ML17263A293

Insp Rept 50-244/93-06 on 930310-0422.Violations Noted. Major Areas Inspected:Plant Operations,Radiological Controls,Maint/Surveillance,Security,Engineering/Technical Support & Safety Assessment/Quality Verification

ML17263A293
Person / Time
Site:	Ginna
Issue date:	05/13/1993
From:	Lazarus W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML17263A291	List: ML17263A290 ML17263A292 ML17263A293
References
50-244-93-06, 50-244-93-6, NUDOCS 9305240138
Download: ML17263A293 (20)

See also: IR 05000244/1993006

Text

U. S. NUCLEAR REGULATORY COMMISSION

REGION I

Inspection Report 50-244/93-06

License: DPR-18

Facility:

R. E. Ginna Nuclear Power Plant

Rochester

Gas and Electric Corporation (RG&E)

Inspection:

Inspectors:

Approved by:

March 10 through April22, 1993

T. A. Moslak, Senior Resident Inspector, Ginna

E. C. Knutson, Resident Inspector,.Ginna

~W~ /PD

W.

s,

'ef, Reactor Projects Section 3B

INSPECTION.SCOPE

Date

Plant operations, radiological controls, maintenance/surveillance,

security, engineering/

technical support, and safety assessment/quality

verification.

INSPECTION OVERVIEW

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shutdown, and plant condition changes,

including reduced inventory operations.

Operator

response

to failure of both source range nuclear instruments during reactor shutdown was

appropriate.

PORC provided effective oversight of plant operations through major outage

activities, as well as thorough assessment

and conservative guidance in response to off-

normal plant incidents.

d fgdg -,

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"gg

was the lowest in the plant's history and reflected effective integration of engineering support

to resolving ALARAissues.

A record low number of personnel contaminations for the

outage (95 cases) further demonstrated

conscientious application of routine health physics and

ALARApractices.

'n

n

il: Stem/disc separation

was discovered as an existing condition in

two valves during extensive refurbishment of the service water system; neither of these valve

failures resulted in loss of equipment function, service water cooling capability, or long-term

loss of service water redundancy.

A containment integrated leak=rate test was satisfactorily

completed.

A fire in the containment building occurred when hot slag from a nearby cutting

job ignited badly loaded roughing filters in the "C" containment recirculation fan.

Failure to

assess

the condition of the roughing filters represented

a shortcoming in the licensee's

maintenance program and is a violation of 10 CFR 50 Appendix B corrective action

requirements.

9305Z40138

930514

PDR

ADOCK 05000244

6

PDR

(OVERVIEW CONTINUED)

- Qg~ri: The blizzard of March 13-14, 1993 produced damage to the site perimeter fence

and intrusion detection system.

Appropriate compensatory

measures

were promptly

implemented and repairs were expeditiously completed.

r

n

r n;

The annual siren activation test was completed with 97 percent of

the sirens demonstrated

to be fully operational.

En in

rin /T hni: Video inspection of the reactor coolant pump internals

verified that a bolt, discovered

unattached in the reactor vessel during the 1990 refueling

outage, had come from the "A" reactor coolant pump diffuser adaptor ring.

TABLEOF CONTENTS

OVERVIEW

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TABLE OF CONTENTS

1.0

PLANT OPERATIONS (71707)..............; ..

1.1

Operational Experiences

1.2

Control ofOperations...................

1.3

Failure of Source Range Nuclear Instruments During

1.4

Control of Operations with Reduced Reactor Coolant

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Shutdown .....

System Inventory

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2.0

RADIOLOGICALCONTROLS (71707)........

2.1

Routine Observations ...............

2.2

Outage ALARAReview ...........-..

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3.0

4.0

MAINTENANCE/SURVEILLANCE(62703, 61726)

3.1

Outage Maintenance Summary

3.2

Service Water System Refurbishment

3.2.1

Failure of Valves 4669 and 4738

3.2.2

Desludging of Screen House ......

Surveillance Observations

Containment Integrated Leak Rate Test

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Pire Involving "D" Containment Recirculation

Containment Material Condition

3.3

3.4

3.5

3.6

SECURITY (71707)

4.1

Routine Observations ...............

4.2

Storm Damage to Site Perimeter Fence ....

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5.0

EMERGENCY PREPAREDNESS

(71707)

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'1Snen Test ...

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6.0

ENGINEERING/TECHNICALSUPPORT (71707, 92701)

6.1

Missing Reactor Coolant Pump Diffuser Bolt

6.2

Containment Recirculation Fan Cooler Replacements......

6.3

Safety Injection System Accumulator Drain Valves Seismic

Quallficatlon

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7.0

SAFETY ASSESSMEN'I'/QUALITY VERIFICATION(90712,

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7.1

Periodic Reports ........................

7.2

Licensee Event Reports....................

90713, 92701,

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8.0

ADMINISTRATIVE(71707, 30702, 94600)

8.1

Backshift and Deep Backshift Inspection

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ExitMeetings..........................

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DETAILS

1.0

PLANT OPERATIONS {71707)

1.1

Operational Experiences

At the beginning of the inspection period, the plant was operating at 89 percent power,

coasting down at approximately one percent per day to the start of the cycle 23 refueling

outage.

At 7:00 a.m. on March 12, 1993, shutdown for the outage commenced with a 10

percent per hour power reduction.

The main generator output breakers were opened at 2:09

p.m. and the main turbine was tripped at 2:14 p.m.

The reactor trip breakers were opened

at 2:30 p.m., March 12.

The plant entered cold shutdown mode at 6:48 a.m., March 13.

