ML20058H255

Insp Rept 50-309/90-19 on 900919-1031.Violations Noted.Major Areas Inspected:Operations,Radiological Controls,Maint, Surveillance,Security,Engineering/Technical Support & Safety Assessment/Quality Verification

ML20058H255
Person / Time
Site:	Maine Yankee
Issue date:	11/06/1990
From:	Mccabe E NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20058H253	List: ML20058H252 ML20058H255
References
50-309-90-19, NUDOCS 9011150036
Download: ML20058H255 (16)

See also: IR 05000309/1990019

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U.S. NUCLEAR REGULATORY COMMISSION

Region 1

Report 50 309/90-19

License No.: DPR 36

Maine Yankee Atomic Power Plant

Wiscasset, Maine

September 19,1990 to October 31,1990

Inspectors:

Charles S. Marschall, Senior Resident Inspector

Richard J. Freudenberger, Resident Inspector

Maitri Banerjee, Resident Inspector, Oyster Creek

Thomas Koshy, Senior Resident inspector, Yankee Rowe

Approved:

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E. C. McCabe, Chief, Reactor Projects Section 3B

Date

OVERVIEW

Operations: The plant shutdown for main lube oil pump seal ring replacement was conservative

and appropriate. Three open items were closed,

Radiological Controls: A weekly meetin.} to update plant employees on the radiological effects

of fuel leakage was assessed as a

sitive initiative for minimizing exposure to workers. An

unlatched high radiation area door c.s not cited because Section V.G. of the Enforcement Policy

was satisfied. An open item concerning termination exposure letters was closed.

Maintenanec/Survel11anct A decision, by operations and maintenance personnel, to rerun a post-

maintenance test on an Emergency Feedwater Pump was considered appropriate. Component

substitution control during maintenance is to be reviewed by the licensee to ensure that 10 CFR 50.59 is appropriately considered. An open item concerning valve maintenance was closed.

Security: Maine Yankee is evaluating measures to provide additional security key control.

Engineering / Technical Suonort:

Apparent violations were identified for an unqualified

environmental configuration for limit switches and for failure to promptly correct that problem

when a switch failure occurred in February 1990. Apparent violations were identified for not

implementing TMI related changes and for failure to provide correct information required by the

associated NRC Order. Four open items were closed.

Sakty Assessment /Oualltv Verification: A need to review nuclear instrument setpoint changes

and reconcile connicting technical specifications was identified. Housekeeping showed signs of

lack of management attention. A weakness in establishing licensing priorities and a need for

management review of resolution of safety issues was evident during Maine Yankee pursuit of

an NRR Temporary Waiver of Compliance. The need for periodic surveillance of the 15% trip

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bistable is unresolved,

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TABLE OF CONTENTS

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Plant Operations (IP 71707, 92701) . . . . . . . . . . . . . . . . . . . . . . . . , . . .

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Radiological Controls (IP 71707, 92701) . . . . . . . . . . . . . . . . . . . . . . . .

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Maintenanec/ Surveillance (IP 6'e /03, 61726)

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Physical Security (IP 71707) . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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Engineering / Technical Support (IP 71707, 92701, 92702) . . . . . . . . . . . . . . .

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~ Safety Assessment / Quality Verification (IP 71707) . . . . . . . . . . . . . . . . . .

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Administrative . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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7.1

Persons Contacted . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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7.2

Summary of Facility Activities

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7.3

Interface with the State of Maine . . . . . . . . . . . . . . . . . . . . . . . .

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Exi t M ee ti n g ' . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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- In spection Hou rs . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . .

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DETAILS

1.

Elapt Operations GP 71707. 92701)

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During routine daily facility tours the following were checked: manning, access control,

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adherence to procedures and LCOs, instrumentation, reccrder traces, protective systems, control

room annunciators, radiation monitors, emergency power source operability, operability of the

Safety Parameter Display System (SPDS), control room logs, shift supervisor logs, and operating

orders. Weekly, selected Enginected Safety Feature (ESP) trains were verified to be operable.

The condition of plant equipment, radiological controls, t.ecurity and safety were assessed. Bi-

weekly, the inspector reviewed a safety related tagout, chemistry sample results, shift turnovers,

portions of the containment isolation valve lineup, operability of selected Engineered Safety

Feature trains, and posting of notices to workers. Plant housekeeping and cleanliness were

evaluated also.

