ML20138J029

Safety Insp Rept 50-289/85-26 on 851026-1112.No Violation Noted.Major Areas Inspected:Letdown Sys Operation,Nuclear Svc Closed Cooling Sys Reactor Bldg Penetration Design Adequacy & Turbine Bldg Fire Loading

ML20138J029
Person / Time
Site:	Crane
Issue date:	12/12/1985
From:	Conte R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML20138H992	List: IR 05000289/1985026 ML20138H989 ML20138J029
References
50-289-85-26, NUDOCS 8512170415
Download: ML20138J029 (36)

See also: IR 05000289/1985026

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U.S. NUCLEAR REGULATORY COMMISSION

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REGION I

Report No.'

50-289/85-26

Docket No.

50-289

License No.

DPR-50

Priority --

Category C

Licensee:

GPU Nuclear Corporation

Post Office Box 480

Middletown, Pennsylvania 17057

Facility At:

Three Mile Island Nuclear Station, Unit 1

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Inspection At:

Middletown, Pennsylvania

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Inspection Conducted:

October 25-November 12, 1985

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~ Inspectors:

- D. Falconer Jr. , Lead Reactor Engineer, Region II

D'.' H'averkamp, ' Technical Assistant for"7MI-1 Restart,

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Region I

D. Johnson, Reactor Engineer, Region I

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L. Reidinger, Instructor,. Reactor Training Center

T. 'Stetka, Senior Resident Inspector (Crystal River),

Region II

R. Urban, Reactor Engineer, Region I

D. Vito, Senior Emergency Specialist, Region I

J. White, Senior Radiation. Specialist, Region I

F. Young, Resident Inspector (TMI-1), Region I

' Contractor Personnel:

W. Apley, Associate Manager, Energy-Systems,

Battelle Pacific Northwest Laboratories (PNL)

T. Morgan, Operator Examiner, EG&G Idaho, Inc.

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-Approved By:

R. Conte, TMI-1 Restart Manager

-Date-

TMI-1 Restart Staff

Division of Reactor Projects

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8512170415.851212

PDR

ADOCK 05000289

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Inspection' Summary:

The TMI-1 Restart Staff conducted routine and special (NRC shift coverage)

safety inspections. (501 hours0.0058 days <br />0.139 hours <br />8.28373e-4 weeks <br />1.906305e-4 months <br />) of power operation focusing on operator and man-

agement perfo.rmance.

Specifically, items' reviewed in the overall facility op-

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eration. area were:

letdown system operation, nuclear service closed cooling

system reactor building penetration design adequacy, turbiac building fire

loading, and procedure implementation issues noted during shift inspector re-

views. Other review items included: makeup system leak, implementation of

administrative controls in the area of plant staff working hours, diesel gener-

ator maintenance an'd operability, and safety grade emergency feedwater inplant

review.

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Inspection Results:

Facility personnel continued to conduct themselves in a professional manner.

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Licensee management and quality assu_rance department personnel continued their

attentiveness and involvement in daily' activities.

The training department is

an apparent strength contributing to the overall positive performance of per-

sonnel.

In addition to personnel. performance, the material condition of the

plant was~also conducive to safe operation, including avoidance of plant trips

and challenges to safety-related systems. The radiological controls department

continued-to exhibit itself as a strength ir. the licensee's organization.

Some

communications lapses occurred but are considered minor.

In general, facility procedures were properly. implemented.

However, certain

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problems in adhering to procedural steps (some of which were compounded by.a

lack of clarity) were indicative of a potentially adverse trend.

Increased

licensee management attention to this area may be warranted, and the TMI-1 Re-

start Staff will continue to review this matter.

Personnel involved locally with a leak in a makeup pump cubicle performed poor-

ly.in using a standing radiation work permit for draining a radioactive system

and'in not isolating the leak before evacuating the cubicle. Control room per-

sonnel were responsive to the event and licensee post review critique and cor-

rective actions were appropriate for the circumstances.

-The following items require further review and will, therefore, be addressed in

subsequent inspections:

(1) design adequacy of the nuclear services closed

cooling water system reactor building penetration, and (2) maintenance and

post-maintenance test data for the recent diesel generator outages.

The licensee needs to provide some additional information for the staff regard-

ing the safety grade design considerations of the emergency feedwater system.

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DETAILS

1.

Introduction and Overview

1.1 General

Throughout this inspection period, the TMI-1 Restart Staff provided

onshift inspection coverage 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> a day to assess restart operat-

ing activities. This coverage was consistent with the reduced level

of testing activity and steady-state facility operation at the 48%

power plateau.

The staff's observation of plant activities was main-

tained by NRC personnel from Region II and the Reactor Training

Center and by reactor operator examiners from Battelle Pacific North-

west Laboratories and EG&G Idaho, Inc., both NRC contractors. Also,

Region I inspectors continued periodic coverage of testing activi-

ties. Additional Region I personnel were on site during portions of

the period-to augment the resident inspection staff.

'1.2

Facility Restart Operations

During the period of October 25-November 12, 1985, the significant

TMI-I restart operational milestones included:

(1) continued main

turbine generator operation at the 48% plateau, and (2) one minor

power level change for integrated control system tuning.

The chrono-

logical- summary of plant operations during the period is presented

below.

Date

Time

Operational Highlight-or Milestone

10/25/85

7:00 a.m.

Reactor at 48% of rated power, reactor

coolant average temperature at 578 de-

grees F and pressure at 2150 psig

11:15 a.m.

Main turbine transferred to manual due

to electro-hydraulic control signal

cycling;. reactor power reduced to 45%

11:55 a.m.

Returned reactor to 48% power

11/8/85

1:00 p.m.

Reduced power to 40% at 2% per minute

in accordance with test procedure TP

800/1

1:20 p.m.

Returned reactor power to 48% at 2% per

minute in accordance with TP 800/1

11/12/85

7:00 a.m.

At the end of this inspection period

the reactor was at 48% of rated power,

reactor coolant average temperature was

579 degrees F, and pressure was 2155

psig

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1.3 Operational Events

Several events occurred during this inspection period that were con-

sidered either operationally significant or were matters of special

interest to the TMI-1 Restart Staff.

These events are discussed

below.

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On October 25, 1985, electricians were measuring decay heat pump

vibration signals- from a cabinet in the relay room. The workers used

a power source on the integrated control system (ICS) panel to pro-

vide power to the test equipment.

This caused an electrical inter-

ference signal with the main turbine electro-hydraulic control (EHC)

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system. The EHC system tripped to manual and the system closed the

turbine stop/ control valves. The plant responded as expected.

Tur-

bine header pressure increased to 1035 psig, and two atmospheric dump

valves, two turbine bypass valves, and a main steam safety valve were

actuated. The operators quickly took manual control of a mein feed-

water pump and the turbine generator and stabilized the plant transi-

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ent. Within an hour, the ICS was placed into automatic control and

the plant was returned to steady-state conditions at 48% power.

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On October 29, 1985, the TMI-1 Restart Staff was informed that a leak

of 150 gallons of primary water from the makeup and purification sys-

tem occurred between 4:20 and 4:30 p.m. on October 28, 1985. At the

time, licensee personnel were conducting corrective maintenance on a

minor leak from one of the' three n'akeup pumps. The pump was isolated

by closing valves on each side to permit depressurization and drain-

i.ng of the pump.

Since one of the isolation valves leaked, draining

of the pump continued for an extended period of time. A local radia-

tYon monitor alarmed causing an evacuation of personnel from the pump

cubicle.

Escaping radioactive noble gases caused ventilation process

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monitors to respond but not alarm.

Licensee personnel reentered the

punp cubicle and secured the leak.

The licensee estimated that ap-

proximately 0.7 Ci of noble gases (primarily Xe-133) were released to

the environment and dose calculations indicated that the release was

a small. fraction of regulatory limits. The licensee conducted a cri-

tique of the' event (see paragraphs 2.1 and 4).

On November 2, 1985, at 9:30 a.m., the licensee attempted to repair a

small water-leak on a flange in the secondary plant moisture separa-

tor drain system which is located in the turbine building. Because

the section of piping was not effectively isolated, a steam leak oc-

curred at the flange as the flange bolts were loosened. The leak was

fed by relatively hot water from the steam and feedwater system.

Some of the workers involved in the repair received superficial burns

that did not require offsite medical attention. The steam: leak last-

ed approximately 20 minutes. Although a minor perturbation was no-

ticed in feedwater flow, the plant remained at steady-state 48% power

conditions.