The blizzard of March 13-14 had no significant effect on the safety of plant operations.

The

reactor coolant system (RCS) was drained to mid-loop level on March 16 for installation of

steam generator nozzle dams in support of U-tube inspections and repairs.

A complete

reactor core off-load to the spent fuel pool was conducted to support service water system

refurbishment and reactor coolant pump inspections.

Core reload commenced on April 3 and

was completed on April 5.

Following completion of steam generator U-tube repairs, the

RCS was again drained to mid-loop level on April 8 for removal of steam generator nozzle

dams.

RCS filland vent was completed on April 18 and forced circulation was restored on

April21.

Heat input from the operating reactor coolant pump raised RCS heatup

temperatures

to just below 350'F to support RCS hydrostatic testing.

This test was

completed on April22. At the close of the inspection period the licensee was preparing to

continue plant heatup to normal operating temperature for completion of testing prior to

reactor startup.

1.2

Control of Operations

Overall, the inspectors found the R. E. Ginna Nuclear Power plant to be operated safely.

Control room staffing was as required;-

Operators exercised control over access to the

.

control room.

Shift supervisors consistently maintained authority over activities and provided

detailed turnover briefings to relief crews.

Operators adhered to approved procedures

and

were knowledgeable of off-normal plant conditions.

The inspectors reviewed control room

log books for activities and trends, observed recorder traces for abnormalities,

assessed

compliance with technical specifications, and verified equipment availability was consistent

with the requirements for existing plant conditions.

During normal work hours and on

backshifts, accessible

areas of the plant were toured.

No operational inadequacies or

- concerns mere identified.

Several management initiatives were implemented during the 1993 outage to improve the

overall control of maintenance/modification

activities.

These included establishing a work

control center (WCS) outside of the control room for carrying out mechanical/electrical

isolation work area boundaries for the various jobs.

The.WCS was maiiaged by a Work

Control Supervisor {a licensed senior reactor operator) with a supporting staff of licensed

reactor operators and auxiliary operators to set and maintain plant configuration.

Final

authorization for requested

actions was by the control room Shift Superv'isor prior to

implementation to provide an added check to assure that the safety of an on-going task would

not be compromised by a requested configuration change.

The shifting of switching/tagging

responsibilities away from the control room to the WCS decreased

the volume of control

room traffic and accompanying distractions to the on-shift control room staff.

The insp'ector

concluded that these administrative changes

enhanced

the overall coordination of outage

activities and plant safety.

Throughout the outage period, frequent Plant Operations Review Committee (PORC)

meetings were held to assess

the safety and regulatory significance of off-normal plant

incidents, the turnover status of plant modifications, and plant readiness for startup.

Strong

management oversight was demonstrated

in the control of plant activities.

1.3

Failure of Source Range Nuclear Instruments During Shutdown

The principle measure of reactor power is derived from neutron flux immediately outside of=

the reactor vessel.

Since neutron flux is directly proportional to the rate of nuclear fission,

this provides real-time indication of reactor power for operational control as well as for the

automatic reactor protection system.

The level of neutron flux varies greatly (approximately

10 orders of magnitude) between shutdown and full power operations.

Because the range of

power is so large, different neutron detection mechanisms

must be employed to cover the full

range of reactor power; a detector that is sensitive enough to monitor shutdown neutron

levels becomes

saturated when the reactor is at power, while a detector that is rugged enough

to operate in the heavy neutron flux present during power operations lacks sensitivity at low

reactor power.

Excore nuclear instruments use three types of detectors,

and reactor power

is, by convention, considered in terms of the three corresponding instrument ranges:

1) the

source range (so called because

the neutron population consists of neutrons originating from

the source fissionable material); 2) the intermediate range; and, 3) the power range.

To

ensure continuity of indication as well as to provide an operability cross-check,

the

intermediate range overlaps both the source and power ra'nge.

To avoid detector saturation

and resultant damage due to high neutron and gamma flux, power is only applied to the

source range detectors when reactor power is within the range of the instruments; this occurs

automatically when reactor power is low in the intermediate range.

During the reactor shutdown on March 12, 1993, the two source range nuclear instruments

(SRNIs) failed to indicate power level when reactor power reached the automatic energization

(P-6) setpoint.

Operators attempted to manually energize the SRNIs, but this had no effect.

The reactor shutdown continued in parallel with these events and reactor power decreased

below the indicating range of the intermediate range nuclear instruments at 2:25 p.m.; from

this point until the reactor trip breakers were opened at 2:30 p.m., the plant was operating

with control rods latched and with no immediate indication of reactor power.

Operators then

stabilized plant conditions in hot shutdown rather than to proceed immediately with plant

cooldown.

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Technical specification (TS) 3.5.1.1 requires that the SRNIs be operable when reactor power

is in the source range, the reactor trip breakers are closed, and the control rod drive system

is capable of rod withdrawal.

Although allowance is made for operation with a single SRNI

(TS table 3.5-1, action statement (4)), operation with no SRNIs is not specifically addressed.

For such conditions, TS 3.0.1 applies and requires that the reactor be shutdown within one

hour.

Beyond satisfying this requirement, the inspector considered that licensee response

to

the situation was appropriate for the following reasons:

Reactor shutdown was in progress in accordance with operations procedure 0-2.1,

"Normal Shutdown to Hot Shutdown."

At the point in this procedure that the SRNIs

energize, most control rods have been almost fully inserted into the core; with no

RCS

cooldown in progress and the remaining rods continuing to be inserted, this

provided adequate

shutdown margin.