1.1

Main Turbine Lube Oil Pump Seal Rine Replacement

On September 29, with the plant at full power, operaicts noticed that main turbine tube oil pump

discharge pressure dropped from 377 psig to 372 psig. On September 30, maintenance and

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operations personnel eliminated several possible causes for the reduced discharge pressure while

pressure continued to drop slowly. The inspection cover was removed from the south end of the

tube oil pump housing, and oil was seen spraying from the seal ring. At 1:45 p.m. plant

mangement directed operators to commence a shutdown. The unit was taken offline, the seal

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rings were repaced, and operators made the reactor critical on October 5. When turbine speed

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was increased to 1800 rpm (rated speed), lube oil pump discharge pressure reached only 225

psig. Additional trouble-shooting confirmed that the sensing line associated with the discharge

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pressure instrument wa3 leaking. Based on the lack of available information on seal ring

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performance from the vendor, plant management's decision to shut the plant down was assessed

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as conservative and appropriate.

1.2

(Closed) Violation 89-18-01: Recurrent Overfill of the Resin Storace Tank

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The inspector reviewed the licensee's response to Inspection Report 50 309/8918 (MN 90-08),

the follow up response to the same inspection report (MN 90 37) and an inspector observation

of waste handling system activities documented in Region I Inspection Report 50 309/90-02,

Detall 2.a. These responses appropriately addressed the violation as well as two related issues

involving the use of personnel who were unfamiliar with the task and the influence of schedular

pressure.

In this case, the licensee concluded tha' neither of these issues were factors

contributing to the violation. Long-term human factor and technical improvements to the waste

handling system to enhance reliability and operability were discussed with Plant Manager during

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a meeting on October 26.

The licensee plans to implement CPA (Conceptual Project

Assessment) Number 89114a, " Installation of a Radioactive Waste Processing Skid," which

includes improvements to the waste handling systems prior to the 1991 refueling outage.

In the interim, the inspector concluded that sufficient action has been completed to allow safe

system operation. This item is closed.

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1.3

(Closed) Unresolved item 50-309/89 22 01: Corrective Actions Associated with T-1H

Switching Error

The intpector reviewed Event Review Board Report Number 009, *T-lH Switching Error.* The

root causes determined by the Event Review Board were lack of proper command and control

by the PSS (plant Shift Supervisor) and personnel error by the AO (Auxiliary Operator). The

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recommendations for corrective action from the board report were appropriate to address the

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causes identined as well as a number of issues that were contributors to the event. Part of the

corrective action included the placement of the PSS in a performance improvement program

which addressed improved communications formality and proper use of the chain of command.

In this case, the Event Review Board provided a strong self evaluation which became the basis

of appropriate corrective actions. This item is closed.

1.4

(Closed) Violation 50 309/90-10-01: Pressurizer Level Drain

The violation involved inadvertent draining of the pressurizer and reduction of the inventory in

the reactor coolant system during reactor coolant system vent and fill during the 1990 refueling

outage. Corrective actions were addressed in NRC Inspection Report 50-309/90-10, Detail 1.a.

The inspector reviewed additional corrective actions outlined in the licensee's response dated

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August 23,1990 (MN 90 81). These additional corrective actions included revising procedures

associated with calibrating pressurizer pre:sure and level transmitters to ensure sensing lines are

backfilled upon completion of calibration. At the close of the inspection, management was

reviewing changes to Procedure 1 26-4, Responsibilities and Authorities of Operating Personnel.

The changes emphasize the importance of believing and acting upon instrumentation, and the

importance of halting evolutions when unexpected conditions arise. The licensee also plans to

review applicable procedures for other vital differential pressure transmitters for adequate

backntting. The inspector assessed the actions taken as adequate to prevent recurrence, if

implemented as required. This item is closed.

2.

Radiological Controls (IP 71707. 92701)

Radiological controls were routinely observed. Areas reviewed included Organization and

Management, external radiation exposure control, and contamination control. Standard industry

radiological work practices, conformance to radiological control procedures and 10 CFR Part 20

requirements were observed.

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2.1

Locked High Radiation Area Door Found Unlatched

On October 9, a security guard on tour determined that a locked high radiation area door was

not latched as required. The door came open when the guard attempted to verify that it was

latched. Guard force verineation of locked high radiation area doors is intended to identify

unlocked high radiation area doors and is a corrective action from a previous violation. The

guard identified a problem with the door closer, immediately latched the door, and reported the

incident to the PSS (Plant Shift Superintendent) and an HP (Health Physics) technician. The PSS

generated a Denciency Report to fix the door closer.

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Technical Specification 5.12.2 requires that locked doors shall be provided to high radiation areas

to prevent unauthorized access. This item is not being cited because of licensee identincation,

low safety significance, nonreportability to the NRC, adequacy of corrective actio ., c.~i the NRC

assessment that previous corrective actions could not reasonably have been expected to climinate

this occurrence (NCV 50 309/9019 01).

