ML20058P794

Insp Rept 50-443/90-15 on 900625-0729.Violations Noted.Major Areas Inspected:Operations,Radiological Controls,Maint/ Surveillance,Emergency Preparedness,Security,Engineering/ Technical Support & Safety Assessment/Quality Verification

ML20058P794
Person / Time
Site:	Seabrook
Issue date:	08/15/1990
From:	Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20058P788	List: ML20058P785 ML20058P790 ML20058P794
References
50-443-90-15, IEB-88-004, IEB-88-4, NUDOCS 9008200086
Download: ML20058P794 (16)

See also: IR 05000443/1990015

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V.S. NUCLEAR REGULATORY COMMISSION

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REGION I

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Docket / Report No.:

50-443/90-15

License No.: NPF-86

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Licensee: Public Service Company of New Hampshire

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Facility:'Seabrook Station, Seabrook, New Hampshire

Dates:-

June 25 - July 29,1990

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,. Inspectors:

Nl Dudley, Senior Resident Inspector

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.R. Fuhrmeister,. Resident Inspector-

A. Cerne_,: Senior Resident Inspector - Construction

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Approved By:

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Ebe C. McCabe, Chief,' Reactor Projects Section 3B

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-OVERVIEW

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,0,peration s:

Operators responded in an excellent manner to unplanned transients'-

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and performed power ascension. tests in a controlled, professional _ manner. ; A -

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violation was issued for failure to adequately control containment electrical

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distribution panel-breakers. An engineered safety features actuation was initi-

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ated by an operator error. Minor' industrial safety concerns (work practices)

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were referred to'the. licensee as i potentially recurrent problem. ~

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-Radio _ logical Controls:

Because'oi licensee-identification and appropriate'cor-

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.rection, no violation was cited f(r failure of an-auxiliary operator:to properly

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monitor his entry into a locked high radiation area.

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Maintenance / Surveillance:

Inadequate control of work resulted in:a turbine-

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generator " setback" to 40*4 power and in an unneeded replacement' of a' secondary

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! system pipe elbow.

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Emergency Preparedness:

Corrective actions were taken on a potential. weakness

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'dentified by the test of the Public Alert and Notification System.

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.. Security:. A supervisor tested positive for substance abuse during a random'

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' drug, screening and was suspended for 14 days.

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Evaluation of the turbine generator 180 Hz1

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Engineering / Technical-Support: :

ground fault. relay continued.

Safety-Assessment / Quality Verification:

The Self-Assessment Team report demon-

strated an ability to be self-critical and.was- consistent with NRC f's .ngs.

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However, the report lacked details supporting each assessment and explaining

the genesis of each-recommendation.

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TABLE OF CONTENTS

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1.0 S umma ry o f Ac t i v i t i e s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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2.0 . 0perations ( 71707, 90712, 92702, 93702, 92701) . . . . . . . . . . . . . . . . . . . . . . .

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2.1 Plant' Tours.....................................................

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2 . 2 ' P l a n t Op e ra t i o n s . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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2.3 -(90-08-04) Mi ssed Ai rlock Survei11ance. . . . . . . . . . . . . . . . . . . . . . . . . .

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2.4'(90-15-01) Actuation of Feedwater I solation Signal . . . . . . . . . . . . . .

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2.5 (90-15-02) Failure to Open Containment Distribution Panel-

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' Breakers.......................................................

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2.6 LER 90-016: ESF Actuation - Containment Ventilation Isolation...

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3.0 Radiological Controls

Unmonitored Entry Into A Locked High-

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Radiation Area.(71707, 92702) (90-15-03)...........................

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4;0 Maintenance / Surveillance (61726, 62703, 90712).......................

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4.1 Maintenance.....................................................

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4.2 Surve111ance.............................................

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' 4.3 Turbine Generator Setback.......,.....................,.........

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4.4 False Ultrasonic Test Indication............

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4.5 LER 90-003: Surveillance'on WRGM.................................

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4.6 LER 90-004: Surveillance on WRGM................................

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5.0 Emergency Preparedness (92701).......................................

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, 6.0 Security l(71707,2515/106).............................................

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6.1 Plant Tours..............................................-.......

