (EDGs).- On August 2 -

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27,1990, the resident inspectors conducted a special inspection of the EDGs (50-29/90-T

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held in the NRC Region I office on September 21,1990, q

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On September 28, 1940, an Unusual Event was declared due to pressure boundary

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leakage at an incore detector conoscal connection. At the time, the plant was in Mode

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3 (Hot Standby)! The heatup was terminated and the plant was returned to Mode 5 (Cold

Shutdown). At the conclusion of the inspection, leak testing for repairs was ongoing.

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On August 27-30,' 1990, a NRC Region I specialist inspector conducted a radiation

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r protection, radwaste and transportation inspection (50-29/90-15).

On September 20, 1990, NRC Commissioner James R. Curtiss conducted a visit at j

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YNPS. He met with the resident inspectors, attended a daily maintenance planning

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meeti~' 9nd attended a site tour conducted by the Manager of Operations. Following the

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' tota 3 mssions were held with station management on NRC and YNPS issues.

On August 24 - 28,1990, two NRC Region I specialist inspectors, a NRC contractor and

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the resident inspectors conducted an inspection of the YNPS emergency preparedness

program and annual exercise (50-29/90-82). Also, during that week, four NRC Region I specialist inspectors conducted an inspection of the EDG replacement (50-29/90-18).

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Plant Oncrations (IP 71707. 71710. 93702)

3.1 Plant Ooerations Review

The inspector observed plant operations during regular and backshift tours of the following areas:

Control Room Safe Shutdown System Building i

Primary Auxiliary Building Fence Line (Protected Area)

Diesel Generator Rooms Intake Structure i

Vital Switchgear Room Turbine Building Cable Tray House Spent Fuel Pit (SFP) Building Safety Injection Building Vapor Container (VC)

The following items were checked during daily routine facility tours; shift-

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staffing, access control, adherence to procedures and _ limiting conditions of '

operation (LCOs), instrumentation, recorder traces, protective systems,' control

,D room annunciators, area radiation and process monitors, emergency power sortce

~ operability, operability of the Safety Parameter Display System (SPDS), control

I room logs, shift supervisor logs, and operating orders. On' a weekly basis,-

- selected Engineered Safety Feature'(ESP) trains were verified to be operable. The ~

condition of plant-equipment, radiological controls,- security and safety were (

assessed. On a biweekly frequency, the inspector reviewed safety-related tagouts,

. chemistry sample results, shift turnovers, portions of the containment isolation -

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valve lineup and the posting of notices to workers. Plant: housekeeping and fire

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y protection were also evaluated. -

Inspections of the control room were performed on weekends and backshifts as

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follows: August 22,26,27, and September 10,11,17,18,24,25, and 26. Deep-y back;hift inspection was performed on August 26 from 10:45 a.m. to 7:00 p.m.

Operators and shift supervisors were alert, attentive and responded appropriately

to annunciators and plant conditions.

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3.2 Engineered Safety Feature System Walkdown

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The inspector performed a complete walkdown of the safe shutdown system (SSS)

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to verify proper alignment and operational status.

The review included verification that system lineup procedures match plant drawings and as-built configuration, accessible major flowpath valves were correctly positioned, power 1.

supplies were energized with breakers and control switches properly positione:i lubrication and component cooling was proper, no equipment conditions exist

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which may degrade performance, no unauthorized ignition sources or flammable materials were present, and levels of cleanliness were adequate.

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i The SSS provides seismic safe shutdown capability by providing borated primary makeup water to the main coolant system (MCS) and feedwater to the steam

- generetors. The system is_ designed to bring the plant from normal operating conditions to cold shutdown following a seismic event, flood, tornado or other

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event' requiring a remote shutdown of the plant.