2.2

Weekly RadiologicaLProtection Presentation

On October 5, the inspectors observed a weekly meeting conducted by Radiolcgical Protection

management to update plant employees on radiological conditions in the plant. The meeting

addressed the effects of identified fuel leakage, discussed the operational concerns, presented the

effects on workers, and dealt with plans to minimize contamination and exposure to workers.

The inspector considered the weekly meetings a positive initiative for minimizing exposures.

2.3

(Closed) Unresolved item 50-300/89-25 01: Termination Exposure Letters

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As documented in NRC Inspection 50 309/89 25, licensee personnel identified ihat the report of

personnel monitoring required by 10 CFR 20.408 was not completed within the required time

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frame. Inspector review of this issue found no instances oflate termination letters since January

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1990. This was achieved mainly by increased sensitivity to the issue on the part of licensee

personnel.

Licensee long term corrective action plans include the comparison of security

termination information with the dosimetry termination information. The licensee committed to

implement this action by January 1,1991. The inspector found these actions to be acceptable

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to prevent recurrence. This item is closed.

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Maintenance / Surveillance GP 62703. 61726)

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The inspector observed and reviewed maintenance and problem investigation activities to verify

compliance with regulations, administrative and maintenance procedures, codes and standards,

proper QA/QC involvement, safety tag use, equipment alignment, jumper use, personnel

qualincations, radiological controls for worker protection, retest requirements, and reportability

per Technical Specifications.

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Also, the inspector observed parts of surveillance tests to assess performance in accordance with

approved procedures and Limiting Conditions for Operation, test results, removal and restoration

of eaulpment, and deficiency review and resolution. The following activities were considered

noteworthy.

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3.1

limergency Feedwater Pump Maintenance

On October 16, during a quarterly surveillance on Pump P-25C, a motor-driven Emergtncy

Feedwater Pump, the inspector observed adjustment of packing on the inboard and outbotrd

pump seals. Maintenance was accomplished using guidance in Procedure 5-34 1, Pump Packii.o,

Adjustment, Revision 1, effective 2/89,

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The mechanic adjusted the packing with the pump running, as required by the procedure. After

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the pump was stopped, seal leakage from the outboard bearing increased significantly. The

mechanic subsequently readjusted the packing, with the pump stopped, until leakage was

acceptable. The mechanic apparently realized that the pump should be run again to ensure

adequate seal cooling. After discussion with the AO (Auxiliary Operator), the SOS (Shift

Operating Supervisor) was consulted to determine whether additional post maintenance testing

was required. The SOS decided to nm the pump to verify adequate seal cooling, and the PSS

(Plant Shift Superintendent) was present during the pump run to ensure adequate seal flow during

and after the subsequent pump run.

The inspector concluded that post-maintenance testing actions by the mechanic, AO, SOS and

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3.2

Diesel Generator Maintenance

The inspector observed of the following corrective maintenance on Diesel Generator l A:

Work Package 4549-90 - Diesel Generator I A Governor Leak

Work Package 2682 90 - Diesel Generator I A Air Compressor

Diesel generator routine Corrective Maintenance Procedure 5 38-2, Revision 14 was used in both

cases. Work Package 2682-90 was prepared in respense to Deficiency Report 2682-90, which

addressed wear of parts in the diesel generator air compressor and associated relay valve setpoint-

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changes. _ Vibrations during air compressor operation were deemed to be the cause. To reduce

vibration, the work package required replacement of a portion of the air compressor discharge

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pipe with a flexible hose. Technical Evaluation 273 90 addressed the design requirements of this

flexible hose including ' hose pressure / temperature ratings and support requirements. Two

additional supports were included and a drawing change request was completed. The inspector

reviewed the work package to determine if the necessary controls were in place and if a 10 CFR

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50.59 evaluation was necessary. The plant engineer indicated that no 10 CFR 50.59 cvaluation

was deemed necessary as the air compressor is a non nuclear safety component, and the pipe

replacement was considered to be a component substitution.

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The inspector then reviewed the Updated FSAR and the following licensee procedures:

17 21-1, Rev 3, Plant Modifications17-212, Rev 6, Engineering Design Change Request - MY

0-06-4, Rev 3,10 CFR 50.59 Determination

Procedure 17-21 1 defines an engineering design change as a permanent plant modification that

requires changes to plant issued controlled drawings. The description of the diesel air start

system in the Updated FSAR, Section 8.3.2, includes the air compressor in addition to the air

accumulators.

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The inspector discussed the air start system component substitution with licensee management.

=The licensee indicated it was their practice to control work such as this piping replacement as

a component substitution, without invoking the design change procedure.