The TS concern for having SRNI indication while the control rod drive system is

capable of rod withdrawal (that is, the reactor trip'reakers are shut) is to provide,

indication of criticality; in the case of a control rod drive system malfunction, an

inadvertent criticality due to uncontrolled rod withdrawal.

However, the control rod

position indication system was operable during all subsequent control rod

manipulations and confirmed that only inward rod motion was occurring.

RCS boration in preparation for refueling was in progress,

thereby increasing

shutdown margin.

The SRNIs provide no automatic reactor protective actions that are required for

reactor safety, and the remainder of the reactor protection system was operable

throughout the event; therefore, independent of the above-listed factors, reactor safety

was not compromised by loss of the SRNIs.

The Plant Operations Review Committee (PORC) immediately convened to review the event

and to develop guidance for subsequent plant operations.

PORC directed that no cooldown

be conducted until one SRNI was made operable.

Additionally, containment access was

prohibited and RCS boron sample frequency was increased until at least one SRNI was

available.

Troubleshooting determined that detector failure was the most likely cause of the loss of both

SRNIs.

To restore one channel to operation, an installed spare detector was connected to

SRNI channel N-32.

Post-maintenance

testing was completed satisfactorily and N-32 was

declared operable at 5:30 p.m., March 12, 1993.

Both SRNI detectors were subsequently

replaced and both SRNI channels were restored to service in their normal detector

configuration by March 17, 1993;

Investigation determined the cause of detector failure in both cases to be failure of the center

tungsten electrode.

SRNI detectors are susceptible to this type of failure due to gamma

corrosion of the tungsten electrode.

As long term corrective action, the licensee willbe

replacing the existing SRNI detectors with detectors of a new design that operate at lower

voltage and employ gold-plated tungsten electrodes

(features which reduce gamma corrosion).

Detector replacement criteria willbe established

based on time in service and number of

energizations.

Additionally, predictive measures,

such as detector operating voltage trending

and time domain reflectometry, willbe used in the future to aid in determining the need for

detector replacement.

In summary, the rod control system remained energized for a period of approximately five

minutes without indication of reactor power due to failure of both SRNI channels.

Through

observation of the event, attendance at the subsequent

PORC meeting, discussions with

personnel,

and review of plant operating conditions, the inspector determined that licensee

response

to the event had been appropriate.

Operations department personnel acted in

accordance with in-use and equipment restoration procedures

and promptly stabilized plant

conditions in hot shutdown.

Management responded immediately to evaluate the condition

and provided conservative operating guidance.

Corrective action rapidly restored source

range indication and willbe effective in minimizing the possibility of recurrence.

NRC

notification in accordance with 10 CFR 50.73 was completed within the specified time.

The

inspector had no additional concerns on this matter.

1.4

Control of Operations with Reduced Reactor Coolant System Inventory

t

Steam generator nozzle dams aOow the steam generator primary side to be maintained empty

and available for maintenance

independent of reactor vessel water level.

Installation of these

nozzle dams requires that reactor vessel water level be lowered to approximately the mid-

loop level.

The potential for loss of core cooling capability is significant during such mid-

loop operations due to reduced net positive suction head available to the residual heat

removal (RHR) pumps and unavailability of redundant components due to outage

maintenance.

The inspector assessed

that the licensee's preparations for establishing mid-loop operations

were thorough.

Operating (0) procedure 0-2.3.1, "Draining'nd Operation at Reduced

Inventory of Coolant System," was implemented to ensure redundancy, availability, and

reliability of on- and off-site sources of electrical power and coolant make-up source/paths.

'-2.3.1A,

"Containment Closure Capability in Two Hours During RCS Reduced

Inventory/Operation," was in place to accomplish rapid containment closure ifrequired.

No

operations were conducted during mid-loop operations that would cause unnecessary

perturbations to the RCS.

Independent channels and mechanisms of RCS temperature and

level indication were available.

Multiple sources of water inventory were available and

adequate RCS venting was established prior to blocking the RCS hot legs.

On March 16 and April 8, 1993, the inspector observed operations to establish RCS water

level at mid-loop. In both instances,

the inspector noted that operations department

personnel performed this significant change of plant conditions in a highly professional

manner.

The procedure to establish mid-loop level was accomplished deliberately and was

well controlled.

Operators were knowledgeable of expected indications as well as of

symptoms of potential problems.

No significant deviations from expected system response

occurred and no operational deficiencies were noted.

Quality Assurance personnel

conscientiously monitored the conduct of these mid-loop operations.

Throughout the outage, daily safety assessments

were performed by the outage coordinators

to evaluate the availability of reactivity control systems, core cooling systems, electrical

supplies, containment integrity, and reactor coolant inventory make-up paths during system

configuration changes.

The inspectors found that this graded approach provided an effective

management

and communication tool to assure that plant safety was not compromised,

particularly during mid-loop operations.

2.0

RADIOLOGICALCONTROLS (71707)

2.1

Routine Observations

The inspectors periodically confirmed that radiation work permits were effectively

implemented, dosimetry was correctly worn in controlled areas and dosimeter ieadings were

accurately recorded,

access to high radiation areas was adequately controlled, survey

information was kept current, and postings and labeling were in compliance with regulatory

requirements.

Through observations of ongoing activities and discussions with plant

personnel,

the inspectors concluded that radiological controls were appropriately

implemented.

2.2

Outage ALARAReview

During the 1993 outage, significant improvements were noted in reducing collective radiation

exposure and personnel contaminations from those of past outages.

For the current outage,

the total dosage was 155 person-rem,

compared to 228 person-rem and 296 person-rem for

the 1992 and 1991 outages,

respectively.