It has a dedicated diesel.

generator and fuel oil supply to support the SSS systems if electrical power is i

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unavailable from normal sources, a

inspector review noted good agreement of procedure OP-3018, Rev.12,. Major

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Plant Shutdown with the SSS, with the station safety system manual and as built i

drawings. - Proper system alignment was verified.' Component labeling was

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generally good. The electrical motor control center (MCC) and instrument panels were properly fastened. The inspector verified the installation of a new diesel generator battery storage rack and the associated fasteners. The previous battery

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fasteners were a noted deficiency during the prior SALP cycle assessment.

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. The licensee: adequately addressed inspector concerns. One exception was the

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support fixture for the mixing motor to the primary makeup boric acid mix tank.

Specifically, the inspector questioned the seismic qualification of the mix tank motor: mount. i The motor mount is secured to the top of the. mix tenk with a-threaded tightening clamp..The motor is located directly above the positive

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displacement primary makeup pump. The licensee stated that the motor support

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nwas analyzed for seismic qualification. However, cognizant personnel indicated that the-analysis was apparently not documented. The licensee is currently evaluating the support for seismic qualification. _ This item is unresolved pending completion of the licensee evaluation and NRC review (50-29/90-16-01).

3.31 Review of Temnorary Chances. Switchine and Tagging Tempo' ary change requests (TCRs), which were licensee approved in support of r

implementing lifted leads and jumper requests and mechanical bypasses, were

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reviewed to verify that: controls established by ~AP-0018, " Temporary Change g

Control," were met; no conflicts with the Technical Specifications were created; the requests were properly approved prior to installation; and a safety evaluation

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in accordance with 10 CFR 50.59 was prepared if required. Implementation of i

the requests was reviewed on a sampling basis.

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The switching and tagging log was reviewed and tagging activities were inspected

to verify plant equipment was controlled in accordance with the requirements of AP 0017, " Switching and Tagging of Plant Equipment."

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Licensee administrative control of off-normal system configurations by the use of

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TCR and switching and tagging procedures as reviewed above, was in compliance with ; procedural l instructions and - was consistent' with plant safety.

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unacceptable conditions were identified.

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3.4 Unusual Event During Plant Heatun At approximately 4:30 p.m. c,1 September 26,1990, the licensee identified main t

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coolant system (MCS) leakage at an incore detector conoseal connection. At the

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time, the plant was in Mode 3 (Hot Standby) with MCS temperature and. pressure

at 397 degrees Fahrenheit and 800 psig, respectively. The heatup was terminated

at 6:00 p.m.. ar.d the plant pressure was reduced to 600 psig for inspection.

u Initially, the licensee determined the leakage was from water trapped-in-the conoseal housing. Subsequent licensee inspection determined the leakage was

coming through the incore instrumentation tubes at a rate of approximately 1 cubic

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c centimeter per minute. The licensee considered this to_ be pressure boundary leakage and declared an Unusual Event at 8:32 a.m. on September 27,1990.. Th'e

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licensee brought the plant to Mode 5 (Cold Shutdown) at 9:28 p.m. and made an

ENS call to notify the termination of the Unusual Event.

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Out of the ll. instrumentation tubes that pass through this conoseal, all but two

, tubes were eday current tested during the recent refueling outage. Two of these

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. tubes were omitted in inspection due to the installed fixed incore detector

'l he licensee is in the process of developing a. plug to seal the leakage a these l_

instrumentation tubes. Such a plugging device would make the tube 3 unavailable

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for instrumentation, however, sufficient tubes remain operable to satisfy Technical Specification requirements. Further inspection of the licensee's carrective actions

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will be discussed in the next resident inspection report.