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On November 4, 1985, emergency diesel generator IA was, removed from

service, as permitted by the plant's technical specifications, for

annual preventive maintenance and inspections. The diesel was re-

turned to service on November 8, 1985, and a 24-hour operational test

with the diesel generator loaded to 3000 KW was completed satisfacto-

rily as well as other required post-maintenance surveillance testing.

After that diesel was officially declared operable, the redundant

emergency diesel generator 1B was removed from service on November

11, 1985, for similar annual maintenance and testing.

'On November 5, 1985, a 120V ac bus was deenergized by a licensee

technician who opened a supply breaker in response to arcing and

sparking at an adjacent breaker. Technicians were performing breaker

preventive maintenance on an engineered safeguards bus at the time.

'Although no significant loads were lost as a result of the bus being

deenergized, the electrical incident was apparently due to technician

error. The deenergized bus was an alternate power source for the

integrated control system .(ICS)/non-nuclear instrumentation (NNI)

bus.

The ICS and NNI remained energized.

The reactor protection and

engineered safeguards actuation systems were unaffected because they

were powered from separate vital buses. The licensee has taken dis-

ciplinary action against the technicians.

On November 5, 1985, the licensee removed one of two letdown coalers

from service because of an indicated leakage at the rate of 1.3 ,

liters / day. The leakage of reactor coolant system water was into the

intermediate closed cycle cooling water system, a non-safety system.

The leakage was first detected on October 24, 1985, by'racioactive

noble gas sampling and was trended on a daily basis. Although the

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cooler is still considered to be operable by the licensee, it is not

in service. The licensee has a spare letdown cooler available on

site if needed.

1.4 Summary

This inspection included restart testing activities at the 48% power

plateau. During this period there were no interruptions of the re-

start testing program due to equipment problems or other reasons.

The' shift inspectors referred only implementation matters or status

questions to shift supervisory personnel and referred programmatic

matters (event followup, design or procedure adequacy problems) to

resident and region-based NIC personnel.

Resident and region-based

personnel interfaced with licensee support groups in followup to

shift inspector referrals /cor.cerns. The staff's observations and

findings regarding plant operation and testing and licensee response

to operational events is discussed in the report sections that

follow.

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2.

Shift Inspection Activities

2.1 Scope of Review and Observations

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During the period of Octobr~ 25-November 12, 1985, the TMI-1 Restart

Staff continued its augmented shift-inspection coverage. The NRC

shift inspectors assessed the adequacy and effectiveness of operating

personnel performance based on the inspectors' observations of oper-

ating and startup activities to determine that:

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operators are attentive and responsive to plant parameters and

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conditions;

. plant evolutions and testing are planned and properly

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authorized;

procedures are used and followed as required by plant policy;

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equipment status changes are appropriately documented and commu-

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nicated to appropriate shift personnel;

the operating conditions of plant equipment are effectively mon-

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i.tored and appropriate corrective action is initiated when

required;

backup instrumentation, measurements, and readings are used as

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appropriate when normal instrumentation is found to be defective

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or out of tolerance;

logkeeping is timely, accurate, and adequately reflects plant

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activities.and status;

operators follow good operating practices in conducting plant

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operations; and

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operator actions are consistent with performance-oriented

training.

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The shift inspectors' observations included, but were not limited to,

those reactor plant operation and testing activities, periodic sur-

veillance activities, and preventive and corrective maintenance ac-

tivities listed below.

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Reactor Plant Operation and Testing Activities

routine con. trol room operations including annunciator alarm re-

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sponse and control room logkeeping

operating and emergency procedures discussions with shift. super-

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visors,. shift foremen, control room operators and shift techni-

cal advisors

periodic inspection observation tours of areas outside the con-

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trol room, including diesel generator rooms, emergency feedwater

rooms, control building, turbine building, auxiliary building,

intermediate building, electrical switchgear rooms, and outside

buildings and yard areas

hydrogen addition'to core flood tank 18 to increase tank

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pressure

operating crew response to main turbine electro-hydraulic con-

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trol system signal causing unexpected turbine control valves

closure

routine-operations and maintenance planning briefings between

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operations plant manager and shift supervisors

operator' response to incorrect securing of reactor coolant bleed

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and feed operations when MU-V-8 was not placed in.the proper

. po s.i ti on

periodic blowdowns of instrument air compressor aftercooler wa-

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ter trap

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fire fighting training drill in burn building for Londonderry

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Township' Fire Company-

shift crew respcnse to kinked fire hose found during periodic

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surveillance

shift crew training session during backshift

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changeover and recharge of powdex vessel IC.

licensee personnel critique regarding leakage and spill from

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makeup' pump MU-P-1A that' occurred on October 28, 1985

shift crew briefings regarding plant incident reports 1-85-13,

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"PORV Setpoint Check," and 1-85-14, " Leakage and Spill from

Makeup Pump"

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shift crew actions in response to inoperable reactor building

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fire system annunciator

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radiation work permit posting practices

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operating crew response to inadvertent draining of reference leg

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to differential flow transmitter DPT-922 for safety valves dis-

charge piping

operator response to chlorine leakage in circulating water pump

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house

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liquid release permit L8511092, documentation completion and

initiation of release from waste evaporator condensate storage

tank 1B.

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securing of waste gas release from waste gas decay tank 1A and-

release permit G8511075 documentation completion

operating crew response to steam leak from moisture separator

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drain pump 1A discharge flow control valve MO-V-1A, which re-

suited during' loosening of flange bolts in valve

waste gas release permit G8511076 documentation completion and

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initiation of release from waste gas decay tank 18

operating crew response to loss of regulating bus TRA caused by

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incorrect power supply breaker operation by electricians after

arcing occurred in emergency safeguards bus IA motor control

center unit IDL

transfer of main feed pump 1B from manual to automatic mode of

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operation and then back to manual when pump appeared to start

hunting again

operating crew removal of the annunt.ator card for heating and

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ventilation panel' alarm B-4-5, " Air Supply Tunnel Combustible ~

Vapor"

bypassing of reactor protection system channel 1B due to appar-

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ent out-of-specification variable pressure / temperature module

operating crew response to moisture separator high level alarm

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and low level alarm conditions

radiological controls personnel response to posted radiation

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work permit No. 031257 that shift inspector found to have ex-

pired on previous day (November 6, 1985)

operator performance of valve lineup and switch position check.

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after maintenance of emergency diesel generator 1A

plant-maneuver from 48% of rated power to 40% and back to 48%

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for integrated control system tuning

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shift crew fire brigade meeting conducted by shift maintenance

foreman p'rior to normal shift briefing

periodic Surveillance and Maintenance Testing

. emergency safeguards channels surveillance testing associated

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with emergency diesel generator IB

reactor protection system / nuclear instrumentation channel 1A

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calibration per surveillance procedure 1303-4.1

reactor trip breaker periodic testing

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reactor building spray pump 1A surveillance testing

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reactor coolant system leak rate calculation per surveillance

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procedure 1303-1.1'

hydrogen recombiner IB surveillance operational testing

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emergency diesel generator IA protective relays calibration per

surveillance procedure 1302-5.30

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emergency diesel generator IA voltmeter calibration per surveil-

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lance procedure 1301-8.2 and preventive maintenance procedure

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control rod movement surveillance per SP 1303-3.1

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emergency diesel generator IA post-maintenance surveillance

testing per surveillance procedures 1301-8.2 and 1303-4.16 and

operating procedure 1107-3

setting of emergency diesel generator'1A governor high limit per

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procedure 1420-EG

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post-maintenance surveillance testing of repaired square root

extractor for reactor protection system loop 1B flow

- performance of local leak rate testing of reactor building

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equipment hatch per surveillance procedure 1303-11.18

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surveillance testing of ' reactor building isolation channel 3 per

procedure 1303-4.13

-- _ reactor. building local leakage measurements-for access hatch

door seals per surveillance procedure 1303-11.25

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Preventive and Corrective Maintenance Activities

~ tightening of building spray pump isolation valve BS-V-1A pack-

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ing to stop leakage of 12 drops / minute

troubleshooting on fire panel in control tower being performed

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by instrument and controls technicians

emergency diesel generator 1A annual ~ preventive maintenance and

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inspection per procedure 1301-8.2

troubleshooting of ground on station battery 1A being performed

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by electricians

repairs to leaking heater drain pump flange

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phase bus heating test to determine ~ temperature rise with fan