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6. 2 Fi tn e s s- f o r-Duty S u s pe n s i on . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . * 0

7.0 Engineering / Technical Support - LER 90-015:

Turbine Trip With

Reactor Trip Due to Ground Fault- Relay Actuation (72701). . . . . . . . . . .- 11

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- 8'0 Safety Assessment / Quality Verification (40500).......................

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' 8.1 Review of Self-Assessment of Power Ascension Test Program 50%

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Powe r. Pl a t e a u . . . :. . . . . . . . . . . . . . . . . . .

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8.2" Review of-Welding Quality Records...............................

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19.0L Meetings (30703).....................................................'14

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DETAILS

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1.0 Summary of Activities

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1.1 Resident Inspector Activities

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Three resident inspectors were assigned to the site.

The 88 on-site

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inspection hours included 39 backshift hours, 21 of which were deep

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backshift hours.

1.2 Visiting Inspector Activities

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A team of region-based and headquarters-based inspectors provided

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24-hour per day coverage of operations.

Inspection of the power

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ascension test program will be documented in NRC Inspection Report

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50-443/90-83.

On July 24-26, 1990, a> region-based inspector reviewed New. Hampshire-

Yankee's response-to NRC Bulletin 88-04, " Potential Safety.Related

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Pump Loss'."

The results of the inspection will be-documented.in NRC.

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Inspection Report 50-443/90-16.

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On July 26, 1990, the Operational Programs Section Chief from the

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Region I Division of Reactor Safety toured the site and met with

licensee personnel.

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1.3 Plant Activities

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At the beginning of the inspection period,'the plant was in Mode 3

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Hot Standby, following a plant trip-from 30*4 power-upon actuation of

a 180 Hz ground voltage relay-on the main generator. .Following an

event evaluation, the reactor was taken critical on June 29, and power-

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was raised to 67*4.

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On June 30, power was reduced and the generator was disconnected from

the grid for repair of a steam leak on the heater drain tank condenser -

dump line.

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On July 2, af ter power was raised to 75*f, a " setback" to 40*4 power

was caused-by improper shifting of.the-main generator step-up trans-

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former cooling fans.

Power was further reduced to disconnect the

generator from the grid nd replace an elbow in the heater drain tank-

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condenser dump line,

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On July ~ 5, af ter testi g was resumed at 75'o power, the reactor was-

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tripped by a turbine electro-hydraulic control system low oil pressure

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signal which was caused by vibration of pressure switches mounted on

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the turbine stop valves.

The pressure switches were relocated to a

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floor support beam and power was returned to 75*4 on July 8.

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On July 22, the 50% load reject test from 75% power was successfully

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conducted, completing the 75% plateau tests.

Following review of

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test results for the 75% plateau, power was raised to 90% on July 15.

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On July 16, power was reduced to 75% to place the moisture separator

reheaters in service. On July 19, reactor power was raised to 100%.

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On July 26, a 10% load swing and a 50% load reduction were success-

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fully completed as part of the power ascension test program. On July

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28, af ter power. had been returned to 100%, a planned reactor trip was

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performed, followed by successful completion of the natural circula-

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tion test.

2.0 Operations

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2.1 . Plant Tours-

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Daily control rocm tours and sustained-control room observations were.

conducted.

Reviews were performed of operator log books, technical

specification action statement tracking. logs, tagout logs, and night

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orders. Assessments were made of technical specification action

statements in effect, control room staffing, management oversight,

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operator awareness of plant conditions, and responses to abnormal

events. No defi

encies were noted.

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Plant tours included the primary auxiliary. building, containment,

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pipe chases, turbine building, switchgear rooms, diesel building,

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circulating water building, service water building, cooling towers,

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and intake structures.

No equipment problems were identified. Minor

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discrepancies were turned over to the licensee and adequately resolved.

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Examples of poor industrial safety practices (e.g., standing on a

flexible chair to rack out a breaker, not wearing eye protection in

an area where it-is required) were discussed with. plant management.

Poor industrial safety practices were previously identified in NRC

. Inspection Reports 50-443/89-83 and 50-443/90-05.

The inspector con-

cluded that continued licensee attention is needed in this area.

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-The inspector noted the presence of equipment labels.

Instruments

.and.their isolation and test valves were observed to have stainless

steel plates stamped or vibra-etched with.the component identifica-

tion attached.