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p 3.5 Vanor Container Hot Leak Insocction

I During plant cooldown for the core XXI refueling outage on June 23,1990, the

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inspector performed a tour of the vapor container (VC) in conjunction with the p

licensee performance of their hot leak inspection per OP-4200, Rev.15, Main l'

Coolant System Leak Inspection or ISI Pressure Test. During the tour, the l

inspector identified numerous components with boric acid buildup evidencing

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L some degree of main coolant system (MCS) leakage. The licensee similarly identified examples of component leakage. The inspector verified the licensee observations of this leakage to be thorough. In the pressurizer cubicle, the inspector observed a temporary scaffold suspended from the overhead approximately eight feet above the pressurizer heater electrical penetrations. The h

scaffold board was not secured to the associated supports. Discussions with L

operations management confirmed that the scaffold had remained in the J

pressurizer cubicle through the operating cycle. The licensee stated that operators

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. had used the scaffold to perform valve lineups during startup. The licensee was l

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unable to explain why the scaffold was not removed or why a permanent scaffold

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had not been installed.-

-i Licensee corrective action included issuing maintenance requests to repair the m

observed leaking components and installing a permanent metal scaffold to replace the wooden platform in the pressurizer cubicle. The licensee counsel;ed the staff j

to be more thorough in performing VC inspections in preparation for operation.

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. Inspector review verined the maintenance requests were comple"' within the

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scope of outage activnies. Although the licensee promptly installed a permanent l

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scaffold in the pressurizer cubicle, operating the plant with an unsecured wocd

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platfam does not reflect a good operating philosophy.

The inspector also noted unsecure transient equipment stored on top of the steam l

enerator cubicles and on the broadway outside the bioshield. Some of these items included a portable welding unit, scaffold materials and I&C equipment in a storage facility. The inspector expressed the concern that unrestrained material l

syred on the charging Door could cause unanalyzed damage in the event of a loss

of coolant accident (LOCA) or steam break inside containment as well as during

-.i a seismic event. The licensee stated that the equipment will be adequately secured :

l and it remains inside the VC due to limited space in plant storage areas. _The l

licensee is currently evaluating the inspector identined concerns, j

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, Safet.y Parameter Disola_v System' UonradrJi :

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During the core XXI refueling outages the licensee performed an upgrade to the Safety Parameter Display. System-(SPDS) to include' Critical Safety Function (CSF) status blocks.

The CSF status blocks Lwere incorporated through

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engineering design change request (EDCR)89-303. The CSF status blocks j

indicate when key plant parameters monitored during an accident are either i

satisfactory or are sufficiently far out 'of specincation to warrant continuous

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monitoring and the performance of procedural steps within the EOPs to correct the situationc

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During a.,Aitine control room review an October 1,1990, the inspector noted an j

orange illuminated CSF status block indicating the alphabetic designation P-0.

i The inspector questioned the primary reactor operator what the CSF status block j

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meant. The operator recognized that it was an orange path but was enable to

explain the alphabetic designation.

The operator examined the emergency

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operating procedure (EOP) status trees and determined it was an orange path for

main coolant system pressure. Inspector review noted the P-O was an orange path

for system integrity. At the time, the CSF wr providing invalid data as indicated i

'by a white border around the orange bloc:.. The inspector questioned the j

individual as to what training had been provided regarding SPDS upgrades. The operator sta:ed that he had received no training on the CSF status blocks.

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The inspector questioned the shift supervisor and secondary reactor operator.what i

the status block P-O meant._ The operators responded correctly. However, the

individuals were unable to determine what another status block with the alphabetic

- designation AE-G meant.- The AE-G is a green path for adverse environmental

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conditions in the vapor container (VC). The operators similarly responded that they had not been trained on the SPDS upgrades. Further, the shift supervisor stated that he had never been given SPDS training of any kind.

The inspector questioned the reactor engineering (RE) staff responsible for

' overseeing the SPDS modifications regarding what training was provided to the i

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operators and shift technical advisors (STAS) on CSF status blocks. RE personnel J

- stated that the SPDS modification should have been presented along with-other outage EDCRs during pre-startup training.

Discussions with the training staff indicated that the subject operators.had been-given training on the'SPDS upgrades. However, this training did not include a discussion of the alphabetic designations used to represent the associated CSF.

