ML17158C191

Insp Repts 50-387/97-03 & 50-388/97-03 on 970408-0519. Violations Noted.Major Areas Inspected:Operations, Engineering,Maint & Plant Support Activities Re Security Plans

ML17158C191
Person / Time
Site:	Susquehanna
Issue date:	06/23/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML17158C189	List: ML17158C188 ML17158C190 ML17158C191
References
50-387-97-03, 50-387-97-3, 50-388-97-03, 50-388-97-3, NUDOCS 9706300239
Download: ML17158C191 (58)

See also: IR 05000387/1997003

Text

U. S. NUCLEAR REGULATORY COMMISSION

REGION

I

Docket Nos:

License Nos:

50-387, 50-388

NPF-14, NPF-22

Report No.

50-387/97-03, 50-388/97-03

Licensee:

Pennsylvania

Power and Light Company

2 North Ninth Street

Allentown, Pennsylvania

19101

Facility:

Susquehanna

Steam Electric Station (SSES)

Location:

P.O. Box 35

Berwick, PA 18603-0035

Dates:

April 8, 1997 through May 19, 1997

Inspectors:

K. Jenison,

Senior Resident Inspector

B. McDermott, Resident Inspector

B. Welling, Resident Inspector, Peach Bottom

Approved by:

Richard R. Keimig, Chief

Projects Branch 4

Division of Reactor Projects

9706300239

970623

PDR

ADOCK 05000387

8

PDR

EXECUTIVE SUMMARY

Susquehanna

Steam Electric Station, Units

1 5 2

NRC Inspection Report 50-387/97-03, 50-388/97-03

This integrated inspection included aspects

of licensee operations,

engineering,

maintenance,

and plant support activities.

The report covers

a 6-week period of resident

inspection.

~Oerations

Unit 2 refueling and inspection outage activities were inspected to determine proper

movement and placement of fuel assemblies.

Two minor events occurred during

the fuel movement activities.

The two events were:

1) an impact between

a single

blade guide and a reactor vessel flange cover and 2) the failure of a safety chain clip

resulting in a portion of the clip being dropped into the fuel pool.

Each of the

events was responded to aggressively by the licensee.

The licensee also identified

and resolved generic issues with human performance related to refueling activities.

Despite the two events and the refueling related issues, refueling activities were

well supervised

and were conducted

in a safe and conservative

manner.

On May 2, 1997, with Unit 2 in cold shutdown,

a SPING alarm was received.

The

operators responded

appropriately, the Shift Supervisor made conservative decisions

which ensured the safety of the unit and the public.

The inspector determined that

one of the plant alarm response

procedures

was inadequate

in that it did not contain

reasonable

validation criteria, and it did not agree with and delayed entry into the

Emergency

Plan.

Approximately seven lighted control room annunciators

correspond to chronic

conditions in SSES ventilation systems.

The licensee currently has a corrective

action plan in place to correct these conditions.

Previous action plans were

determined to be ineffective in that there was very little progress

made from

approximately 1988 until the present to correct these specific alarmed annunciators.

There is no present impact on Technical Specification operability requirements,

compliance with NRC requirements

or compliance with selected

SSES design

standards.

~

The licensee has directed considerable efforts toward the resolution of plant

conditions that are documented

in condition reports (CRs).

Upper levels of PPKL

management,

including the Vice President - Nuclear Operations and the General

Manager - Nuclear Engineering, were observed routinely addressing the resolution of

individual CRs.

The licensee demonstrated

during this inspection report period. that

they were capable of addressing

and resolving CRs adequately

in the short term.

Maintenance

In general, maintenance

activities observed

during this report period were

adequately controlled and performed in accordance

with station procedures.

In the

case of the spent fuel pool temperature

monitor maintenance

activities, the work

was well performed and controlled.

A Unit 2 'C'eactor feed pump turbine bearing was replaced for corrective

maintenance.

The replacement

bearing did not meet the vendor specified

clearances

and the System Engineer unilaterally revised the procedural clearance

limits in the field. The System Engineer did not effectively communicate to the Unit

Supervisor

(US) that he had not met the bearing clearance limits when he was

requesting

and performing a post maintenance

acceptance

test.

The test was

terminated by the US due to high temperature.

The bearing was brought into

tolerance and a successful

post maintenance

acceptance

test was performed.

The

reactor feed pump is not safety related and its failure would result in bounded plant

transients.

There was no impact on the safe operation of the unit. Therefore, no

violations of NRC requirements

occurred.