These markings included high and. low pressure con-

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nections for differential pressure ~ detectors and inlet and outlet

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connections for flow totalizers.

The inspector considered this to be:

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an adequate equipment identification initiative,

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2.2 Plant Operations

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Unplanned transients during the period included a reactor trip from

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75% power, several feedwater system flow oscillations, and minor tur-

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bine load swings. The operators responded in an outstanding manner.

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They were alert, well-informed and reacted properly. During the trip

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from 75% power,= the Unit Shif t Supervisor provided the necessary

leadership, and the crew stabilized the plant in Mode 3.

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The inspector witnessed crew training for the 50% load reduction from

100% power, the unit trip from 100% power, the natural circulation

test, and the loss of offsite power test.

That training included

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both classroom and simulator portions. The crews selected for each

test actively participated in the training and became involved in

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developing the best method for controlling expected plant response.

'The inspector witnessed the major plant transient tests conducted ai

100% power.

The crews were well-disciplined and communicated well

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with personnel from other departments.

Plant equipment responded as

designed.

The operators efficiently stabilized plant conditions and

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. restored normal operating conditions. The inspector concluded that

the operators were well-trained and performed the tests in an excel-

lent manner.

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During observed power changes, the licensee's staff was professional

and understood their. responsibilities.

During shif t turnovers . the

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crews communicated well with their counterparts and among each other.

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Communications between licensed and auxiliary operators were good

both during shift briefs at the beginning of.and throughout the shift.

'The inspector concluded that the operations staff was very knowledge-

able, understood their plant, was cognizant of their own and each

other's responsibilities, and was very professional in operation of

the facility.

2.3 f(Closed) Violation

90-08-04:

Missed Airlock Technical Specification-

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15) Surveillance

The inspector reviewed the actions described in New Hampshire Yankee's

June 6, 1990, letter which responded to a letter from the NRC dated

May 8, 1990.

The actions described were to prevent recurrence of

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missing required Technical Specification surveillances of containment

air locks.

The inspector verified signs were in place at the air locks, reviewed

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the lesson plans for licensed operator requalification training

(L5005C) and for the Auxiliary Operator Continual Training Program

(N5003C), and observed the Security Department notifying the main

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control room of containment air lock outer door openings. The in-

spector concluded that adequate steps were taken to prevent recur-

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rence of this event.

This violation is closed.

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2.4 Actuation of a Feedwater Isolation Signal

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On July 5,1990, following an unplanned reactor trip from 75's power,

a feedwater isolation signal was produced by several momentary er-

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roneous actuations Of stcam generator (SG) high level trip signals

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over a three second time period.

No indication of actuation of the

SG high level trip signal was produced on the main control board.

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order _ to reset the feedwater isolation signal, steam generator level

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was raised above the low level trip setpoint and the reactor trip

breakers were shut.- This-delayed securing the turbine-driven emer-

gency feedwater pump and thereby added to the cooldown of the reactor

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coolant system following the reactor trip.

Through discussions with Westinghouse, New Hampshire Yankee determined

that the high steam generator level signals were caused by " ringing"

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of the two SG level Rosemount 1153 transmitters which share the upper

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. steam generator taps with the steam flow transmitters. The ringing

was postulated to be caused by pressure pulses created by rapid

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closure of the turbine control valves after the reactor tripped.

Westinghouse recommended installation of an indication circuit modi-

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fication similar to the ones installed at three other Westinghouse

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fa:ilities in the United States.

T3e inspector discussed the event with the licensee and noted the

dmplications this event has for controlling plant cooldown following

a reactor trip and for restoration of feedwater af ter a loss of feed-

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water event.' The engineering department is evaluating the need for

the modification and the operations department is evaluating addi-

tional procedural guidance.

Pending NRC evaluation of tne adequacy

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of licensee corrective actions, this issue is unresolved (UNR 90-15-01).

2.5 Failure to Open Containm nt Distribution Panel Breakers

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On July 21, 1990, with the plant in Mode 1, the operators discovered

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that the circuit breakers feeding the primary containment power dis-

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tribution panels 1-ED-PP-7A and 78 had been shut for over 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />,

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contrary to the requirements of Technical Specification 3.8.4.1. . The

breakers were op'ined to meet the requirements of the Technical Speci-

fication action atatement and were immediately reclosed for eight and

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a half hours while work was conducted in the containment.