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secured in preparation for replacement of_a fan bearing

repairs to moisture separator drain tank dump valve ~ controller

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troubleshooting and repair of "B" loop square root extractor for

reactor coolant system flow

spray welding of phase bus cooling fan shaft

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pre-maintenance tagging activities for. emergency diesel genera-

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tor 18 annual inspection, including racking out of generator

output breaker

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electrical checks of thermal breaker trip setpoint for the

prelube oil pump breaker in panel IB per preventive maintenance

procedure E-62

performance of various maintenance activities related to annual

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inspection of emergency diesel generator IB

In addition, shift inspectors conducted or contributed to the follow-

ing special reviews of facility design and operational matters or of

the licensee's administrative controls programs.

control room panel indicators operational review, including use

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of temporary labelling and instrument accessibility

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containment-isolation. valves installation review for nuclear

service closed cooling water to reactor building cooling unit

fan motors

surveillance. procedure lineup requirements for containment pres-

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sure instruments test "T" connections

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reactor building pressura instruments for emergency safeguards-

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actuation channels installation review

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decay heat closed cooling system surge tank level instrumenta-

tion calibration review

followup review to leakage and spill from makeup pump MU-P-1A

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that occurred on afternoon of October 28, 1985

followup review of main turbine control valve closure including

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review of plant incident report 1-85-14

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interviews with selected licensed-operators regarding restart

qualification training evolutions

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calibration procedures and data review for computer po:nts used

to perform the reactor coolant system leak rate calculations

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administrative controls for out-of-service equipment

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printer /CRT display of computer group 55 data (reactor power,

imbalance, quadrant power tilt, rod index)

locked valve and component listing and sampling ch ck of locked

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valves and breakers for conformance with administrative proce-

dure AP .1011, " Locked Valve and Key Control"

borated water concentration in boric acid mix tank conformance

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with operating procedure 1140-47B:

methods for calculating sodium hydroxide tank and borated water

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storage tank levels

-- -hand-and-foo't monitor calibration practices per procedure

9000-PMI-4221.16

2.2 Assessments of Shift Inspectors

2.2.1-

General'

The shift inspectors assured that any potentially adverse

safety concern or regulatory finding was identified prompt-

ly to both the. licensee's shift supervisor and the TMI-1

Restart Manager.

Those items requiring additional staff

review or followup are described in paragraph 3 of this -

report. Also, at the end of their assigned period of shift

inspection activities, the inspectors provided their gener-

al assessment of facility operational readiness and person-

nel. performance. These general assessments included, as

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applicable, each inspector's overall views related to oper-

=ating staff performance, fire protection, maintenance, sur-

veillance, radiological controls, training, emergency

planning, and physical security. The inspectors' assess-

ments are presented below.

2.2.2

Operating Staff Performance

Shift inspectors continued to provide many positive com-

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ments on.the knowledge level and overall quality of perfor-

mance of-facility operating, maintenance and technical

staff personnel similar to that described'in previous in-

spection reports issued since criticality. -This period

. reflected _the first opportunity for NRC staff to observe

personnel during normal steady-state power evolutions.

It

was during this period that shift inspectors observed a

relatively large sample of mechanical, instrument and con-

trols, and electrical maintenance and testing similar to

that which would be conducted during.100% power operations.

The focus in this period was plant operations, not restart

testing. Overall, personnel interfaced well with one an-

other in support of the operations department. However,

some indications of performance problems were revealed.

As an example, shift inspectors intensely reviewed proce-

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dure implementation by the various departments within the

TMI-1 division. . They observed a number of instances where

procedures could have been better written or where certain

individual steps were not ~ properly followed. They referred

.these problems to the region-based staff (see paragraph

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3.2.5).

In general, procedures were adequate and properly

implemented.

During observations of control room personnel, shift in-

spectars, in general, found them to be professional in

their approach to shift operations. Communications were

handled effectively-and during this time no problem attrib-

utable to unclear communications occurred.

During a steam

leak in the secondary plant on November 3 (see paragraph

1.3), licensee. personnel response was controlled and they

remained-calm and rational during the event. The leak was

identified by an auxiliary operator and maintenance person-

nel. The shift foreman was dispatched to the scene and.the

leak was isolated quickly.

Subsequently, the auxiliary

operators were called to the control room and the opera-

tions manager outlined for the crew how the system would be

returned to normal.

They then followed a step-by-step pro-

cedure in returning the system to normal status.

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Similar comments can be made about the loss of the TRA bus

on November 5, 1985.

The licensee will be issuing plant

incident reports on both events.

The adequacy of the

licensee's review of this event is unresolved pending com-

pletion of~ licensee action as noted above and subsequent

TMI-1 Restart Staff- review (289/85-26-01).

Shift foremen were using the control room operator's (CRO)

log to complete the right side of their log for

event / evolution chronology before shift turnover.

It ap-

pears that they rely on the CR0 log for detailed secretar-

fal type recording of routine evolutions and events. Non-

routine events are recorded on the left-hand side of the

log and they appear to be original entries.

These entries

encompassed technical specifications action statements,

reportable events or items of potential public interest,

and major events.

The inspector noted that the copying of

the CR0 log in the shift foreman's log diminished the use-

fulness of the shift foreman's log for independent event

analysis.

Overall, operators.and technicians continued to perform in

a competent and professional manner.

2.2.3

~ Training

In general, shift inspector observations continued to sug-

gest that the training department provided sufficient

knowledge and assured demonstrated skills to support the

performance of activities in a' safe and professional man-

.ner.

However, the procedure implementation problems de-

scribed in paragraph 3.2.5 suggest additional management

attention to personnel making changes to procedures for

improvement.

A primary focus during this inspection period was the im-

plementation of the licensee's restart qualification pro-

gram (NRC Inspection. Items 289/81-33-04 and 289/84-19-01).

The TMI-1 Restart Staff initial review of the program con-

sisted of reviewing the qualification cards. The cards

were set up according to operating crews. The system was

established so that most of the operating crews observed or

performed most of the items listed. This meant that no

individual qualification card was required, and, as stated

in a licensee inter-office memorandum, every person was not

required to have been involved in every evolution.

It ap-

peared as though the basic purpose of the cards was to en-

sure that at least a specified level of operating experi-

ence existed prior to the plant operating and returning to

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power after the 100% trip test.

Licensee management con-

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firmed that the qualification cards were not training docu-

ments, but.they were a means to record crew experiences

versus simulations to identify any significant shortcomings

in experience level at the end of the power escalation

program.

Interviews with shift personnel revealed a general at'titude

that the restart qualification card program was not effec-

tive in achieving a training purpose. One contributing

factor was that most of the items were either simulated or

walked through and, therefore, did not provide the operat-

ing experience desired. Another reason is that the majori-

ty of the-items were completed in 1983, two years prior to

the plant startup, and no program existed for followup re-

view to assure retention of the information.

Licensee man-

agement indicated its intentions to update the qualifica-

tion card to reflect recent experience gained during the

power escalation program

Shift inspectors conducted interviews of four operating

crews (A,C,E,F) including a total of eight reactor opera-

tors and three senior reactor operators.

The questions

asked were performance oriented taken from two qualifica-

tion card sections, primary and power escalation, since

these had the most applicability to present plant

configuration.

The overall knowledge level;in these areas was considered

average or above, as determined by the NRC examiners' ex-

pectations. There appears at this point to be no areas of

unexpected poor knowledge with respect.to license. examiner

standards. The inspectors initially concluded that the

qualification card program at this point was of little val-

ue in assessing the operators' knowledge level or in mea-

suring performance results to date. Two years of

additional training provided by the training department

prior to startup was the key factor in enhancing operator

performance.

This area will continue to be routinely re-

viewed during the 48% and 75% power plateaus.

-2.2.4

Startup Testing

During the implementation of the residual ICS tuning tests

from previous power levels,.the shift inspectors observed

good personnel communications and proper implementation of-

the applicable test procedures.

Licensee personnel demon-

strated. good overall control of the evolutions,

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2.2.5

Maintenance

Preventive and corrective maintenance continued to be ag-

gre'ssively pursued and promptly effected. Overall control

of the'"A" diesel generator outage was good (see paragraph

6).

Problems with inexperienced backshift personnel continue to

be evident although this has not as yet caused a challenge

to a safety-related system or adversely affected safe.oper-

.ation of the facility.