Personnel contaminations for this outage numbered

95 cases,

versus 200 cases and 260 cases for the 1992 and 1991 outages, respectively.

The

inspectors determined that these improvements have resulted from aggressive radiological

controls management

that has effectively integrated engineering support to resolving ALARA

issues and has assured

that routine health physics practices were carried out.

Throughout the

outage, the inspectors observed good adherence

to Radiation Work Permits and related

procedures,

detailed pre-job ALARAbriefings, good housekeeping

practices, and prompt

identification of changing radiological conditions with timely implementations of control

measures.

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3.0

MAIN'IENANCE/SURVEILLANCE(62703, 61726)

3.1

Outage Maintenance Sinnmary

Significant maintenance

conducted during the 1993 refueling outage included:

reactor

refueling, including a complete core offload to support service water system and reactor

coolant pump maintenance;

steam generator U-tube inspections and repairs; replacement of

all containment recirculation fan cooler air coolers; "B" reactor coolant pump shaft seal

inspection; reactor coolant pump diffuser inspections; service water system valve

refurbishment; preseparator

drain tank replacement;

and number two low pressure turbine

inspection.

3.2

Service Water System Refurbishment

3.2.1

Failure of Ualve's 4669 and 4738

During the 1993 Refueling/Maintenance

Outage, the licensee implemented phase V of the

Ginna Station Valve and Valve Actuator Improvement Program that began in 1989.

Past

refurbishment/replacement

efforts have been directed at priority valves in the reactor coolant

system (RCS), residual heat removal (RHR) system, safety injection (Si) system, component

cooling water (CCW) system, auxiliary feedwater (~V), and various balance-of-plant

systems.

During the current phase,

service water system (SWS) valves and valve actuators

considered vital to plant safety and reliability were inspected and refurbished or replaced.

Particular attention was directed at inspecting and replacing several 3" and 4" gate valves,

Crane Model No. 101XU, that are no longer manufactured.

An upgraded design, the Crane

Model No. 47 1/2XU, was selected

as the replacement style.

Additional information on the

as-found valve conditions, the licensee's root cause analysis, and resulting SWS

Refurbishment Program scope increase are contained in Inspection Report 50-244/93-08.

During the SWS valve inspections,

several valves were found to be significantly corroded.

Two valves (4669 and 4738) were found to have the valve stems separated

from the valve

discs, with the discs in the flow stream (functionally, fully closed) and the valve stems fully

retracted (and from external observation,

assumed

to be fully locked open).

Valve 4669 in

combination with valve 4760 provides cross connection between the "A" and "B" headers

allowing the Emergency Diesel Generators

(EDGs) to be supplied from either header, in

addition to the normal redundant flow paths.

Valve 4738 is the isolation valve from the "B"

service water header to the safety injection pump thrust bearing coolers, the SI, RHR, and

charging pump mom coolers, and the containment penetration cooler.

Since the spring of 1988, the service water system normal configuration has been cross

connected between the two loops.

This cross connection, in effect, provides for one single

loop where service water pumps from either header can provide all service water loads.

Even with the two affected valves closed, the loops are connected by a normally open 14"

line. Therefore, with this normal lineup, system redundancy was maintained in that any

SWS pump could provide all service water loads even with the valves 4669 and 4738 closed.

Prior to August 1992, during brief periods (typically less than two hours) for monthly service

water pump testing (PT-2.7), the loops were separated.

Redundant cooling to the EDGs was

not affected by this temporary re-alignment.

However, during testing of the "A" loop, valve

4739 (SWS loop "A" isolation valve, that is, the counterpart to valve 4738), was closed.

During these periods, with 4738 unknowingly failed closed, service water would have been

isolated to the room coolers and SI thrust bearing coolers.

Being out of a normal alignment

would only have occurred during performance of surveillance testing and the loops would

have been manually realigned to the normal cross connected configuration, providing

redundant flows, upon the initiation of a significant event during these short periods.

Because of the short duration of the surveillance testing, the low cooling loads (the SI pumps

are not operated, i.e., tested, coincident with the PT-2.7 test) from affected components,

and

the absence'f in-line instrumentation to detect reduced SWS flow to these components,

no

abnormal indications from the isolated loads were readily evident.

This practice of cycling valves 4738 and 4739 during surveillance testing was discontinued in

August 1992.

At that time, the monthly performance of PT-2.7 was replaced with a

quarterly surveillance test, PT-2.7.1, that does not isolate an individual loop to the thrust

bearing coolers and associated

loads.

This change was made to conform with the guidance

contained in Section XI of the 1986 ASME Boiler and Pressure

Vessel (BPV) Code, in that

quarterly inservice testing is an appropriate test frequency.

Ginna Technical Specifications

do not directly address

surveillance testing of SWS components, but refers to the inservice

test requirements

specified in the ASME BPV Code Section XI, and applicable addenda.

Upon identifying the failed valves, site management,

operations personnel,

and site/corporate

engineering staffs developed a comprehensive corrective action plan.

From plant incident

report A-2S.1, No.93-040, Corrective Action Report 2077 was initiated, tasking the

responsible departments within RG&E to evaluate the safety significance, reportability, and

the extent to which the program scope would be increased to verify SWS valve positions and

stem/disc conditions of other Crane 101XU valves.

Additionally, the adequacy of SWS flow

to SI thrust bearing coolers would be evaluated.

The inspectors reviewed selected activities to ascertain whether the SI pump thrust bearing

coolers were receiving unobstructed

SWS flow. During annual outages,

each SI pump is

inspected which includes a thrust bearing and lubrication inspection, in accordance. with M-

11.12.1, "Safety Injection Pump, Mechanical Inspection."