However, this

replacement of a hard pipe with a flex hose does not appear to be a component substitution

because the reason for the change was to take advantage of the different characteristics of the

new' component. Because the air compressors are described in the Updated FSAR and the

function of the replaced component was changed, a 10 CFR 50.59 analysis may have been

appropriate. The licensee agreed that a review of the component substitution process was

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warranted. This item is unresolved pending NRC determination of whether 10 CFR 50.59

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requirements were met (90-19-03).

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3.3

(Closed) Violation 89-18-03: Maintenance Performed on Wrong Valve

The inspector reviewed the licensee's response to inspection Report 50-309/89-18 (MN 90-08),

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and the follow-up response to the same inspection report (MN-90-37) Based on this review the

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inspector determined that the responses appropriately addressed the violation as well as two

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related issues raised by the inspector. These issues were the use of personnel who .were

unfamiliar with the task and the influence of schedular pressure ca the quality of work.' In this

case, the licensee concluded that both of these factors contributed to the violation. Corrective

actions discussed in the responses appropriately addressed these issues. No repetition has been

noted. This item is closed.

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Surveillance Observations

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3.4.1 Procedure 3-6.2.2.5, Steam Generator Level, Revision 16, was performed . by

knowledgeable personnel on September 27,1990. No inadequacies were noted.

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3.4.2 ' Procedure 3.1.5, Emergency and Auxiliary Feed Pump Test (CRS-1), Revision 33, was

performed on October 16,1990. The inspector observed careful attention to detail on the

part of the Auxiliary Operator. A discussion of associated maintenance activities appears

in Section 3.1, preceding.

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3.4.3 The inspector observed the licensee performing Emergency Diesel Generator Surveillance

Procedure 3.1.4, Revision 32, Emergency Diesel Generator Surveillance Testing. This

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involved local manual start of the diesel generator and loading the diesel generator to its

designed rated load of 2500 KW, unloading, and manual trip after a two-hour run.

The inspecter reviewed the work package for necessary approval and interviewed the

auxiliary operator performing the test. The procedure required prior approval of the PSS

(Plant Shift Supervisor). Although the operator indicated that a signature of the PSS

documenting this approval was not obtained prior to beginning of the test, the operator

recognized this deficiency and obtained the required signature within a short time frame.

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Starting, loading and unloading of the diesel generator was uneventful. The operator

demonstrated attention to detail and understanding of the procedural steps and

requirements. The inspector had no other questions.

4.

Physical Security UP 7170h

Checks were made to determine whether security conditions met regulatory requirements, the

physical security plan, and approved procedures.. Those checks included security staffing,

protected and vital area barriers, vehicle searches and personnel identification, access control,

badging, and compensatory measures when required.

4.1

Security Key Control

On September 29, security personnel performing a key inventory discovered that a key was

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missing. Security notified the PSS, took compensatory measures, and conducted a records

review to determine the location of the key. The key, which had been checked out by a security

- supervisor on a previous shift, was subsequently found in a locked desk drawer.

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In this instance, loss of control of the uurity key was not a violation due to the nature of the

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weakness. In response to these incidents, the licensee is conducting a programmatic review of

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the contrl of plant keys. Key control will be routinely re-assessed by the NRC incident to

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inspection program implementation

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Engineering / Technical Suonort UP 71707. 92701. 92702)

5.1

Moisture Intrusion into Environmentally Qualified Limit Switches

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In 1985, in response to a Commission Order dated June 14,1984, Maine Yankee was required

to install environmentally qualified limit switches to provide position indication of the

Containment Isolation Valves. To establish EQ (environmental qualification), the licensee

qualified a conduit seal for preventing moisture intrusion into the valve position limit switches.

These limit switches are required to be functional during and following design basis accidents.

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The seal was made by placing a damming material around the cable inside a special conduit

fitting, and pouring a scaling compound into the fitting. This configuration was extensively

used. NRC Inspection Report 50 309/87-16 questioned the presence of moisture in one of the

limit switches that was inspected, and identified 16 limit switches with questionable conduit seal

performance. The licensee evaluated this problem, concluded that the conduit seal was installed

inadequately, and reinspected all the conduit seals. Questionable seals were resealed by adding

more scaling compound.

NRC Inspection Report 50-309/8716 also noted that, although the licensee's qualification file,

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QDR-3071, required a low point drain hole to be drilled immediately before the seal to divert

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any moisture accumulation in the conduit, the drain hole was not incorporated in numerous

installations.

In response to this concern, the licensee's engineering staff performed an

evaluation which concluded that the quantity of accumulated moisture resulting from an accident

was insignificant and the drain hole was not required in the installations.