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2.2.6

Surveillance

Surveillances required by the technical specifications (TS)

were conducted at the specified frequency without excep-

tion. This appears to be due to the implementation of-a

strong administrative program which assures performance c?

TS surveillances at the specified frequency.

The surveillance tests were performed in a slow deliberate

manner ensuring that each step was completed prior to pro-

ceeding to the next step. During the performance of the

reactor protection system calibration, a question arose as

'

-to the requirement for shift supervisors' signatures on

attached data sheets. After talking to the shift supervi-

sor, an operations engineer, and an instrument and controls

representative, the shift inspector considered that it ap-

peared as though they understood when the signature was

required, although the data sheet was-vague in its wording

(see paragraph 3.2.5.3).

2.2.7

-Radiological Controls-

Radiological controls, including. personnel monitoring and

area posting, appeared adequate. The licensee has done a

commendable job in reducing contaminated areas.

Radiation

levels throughout the facility were very low, but-this

could be due to the relatively short operating time for the

plant since the October startup.

On November 11, 1985, a problem was noted in that a

hand-and-foot monitor was in use and beyond its calibration

due date by three days.

Radiological controls management

acknowle'dged that condition. This finding was contrary to

Radiological Controls Procedure 9000-PMI-4221.16, which

requires quarterly calibration of the instruments. They

-indicated that the instrument became inoperable due to the

out-of-calibration period expiring ~at midnight of the due

day and that no allowances are made to exceed this date.

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The' field operations manager checked with the Radcon in-

strument person _nel 'and the computer printout of October 26,

1985, indicated that calibration of this instrument was

~ due.

Radcon management attributed the missed calibration

to a move.in the facility where radcon instruments were

Estored. The manager also stated that this was no excuse

.and it-should still have been calibrated. -Another person-

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.nel monitoring station was available at,the Unit 1 Process-

.

ing Center. As a result of the use of the

out-of-calibration instrument, there was no incident of

_

contaminated personnel leaving-Unit 1.

The inspector had

no further questions on this' matter.

Review of applicable radiation work permits (RWPs) on

November 1, 1985, revealed that all general RWPs in effect

.for October.were renewed and properly placed on the bulle-

,

tin board. .On'e minor problem was noted in.that RWP No.

031257 expired on November 6,1985, but it was still posted

on November 7, 1985. Apparently,-no individual used that

RWP during its' expired time period.

Review of RWP require-

ments for'the building spray surveillance test and personnel

performing the-test showed-full compliance with the RWP.

During periodic tours, all high radiation area doors were

clearly marked and locked. No discrepancies were found.

2.23

Physical Security

.,

Based on routine observations of security systems _ operation

and guard performance, no adverse conditions were

identified in this area.

2.3 Conclusion

Overall, facility personnel continued to conduct themselves in a pro-

fessional manner. The shift inspectors attributed the procedural-

implerrentation problems to a lack of- attention to detril on the part

of plant personnel ~(see paragraph 3.3 for more_information). Opera--

tor response to minor events was conducted in a controlled and calm

manner. Overall, licensed operators continued to demonstrate good

knowledge of facility-design and of daily. plant status.

__The routine training program, in distinction to the restart qualifi-

-cation card program, apparently contributed to the quality of person-

nel performance since the plant startup.

The radiological controls program continued to evidence itself as a

-strength in the licensee's organization.

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3.

Plant Operations

3.1 Scope of Review

TMI-1 Restart Staff inspectors periodically inspected the facility to

determine the licensee's compliance with the general operating re-

quirements of Section 6 of the Technical Specifications (TS) in the

following areas:

--

review of selected plant parameters for abnormal trends

plant status from a maintenance / modification viewpoint including

--

plant housekeeping.and fire protection measures

control of ongoing and special evolutions, including control

--

room personnel awareness of these evolutions

--

control of documents including log-keeping practices

implementation of radiological controls

--

implementation of the security plan including access control,

--

boundary integrity, and badging practices

'

The inspectors also focused their attention on'the areas listed

below.

control room operations during regular and backshift hours, in-

--

cluding frequent observation of activities in progress, and

periodic reviews of seiected sections of the shift foreman's log

and control room operator's log and other. control room daily

. logs

followup items identified by shift inspector activities (see.

--

paragraph 2)

--

areas outside the control room

--

selected licensee planning meetings

As a result of this review, the inspectors reviewed specific events

in more detail as described in the sections that follow.

3.2 Findings

3.2.1

General

In general, housekeeping and fire protection measures re-

mained consistent with previous high standards in

safety related areas. The licensee has done a commendable

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job in ensuring safety-related spaces are well kept and

free of litter. The practice of assigning each space to a

particular department and a particular person in that de-

partment ensures that someone is ultimately responsible for

the cleanliness of that space. All anti-C clothing was

stored in designated areas in a neat manner. Containers

were clearly marked as to the type of trash or clothing

that was to be deposited in each.

In the turbine and control buildings, the practice of plac-

ing absorbent materials under machinery that may have oil

leaks ensured that the oil was contained within the bounds

of the machinery.

Fire loading in the turbine building was

addressed (paragraph 3.2.4).

Overall, the material condition of the plant remained quite

good.

Plant performance continued to be as expected.with

only minor safety related equipment problems.

Management involvement and attentiveness to daily activi-

ties continued. The maintenance department's daily plan-

ning meetings continued to keep management-abreast of plant

problems and to provide overall direction to proceed delib-

.erately and cautiously during steady-state power

operations.

3.2.2

Letdown System Operations

On October 25, 1985, the shift inspector witnessed the con-

trol room operators secure portions of the lineup associat-

ed with transferring (letting down) water from the reactor

coolant system via the letdown system to a reactor coolant

bleed tank. While performing the evolution, the shift in-

spector noted some confusion between the control room oper-

ators on the proper sequence of valve repositioning.

The TMI-1 Restart Staff reviewed and discussed with a se-

.

nior plant engineer the sequence of valve repositioning

that should occur during this evolution. This review was

performed initially using current plant drawings in order

to verify the accuracy of the procedur'e.

The inspector

then discussed the procedure with the operating crew that

had performed the water transfer on October 25, 1985.

The

control room operator walked.the inspector through the

steps that they had performed.

From the walkthrough and

review of the drawing,.the inspector concluded that the

applicable procedu're was correct as written. The problem

arose because the two operators involved did not properly

communicate their actions to each other.

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In the sequence of valve repositioning, the operators are

required to reposition the bleed valve (MU-V-8) from the

~

" bleed". position to the "through" position and then isolate

the bleed tank to which it is being drained.

If the isola-

-tion valve associated with the blet

tank is shut before

MU-V-8 repositions, the letdown sys tm would be pressurized

to reactor coolant system pressure and challenge the let-

down system relief valve.

Control of MU-V-8 is from the

.

center console in.the control room while the bleed tank

isolation valves are controlled from the liquid waste dis-

posal system panel in the back of'the control room.

The-operator at the center console requested that a second

operator walk over and shut the bleed tank isolation valve.

The operator at the console noted that pressure in the

~

makeup tank was high in its operating band. He paused for

a moment to allow pressure to drop before cycling MU-V-8;

however, the operator had already started for the panel at

the back of the control room. The shift supervisor who

realized that the valves were going to be repositioned in

the wrong order, which would cause the letdown system re-

lief valve to actuate, took control of the evolution and

remedied the situation.

As a precautionary measure, the

shift supervisor had the control room operators check the

indications in the control room to ensure that the relief

valve did not lift.

Discussions with the operators indicated that-the operator

at the center console had miscommunicated his directions as

to when to shut the bleed' tank isolation valve.

Prior to

these discussions, the shift supervisor had personally re-

viewed the situation with his crew.

The inspector noted

that the operators on this shift were fully knowledgeable

of the interactions of the two systems involved.

From re-

view of the control room records.and discussions with the

operators, the inspector concluded that the relief valve in

question had 'not lifted and the shift supervisor had main-

tained positive control of evolutions occurring in the

plant.

3.2.3

Nuclear Services Closed Cooling Water Relief Valve Design

Adequacy

The inspector reviewed the reactor building isolation valve

arrangement of the nuclear services closed cooling water

system. This review ~was in response to a shift

inspector's concern raised during a tour of the intermedi-

ate building during'the previous inspection period. The

arrangement consisted of relief valves (NS-V-36 A, B & C)

!

between.the reactor building wall and the isolation valve.

As part of.this review, the following documents were

examined.