This information was similarly absent from the pre-startup training manual. The.-

i training and RE staff s queried other operators to determine the extent of personnel

. knowledge. The licensee stated that the operators were knowledgeable in the use.

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of EOP status trees but 'noted some weakness in alphabetic designations used on

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The operations department issued Special Orders90-171 and 90-172 to detail the

identifial weaknesses and to: require the operators to review a revised training excerpt delineating the alphabetic designation relationship with the associated CSF.. Licensee actions were adequate in correcting the training deficiency.

'4 Radiolonical Controls GP 71707)

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Radiological conOls mre reviewed on a routine basis relative to industry radiological

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standards, admL.ittrative and radiological control procedures, and regulatory

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requirementsc Selected work evolutions were observed to determine the adequacy of

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program implementation commensurate with the radiological hazards and importance to

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safety. Independent surveys were performed by the inspector to verify the adequacy of j

radiological. controls and instructions to workers.

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- Radiation Work Wrmits (RWPs) accurately reflected inplant radiological conditions.

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Radiation protection (RP) personnel provided effective exposure controls for radiography activities in the vapor container (VC). No oeficiencies were noted in radiological posting and labeling. The inspector observed personnel consistently wearing the requisite

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dosimetry specified on RWPs. Station and contractor personnel demonst'ated Idequate knowledge in the use of station procedures for personnel contamination monitoring ami the release of personal items from the radic6gis! control area (RCA).

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Main Coolant Pump Maintenance On July 30 - August 13, 1990, the licensee performed maintenance to correct observed motor-to-pump leakage on the No. 4 mam coolant pump (MCP). The-leak was identified during the initial outage vapor container (VC) inspections.

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The. repair activity entailed removing the MCP motor to the VC charging floor where the seating surface was lapped. Similarly, the pump seating surface was lapped to establish a proper mating surface. Following the lapping, the motor was installed on.the pump and the fasteners were appropriately torqued.~ The MCP is a canned rotor, canned-stator Westinghouse design.

_ Overall, radiological control for the MCP maintenance was generally good, i

Radiation protection.(RP) personnel demcastrated noteworthy occupational n

exposure control for workers. However, the resultant exposure of 18 person-rem -

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was considered high. This was due, in part, to a manual lapping device used by the maintenance department. A total of 11.3 person-rem was expended using the'

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manual lapping tool. The licensee stated tnat an automatic lapping device was -

commercially available and would be purchased. Additionally, removing and.

i reinstalling the. motor on the pump expended 4.76 person-rem. The licensee l

attributed this to the tight fit of motor-to-pump interfaces where removal and'

reinstallation required several attempts.

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Although radiological planning and program implementation was _ good, maintenance for this evolution did not provide personnel with the tools to do the j

job consistent with maintaining occupational exposure as low as is reasonably achievable (ALARA). Further licensee review is warranted to limit the difficulty

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in removing and reinstalling the motor.

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Maintenance and Surveillance GP 61726; 62703)

The inspector observed and reviewed maintenance and surveillance activities relative to

' industry standards, administrative controls, and regulatory requirements. Selected work evolutions and surveillance tests were observed to verify safety and compliance. Specific -

areas examined were licensee use of station procedures, codes and standards, QA/QC involvement, management oversight, safety tag use, jumper use, equipment alignment and post-maintenance testing (PMT). In addition, the inspector evaluated radiological controls for. worker protection, fire protection, limiting conditions for operation (LCOs),

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deficiency review, resolution and reporting per Technical Specification.

5.1 Maintenance of Emergency Diesel Geacrators During the core XXI refueling outage, Yankee Rowe performed major overhauls on each of the emergency diesel generators (EDGs). EDG-1 and EDG-2 were returned to service following satisfactory monthly Technical Specification (TS)

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operability testing at 200 kW for two hours. Following overhaul, EDG-3 failed the 18-month TS operatiility test which loads the EDG to 400 kW for one hour.