On February 7,1990, during normal plant operation, moisture intrusion caused failure of PS A-

20 (LER 90-001). In response, Maine Yankee procured a prefabricated conduit seal of better

quality and replaced 36 of the 72 such seals. During this installation process the licensee

discovered a limit switch (on Valve DR A 6) that had no epoxy in the conduit fitting due to the

absence of damming material when the compound was poued. In addition, the limit switch on

Valve PCC-A 270 was found one quarter filled with potting compound; corrosion from moisture

intrusion was evident in this limit switch. The remaining limit switch boxes were inspected for

the presence of moisture, but the licensee did not rectify the problem for all of the remaining EQ

seals. The licensee concluded that the seals were functional if the normal environment did not

introduce moisture to the limit switch. Maine Yankee committed to upgrade the remainder of

the seals prior to the end of the following refueling outage.

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On October 17,1990, the EQ limit switch associated with Containment Isolation Valve PD A-

122 failed, also due to moisture intrusion under normal operating conditions. As a result of this

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failure, the Plant Manager decided to shut the plant down to upgrade all EQ limit switches using

potting compound and Scotchcast epoxy scalant. During the shutdown, the licensee discovered

that the piping and cibow joint used to connect the limit switch box to the conduit fitdng

containing ScotcDeast were not part of the tested configuration, and thread sealant to prevent

moisture intrusiot. past the pipe threads had not been incorporated,

in addition, licensee

engineering found ihat, while modifying rotation of approximately twenty of the limit switches

during installation il 1985, the licensee did not document (and may not have accomplished)

proper restoration of the breached environmental enclosure, and environmental qualification of

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the twenty switches vas thereby compromised.

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Although the licensee had analyzed moisture intrusion past the Scotchcast seal under accident

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conditions, the lack of low point drain holes deviated from the originally qualified configuration

of seal installation. During limit switch modifications, the licensee failed to document proper

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restoration of the EQ cnclosure, which also invalidated environmental qualification. Moreover,

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cibow fittings between the conduit seal and the limit switch box, installed as part of the original

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design, deviated from the qualified connguration and were not installed with a sealant to prevent

moisture intrusion. The lack of low point drain holes, failure to properly restore the EQ

cnclosure after switch modifications, and the elbow Stting modifications invalidated the

quali6 cation of the limit switches involved.

10 CFR 50.49 requires environmental qualification of the post accident monitoring equipment

specified in Regulatory Guide 1.97. As a result of deviations from the qualified configuration,

more than forty valves were not environmentally qualified from the date of required

implementation, October 1,1985, until replaced with a qualified configuration subsequent to the

plant shutdown on October 19, 1990.

The failure to meet requirements for environmentally

qualified limit switches by deviation from the design configuration is an apparent violation of

10 CFR 50.49 (90-19 04).

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10 CFR 50, Appendix B, Criterion XVI requires measures to assure that the cause of conditions

adverse to quality are promptly determined and that corrective actions are taken to preclude

repetition. The limit switch failure on February 7,1990 indicated that normal plant conditions

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introduced moisture through the conduit scal; therefore the moisture intrusion calculation for

accident conditions and the performance of the conduit seal were questionable. However, the

licensee failed to call the moisture intrusion calculation into question, and elected to improve

only half of the conduit seals at that time. On October 19,1990, the limit switch on PD A 122

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also failed due to moisture damage under normal plant conditions. Later inspection confirmed

the cause of this failure to be an inadequate conduit seal also. The failure of PD A 122 was duc,

in part, to failure to take prompt corrective action for the limit switch failure on February 7,

1990, and is an apparent violation of 10 CFR 50 Appendix B criterion XVI (9019 05).

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The inspector interviewed Equipment Qualification staff members and asked whether there are

similarly designed seals of questionable performance. The licensee staff did not identify any

other potential areas of similar concern. Subsequent to the identified concern on October 19,

1990, licensee management decided to promptly replace all the conduit seals with prefabricated

seals. Qualification of this seal is established through a new test on the NAMCO limit switch

assembly, inspector review of Qualification File QDR-1135-3 concluded that the licensee

installation instructions were in accordance with the manufacmrer's guidelines. The inspector

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witnessed the connector assembly process for PR-A 40 and CH-A 33 and observed QC

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verification of the significant attributes. No discrepancies were obsuved.

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5.2

ESF (Engineered Safeguards Feature) Light Box Malfunction

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A June 14,1984 Commission Order required Maine Yankee to implement commitments made

to the NRC in response to NRC Generic Letter 82-33. The Commission Order, in part, required

- Maine Yankee to submit a report to to the NRC by March 15, 1985, describing how the

requirements of Regulatory Guide 1.97 (RG 1.97) would be implemented, in addition, the Order

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required completion ofinstallation or upgrade requirements of Regulatory Guide 1.97 by October

1,1985.