The in-

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spector concluded that there was inadequate review, prior to reshut-

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ting the breakers, of the consequences of operating with the breakers

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closed for greater than 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />. Also, the inspector noted that the

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operators exhibited no knowledge of the ru sen for this breaker control

requirement.

The initial closure of the circuit breakers was properly logged in

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the Control Room Operator's log. When the breaker closure extended

to the next shift, the closure was not entered in the- Action Statement

Status 1.og as required,

The fact that the breakers were closed was

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not noted until another containment ontry was initiated 78 hours9.027778e-4 days <br />0.0217 hours <br />1.289683e-4 weeks <br />2.9679e-5 months <br /> after

the initial closure of the breakers. The inspector verified that the

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five closures of the breakers which occurred after this occurrence

were controlled by entries in the Action Statement Status log and

that each lasted no more than 8 hours9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br />, compared to the 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />

allowed by the Technical Specifications.

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Operations Management Manual procedure OP 10.6, " Action Statement

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Tracking," tracks and controls the shutting of the circuit breakers.

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After this event.,-the licensee initiated development of a procedure

for controlling program requirements connected with containment en-

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tries. That procedure was not completed by the end of the reporting

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period.

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Inspector evaluation concluded that OP 10.6 was ineffective in meeting

the Technical Specification 3.8.4.1 exception requirement for an

operating procedure to control breaker position.

Failure to lock

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open the circuit breakers violated Technical Specification 3.8.4.1

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requirements which assure containment integrity (NV4 90-15-02).

2.6 _(Closed) Licensee Event Report (LER) 90'-016:

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Containment Ventilation Isolation

On June 24, 1990, an auxiliary operator inadvertently deenergized .

vital 120 VAC power panel PP-1E.

That'resulted in the de-energization

of the Train "A" radiation monitor for the containment on-line purge

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(COP) system and in the initiation of a Containment Ventilation Isola-

tion signal. The C0P system, which was in service purging the con-

.alnment, responded as designed.

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'This event was attributed to personnel error. The uninterruptable

Power Supply to panel PP-1E had failed due to a capacitor: failure and'

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a' blown fuse. The power supply to PP-1E was automatically transferred

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to the maintenance power supply, While investigating the problem,

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the auxiliary operator opened the maintenance power supply breaker to-

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PP-1E as a result of: confusing directions from the main control room;

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inaccurate labeling of the maintenance supply breaker; and insufficient

understanding of the power supplies to-the Vital 120 VAC panels.

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Immediately after the event, the inspector verified, by walkdown of

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the electrical switchgear and discussions with auxiliary operators,

an electrical engineer, and an electrical technician, that the loca-

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tion of supply breakers zo the vital 120 VAC panels was not well

understood,

The auxiliary operator was counseled by the licensee and auxiliary

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operators on other shifts were briefed on the importance of verifying

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the ~ proper equipment and the required actions before performing an

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evolution.

Requests for Engineering Services were submitted for re-

labeling escutcheon plates to more accurately identify feeder breakers

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and for developing placards of one-line drawings of the inverter power

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supply feeders to be placed on the individual inverters.

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The inspector reviewed the, Station Information Report (90-044) and

the Licensee Event Report and concluded that the licensee's review of

the event and the corrective actions taken were adequate.

3.0 Radiological Controls - Unmonitored Entry Into A Locked High Radiation Area

On June 29, 1990 an auxiliary operator (AO) entered the "Demin Alley" in

the primary auxiliary building, a locked high radiation area, without a

survey meter.

This was identified by the licensee as a violation of Tech-

nical Specification 6.11.1.

Demin Alley 'was a locked high radiation area due to radiation fevels -of

2 R/hr (roentgens / hour) at the bottom of the Demineralizer Filter Vaults.

General area radiation levels in the passageway were less than 0.2 R/hr.

Access to locked high radiat' ion areas is controlled through use of security

key cards and Radiation Work Permits (RWPs).

The Health Physics (HP) Super-

visors adinistratively control access through use of a computer at the-

Radiation Control Area access point. At the time of the violation,-118

people were provided unlimited access to locked high radiation' areas.