Subsequent licensee evaluation and modification determined that each EDG was incapable of supporting its design basis loading of 400 kW. The NRC questioned the adequacy of testing practices following significant maintenance. Also, the NRC questioned the practice of using one of the EDGs to supplement normal station power, thereby rendering multiple emergency electrical buses vulnerable

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to a' single fault. These observations and concerns were reviewed and detailed in

.a special NRC inspection (50-29/90-18)._

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The licensee decided to replace the EDGs. The installation and testing of the new EDGs was reviewed during a special NRC inspection (50-29/90-19).

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Excore Nuclear Instrumentation Uogrades The original excore nuclear instrumentation (NI) system supplied by Westinghouse

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utilized magnetic: amplifiers and vacuum tube electronic circuitry.

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unavailability of replacement parts prompted the licensee to replace the original system with'new NI instrumentation.

The NI: panel installed during the core XXI refueling accommodates new instrument channel drawers with General Electric (GE) logic, relays, power supply cordections, status lights, switches and other devices for eight'new NI

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chanocis with interfacc items for the existing Reactor Protection System. The j

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following eight NI channels -were included in the scope of the maintenance j

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modification:

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L channels I and 2: Source Range (SR);

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channels 3,4 and 5: Intermediate Power Ran;>e (IPR)

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channels 6,7 and 8: Power Range (PR)

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channels 4 and 5: Intermediate Range (IR)

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The licensee performed the engineering associated with this design change in-Laccordance with procedure EDCR 88-310 (ECN-5). The panel was fabricated at the' site and installed in the control room in accordance with structural modification EDCR 88-307. The inspector noted that changes to the front of the

.l Main Control Board (MCB) were minor.

The inspector performed a review of the control room NI panel installation and its interconnection to the RPS and MCB to determine if the installation was performed in accordance with the design package drawings and industry standards.

The NI instrumentation meters are located in the front and rear of the panel with channel logic card drawers. The interconnection. ' the panel to the RPS and MCB was incomplete at the time of the inspection Cable pulling for the MCB

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was in progress.

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The inspector observed the cable and wiring connection inside the panel with channelization, six inch separation and barrier provisions. The installation was noted to be in accordance with separation criteria specified in IEEE-384. The inspector noted the location of. ASEA relays and bistable units between the

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drawers and cable connections. It was observed that these items are not readily l

accessible for' maintenance or replacement unless wiring connections are

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. disconnected and tagged for later reconnection. The inspector determined that the

licensee was aware of this condition. Future maintenance or replacement is n

planned to address this condition, k

The inspector noted that vendor QA was performed at GE through Tannon, Inc.

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General Electric reliability data and performance was used in the procurement and :

design of the' panel. Site QA efforts were recently initiated. The licensee stated

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these efforts will_ continue until the panel is declarei operational. The inspector

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reviewed the QA findings with the cogaizant engineers. All of the QA findings.

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were resolved and the items were clos < ' and docun'ented by the licensee.

The inspector noted that an independent review of the n'odification was performed and that a 10 CFR 50.59 review and safety evaluation was conducted. The safety j

evaluation was descriptive and supported the conclusions. The modification and j

y the installation procedures were reviewed and approved by Plant Operations i

Review Committee (PORC).. The maintenance actions on this modification were deemed adequate. ' No unacceptable conditions were identified.

5.3 Station Emergency Lighting

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At 6:55 a.m. on June 29,1990, the Z-126 offsite power supply was removed-from service. The' licensee removed the No. 3 480 Volt Bus 5-2 and motor.

control center No. 4, Bus 1 (MCC-4, Bus 1). This resulted in the planned removal of normal smtion electrical lighting in some plant areas.