Based on discussions with

mechanical maintenance personnel responsible for performing this activity, the inspectors

determined that never in plant life have thrust bearings been degraded

as a result of

insufficient cooling.

Inspections of bearing lubricating oil in the current outage did not

indicate abnormal cooling.

Additionally, quarterly SI pump surveillance test results did not

identify elevated vibrations, that would be indicative of bearing degradation.

On April 14,

1993, the inspectors observed performance of M-11.12.2, "0'1-A, 1-B, and 1-C Safety

0

Injection Pump Outboard Jacketed Bearing Service Water Cooling System Maintenance."

From observing this activity, the inspectors confirmed that SWS flow was unobstructed

through each of the thrust bearing coolers.

To accurately quantify the flow rate, the site engineering staff attempted to measure SI

bearing cooler flow using an external ultrasonic flow monitor, however, the results were

unreliable.

As a result of these unsuccessful efforts, the licensee has replaced the elbow

connections on the inlet and discharge of each bearing cooler with tee fittings, to permit

installation of pressure or flow instrumentation at some later date.

A procedure to establish

this flow monitoring is under development.

Inspection endings

Through close observation of site management's

response to identifying failed valves 4669

and 4738 in the SWS, the inspector concluded that prompt, comprehensive

actions were

taken to address

the problem and to evaluate the effect that the as-found condition had on

plant safety.

Responsible departments within the RG&E organization were expeditiously

tasked with: evaluating the reportability per 10 CFR 50.72, 50.73, and 10 CFR 21;

assessing

the safety impact of the valves found not in their expected positions; identifying all

Crane 101XU valves in use at the site and confirming that those used in safety-related

SWS

applications were intact and in the proper position; and evaluating the preventive maintenance

program scope and schedule for SWS valves.

Through review of safety evaluations

addressing

the impact of the failed valves on plant safety, the inspector concluded that with

the normal single SWS loop configuration and the ability to supply loads from either header,

there was no impact upon the ability to mitigate the effects of an accident with valves 4669

and 4738 closed.

The inspector found that the licensee had the ability to quickly realign the

SWS during the performance of PT-2.1 ifan incident occurred requiring SI pump thrust

bearing cooling, during past SWS monthly testing.

3.2.2

Desludging of Screen House

During this outage period, the licensee inspected underwater components in the screen house.

The screen house is a building containing the traveling screens,

circulating water and service

water pumps, and associated

bays.

The inspection, performed annually in accordance with

M-92, "Underwater Inspection of Mechanical Equipment and Structures in the Screen

House," revealed a buildup of zebra mussels on stationary structures and an excessive

amount of silt deposited.

Walls were subsequently

scraped and the loose material (about 31

tons) was removed from the fore bay, circulating water bay, and service water bay by

vacuuming.

Video inspections recorded the as-found and as-left condition.

Through observing the work in progress,

the inspector concluded that the licensee is taking

prudent actions to assure the continued reliability of the service water system.

3.3

Surveillance Observations

Inspectors 'observed portions of surveillances to verify proper calibration of test

instrumentation,

use of approved procedures,

performance of work by qualified personnel,

conformance to limiting conditions for operation (LCOs), and correct system restoration

following testing.

The following surveillances were observed:

~

Refueling Shutdown Surveillance Procedure (RSSP)-2.2, "Diesel Generator Load and

Safeguard

Sequence Test," revision 44, effective date March 26, 1993, performed

April20, 1993

~

PT-7, "ISI System Leakage Test, Reactor Coolant System," revision 43, effective date

May 7, 1992, performed April22, 1993

The inspector determined through observing this testing that Results and Test personnel

adhered to procedures,

equipment operating parameters

met acceptance

criteria, and

redundant equipment was available for emergency operation.

3.4

Containment Integrated Leak Rate Test

Documents Reviewed

RSSP-6.0,

"Containment Integrated Leakage Rate Test," Revision 19, dated 4/2/93

RSSP-6.1,

"Integrated Leakage Rate Test Valve Alignment," Revision 17, dated

4/7/93

RSSP-6.2,

Pressurization Monitoring of Penetration Free Volumes During CILRT,"

Revision 13, dated 4/2/93

RSSP-6.4,

"Integrated Leak Rate Test Instrument Integrity Check," Revision 3, dated

4/2/93

RSSP-6.5,

"Integrated Leak Rate Test Containment Structural Inspection," Revision

4, dated 3/26/93

RSSP-6.7, "ILRTInstrumentation Preparation," Revision 3, dated 4/5/93

Scope of Review

The inspectors reviewed the Containment Integrated Leakage Rate Test (CILRT) procedures

listed above and supporting procedure change notices for technical adequacy and to determine

compliance with Technical Specification 4.4 and 10 CFR 50, Appendix J.

The inspectors

noted that test procedures

were in general conformance with the guidance of ANSI-ANS

10

56.8-1987, "American National Standard Containment System Leakage Testing

Requirements."

The inspectors discussed

various aspects of the CILRT with licensee

representatives

and contract personnel conducting the test.

The inspectors also reviewed the

test procedure initial conditions and prerequisites,

pressurization equipment setup, monitoring

instrumentation placement, data recording and processing

methods, and test personnel

training.

The licensee utilized 'a contractor, TER Services, Inc., to perform the data

collection and processing.

Results

The inspectors witnessed various phases of the CILRT. Initial alignments were completed

and pressurization

commenced at 2:41 a.m., April 12, 1993.

During the pressure

stabilization phase,

leakage paths were identified in the flange connections to the "C" and

"D" containment recirculation fan cooling system that were installed during this outage.