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RWP used.for routine A0 tours provided directions on using a survey meter

when entering a locked high radiation area. Once provided access, entry

into a locked high radiation area is by use of the key card.

As a result of this event, the licensee removed all HPs and auxiliary

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operators from general access to all high radiation areas and now requires

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auxiliary operators to check with HP supervisors at the access point, to

pick up a radiation meter, to review survey maps, and to be logged into

the HP access computer before entering a high radiation area.

The inspector reviewed the corrective-actions and verified, by reviewing

the dosimetry logs, that the individual entering the locked high radiation

area received no significant dose. The inspector concluded that adequate

corrective actions were taken to prevent recurrence and that this specific

violation had a low safety significance. Based on the NRC Enforcement

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Policy (10 CFR Part 2, Appendix r Section V.G.L), this nonrecurring lic-

ensee-iden J fied violation is classified as a Severity Level IV, non-cited

violation (NCV 90-15-03).

4.0 Maintenance / Surveillance

4.1 Maintenance

The inspector observed activities related to the replacement of piping

system components in the heater drain system.

The inspector reviewed

Maintenance Work Order MWO-90W003638 and supporting documentation for

replacement of a 6" elbow.

The inspector also reviewed MWO-90 WOO 3637

and supporting documentation for replacement of a concentric reducer

which had developed a through-wall leak.

Both MW0s identified the

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location of the work in reference to plant stations and with respect

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to easily identifiable system components. Appropriate procedures for

the work to be performed were included in the werk packages, and work-

men adequately identified the equipment and necessary safety precau-

tions.

The inspector also noted the presence of quality control per-

sonnel at the work site,

The inspectur also observed the resetting of the gain settings for

steam generator feedwater control valves. The valves were then pro-

perly stroked in accordance with work request 90 WOO 3729.

The inspector concluded that the observed maintenance was conducted

properly.

4.2 Surveillance

The inspector observed the following surveillonce activities and con-

cluded that they were satisfactorily performed and documented.

The

surveillances were performed to approved and current procedures and

the results were properly reported by qualified technical personnel.

Containment Personnel Airlock

A semiannual verification of the interlocks on the Containment Per-

sonnel Airlock, required by Technical Specification 4.6.1.3.c, was

performed in accordance with EX1803.

The interlocks between the inner

and the outer doors were operated to assure that only one door in the

air lock can be opened at a time.

Residual Heat Removal Pump Vibration Readings

Vibration readings were taken during a quarterly surveillance of

Residual Heat Removal (RHR) Pump "B."

The surveillance is required

by the Technical Specification 5.4.2.f and was satisfactorily con-

ducted in accordance with OX413.01B.

Auto-Start and Operability of Emergency Diesel Generator

A monthly auto-start surveillance of Emergency Diesel Generator EDG-1B,

required by Technical Specification 4.8.1.1.2, was satisfactorily

performed in accordance with OX1426.19.

Concurrently, a quarterly

surveillance of the diesel generator cooling water and air start system

valves was satisfactorily performed in accordance with OX1426.14.

This quarterly surveillance is required by Technical Specification 4.0.5 and the ASME Boiler and Pressure Vessel Code, 1983 Edition.

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A 30-day and a monthly surveillance were satisfactorily performed

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concurrently on the motor-driven Emergency Feedwater Pump "B" in

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accordance with OX1436.03A.

These surveillances included taking

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vibration readings and are required by Technical Specification 4.0.5

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and 4.7.1.2.1.

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4.3 Turbine-Generator Setback

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On June 26, 1990, the turbine generator received a setback signal

'which reduced power from 75?4 to 40%.

The signal was caused by loss

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of the generator transformer fans during a weekly repetitive task:

sheet (RTS) surveillance for shifting the, fans to equalize equipment

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run time.

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- A similar problem occurred on June 12 when power could not be raised

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.above 40*; due to a locked in setback signal produced under the same

RTS. At that time, a caution tag was hung on the control selector

switch and a caution was added to the RTS to only perform the sur-

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ve111ance below 40?f power. The bank selector switch was tested and,

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since no-setback signal was produced, no caution tag was.ettached to

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the switch.

Further licensee investigation after the second occur-

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rence determined that the bank selector may or may not start one bank

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of fans before the other bank is stopped.