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--

GAI Drawing C-302-610, " Nuclear Services Closed Cycle

Cooling Water System," Revision 31

--

SP 1300-3J, "NSCCW Pump and Valve Functional Test,"

Revision 9, dated July 23, 1985

TMI-1 FSAR Section 5.3, " Isolation System," dated July

--

1982

Operations Plant Manual, Section B-11, " Nuclear Ser-

--

vices Closed Cooling Water," Revision 0

OP 1104-11, "NSCCW System," Revision 23, dated Septem-

--~

ber 26, 1985

--

10 CFR 50, Appendix A, Criterion 57, " Closed System

Isolation Valves"

--

Safety Evaluation by the United States Atomic Energy

Commission, dated July 11, 1973

This as-built configuration of the reactor building isola-

tion valve arrangement for the NSCCW system is:

--

entrance isolation valves NS-V-52 A, B &.C

--

reactor building penetration-

--

fan motor coolers IA, IB & IC

reactor building penetration

--

relief valves NS-V-36A, B & C

--

9

--

exit isolation valves NS-V-53 A, B & C

The purpose of the NS-V-36 relief valves is to protect the

fan motor coolers against. rupture in case an overpressuri-

~zation condition were to occur. The relief valves were set

at 175 psig.

The NSCCW. system pressure is kept above the maximum pres-

sure attainable in the reactor building following a LOCA.

If a break were to occur, leakage would be into the reactor

building. -However, the NS-V-36 relief valves are located

in a position such that an unisolable leak may develop be-

tween the reactor building and the intermediate

building. This could occur if the NSCCW system isolated

causing the relief valves to lift and, subsequently, not

close. When system pressure dropped below reactor building

pressure, a line break would open a path outside of

containment.

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21

Criterion 57 of Appendix A, 10 CFR 50 states: "each line

.that penetrates primary reactor containment and is neither

part of the reactor coolant pressure boundary nor connected

directly to the containment atmosphere shall have at least

one containment isolation valve which shall be either auto-

matic or locked closed or capable of remote manual isola-

tion.

This valve shall be outside containment and located

as close to the containment as practical."

The TMI-1 FSAR, Section 5.1.2 classifies the NSCCW system

as a Type III fluid penetration.

Therefore, it must have

at least.two isolation barriers; the first must be at least

one valve, either a check valve or a remotely operated

valve located external to the reactor building. The closed

loop is considered the second barrier. . Table 5.3-2, "Reac-

tor Building Isolation Valve Information," in conjunction

with Figure 5.3-1,." Reactor Building Isolation Valve Ar-

rangement," depicts a configuration that does not truly

~

reflect the as-built condition of the NSCCW system.

Licensee representatives stated that the relief valves were

located outside primary containment so that water would not

be added to containment during a LOCA if the relief valves-

were to lift. They also stated that TMI-1 was 60% complete

when the FSAR was submitted to the Atomic Energy Commission

(AEC).

This was~before the present version of the General

Design Criteria (GDC) was published in July 1971.

The

plant was designed and constructed to meet the intent of

the original AEC's GDC proposed in July 1967 and other as-

pects were reflected in the staff SER for facility opera-

tion issued prior to the 1974 criticality.

The TMI-1 Restart Staff is questioning the design adequacy

of the relief valve configuration present in the NSCCW sys-

tem.

The implications of having a stuck-open relief valve

outside of containment concurrent with a break of the NSCCW

piping inside containment.could violate containment integ-

rity.

However, the probability of this occurring appears

to be extremely low.

Further NRC review of this matter will be necessary to de-

termine the design adequacy of the valve configuration and

~

to determine if the licensee is meeting GDC 57. The re-

sults of this review will be documented in a future inspec-

tion report (289/85-26-06).

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3.2.4

Turbine Building Housekeeping / Fire Hazard

Based on several walkdowns by shift inspectors, the turbine

building physical appearance led to the question of whether

the turbine building posed a fire hazard and could cause a

significant plant transient that would have an adverse ef-

fect on' plant safety.

The inspector performed an indepen-

dent walkdown.in the turbine building and developed a list

of potential fire hazards. This list was then compared

with a list that'is. prepared by licensee fire protection

engineers on a weekly basis.

The inspector did not identi-

fy any new findings that had not been identified by the

licensee personnel.

Based on discussions with licensee representatives and a

review of previous findings, the inspector determined that

if a significant fire hazard existed in the plant, immedi-

ate corrective actions were taken. The inspector noted-

that many of the NRC shift inspectors' observations were

. scheduled'as part of maintenance work.

The inspector con-

cluded that the licensee was appropriately identifying and

correcting potential fire hazards in the entire plant.

'

Because of the sign'ficant amount of burnable materials in

the turbine building, the inspector questioned wnether a

fire hazard loading review had been perform'ed for the tur-

bine building. The inspector questioned whether the amount

of material might exceed the capabilities of the fire sup-

pression system in the turbine building. Through a review

and discussion of the fire _ hazard analysis performed,

licensee representatives demonstrated that the loading did

-

not exceed the s/ stem design and that they were tracking

l

the changes in the loading of burnable material in the tur-

bine building.

The inspector concluded that the turbine building condi-

tions.did not-create a hazard that would affect plant safe-

ty.

However, the inspector noted that the physical

appearance of the turbine building was not maintained as

well as the rest of the plant. The TMI-1 Restart Staff

expects to continue to receive subjective comments about

the physical appearance of the turbine building from shift

inspectors.

3.2.5

Procedure Implementation

As a result of shift inspector referrals, the TMI-1 Restart

t

Staff reviewed the below listed findings to get a better

understanding of those findings and to assess their overall

safety implications.

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3.2.5.1

During liquid releases from the waste evaporator

condensate storage tanks on November 1 and 6, 1985, the

NRC shift inspector observed that the instrument and con-

trols technician performed A, B, and C steps of paragraph

'

1.3 of the applicable operating procedure out of sequence.

The operating procedure (0P) was OP 1104-29S, Revision 30,

September 5, 1985, " Transfer from'the Waste Evaporator Con-

densate Storage Tank," and it referenced surveillance pro-

cedure (SP) 1302-3.1, Revision 43, October 1, 1985, "RMS

Calibration." Appendix A, paragraph A.9 of that appendix

required essentially the same steps as B.1, B.2, and B.3 of

OP 1104-29S.

Further review by.the TMI-1 Restart Staff

revealed that the proper sequence of steps was:

calcula-

tion of the setpoint for RM-L6 bigh alarm to close the dis-

charge isolation valve; setting the calculated setpoint

into the instrument RM-L6; and performance of a functional

check of the interlock prior. to the discharge.

The OP

1104-295 did not reflect that proper sequence but the tech-

nician performed the evolution in the appropriate sequence

without signoff on SP 1302-3.1.

The licensee initiated a procedure change request to cor-

rect the deficiencies of OP 1104-29S and to correct some

other minor discrepancies.

These changes will make this

procedure compatible with SP 1302-3.1.

The TMI-1 Restart-Staff concluded that the two proce-

dures were inconsistent; and, because of the frequency at

which personnel' implemented these procedures, facility per-

sonnel neglected to formally initiate a change to the pro-

cedure. The actual steps performed were technically

correct.

3.2.5.2

On November 5, 1985, an NRC shift inspector observed an

operator remove a circuit board related to annunciator win-

dow B-4-5 on the heating and ventilation panel. The opera-

.

tor did this to clear an annunciator " nuisance" alarm to

prevent unnecessary distractions while attempting to place

the B main feedwater pump in automatic. The TMI-I Restart

Staff observed this practice throughout the startup testing

period since October 3, 1985.

The NRC shift inspector ques-

tioned why an out-of-service tag was not placed on the an-

nunciator because removal of the circuit board places an

alarm input in an out-of-service (00S) condition.

The TMI-1 Restart Staff reviewed AP 1036, Revision 6,

February 10, 1985, " Instrument Out-of-Service Control."

Paragraphs 3.1.1.a to e of the procedure required that,

if an instrument was out of service for more than a

shift, the instrument out-of-service condition was to be

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6

logged in the out-of-service log with the appropriate

sticker affixed to the instrument.

However, parag'raph

3.1.1.d required that the 00S sticker was to be affixed to

any alarm window in which the alarm circuitry was disabled

for maintenance.