Subsequently,

containment was depressurized,

blind flanges were installed on the suction.and

discharge piping the "C" and "D" coolers, and the piping hydrostatically tested for leaks.

At

4:55 p.m. on April 13, 1993, containment pressurization

was reinitiated.

Upon stabilizing

test pressure at -35 psig, the 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> leakage test period commenced at 2:15 p.m. on April

14, 1993 and ended at 2:15 p.m. on April 15, 1993.

The observed leakage rate was verified

to be less than the 0.75 L, (L, = maximum allowable lmPage rate).

Between 3:30 p.m. and

7:30 p.m. on April 15, 1993, the leakage verification test was conducted.

The results of the

controlled leakage test were satisfactory and the containment was depressurized

in accordance

with the test procedure.

No discrepancies

were identified in the licensee's conduct of the test, review of test

procedures,

or review of results achieved.

In accordance with T.S. 4.4.1.6, RGkE will

submit a technical report for NRC staff review summarizing the conduct of the test and test

results.

3.5

Fire Involving "D" Containment Recirculation Fan Cooler

Overview

On March 19, 1993 at 11:35 a.m., a fire occurred in the roughing filter bank to the "D"

containment recirculation fan cooling system (CRFCS).

The fire resulted when a hot slag

particle from a nearby cutting job made contact with badly loaded roughing filter media,

igniting the filters.

The fire occurred despite the implementation of various fire prevention measures prior to

beginning the job that included a job site area inspection for removal/covering combustible

materials, shielding of the roughing filters with flameproof covers, and posting a fire watch.

11

The fire was promptly reported. by the fire watch to the control room.

Subsequently,

the fire

brigade was dispatched to the scene and quickly extinguished the fire in less than ten

minutes.

Upon extinguishing the fire, a reflash watch was stationed in the area until 5:30

p.m.

No further incidents occurred.

As administrative follow-up to investigate the cause for the fire and identify corrective

actions needed, a plant incident report (A-25.1-93-28) and a Corrective Action Report (CAR-

2076), were developed.

Immediate actions included visually inspecting and meggaring the

"D" CRFCS fan motor to identify possible damaged components,

and inspecting associated

cabling, filter housing and CRFCS structural supports.

Damage was found to be limited to

the "D" CRFCS roughing filters. Since the condition of the "D" bank high efficiency

particulate absolute (HEPA) filters, downstream of the roughing filters, was not readily

apparent,

these filters were scheduled for replacement.

Subsequent

inspection of roughing

filters in unaffected CRFCS units found them to be also badly loaded, with an oily surface

coating of unknown origin.

The roughing filters (sometimes referred to as prefilters, moisture separator pads, or

demisters) are designed to enhance air quality by removing moisture and particulates

entrained in the air stream before it enters the HEPA filters in the CRFCS units. A set of 25

moisture separator pads, in a 5 x 5 array, are located downstream of a single set of inlet

louvers, the CRFCS coiling coils, and a second double set of moisture separating louvers

composing the moisture separator section inside each unit. By design, the HEPA filters

perform their safety-related function even in the presence of entrained water, and as such the

roughing filters do not perform a safety-related function and were not loaded sufficiently, in

this case, to significantly reduce flow.

The inspector reviewed the circumstances

that lead to the roughing filter fire and the

licensee's response.

In performing this review, the inspector evaluated the adequacy of the

licensee's fire protection program by verifying that appropriate preventive measures

were in

place and implemented as required by established procedures.

Through inspection of the "D"

CRFCS roughing filter area, examination of post-fire photographs of the affected area,

review of relevant documentation,

and discussions with license representatives,

the inspector

concluded'hat pre-job fire prevention measures

were implemented as required.

Specificaily,

in accordance with Administrative (A) procedure A-905, "Open Flame, Welding, and

Grinding Permit,

revision 18, Fire Permit No.93-147 was issued prior to beginning the

cutting job to assure that a pre-job area inspection was performed identifying combustible

materials, covering nearby combustibles, and stationing a fire watch with a fire extinguisher

at the job site.

The inspector also concluded that the appropriate contingency measures

as specified in A-

202, "The Fire Protection Program and Ginna Station Staff Responsibilities for Fire

Protection," revision 6, were in place to quickly extinguish the fire. The inspector further

12

concluded that despite the implementation of the various fire prevention measures,

the

licensee was not aware of the flammable condition of the roughing filters. From a review of

licensee procedures,

Maintenance (M)-1306, "Ginna Station Material Condition Inspection

Program," revision 6, no routine inspections were performed of the roughing filter condition.

Additionally, the roughing filters were not routinely visually inspected as part of periodic test

(PT)-47.5, "Containment Fan Recirculation Unit-HEPA Filtration System Efficiency Test,"

revision 3, or other relevant ventilation procedure.

ANSUASME N510-1980, "Testing of

Nuclear AirCleaning Systems," on which PT-47.5 and other relevant ventilation testing

procedures

are based,

recommends

that prefilters be inspected for dirt loading to assure

optimum air flow to the HEPA/charcoal sections of the filtration unit. From a review of

relevant documentation,

the installed media have not been changed or inspected in 5 to 8

years; an exact time period could not be determined.

Based on this review, the inspector concluded that the licensee's failure to periodically assess

the condition of the roughing filters represented

a maintenance program shortcoming.

As

such, this represents

a violation of Appendix B of 10 CFR 50, Criterion XVI, "Corrective

Actions," which states in part... "Measures shall be established

to assure that conditions

adverse to quality, such as failures, ...defective material and equipment, and

nonconformances

are promptly identified and corrected."