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m During a routine plant tour at approximately 11:45 a.m., the inspector observed that the emergency lighting in the safe shutdown system (SSS)-building control-and pump rooms was not illuminated. Emergency lighting is required for the SSS building per 10 CFR 50, Appendix 'R,-III.J. which specifies that emergency o

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g fi lighting units with at least an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> battery power supply shall be provided in all areas needed for the operation of safe shutdown equipment and in access,and L

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egress routes thereto. Other plant areas lacking illumination from normal or w

cmergency electrical sources were the spent fuel pool building, the No. 3 battery

room and the safety injection accumulator cubicle. Although the other plant areas observed do not fall under Appendix R requirements, thes do constitute a potential personnel safety hazard.

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. The inspector discussed the above described observations with cognizant licensee

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-. personnel. The inspector requested to review the most recent 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> emergency

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lighting discharge test. Additionally, the inspector questioned the adequacy of emergency lighting in the other areas observed. '

The licensee responded to this discussion via Technical Services memorandum TS

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247/90 dated August 5,1990.. The memorandum detailed licensee plans to test-SSS and Switchgear Room Appendix R lighting for 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> and replace batteries not meeting acceptance criteria; plans to raodify the battery replacement frequency -

to. four years; an explanation of proteedve circuitry to limit batteries from

discharging below a specified voltage; previous annual discharge surveillance test

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results; and plans to evaluate emergency lighting for other-than Appendix R q

purposes. The licensee also noted another nuclear facility which uses portable.

lighting to satisfy some. Appendix R requirements. Yankee has three flash lights

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in the shift supervisors emergency kit in the control room.

Attached ito TS.247/90 memorandum was a maintenance support department memorandum MSD 156/89 which detailed problems with the most recent surveillance test OP-5635, Rev. 2, Annual Surveillance of Appendix R. Lights g

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- completed December 7,1989. Four consecutive emergency lights, tested in pairs

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in accordance with the procedure, failed the 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> discharge test. The licensee

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subsequently completed the surveillance test by successfully testing two more emergency lights. The batteries were replred in the failed devices; however, no

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ferther actions to correct this situation were taken. After consulting the battery w

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f manufacturer, the. cognizant _ engineer recommended changirg the battery replacement periodicity to a four year frequency.

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10 CFR 50, Appendix B, criterion XVI requires that measures shall be established

[a to assure that/ conditions adverse to quality, such as failures, malfunctions,

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lj deficiencies, deviations, defective material and equipment, and nonconformances

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determined and corrective action taken tc> )teclude repetition.

Contrary to the above, licensee actions in response to the failed emergency lights u;

in December 1989 did not assure that defective equipment was promptly identified E

and corrected in that corrective actions did not include sufficient additional'

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testing Licensee corrective actions were not adequate to preclude repetition in

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J June 1990. This constitutes a violation of 10 CFR 50, Appendix B,' Criterion b

XVI (50-29/90-16-02).

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. Emergency Prr3secass

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6.1 Emergency Prcoaredness Drill-

On September 11.1940, the licensee conducted an emergency preparedness drill to test facility communications. The drill scenario simulated a steam generator

.i tube rupture event. Following the drill, the licensee conducted a critique to assess performance. The licensee determined the following areas needed improvement:

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communication of chemistry data in units appropriate for emergency action level '

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(EAL) determinations, staffing the emergency operations facility (EOF) in a more -

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. timely? manner, and the assignment of personnel dedicated to continuously.

maintain telephone communications.

I-Inspector review noted no significant deficiencies. The drill was conducted in an orderly, professional manner. Technical Support Center (TSC) and Operations i

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Support Center (OSC) status boards were updated consistent with the progress of

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s 6.2 Emergency Preparedness Exercise On1 September 25,. 1990,- the licensee conducted the annual emergency i

preparedness exercise. It was a full scale exercise with the Commonwealth of

. Massachusetts and the State of Vermont participating. Personnel from the Federal

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i EmergencyLManagement Agency (FEMA) observed offsite activities.