The operator's action noted'above was not. exactly for main-

tenance; and in actuality the procedure did not address

preventing a nuisance alarm. When questioned, licensee

representatives indicated that past practice was to use the

more than a shift criterion for placing the instrument in

the 00S log.

The inspector reviewed controls that would remind operators.

that an instrument was 00S. The controls should result in

the proper recording of the 00S condition and the turnover

of that information to the next shift. The inspector con-

firmed that the control room operator turnover checklist

provided for 00S entries.

The inspector concluded that the licensee's procedure was

not clear on the handling of nuisance alarms and licensee

representatives had not initiated a change to the procedure

prior to this time to update the procedure to current

practice.

3.2.5.3

On October 30, 1985, the shift. inspector observed the high

neutron flux reactor trip calculation and instrument set-

ting (based on heat balance) in accordance with SP

1303-4.1. Page 17 of the procedure. required that the shift

supervisor indicate approval of the calculation by signing

i

approval on the space provided. When questioned by the

. shift inspector, the shift supervisor (SS) was not sure

when the SS was to sign regarding the following options:

(1) after the calculation and before instrument setting,

or (2) after calculation and instrument setting. The

licensee's operations staff indicated that option (2) was

required.

Based on the inspector's review of the procedure, it ap-

peared that option (1) was required.

The. inspector concluded that the procedure was not clear'as

to'when the shift supervisor's signature was required.

Further,~in spite of the numerous. times this procedure sec-

tion was implemented, licensee representatives had not ini-

tiated a change to the procedure to enhance its clarity.

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3.2.5.4-

Between November 7 and 10, 1985, the.NRC shift inspector

selectively verified that proper valve / breaker positions

were in accordance with the. licensee's locked valve list,

as required by AP 1011. The inspector found no valves mis-

positioned.

However, he found core flood valve CF-V-3B

breaker not properly locked. The operator properly locked

the breaker.

The TMI-1 Restart Staff reviewed administrative procedure

AP 1011, Revision 24, July 7,1983, " Controlled Key Locker

Control." The procedure provides no guidance on how to

check valves or breakers to determine if they are properly

locked.

Further, there was no provision to periodically

verify the list during power operations.

The mechanical

and electrical lineup operating procedure for CF-V-3B does

not require the breaker for.CF-V-3B to be locked. This

operating procedure and others do not required independent

verification of valve lineups if performed such as for

startup after a major outage.

Based on previous NRC review

in this area, restart valve lineups were properly

conducted.

The inspector concluded that AP 1011 lacked clarity, over-

all, on exactly how to implement locked valve controls

'

aside from isolated valve / breaker repositioning for infre-

quent special evolutions / tests or preventive / corrective-

maintenance.

Further, existing administrative controls on independent

verification of proper valve / breaker positions were mini-

mal;

i.e., removing and restoring equipment.to service in

meeting TAP I.C.6 requirements, as described in NUREG-0578

and NUREG 0737. Had additional or clearer controls been in.

place, .the improperly locked CF-V-3B breaker might have

been detected upon independent verification (see paragraph

'

3.2.5.5).

3.2.5.5

During the setting of the liquid radiation monitor RM-L6

interlock for discharge isolation on November 1, 1985, the

shift inspector observed that the I&C technician reposi-

tioned a " slide link" (or essentially lifted or discon-

nected electrical leads). The shift inspector questioned

whether or not an independent verification of that action

(on removing or restoring to service) was warranted.

The

TMI-1 Restart Staff reviewed administrative procedure (AP)

1013, Revision 17, October 23, 1985, " Bypass of Safety

Function and Jumper Control." Paragraph 1.3.A.2 required

that, for evolutions involving electrical jumpers, lifted

leads, or temporary mechanical modifications, entries into

the applicable log were not needed provided that the shift

supervisor was informed and a properly approved procedure

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26

which initiated a 10 CFR 50.59 review was used. This ap-

plied to important-to-safety (ITS) systems and manipula-

tions to an ITS system were required by~0P 1104-29S.

However, that procedure did not require independent verifi-

cation of the movement of the slide link for setting the

RM-L6 setpoint.

'The inspector concluded that OP 1104-29S was inconsistent

with AP 1013 requirements.

Further, licensee management

initially responded that AP 1013 was not reflective of com-

mitments to NRC on independent verification because ITS

systems are a broader classification of systems as com-

pared to safety-related systems, as defined by the

licensee's operational quality assurance plan.

Licensee

representatives agreed to review this-matter.

Licensee

commitments and implementation for independent verification

is unresolved pending completion of licensee action as

stated above and subsequent NRC Region I review

(289/85-26-02).

,

3.2.5.6

Between November 1 and 8, 1985, the shift inspector

verified proper calibration of instrumentation used to per-

form the RCS leakrate calculation.

During this review,-the

shift inspector identified that the manual method (if the

plant computer is unavailable) could use a general form for

makeup tank and reactor coolant drain tank (RCDT) tempera -

ture and one procedure step called for the recording of

makeup tank or RCDT temperature.

Steps later in the form

do not use RCDT temperature since.the calculation assumed

120-140 degrees F in the RCDT. The step calling for RCDT

temperature could be confusing to the operator but the in-

spector did not consider it significant in that it would

not preclude satisfactory completion of the manual calcula-

tion. The' inspector had no further comments on the

-

procedure.

Similarly SP 1301-8.2, " Diesel Generator Annual Mainte-

nance," paragraph 6.3.5 and 6.3.6 erroneously referenced

preventive maintenance procedure E-38 as the frequency me-

ter calibration procedure and E-39 as the voltmeter cali-

bration procedure, respectively.

In actuality, they were

reversed but it did not preclude the satisfactory comple-

tion of the required procedure.-

3.2.5.7

On November 2, 3, and 10, 1985, during a radioactive waste

gas tank discharge, the shift inspector identified a fail-

ure to meet a procedural step.

Paragraph 3.7.2.5 of OP

1104-27, Revision 34, October 31, 1985, " Waste Disposal-

Gaseous," required that_an operator position an appropriate

sign, " Gas Release in. Progress," at the auxiliary building

and control room liquid waste disposal panels. They posted

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27

the sign at the radwaste panel in the auxiliary building-

as required by the procedure but not in the control room.

Two shifts were involved in the three incidents observed

and in each case the NRC inspector identified the failure

~

to comply with the procedure step. The procedure step was

also signed off as complete, apparently only because the

auxiliary building ' sign was posted.

'

The' posting of the sign in the control room was a licensee

initiative to highlight to operators that the evolution was

in progress. The inspectors concluded that the failure to

comply with the procedural. step was minor in nature.

How-

ever, the repetitive failures to comply could not be

considered an isolated case.

.The inspector concluded that, for the above instances,

shift personnel had not demonstrated the initiative or at-

tention to detail to take effective corrective action to

assure-implementation of the procedurally directed steps

desired by licensee management.

3.2.5.8

In summary, many of the procedural steps reflect licensee

initiatives to properly control the evolutions dictated by

these procedures.

Each of the above items can be grouped

into the following categories:

failure of personnel to

strictly adhere to individual procedural steps; failure of

personnel to assure that each procedural step adequately

meshes within its procedure or is consistent with a refer-

enced procedure; and/or failure of field personnel to con-

scientiously provide feedback to licensee management for

procedure improvement. Although each personnel performance

shortcoming was minor in nature, collectively they raised

concerns by NRC staff members as to whether or not an un-

derlying problem existed within the licensee's organiza-

tion.

Licensee representatives stated that, in their

opinion, personnel performance has been good overall;

however, corrective actions would be taken. The TMI-1 Re-

start Staff considered this area to be unresolved pending

additional review by the staff (289/85-26-03).

3.3 Conclusion

Licensee management and their quality assurance department continued

their detailed attentiveness and involvement in daily activities.

'Although some communication lapses occurred, they were considered

minor and they had no adverse affect on the facility. The material

condition of the facility remained quite good and the licensee demon-

strated control of fire hazard loading as noted in the review of that

area in the turbine building.

It was noteworthy that the licensee

was essentially on schedule with their power escalation program and

that the per formance of equipment and personnel had not resulted in a

pla.nt trip or challenge to other safety-related equipment.

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Ine procedure implementation problems were indicative of a poten-

tially adverse trend.

Considering the number of procedural steps

implemented at the facility on a daily basis, the NRC-observed prob-

lems might be considered statistically insignificant, and they were

individually minor problems.

However, the TMI-1 Restart Staff con-

siders that these problems may be indica'tive of a more fundamental

problem.