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The RTS used in the field did not have a caution for performing the

RTS above 40% power. To prevent recurrence of a similar event, the

software program for changing RTSs was modified to list all related

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RTSs which had been issur.d but not closed out when a change is entered.

4.4- False Ultraconic Tes? Indications .

Due to ultrasonic test (UT) indications of excessive wall t'hinning on

a six-inch elbow on the heater drain tank system condenser high level

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dump line, reactor power was decreased and the elbow was replaced.

Upon removal, the elbow showed no wall thinning.

A subsequent inves-

tigation was conducted and presented in a report issued by the Nuclear-

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Quality Group on July 24, 1990.

The inspector reviewed the report and conducted an independent UT of:

the affected piping after it was removed. One New Hamp; hire Yankee

technician-performed the initial-thickness measurement and backup-

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measurements using a UTM110 bitrasonic Digital Thickness Meter which

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indicated a wall thickness as low as 0.003" in the outer bend of the

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6" elbow.

An engineer using infrared thermography confirmed the sus-

pected wall thinning.

Upon removal, the elbow showed no wall thinning.

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The licensee attributed the inaccurate UT readings to the limitations

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of the UT meter and to fluid flow in the pipe.

Erroneous low readings

have been noted at other plants.

The UT meter manufacturer's maximum

test specimen temperature was 122 F; the elbow temperature was measured

at 165 degrees F.

Effects on the meter by system vibrations have not

been quantified or qualified by the manufacturer.

The licensee was able to reproduce the erroneous readings after the

installation of the new elbow only when flow was present in the pipe.

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The erroneous infrared thermography reading was determined to be caused

by the lack of an established program for the use of the detector.

The licensee concluded that an established method for esing detectors

for NDE measurements requires further development.

The report recommendations which were accepted by New Hampshire Yankee

management for implementation included the following.

Establishment of a formal policy for obtaining such NDE informa-

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tion.

Securing components prior to performing such UT or similar tests.

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Purchasing UT flow detection equipment and a high temperature

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probe.

Developing procedures for use of all portable testing equipment

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to ensure that limiting conditions are identified.

False indications of inadequate wall thinning like this one are a

problem because they result in urnecessary work, but they are not a

nuclear safety hazard. The inspector cor.cluded that the licensee's

review of the false UT readings and the planned corrective actions

were adequate.

4.5 (Closed) Licensee Event Report (LER) 90-003:

Noncompliance With

Jechnical Specification Action Requirements for Inoperable Wide

Range Gas Monitor

This LER reported the inoperability of the backup sample pump for the

Wide Range Gas Monitor (WRGM).

This event occurred while the WRGM

was out of service for nedification. The apparent cause was the power

fuse becoming dislodged due to work activities in the area.

The fuse

was-reinserted and the pump was returned to service.

Long-term cor-

rective action has been completed with installation of a stainless

steel guard to prevent inadvertent operation of the power switch or

dislodgement of the power fuse.

The inspector verified the presence

of the guards on the pumps in the field.

This item is closed.

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{. Closed) Licensee Event Report (LER) 90-00_4: Noncompliance With

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Technical Specification;

'de Range Gas Monitor Inoperable

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This event occurred while restoring the WRGM to. service after modifi-

cation. The technician performing the surveillance test for oper-

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ability did not adequately complete thc test due to an oversight.

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This resulted in the use of a flow value other than actual flow.

The

system flow calculation circuit was reset properly and the Repetitive.

Task Sheet for the test was revised to ensure completion of the entire

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calibration procedure.

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This item is closed.

5.0 Emergency Preparedness

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The inspector discussed the siren failures that occurred on May 16, 1990

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with the licensee. As a re ; ult of the failures, new model Antenna Special-

ist' Company directional antennae are scheduled for installation on seven

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poles, including the four poles where the failures occurred. Additional.

antennae replacements are planned on an as-needed basis.

In addition, one.

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antenna was elevated for better reception,

The new antennae are expected.

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to prevent recurrence of the failures.

Testing of signal reception prior

to and after the siren test did not repeat the' siren failures.

6.0 Security

6.1 . Plant Tours

The inspector. observed security personnel conductingiroutine activi-

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ties including personnel access control, plant tours, and Central

Alarm Station (CAS) activities.