(50-244/93-06-01)

As immediate corrective action, the licensee replaced the roughing filters in all CRFCS units.

Additionally, samples of the loaded media were removed for analysis by an independent

laboratory to determine the chemical nature of the contaminants and assess

the effect that

combustion products may have on electrical equipment.

3.6

Containment Material Condition

The inspectors conducted several routine safety inspections in the containment building. The

inspectors

assessed

that material conditions within the containment building were generally

good.

A number of minor material discrepancies

were noted and passed on to the licensee

for corrective action.

Although none of these items were safety significant, the inspector

considered that existing licensee inspection and self-monitoring programs,

such as system

walkdowns, zone inspections,

safety inspections, and supervisory tours, could be more

effective in self-identification of such deficiencies.

The inspector considered

that licensee

corrective actions for identified deficiencies were appropriate and concluded that material

conditions in containment were adequate for the resumption of critical operations.

4:0

SECURI'IY (71707)

4.1

Routine Observations

During this inspection period, the resident inspectors verified that x-ray machines and metal

and explosive detectors were operable, protected area and vital area barriers were well

maintained, personnel'were properly badged for unescorted or escorted

access,

and

compensatory

measures

were implemented when necessary.

No unacceptable

conditions

were identified.

13

4.2

Storm Damage to Site Perimeter

Fence-'he

blizzard that occurred on March 13-14, 1993, caused damage to portions of the site

perimeter fence.

High winds, sub-freezing temperatures,

and water spray from Lake Ontario

formed ice on the fences along the lake front to such an extent that they buckled under the

prevailing forces.

Approximately 300 feet of the outer fence and 100 feet of the inner fence

were so damaged.

Associated damage to the intrusion detection system affected much of the

remainder of the site perimeter.

The licensee responded by implementing full compensatory

measures for affected portions of the site perimeter.

Repairs to the intrusion detection

system and perimeter fence were completed.

The inspector determined that compensatory

measures

established for loss of the intrusion

detection system and degradation of the perimeter fence were adequate and timely. Repair of

storm damage was promptly completed.

The inspector had no additional concerns on this

matter.

5.0

EMERGENCY PREPAID)NESS (71707)

5.1

Siren Test

On April 6, 1993, the licensee's Emergency Preparedness

staff conducted the annual siren

activation test.

The test requires that all of the sirens be actuated and run uninterrupted for a

three minute period.

Results of this test were that three of 96 sirens failed to perform

properly.

With 97 percent of the sirens fully operating, the licensee met the FEMA

acceptance

guidelines of greater than 90 percent operable.

The three failed sirens were

promptly repaired.

From a review of the test results, and through discussions with licensee representatives,

the

inspectors determined that, in addition to successfully performing the annual siren activation

test, quarterly growl tests and monthly silent tests have been performed as specified by

NUREG-0654, "Criteria for Preparation and Evaluation of Radiological Emergency Response

Plans and Preparedness

in Support of Nuclear Power, Plants."

6.0

ENGINFXMNG/TECHNICALSUPPORT (71707, 92701)

6.1

Mssing Reactor Coohnt hnp Diffnser Bolt

During the 1990 refueling outage, part of a bolt was discovered inside of the reactor vessel,

on top of the lower core plate.

Through examiiiation of the bolt, the licensee determined that

it was the type used to secure the reactor coolant pump diffuser adapters.

Although the

source of the bolt could not be established without examining the RCP diffuser adaptors,

the

licensee pursued the worst-case assumption that it had backed out of one of the RCPs (as

opposed to being a spare that had inadvertently been left in the RCS following RCP

maintenance in an earlier outage).

Since any examination of RCP internals would require

'

14

extensive preparation,

the licensee demonstrated,

through analysis, that continued operation

with such a condition could be justified; in parallel, planning was initiated to conduct a'direct

inspection of the RCP diffuser adaptors.

Actions were coordinated and tracked under

corrective action report (CAR) 2001, "Foreign Material Found in the Reactor Vessel."

Analysis conducted by the pump vendor demonstrated

that an RCP would operate

satisfactorily under all conditions with one of the 16 diffuser adaptor bolts missing.

RG&E

technical engineering evaluated the risks associated

with continued plant operation with the

remaining portion of the bolt in question not accounted for, and determined that this did not

create an unreviewed safety question.

Running RCPs exhibited normal operating

characteristics.

Based on these findings, the licensee determined that continued plant

operation prior to RCP inspection was acceptable.

The licensee determined that the RCP diffuser adaptor inspections should be conducted by

remote video camera.

To facilitate access,

the inspection would be done following a

complete reactor core offload. In that a core offload would be required to support service

water system maintenance during the 1993 refueling outage, and contingent on continued

normal operation of the RCPs, the video inspection was scheduled for 1993.

In preparation for the RCP inspection, a full size mock-up of the applicable RCS piping was

constructed to test the inspection equipment.

The equipment consisted of a video camera

mounted on a fiber glass push rod; insertion into the RCS was made through a snorkel

adaptor at the steam generator cold leg manway.

Operator training was conducted using this

mock-up prior to the actual inspection.

No significant procedural or equipment deficiencies

were identified as a result of the mock-up training.

The RCP diffuser adaptor inspections commenced

on March 22, 1993, in accordance with

EM-774, "Remote Visual Examination of Reactor Coolant Pump Diffuser Adaptor Ring and

Bolting," revision 0, effective date March 5, 1993., All 16 bolts were found to be in place

on the "B" RCP, and one was found to be missing from the "A" RCP. Allremaining

installed diffuser adaptor bolts were determined to be fully inserted.