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On February 28,1985, Maine Yankee letter MN 85-43 to the NRC certified that the position

indication power supply for the Containment Isolation Valves was battery backed, non-

interruptable, redundant, and Class lE, as required by the Commission Order.

During the shutdown on October 19,1990, operators observed and documented a problem with

indication on the ESF lightboxes. The ESF lightboxes provide control room indication of all

automatic containment isolation valve positions; the valves are separated into two trains and

indication is split between the light boxes. Trouble-shooting revealed that the degradation of a

single diode caused indication problems with both trains of the ESF lightboxes. Subsequently,

the licensee discovered inat, although credit had been taken for the ESF lightboxes in meeting

the requirements of r.egulatory Guide 1.97, both boxes were supplied with Non-lE Power from

.a single source -nd the requirements of Regulatory Guide 1.97 were therefore not met.

- Preliminary review b,, 'he licensee identified inadequate engineering review as the root cause of

the failure to impleme,n tnis requirement. Licensee review continued past the end of the

inspection period. The inspectors observed that licensee engineering action to investigate and

correct this problem was aggressive, timely and thorough after its discovery on October 19.

Prior to starting up the plant, the licensee developed a Justification for Continued Operation

based on the following premises.

The likelihood of an accident requiring containment isolation combined with a failure of

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valve position indication is extremely remote.

Functioning of the Containment Isolation Valves is independent of the position indication

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system.

Compensatory measures were taken to veiify valve position in the event of a containment

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isolation signal.

Many of the containment isolation valves have independent, Class 1E position indication,

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although 27 valves do not have Class 1E position indication.

In addition, the licensee committed to modify the power source supplying position indication for

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the Containment Isolation System to meet the guidelines of RG 1.97 during the next refueling

outage.

The Justification for Continued Operation was reviewed by the NRC staff, discussed with the

licensee, and determined to be acceptable.

The failure to install a qualified power source for position indication of the Containment Isolation

Valves by October 1,1985, as required by the Commission Order dated June 14,1984, is an

apparent violation (9019 06).

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In addition, the failure to provide accurate information in the February 27,1985 letter from

Maine Yankee to the NRC is an apparent violation (90-19-07),

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5.3

(Closed) Unresolved Item 87-09-01: EOP Setooint Updates

In May 1989, inspector review identified that the harsh containment environment instrument

setpoint cited in the EOPs (Emergency Operating Procedures) did not reflect the licensee's latest

evaluation for overall SMM (Saturation Margin Monitor) instrument uncertainty. The SMM

setpoints were used as the basis for signincant decisions within the EOPs, including termination

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of safety injection and operation of the reactor coolant pumps. An EOP change in November

1989 addressed the inspector identified inconsistencies,

in addition, due to the difficulty

associated with updating setpoints consistently throughout the EOPs, Maine Yankee plans to

implement a computer based system to allow tur.ely and accurate changes to be made. Other

actions to enhance the quality of EOP setpoint control included coordination of EOP setpoints

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with an existing plant setpoint upgrade and plans for the establishment of a PORC (Plant

Operations Review Committee) subcommittee with responsibility for EOP changes specifically.

The inspector noted that the licensee corrected the discrepancy identified by the inspector,

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recognized the weakness in the processing of EOP setpoint changes, and was taking action to

address this weakness. This item is closed.

5.4

(Update) Unresolved item 50-309/88 16-01: Wiring Discrepancies

The wiring discrepancies discussed in the inspection report as well as other wiring problems

identiSed during the 1988 refueling outage were reviewed by the licensee in PRCE (Plant Root

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Cause Evaluation) Number 166. The purpose of the PRCE was to identify any similarities with

these issues, evaluate root causes and provide corrective action recommendations. The PRCE

concluded that there were no common systematic weaknesses, but that there were several

weaknesses. The issues identified in NRC Inspection Report 50 309/88-16 were considered

adequately resolved by the PRCE. Inspection Report 50-309/88-21, Detail 3.c. also addressed

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wiring discrepancies during the 1988 refueling outage. This report references lleensee CAR

(Corrective Action Request) 88-143-0. Two conditions identified in CAR 88-143-0 were found

to not have been accomplished by a recent Quality Assurance Surveillance of corrective actions.

Further inspector review appears warranted; therefore this item remains open,

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5.5-

(Closed) 50-300/88-04-01:

Limit Switch Setooint C atrol and Maintenance for Limit

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Sented Motor Operated Valves

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This item addresses two related issues concerning limit seated MOVs (Motor Operated Valves).