The inspector reviewed the CASJshift

logs and verified appropriate compensatory measures were established.

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The inspector verified an adequate understanding by-plant.personnei-

of the Fitness-For-Duty (FFD) program by conducting interviews with

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maintenance technicians, 1&C technicians, auxiliary operators and a

licensing engineer.

One weakness in personnel knowledge was a lack

of understanding of the details of an individual's right of appeal..of

a positive substance abuse test.

No unacceptable-conditions were

identified.

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6.2 Fitness-For-Duty Suspension

- On July 20, 1990, a non-licerised supervisor tested positive for a

controlled substance during a routine random screening as part of.the-

fitness-for-duty (FFD) program.

The individual's site access was

suspended for 14 days.

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The Nuclear Quality Group initiateG a review of work for which the

individual was responsible for 6e last six months.

Interviews with

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individuals who worked with or for the supervisor were conducted.

No

work'concerning equipment covered by Technical Specifications was

identified. Work associated with equipment covered by quality assur-

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ance requirements is being reviewed. A walkdown of all areas and

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equipment uorked on by the supervisor is planned.

Interviews with

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plant personnel identified no indication of abnormal behavior. A

final report is expected to be issued on July 30, 1990.

The inspector concluded that the requirements of the FF0 program was

properly performed, that required notifications were made, and that

an adequate investigation was in progress. The inspector had no fur-

ther questions.

7.0 EngineerinJ/ Technical Support - Licensee Event Report (LER) No. 90-015:

Turbine Trip With Reactor Trip Due to Ground Fault Relay Actuation

"

This LER provides the status of the root cause analysis of the Juna 20,

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1990. reactor tt;p from 30*o power upon actuation of the 180 Hz main ;&ne-

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rator ground fault relay. The relay is an added (not required for genera-

tor warranty) protection feature manufactured by ASEA Brown-Boveri.

It

provides ground fault protection for 5% of the generator windings by moni-

toring 180 Hz voltage using a neutral transformer secondary circuit with a

bandpass filter arrangement.

The other 95?4 of the windings are protected

by a separate relay (manufactured by General Electric) which monitors the

60 Hz voltage on the nettral phase grouni.

New Hampshire Yankee conducted an event evaluation which included a Doble

test of the generator windings, vertfication of bus and bushing cooling

systems, consultation with the relay and generator manufacturers, and test-

ing of the 120 Hz ground fault relay.

No cause for the trip was identified.

The trip function of the relay was bypassed, voltage across the relay was

monitored, and a reactor startup and power escalation was conducted.

Two

power increases above 30?; power have since been conducted, and the relay

has :brmed once. The monitored voltage did not reach the trip set point.

New Hampshire Yankee is continuing to evaluate the relay and its use as an

alarm or 65 a supplementary trip. No unacceptable conditions were 'dentified.

8.0 Safety Assessment / Quality Verification

8.1 Review o.f Self-Assessment of Power Ascension Test Program 50?s Power

Plateau

New Hampshire Yankee established a Self-Assessment Team (SAT) to

evaluate power ascension testing at Seabroak Station.

This assessment

is being conducted in two phases.

Phase 1 was the assessment of plant

readiness to conduct power ascension testing and Phase 2 is the

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assessment of the implementation and conduct of the power ascension

testing program.

The Self-Assessment Team program, which defined the

areas to be evaluated and the criteria to be used, was reviewed in

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NRC Inspection Report No. 50-443/88-13 and determined to be well-

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directed. New Hampshire Yankee's Phase I self-assessment report was

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reviewed in NRC Inspection Report No. 50-443/88-15, and was determined

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to be formal, well-structured and self-critical. As part of the Phase

2 assessment, New Hampshire. Yankee completed their Power Ascension

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Test Program. Fifty Percent Power Self-Assessment Report on June 18,

1990 and presented their findings to the NRC during a public meeting

at Kingnof Prussia, Pennsylvania en June 19, 1990.

The self-a'ssessment team (SAT) report was reviewed by the NRC staff .

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The: inspector also reviewed the supporting docmmentation used to de-

velop.the assessment findings. The self-assessment was conducted in

nine specif.ic topical areas.

Fifty-three recommendations for program,

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procedure or management control enhancements were identified.