No other material

deficiencies were noted and the inspection activities concluded on March 24, 1993.

The inspector reviewed completed work order 9320011, "Perform Remote Video Inspection,

"A"and "B" Reactor Coolant Pumps," and noted no discrelencies.

The inspector concluded

that the activities to inspect the RCP diffuser adaptors had been well planned and successfully

implemented.

CAR 2001 was an effective tool in progressing a three-year old deficiency

through development and implementation of corrective action.

Final corrective action to

replace the missing bolt willbe accomplished when the "A" RCP requires future

maintenance.

The inspector had no additional concerns on this matter.

'CI

15

6.2

Containment Recirculation Fan Cooler Replacements

The reactor containment building contains four recirculation fan cooler units.

The safety

function of these units is to condense

steam released during either a loss of coolant or steam

line break accident, and thereby limitthe time that the containment building is subject to high

internal pressure.

Each unit contains three primary heat exchangers

that are supplied by the

service water system.

Containment recirculation fan cooler heat exchangers

had been scheduled for replacement in

1995; however, in light of the increasing number of service water leaks, the licensee

accelerated

this schedule,

to replace these heat exchangers

during the 1993 refueling outage.

The replacement primary heat exchangers

are of an advanced design, with greater

erosion/corrosion

resistance

and a physical configuration which willprovide both better

access for maintenance

and greater fiexibilityin selecting repair techniques.

The inspector observed portions of production work to install the containment recirculation

fan cooler heat exchangers.

The licensee utilized four 10-hour crew shifts to provide around-

the-clock support for this maintenance effort.

Continuous engineering support was also

available, with two licensee field engineers,

the licensee liaison engineer,

and one contractor

engineer covering the work. Through discussions with personnel, observation of work in-

progress,

and completion of an aggressive

schedule,

the inspector concluded that good

engineering support was provided for the containment recirculation fan cooler heat exchanger

replacement.

The inspector reviewed station modification procedures SM-5275.1 and .2, "Units "A" and

"B" ("C" and "D") Containment Recirculation Fan Cooler Replacements."

These procedures

served as the coordinating documents for accomplishment of production work, inspection,

testing, and acceptance of the recirculation fan cooler modifications.

No deficiencies were

noted.

Modification testing was completed at the close of the inspection period, but the

results were still under licensee review.

The inspector had no further questions at this time.

6.3

Safety IILiection System Accumulator Drain Vabes Seismic Qualification

During a routine safety inspection in containment, the inspector noted that the safety injection

system accumulator drain valves, AOV-844A and B, did not appear to be properly

supported.

Specifically, the valves each appeared to be missing a support bracket between

the air operator and the base of the SI accumulator.

The inspector had observed that

similarly sized air operated valves in containment had such supports, and a ground-off area

on the base of each of the SI accumulators

suggested

that such brackets had, at one time,

been installed.

The inspector noted that the AOV-844A and B air operators could be set into

cyclic, side-to-side motion about the axis of the drain line, with only slight exertion.

The

inspector questioned the seismic qualification of these valves.

'

16

The licensee reported 'that- diagonal braces had, in fact, been installed on valves AOV-844A

and B in the past, but that they had been removed about 10 years ago as a result of analyses

conducted during the seismic upgrade program.

The inspector 'was in the process of

reviewing the applicable piping stress analysis reports at the close of the inspection period.

The inspector had no further questions at this time.

7.0

SAFETY ASSESSMFAT/QUALITYVXRIHCATION(9071?, 90713, 92701,

40500)

7.1

Periodic Reports

Periodic reports submitted by the licensee pursuant to Technical Specification 6.9.1 were

reviewed.

Inspectors verifie that the reports contained information required by the NRC,

that test results and/or supporting information were consistent with design predictions and

performance specifications,

and that reported information was accurate.

The following

reports were reviewed:

No

Licensee Event Reports

Monthly Operating Reports for February and March 1993

unacceptable

conditions were identified.

A licensee event report (LER) submitted to the NRC was reviewed to determine whether

details were clearly reported, causes were properly identified, and corrective actions were

appropriate.

The inspectors also assessed

whether potential safety consequences

were

properly evaluated, generic implications were indicated, events warranted onsite follow-up,

and applicable requirements of 10 CFR 50.72 were met.

The following LER was reviewed (Note: date indicated is event date):

~

93-001, Loss of Source Range Detector Indication During Energization Due To Faulty

Detectors

The inspector concluded that the LER was accurate and met regulatory requirements.

No

unacceptable conditions were identified.

8.0

ADMINISTRATIVE(71707, 30702, 94600)

8.1

Backshift and Deep Backshift Inspection

During this inspection period, backshift inspections were conducted on the following dates:

March 16 and April 8, 1993.

Deep backshift inspections were conducted on the following

dates:

March 13, 20, 21, 27, April 3 and 22, 1993.

17

8.2

Exit Meetings

At periodic intervals and at the conclusion of the inspection, meetings were held with senior

station management to discuss the scope and findings of inspections.

The exit meeting for

inspection report 50-244/9342 (inservice inspection program, conducted March 23-26, 1993)

was held by Mr. Pralmh Patnaik on March 26, 1993.

The exit meeting for inspection report

50-244/93-07 (health physics, conducted April 12-16, 1993) was held by Mr. James Noggle

on April 16, 1993.

The exit meeting for inspection report 50-244/93-08 (service water

system outage maintenance,

conducted April 12-16, 1993) was held by Mr. Harold Gregg on

April 16, )993.

The exit meeting for inspection report 50-244/9346 was held on April22,

1993.