The exercise scenario simulated a loss of coolant accident (LOCA).

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. NRC Region I specialist inspectors, an NRC contractor and the resident inspectors conducted an-inspection of emergency preparedness program implementation during the exercise. Licensee performance was generally good. The findings of this evaluation are documented in NRC inspection report 50-29/90 82.

7.-

S.ecurity (IP 71707)

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Observations of Physical Security i

Selected aspects of plant. physical security were reviewed during regular and backshift hours to verify that controls were in accordance with the security plan and approved procedures. This review included the following security measures:

guard staffing, vital and protected area barrier integrity, maintenance of isolation p

zones, and -implementation of access controls including authorization, badging, l

escorting, and searches. No inadequacies were identified.

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'12 The security management conducted a quarterly trend review to compare the a

security performance with the industry and other NRC Region I facilities. The.

study indicates improvements in personnel and equipment performance. The licensee initiative to assess performance with the industry is noteworthy.

7.2 Fitness for Duty Program Imolementation At approximately 2:30 a.m. on September 1,1990 three contractor employees

.l arrived onsite and entered the protected area late for their scheduled work i

assignment.

When the individuals reported for. duty, the cognizant work supervisor observed behavior indicating the potential for impairment due to the influence of alcohol.- Licensee management was informed of the situation and the individuals: were escorted out of me protected area: by 3:30 a.m.

~. At approximately 5:30 a.m. the individuals tested positive for alcohol. The licensee suspended the individuals: from work and. terminated their' site access

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authorizations, On August 30,1990, the licensee notified the. inspector that five individuals who -

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n-had been granted unescorted access to protected and vital areas had not been included in the data base for random drug screening. The oversight was identified

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through a licensee audit. The time periods that these individuals were omitted from the random test population ranged from 5 to 56 days. The licensee stated

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'4 that the individuals were unaware that they.were not subject to random drug

l testing.? The licensee determined the root cause to be a communication error between security access control and FFD staffs. Licensee corrective actions

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-included implementing new. access control forms to preclude recurrence of the

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. communication errors and performing monthly audits of the random drug testing

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Inspector review noted' licensee actions were adequate in' identifying the above ui described deficiencies. However, the'overall effectiveness and adequacy of these

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aspects of the licensee's program. (timeliness of alcohol testing and the -

implementation of unescorted access authorization) is unresolved pending NRC specialist inspector review (50-29/90-16-03).

-8.

Engineering and Technical Suoport (IP 71707. 37828. 92700. 92702)

8.1 (Closed) Notice of Violation (88 02 02 Commercial Grade Fuses Installe;'_in Vanor Containment Alarm and Configuration This item is related to the inadequate controls on the quality and size of a fuse that was used for replacement during corrective maintenance. The replacement

fuse, utilized in the vapor container pressu e channel VC-PI-243, had not been

procured through the quality assurance program and the size of the required fuse l

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was'not documented. The licensee responded to this violation in a letter dated May 25,1988. The completed corrective action included a program to dedicate the commercial grade fuses, a system walkdown in refueling outage cycles 19-20,

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20-21, and a final verificat'

alkdown during the cycle 21-22.

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The inspector reviewed licensee actions during cycle 20-21. The licensee has i

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completed the system.walkdown, however, no assessments were made to the i

acceptabil!ty of the existing fuses. The inspectors discussed the' need for a-

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preliminary assessment to ensure the adequacy of the fuses that are important to

the safety of the plant. This item' is unresolved pending NRC review of the -

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licensee action to assess the size and rating and quality of the fuses. (90-16-04).