Licensee management may need to focus additional attention

on this area. The TMI-1 Restart Staff will continue to review this

along with how the licensee conducts independent verifications in the

next . inspection period.

Further, the TMI-1 Restart Staff will continue to review the design

adequacy of the relief valve between the reactor building wall and

the primary containment isolation valves in the NSCCW system.

4.

Radioactive Water Spill in Makeup Pump Cubicle

4.1 Event Chronology

On October 28, 1985, between 4:20 and 4:30 p.m., approximately 150

gallons of primary water spilled in the

"A" makeup pump cubicle.

The

plant was at steady-state 48*s power and the reactor coolant system

was at normal operating temperature and pressure.

Auxiliary operators (A0s) were to isolate and drain makeup pump

MU-P-1A to repair a leaking drain plug in the discharge line. This

task was to be accomplished by using switching order 85-1672. After

the isolation valves were positioned and tagged in accordance with

the switching order, the pump casing drain in the cubicle was opened.

'

Water then began to flow into the floor drain.

At this time, the control room operators noticed a decreasing level

in the makeup tank, MU-T-1.

Before the shift foreman could tell the

A0s to close the drain valve, local radioactivity airborne monitoring

system AMS-3 alarmed in the MU-P-IA cubicle. The A0s exited the cu-

bicle and the drain valve was left open.

Within ten minutes, the A0s re-entered the cubicle to shut the drain

valve. Water was running off the skid, which supports the pump and

motor, on to the floor because the floor drain was apparently

clogged.

The A0s shut the drain valve and the decreasing level in

MU-T-1 stopped.

The shift supervisor dispatched the_ shift foreman to the auxiliary

building where he verified the switching order as being correct and

the isolation valves were recycled. A second attempt to drain the

pump was made at 8:00 p.m.; this attempt failed. The shift supervi-

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sor ordered a suction cross-connect valve to be closed to provide

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double isolation from MU-T-1.

A third attempt to drain MU-P-1A was

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made;'this attempt was successful.

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Auxiliary building airborne radioactivity levels remained higher than

normal but still below RM-A 4, 6 and 8 alarm setpoints from 4:30 to

11:00 p.m.

However, AMS-3 alarms were received in the Unit 2 fuel

handling building.

The reasons for the Unit 2 alarms are due to in-

terconnection of the fuel handling buildings and their ventilation

systems.

Radiological engineers estimated approximately 0.7 Ci of

noble gases (mostly Xe-133) was released; the dose calculations at

the site boundary was estimated to be O'.002% of the federal quarterly

limit for the noble gas released to the environment from the plant's

ventilation system.

4.2 Scope of Review

The inspectors reviewed the details of the spill and the licensee's

review of this event to determine:

details regarding the cause of the event and event chronology;

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consistency of licensee actions with NRC license and procedural

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requirements; and,

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proposed licensee actions to correct the cause of the event.

The inspectors' review of this incident included discussions with

cognizant licensee personnel and review of the following documents

system drawings

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CR0 and SF logs

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Switching Order 85-1672

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Surveillance Procedure 1101-2.1, Revision 16, dated October 28,

1985

. Administrative Procedure 1044, Revision 14, dated June 4, 1985

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various radiological data and calculations

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critique meeting minutes

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The inspectors also attended the radiological investigative critique

that was held on October 29, 1985, and conducted inspections of the

MU-P-1A cubicle.

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4.3 Licensee Findings / Actions

The cause of the event was leakage past closed suction valve

MU-P-72A.

The event was complicated by a clogged floor drain and the

A0s exiting the cubicle 'without closing drain valve MU-V-172A after

AMS-3 alarmed.

The licensee determined that radioactivity levels

- were greater at RM-A4 as compared to RM-A6 because the upper level

fuel handling building ventilation system draws air from the Unit I

side of the building including the exhaust header.

Since radiation

levels did not get high enough to isolate the fuel handling building

dampers, the noble gas also dispersed to TMI-2.

Nine workers were slightly contaminated with short-lived radioactive

particulate daughter products; principally, Rb-88. Personnel exposure

was derived'from both isotopes.

However, within ninety minutes the

contamination had decayed away. Whole body counts of the individuals

involved identified no substantial change from the baseline whole

body count.

As part of the licensee's follow-up actions, this event will be re-

viewed with each of the operations crews. Work requests will also be

initiated to repair MU-V-72A and the clogged floor drain; MU-V-72A

will also be caution tagged. Grab samples in the area during the

event identified noble gas radioactivity as the primary constituent;

i.e., XE-133, Xe-133m, and Kr-88.

Such activity indicated that. valve

leakage was occurring probably from MU-V-72A.

Total release due to the event was estimated to be about 6.77 E5 uCf,

resulting in approximately 7.39 E-5 mrad gamma air dose and 8.69 E-5

mrad beta air dose as compared to the quarterly limits of technical

specifications;

i.e., 5.0 mrad and 10.0 mrad respectively.

The job was completed without specific procedural coverage using a

standing radiation work permit (RWP). The licensee immediately re-

vised standing RWPs to assure that radiological controls personnel

were contacted if systems were to be opened on a standing RWP.

Licensee management indicates that standing RWPs would not be used in

the future to drain open systems containing radioactive material.

4.4 NRC Finding

Licensee action of self review was appropriate and reasonably thor-

ough.

However, based on the above review, the inspector found some

items of concern.

Operations personnel did not close the drain valve

before evacuating the makeup pump cubicle, and their use of a stand-

ing RWP caused radiological controls personnel to be unaware of the

draining operation.

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The licensee's radiological incident report (RIR) was to be finalized

by the operations department along with prepared corrective actions.

This area is unresolved pending issuance of the RIR by the licensee

and subsequent TMI-1 Restart Staff review (289/85-26-04).

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4.5 Conclusions

,

The operators' response at the cubicle was poor in that they failed

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to isolate the leak before evacuating the makeup cubicle.

Contral

room operators were responsive to the symptoms noted in the control

room. Critique review and actions were appropriate for the circum-

stances. The TMI-1 Restart Staff will review the completed RIR and

licensee corrective actions.

>

5.

Nuclear Plant Staff Working Hours

The inspector conducted a review of the overtime expenditure for.certain

operating personnel, including senior reactor operators, reactor opera-

tors, and radio. logical controls personnel for the period September 30

through October 27, 1985.

The licensee is required to limit overtime-use

in accordance with Administrative Procedure (AP) 1031, and Technical Specification (TS) 6.8.1.

This subject was previously reviewed by Region

I as discussed in Inspection Report 50-289/85-08, dated March 8, 1985,

which verified-licensee awareness of the requirements and compliance to

AP-1031 and Generic Letter No. 82-12.

The inspector determined that no onshift SR0s, R0s or radiological control

personnel exceeded AP-1031 policy limits during the above period.

In some

instances personnel were required to work hours that were close to the

weekly limits of 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />'in a 7 day work week. This was due to the

. training required for restart activities for licensed operators. The

licensee does maintain a sufficient staff of licensed personnel (six full

shifts) to. permit compliance with Commission policy that encourages a 40

hour, 5 day work week, with overtime use on only rare occassions.

The

licensee is encouraged in the. future to plan training evolutions in ad-

vance of other activities to' alleviate the use of excessive overtime and

long work periods for licensed operators.

The inspector had no further comments in'this area.

6.

Diesel Generator Maintenance

6.1 Background

The emergency diesel generator (EG-Y-1A) was isolated for the annual

preventative maintenance inspection on November 4, 1985.

The

licensee entered a seven day action statement in accordance with

Technical Specification (TS) 3.7.2.c which requires that the operable

diesel be run immediately and daily during the period when the other

diesel is inoperable.

If the diesel generator cannot be made opera-

ble within the seven day period the reactor must be shut down.

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6.2 Scope of Review

The following procedures, manuals and documents were reviewed by the

inspector during the course of the maintenance inspection.

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Technical Specification Section 3.7, " Unit Electr1tal Power

System"

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Maintenance Procedure (MP) 1405-3.2, Revision 7, " Diesel Genera-

tor Maintenance"

Operating Procedure (0P) 1107-3, " Diesel Generator Operation"

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Surveillance Procedure (SP) 1301-8.2, " Diesel Generator Annual

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Inspection"

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SP 1302-5.30, " Diesel Generator Electrical Inspection"

SP 1303-4.16, " Electrical Power Systems"

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During'the course of the maintenance the inspector observed selected

portions of the mechanical and electrical surveillance activ.ities.