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The. recommendations and assessments were derived from performance-

based. inspections conducted by experienced personnel using criteria

established by the SAT program.

The results of each inspection were

documented.on a check list or in notes which provided a basis for

assessment,..The inspector reviewed the-bases for the assessment of

interf aces in the topical areas.

SAT recommendations were made to improve weaknesses, to formalize and

quantify requirements, and to improve the efficiency of existing pro-

grams.

For example, renommendations 4018 and 4019 suggest developing

goals, objectives and a performance measurement system to enhance the

maintenance program.

Recommendatior. 4049 suggests that line manage-

/ ment conduct a self-critique of their participation in the event pro-

cess -for the turbine trip for the purpose of identifying possible

inefficiencies in thei

review process.

Recommendations were assigned completion dates and tracked on the

Integrated Control Tracking System.

Two of the fifty-three recom-

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mondations were classified as requiring short-term actions. All-other

recommendations were considered long-term actions since resolution of

2the items did not affect continuation of the poser ascension test

program.

Six of the recommendations were completed at the= time of

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report. issuance.

Other observations such as improper deficiency tags

on,the main-control board were corrected immediately and later re-

-verified.

Recommendation 4016B to document troubleshooting activities

on wcrk requests was implemented and later formalized by a procedural

regiirement on July 1,1990.

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In conjunction with a review of the basis for SAT assessments, the

inspector evaluated the program requirements for the responsibilities

of the' Work Control Supervisor, for the conduct of a mode change check

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list meeting, for the implementation of the specific appraisal progics.

and for the responsibilities of the Station Operation Review Committee.

The inspector found the program requirements to be acceptable.

The inspector reviewed the basis for the recommendation for annual

review, for outstanding ,ork requests (WRs) and requests for engi-

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neering' services (RESs). A yearly review of these items coincides

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with the annual upda<.ng of the five year plan and is consistent with

the 48-week system schedule.

Reviews of open WRs and RESs are con-

ducted more frequently by Plan-of-the-Day meetings on a daily basis

and by the Station Modification Review Committee every two to three

weeks.

TheEinspector reviewed the bases for program limits or requirements

presented in the SAT report.

The use of hydrogen peroxide to clean.

the reactor coolant system was verified to have been appropriately

researched prior to being approved for use at the site.

The require-

ment for review of radiological jobs in the range of one to less than.

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23 man-rem at the job supervisor / planner level was verified to be a

recommendation from a utility owners' group.

The use of plant in-

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struments calibrated by the normal program as test instruments was

verified to be in accordance with Regulatory Guide 1.68.

The NRC staff observations made during the Power Ascension Test Program

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corresponded well with the recor. ondations presented in the report.

However, the SAT report lacked detailed' support for tbo cssessments

and did not explain the origins of the recommendations.

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inspector review verified that the assessments-were adequately supported

by well-developed SAT performance based observations, defined criteria,

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and industrial standards.

The SAT report was therefore evaluated as

satisfactory. Overall, the NRC staff concluded that the SAT report

demonstrated an ability- to be ef fectively self-critical.

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8.2 Reviews of Welding Quality Records

As documented in previous NRC Region I Inspection Reports, additional

inspection of weld quality -for piping systems was initiated as a re-

sult of Congressional. int? rest.

During this period, NRC resident

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inspector review of constiuction records continued.

Region..I inspectors have prcvided support and assistance to an In-

dependent Review: Team (IRT) Lut did not participate in the IRT in-

spection.

In addition to providing'suppor to the Independent Review Team, the

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resident inspectors separately reviewed activities to assess the

acceptability of licensee record controls, retrievability, and quality

verification for the welding /NDE areas of-interest. -Construction

procedures were reviewed, quality assurance and engineering personnel

were interviewed, and quality documents were examined to evaluate the

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sampled construction processt.s-for documented verifiability of' quality

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from'ai historical standpoint. As of; the end of_ _this report period,'-

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'the record' review and' quality verification inspection activities.were

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' Resident inspector: review during this inspection period has produced"

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.no' evidence of defective work, improper process controls, incomplete

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Inspection

Lead:

Dates ,

Subjects'

Report No.

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Power; Ascension Interim Exit

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Power Ascension Interim Exit

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