When the closing valve motion is terminated using a signal from the limit switch, residual energy

(the inertia from valve motion) drives the valve to full closure. This requires a precise setpoint

for the limit switch such that there is suf0clent inertial force to seat the valve adequately. As

stated in NRC Inspection Report 50-309/88 04, Detail 4.5, there was no immediate concern with

the present setting of limit seated valves at Maine Yankee because test records indicated sufficient

thrust development on the seat. The inspector reviewed the status of limit seated MOVs

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currently in the MOV program. This encompassed HSI M 50, HSI M 51, CS-M 90, CS M 91,

PCC-M 43, SCC M 165 and eight (8) motor-operated dampers in the control room ventilation

system. The valves for which full closure is important were found to have been tested for

leakage due to other requirements.

The second issue was the effect of wear of the valve seat on the limit switch setting. The

licensee was aware and sensitive to this concern, which is not limited to Maine Yankee's MOV

program only. This is an industry issue that is being addrested through Generic letter No. 89-

10, * Safety Related Motor Operated Valve Testing and Surveillance,' in recommended action

'd."

The licensee has plans to address this issue as part of the MOV program to address the

Generic Letter. The licensee's MOV program will be subject to inspections related to the

Generic Letter. This item is closed.

5.6

(Closed) Utiresolved item 89-018 002: Unqualified Flange on DWST Heater Piping

Inspector review of this item identified an error in the information documented in Region I

Inspection Report 50-309/8918, Detail 4.b.

The Emergency Feedwater system was installed as a safety-class system during initial

construction, and not upgraded to safety class as stated in the report. Therefore, a review of the

quality of the upgrade process was not warranted.

The inspector did review the licensee's actions to address this issue. The unqualified flange was

sent to a lab for a determination of its material properties. The results of the testing indicated

that the material properties of the flange were acceptable, with significant margin for the flange

to fulfill its safety function. Therefore, the integrity of the system was not in question at any

time. This item is closed.

5.7

(Closed) Unresolved Item 50-309/89-25 03: Comoonent Cooling Water isolation Valve

Stroke Time Criteria

The licensee's review of this issue identified the root cause as weak control of the' PCR

' (Procedure Change Report) process as it relates to setpoint and acceptance criteria. The

corrective action stated in the response to NRC Inspection Report 50 309/89-25 dated March 23,

- 1990 involved a revision to Procedure 0-06-2, " Procedure Review, Approval and Distribution,"

to require a technical justification for PCR changes to setpoints or acceptance criteria. - The

inspector verified that the procedure had been changed.

No further examples of similar

inappropriate use of the PCR process have been identified. This item is closed.

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6.

Safety Assessment /O ality Verification HP 71707)

6.1

Power Range Sr,[dy_Cnannel Nuclear Instrumentation

On October 2, a licemme safety issues concern report was initiated to review an identified

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setpoint drift in the dual linear power range nuclear instrumentation (NI) bistables. Setpoint drift

was approximately 2% higher than the specified 15% setpoint.

The power range nuclear instrumentation has four channels. Each channel consists of two ion

chambers, respectively located outside the reactor near the upper and lower half of the core.

Signals from the ion chambers are processed in separate linear amplifiers to provide voltage

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signals and are then summed in a summing amplifier to provide a channel power signal to the

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reactor protection system (RPS). This channel power signal can be adjusted to agree with

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calorimetric power calculation by a gain adjusting potentiometer on the RPS calibration and

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indication panel in the control room. The summed channel power signal is monitored by a 15 %

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power bistable. This bistable closes contacts above 15% power to inhibit the RPS startup rate

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trip and enable the RPS symmetric offset trip.

After technical specification surveillances, the control room operators adjust the power range

safety channel to agree with the calorimetric power calculation. The gain adjusting potentiometer

is adjusted every night to accomplish this. The potential exists for introducing error in the

bistable setpoint since the summing amplifier, when adjusted, affects the setpoint of the 15%

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power bistable, and the bistable setpoint is not adjusted to compensate. As a result, the bistable

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setpoint could become increasingly out of calibration until the bistable is calibrated either during

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a refueling outage or prior to plant startup.

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The licensee identified Unusual Occurrence Report UOR 016 86, prepared on February 26,

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1986, which describes a similar condition and a suggested corrective action. This UOR indicated

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that the symmetric offset was not a limiting condition for accident analysis until well above 20%

power since the out-of specification condition was deemed not to compromi.e plant safety. The

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deficiency was identified while the plant was shut down and was not considered reportable. The

daily calorimetric calibration of the power signal to the trip functions was identified to be the

root cause.