These included removal, testing, reinstallation of injectors,

inspection of cylinder wall for water leaks, pressure switch setpoint

checks, generator brush replacement, and cooling water check valve

inspection. The inspector observed vendor representative and QC/QA

involvement during the inspection process.

6.3 NRC Findings

The inspector verified that the operaM.e diesel (EG-Y-18) was tested

at the required frequencies and that the licensee maintained the re-

quired systems, subsystems, trains, and components that depend on the

operable diesel generator in an operable condition as required by

technical specifications. This was accomplished based on shift in-

spector review of the licensee's engineered safety features /emer-

gency feedwater operability checklist.

The licensee discovered two minor problems during the inspection.

The first was a small engine coolant leak from the cylinder cooling

jacket lower end to the engine crankcase.

This type of leak was con-

sidered acceptable by the licensee and vendor personnel and would not

result in any engine damage during operation, or prevent proper

starting of the diesel. The licensee plans to increase sampling fre-

quency of the diesel lube oil for presence.of water from the normal

six-month period to weekly to assure that there is no excessive water

buildup in the lube oil system. The inspectors concurred with this

action and had no questions on the subject. The inspectors will re-

view licensee action on thi; subject in future inspections.

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The second problem occurred with check valves that were removed from

the diesel cooling systems.

These valves were required to be in-

spected in response to IE Bulletin 83-03.

The licensee determined

that valve EG-V48A, had a rubber valve seating surface that was

slightly rippled and could possibly cause improper seating of the

valve.

In conjunction with the licensee's plant engineering depart-

ment review of this valve, the maintenance department determined that

this was an acceptable condition and the valve would function proper-

ly but no written documentation was issued on this review other than

the as found condition on the procedure data sheet.

The inspector

concurred with this action and had no further questions. The inspec-

tor will examine diesel engine standby heating operation in future

inspections to ensure proper operation of this check valve.

The shift inspector obses.ad that, after removal from the system,

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check valve EG-V48A wrc .* en approximately 20 degrees. The check

valve has butterfly type

iscs that are intended to be 90 degrees

full open or 0 degrees full closed (by spring action) depending upon

. system operating ,:ondition. The licensee organization did not con-

sider this to 'be a problem since the maintenance record, and plant

inspection reports do not reflect this as found condition and there-

fore, there was no engineering evaluation on the condition.

Based on

inspector judgement, it appeared that the valve would seat closed on

reverse flow differential pressure, so no concern was identified for

the operability of the check valve.

The check valve prevents " keep warm" system water from being diverted

to thel radiator section of the cooling water system thereby prevent-

ing inadvertent cooldown of the system.

The as found condition will

be reviewed further by NRC Region I during the next inspection period

related to the B. diesel maintenance (289/85-26-05).

The inspector observed the licensee QA and QC coverage during the

inspection. QA personnel were present and observed that selected

tests and actions performed by maintenance personnel were conducted

in accordance with the applicable procedures.

The QA/QC personnel

are alerted to the performance of maintenance on safety-related

equipment by a memorandum from the maintenance department.

This al-

lows the QA/QC personnel to perform their audit function on a random

basis.

The surveillance procedures do not contain any specific hold

points where QA/QC would be required to verify any critical data that

were to be recorded. This would only be accomplished if the QA/QC

personnel were present at the correct time.

The inspector feels that

the maintenance and surveillance procedures that accomplish poten-

tially critical work on the emergency diesel generators should be

reviewed in advance of performance to identify in the procedure any

critical data points requiring independent verification by QA/QC per-

sonnel, vice the random method that is now employed. Although the

QA/QC coverage appeared to be satisfactory based on inspector obser-

vations, an independent evaluation of critical data should be made by

senior QA personnel and reflected in the procedure.

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6.4 Conclusion

The annual in,spection of emergency diesel generator 1A was accom-

plished in a manner'that would ensure proper operation of the diesel.

Vendor involvement and QA/QC coverage were acceptable. .A review of

the completed maintenance package and test results will be conducted

by the TMI-1 Restart Staff' subsequent to the B diesel outage.

7.

Safety-Grade Emergency Feedwater Inplant Review

On November 7, 1985, members of the TMI-1 Restart Staff inspected spaces

in the intermediate and diesel generator buildings which housed a major

portionuof the emergency feedwater system.

The purpose of the inspection

was to identify to members of the Office of Nuclear Reactor Regulation

items of concern in which there was a potential that the system design

would not meet the safety grade criteria embodied in the General Design

Criteria (10 CFR 50, Appendix A) and related guidance in the NRC staff's

standard review plan. The inspection consisted of a flow path walkdown

along with the identification and system description of major components

and support system components such as the 2-hour backup air supply system.

As a result of this review, a number of questions were formulated by the

staff to be considered by the licensee (Attachment A).

-These and other items were discussed with licensee representatives at an

interim exit meeting on November 7, 1985.

Licensee representatives agreed

to submit a report on'the resolution of the inadvertent actuation of the

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steam ~ relief valves (MS-V22A/B) to the EFW turbine (Inspection Report

50-289/85-22). The Technical Specification Table 4.1-9 for EFW valve

lineup will be updated prior to cycle 6 startup.

(The operating procedure

assures proper EFW valve alignment.) The licensee plans to make the re-

dundant block valves manual valves instead of automatic. The staff indi-

cated that this is a significant change in commitment which will be need

to be evaluated by the staff. The staff accordingly urged the licensee to

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make the submittal on this issue as soon as possible.

.8.

Exit Interview

The inspectors discussed the inspection scope and findings with licensee

management at the exit interview conducted on November 12, 1985. The

following licensee personnel attended the final exit meeting:

J. Colitz, Plant Engineer Director, TMI-1

T. Hawkins, Manager, TMI-1 Startup and Test, Technical. Functions

H. Hukill, Vice President and Director, TMI-1

C. Incorvati, TMI-1 Audit Supervisor, Nuclear Assurance

G. Kuehn, Manager, Radiological Controls TMI-1, Radiological and

Environmental Controls

S. Otto, TMI-1 Licensing Engineer, Technical Functions

L. Ritter, Administrator II, Plant Operations, TMI-1

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M. Ross, Manager, Plant Operations, TMI-1

D. Shovlin, Manager, Plant Maintenance, TMI-1

C. Smyth,.TMI-1 Licensing Manager, Technical Functions

. . Toole, Operations and Maintenance Director, TMI-1-

The exit. meeting was also' attended by S. Maingi, a nuclear engineer repre-

senting-the Commonwealth of Pennsylvania. The inspection results, as dis-

cussed at the meeting, are summarized in the cover page of.the inspection-

report.

Licensee representatives indicated that none of the subjects dis-

cussed contained proprietary information.

Unresolved items are matters about which information is required in order

to ascertain whether.they are acceptable items,. violations, or deviations.

Unresolved item (s), discussed during the exit meeting, are documented in

paragraphs 2.2.2, 3.2.3, 3.2.5.5, 3.2.5.7, 4.4, and 6.3.

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ATTACHMENT A

REQUEST FOR ADDITIONAL ~INFORMATION

TMI-1 EMERGENCY FEEDWATER SYSTEM MODIFICATIONS

1.

. Confirm that ducting, piping.and other components that could potentially

impact the backup instrument air bottles in the diesel generator room are

either seismically supported or, if not, that their failure would not re-

sult in loss of function of the backup air bottles.

For equipment that is

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' seismically supported, provide the criteria used to establish seismic

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qualification (e.g. Regulatory Guide 1.29).

2.

Provide a discussion which justifies the proposal-to change the failure

. mode for the new emergency feedwater flow control valves (EF-V30s) to

-closed rather than open on loss of air.

This discussion should address

'the importance of assuring reliable emergency feedwater flow against other

considerations such as overcooling / overfilling.

3.

Describe those' features (indications) and actions relied on to alert the

operators of flooding in the tendon access gallery in the intermediate

building, for example, as a result of a main feedwater line break,

Speci-

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fy'the design basis for.these features.

This discussion should also ad-

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dress the actions taken in the event of inadvertent indication of flooding

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and'an assurance that these actions will not cause unnecessary challenges

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to safety systems.

4.

What are the additional hazards and/or effects on safety-related systems

in the intermediate building (especially the emergency feedwater system)

with the storage of hydrogen and oxygen calibration gas bottles in the

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vicinity of safety-related equipment?

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