8.2 (Closed) Insocctor Followup Item (87-11-09) Calibration of Hydrogen Analyzer Usine low Soan Gas

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This item relates to the use oflow span gas (5%) to perform TS calibration of the -

high range (0-20%) post-accident hydrogen analyzer. The licensee conducted an

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evaluation of the potential methods of calibratire this instrument, achievable accuracy and the risks associated with the process. The licensee determined by-testing-that a 5% gas provides a maximum' inaccuracy of.1.4 percent of the hydrogen concentration. For example, a reading of 18.6% noted when the actual

' concentration is 20%. According to Regult; Y Guide 1.97, the primary purpose

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of this indication is to' detect and mitigate the potential for containment breach.

The neensee considered this adequate in providing the intended design function.-

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. The licensee ~ control room is equipped with a second indication from a hydrogen

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analyzer with a range of 0-5 percent. The licensee accident analysis indicates a

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duration of 139 days for hydrogen concentration to reach 4%. The use of sample

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R gas at higher than 5% hydrogen gas was considered an undue risk to the plant and

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y personnel. The licensee technical bases for using low span gas and the concern

for personnel safety is adequate. This item is closed.

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8.3 (Closed Insocctor Follow Item (87-11-01) Clarification Needed for TS Sections i.

3.5.5 and 4 5.2 (ECCSY This item is related to the inspector identified need to clarify TS requirements and

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action statements for ECCS recirculation and long term hot leg injection

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subsystem's which were upgraded to utilize the pumping capabilities of the safety

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injection subsystems. The original recirculation and hot leg subsystems relied on

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purification pumps and charging pumps. The licensee detailed plans to submit a TS Proposed Change by the end of 1990. Discussions with licensee personnel resolved technical concerns regarding the adequacy of the planned submittal. This item is clor..

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9.-

Safety Assessment and Ouality Verification (IP 40500. 90712)

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The inspector reviewed selected portions of the licensee's self-assessment program to -

-verify implementation and determine if those programs contribute to the prevention of problems through monitoring and evaluating plant performance, providing assessments

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and findings, and communicating and following up on corrective action recommendations.

The inspector reviewed the following Plant Information Reports (PIRs) relative to station

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L procedures and regulatory requirements:

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9.1 PIR 90-001. Primary Drain Collecting Tank (PDCT) Overflow addresses the

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March 20,1990, incident when a level instrument failed causing an overflow of the PDCT. This resulted in a radiological spill in the chemistry sampling area and an offsite release via the waste gas loop seal and subsequently through the

primary vent stack. The PIR adequately detailed the root cause and corrective action to preclude recurrence. The licensee initiated a design change to provide additional levei inonitoring. This was reviewed in NRC inspection 50-29/90 07.

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No deficiencies in reportability were identified.

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. 9.2 PJR 90-002.' Omission of TS Reauired item from Auxiliarv Oocrator Lonsheet -

W Laddresses the April.14,1990, licensee identified procedure revision deficiency J

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. here..the _ position verification of safety injection-valve SI-MOV-1 was w

inadvertently omitted from the auxiliary operator logsheet. The PIR adequately detailed ' the licensee self-assessment.and corrective actions.

The licensee -

concluded that system operability was not comprised and the incident was not

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_ reportable as an LER._ This item was reviewed in NRC Inspection Report-50-

' 29/90-07. - No unacceptable conditions were identified in the licensee self-assessment ar management review of this incidera.

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' 10.-

Review of Periodic Reoorts (IP 90713) -

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.Upon receipt, the inspector reviewed periodic reports submitted pursuant to Technical Specifications. This review verified, as applicable that, the reported information was valid and included the required data; test results and supporting information were consistent with design predictions and performance specifications; and planned corrective L

actions were adequate for resolution of the problem. The inspector also ascertained L

whether any reported information should be classified as an abnormal occurrence; none

.was noted. The following report was reviewed f

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The September 1990 Monthly Statistical Report for plant operations.

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N-At periodic intervals during this inspection, meetings were held -with senior plant

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management to discuss the findings. A summary of findings for the report period was

.also discussed at the conclusion of the inspection and prior to report issuance.- No

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proprietary informationLwas identified as being included in the report.

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