Although the licensee continued to experience this problem, a plant change

suggestion prepared in 1988 to correct this was deemed as a relatively low priority " good idea"

item and was not implemented.

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Additionally, Technical Specification 2.1, Limiting Safety System Setting - RPS and Table 3.9-1,

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.instramentation Operating Requirements for RPS, state that the symmetric offset trip may be

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bypassed below 15 % rated power, while Technical Specification 3.10.C.7 requires operation of

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excore symmetric offset above 20% rated power. These Technical Specifications appear to need

clarification of the requirements applicable from 15% power to 20% power.

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The licensee is currently reviewing the entire issue to determine appropriate corrective acuon.

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Feasibility of periodic surveillance on the 15% trip bistable to determine its setpoint and make

necessary adjustments is being considered, and the possibility of installing the proposed

modification with the plant on line is being reviewed.

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The inspector concluded that the licensce's assessment of the safety significance associated with

this issue was appropriate. However, as it indicates a potential for a condition outside plant

Technical Specifications, a higher priority in resolving this matter appears warranted. This

condition indicates a potential weakness in the licensce's corrective action program and is

unresolved pending resolution of the need for periodic surveillance of the 15% trip bistable

(UNR 50-309/90-19 02)

6.2

Dron-in Meeting

]

On October 2,1990, the following licensee managers came to the NRC Region I office to discuss

Auxiliary Feedwater (AFW) flow during plant start up with Mr. E. McCabe, Chief, Projects

Section 3B.

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Mr. S. Nichols, Licensing Section Head

Mr. D. Whittier, Manager, Licensing & Nuclear Engineering

During the meeting, the licensee addressed AFW line-up to supply feedwater via the first point

feedwater heaters, single failure considerations, the analysis showing the potential accident

consequences to be bounded by other accidents, and Technical Specification provisions related

to such operation. References include:

Technical Specification 3.22

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EDCR 83-29

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1/23/84 Yankee Atomic - Framingham Memo from K.R. Rousseau /P.J. Gulmond to

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F.L. Anderson

It was concluded that further licensee and NRC addressal of the associated safety and compliance

aspects was appropriate and would be accomplished.

6.3

Housekeeping

During this inspection the inspectors noted that housekeeping showed signs of lack of

management attention or a lack of sufficiently high housekeeping standards. On October 15, the

inspectors noted that the state of housekeeping appeared similar to housekeeping observed during

a refueling outage. A lack of attention to housekeeping was previously noted by the inspectors

in NRC Inspection Report 50-309/90-13 and brought to the attention of the Plant Manager. The

Plant Manager attributed the lack of improvement in housekeeping to management failure to

establish housekeeping as a priority. The plant Manager took steps to immediately improve

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housekeeping and to explore and address the attitudes which permit housekeeping to periodically

decline.

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6.4

Request for NRR Waiver of Compliance

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On October 3, Maine Yankee requested a temporary waiver of compliance to Maine Yankee

Technical Specification 3.22, Feedwater Trip System. The request was in anticipation of a plant

start up later on October 3; NRR and Region I persor.ncl reviewed and approved the waiver of

compliance. It appeared to the resident inspectors, however, thot action by the licensee to pursue

relief from Tecnnical Specification requirements for Feedwater Isolation Trip system was not

timely since, at the time the waiver was requested, the need for resolution of the Technical

Specification had been known to plant personnel for several weeks. The inspectors concluded

that a weakness exists in the method of establishing licensing priorities.

7.

Administrative

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7.1

Persons Contacted

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Within this period, interviews and discussions were held with various licensee personnel,

including plant operators, maintenance technicians and the licensee's management staff.

7.2

Summarv of Facility Activities

At the beginning of the report period the plant was operating at full power. On September 30,

the plant was shut down to allow repairs of the turbine-generator main lube oil pump seal rings.

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The reactor was made critical on October 5, and power operation was resumed on October 6.

A plant shutdown was initiated on October 19 to resolve environmental qualification concerns

related to a limit switch that failed during surveillance testing on October 17. The plant was

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returned to power on October 25, and remained at power for the rc:t of the report period.

7.3

Interfarg_with the State of Maine

periodically, the resident inspectors and the onsite representative of the State ofI.taine discussed

their respective findings and activities. No unacceptable conditions were identified.

7.4

Exit Meeting

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Meetings were periodically held with senior facility management to discuss the inspection scope

and findings, A summary of findings was also discussed at the conclusion of the inspection.

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7.5

Insnection Hours

The inspection involved 177 inspection hours, including 14 backshift and 9 deep backshift hours.

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