The monthly surveillance test for the 'C'DG was generally performed according to

approved surveillance test procedures.

However, the inspector observed

a failure to

comply with alarm response

procedures for a known equipment condition

associated

with an oscillating jacket water standpipe

level indication.

This failure to

follow procedures

is being treated as a non-cited violation.

The performance of the quarterly surveillance test for the Division I core spray

system was generally well-controlled.

However, the inspector concluded that the

practice of venting the core spray pumps immediately prior to starting them for their

quarterly surveillance test was a preconditioning action.

The failure to perform the

core spray surveillance test under suitably controlled conditions is considered

a

violation of NRC requirements.

In addition, the inspector observed that the

methodology for performing independent

verifications within the test procedure was

weak and did not meet licensee expectations.

Two maintenance

activities associated

with restoration from the Unit 2 refueling

outage had the potential for personnel injury. Both issues were adequately

resolved

by the licensee.

No personnel injury occurred, there was no impact on safety

related equipment,

and no violations of NRC requirements

occurred.

During fuel movement activities, a fuel assembly was suspended

(less than one

foot) above the reactor vessel fuel support piece without the ability to raise or lower

it through normal means.

Maintenance activities were initiated on the Unit 2 rod

control system to resolve this condition.

The inspector observed that the

maintenance

was performed using an "information only" SSES Training Department

drawing that was not authorized for use.

This failure to use controlled drawings

was characterized

as a non-cited violation.

As a result of weaknesses

identified in the under vessel maintenance

activities

during the Unit 1 refueling outage,

and condition reports written by the licensee, the

inspector observed/reviewed

under vessel activities during the Unit 2 refueling

outage.

The licensee issued and resolved

a number of condition reports and took

adequate

corrective actions.

No violations of NRC requirements

were identified.

Maintenance

on safety-related

instruments performed by the Emergent Work Action

Crew (EWAC) was observed to be commensurate

with the scope and complexity

proscribed by the administrative procedure for minor maintenance.

Appropriate

documentation

and communication practices were noted.

PPRL's efforts to remove foreign materials from the Unit 2 containment following

the refueling outage were very good.

The final containment walkdown and

inspection by the Operations

and Maintenance department managers was viewed as

strength.

Based on the areas reviewed during the inspector's containment

walkdown, PPSL was effective in restoring equipment (hatches,

insulation, hangars,

etc.) to the condition required for plant operation.

PPS.L's efforts to reduce foreign debris in the Unit 2 suppression

pool during the

Spring 1997 refueling outage were through.

The compensatory

actions requested

by the NRC in'conjunction with deferral of the final resolution of Bulletin 96-03 were

implemented by PPSL.

The Unit 2 suppression

pool cleanout results were

consistent with the assumptions

contained in PP&L's existing operability evaluation

that addressed

suction strainer clogging.

Reactor building ventilation system (RBVS) back draft isolation dampers

are safety

related components within the non-safety related system.

Although the RBVS is

addressed

by the maintenance

rule program at SSES, the function of the BDIDs was

not included in the licensee's

scoping document.

A determination of the

significance of not including the BDIDs in the RBVS maintenance

rule scope will be

tracked as an unresolved item.

~

The Nuclear Assessment

Services (NAS) audit of the Test Control Program provided

an adequate

review of post maintenance

testing as required by the Operational

Quality Assurance

Manual.

However, the inspector considered the audit sample

size (15 packages)

small relative to the number of safety related work authorizations

processed

in a two year period (22,000 packages)

~ The lack of an NAS audit in the

minor maintenance

area was considered

a weakness

in testing program oversight

required by the Quality Assurance

Manual.

~En ineerin

A Unit 2 cycle 9 core reload hydraulic performance evaluation (separate from

thermohydraulic performance) for the Atrium 10 fuel was reviewed and determined

to be adequately

bounded

by analysis.

After a safety related 4160 volt breaker failed to operate when required, the

licensee identified a possible new failure mode involving a personnel protection

device referred to as a tripper lever.

Although problems with tripper levers were

previously identified in NRC Information Notice (IN) 96-50, this failure was different

since it occurred after the breaker had been racked in and cycled.

The licensee's

response

to the IN was very conservative

and aggressive.

The involvement of first

and second

line engineering management

in this issue was laudable.

The generic

aspects of the issue have been forwarded to NRR for review.

~

The licensee maintained the capability to utilize emergency alternate water sources

identified in the SSES Emergency Plan and discussed

in the FSAR.

No violations of

NRC requirements

were identified.

~

In four instances,

PP&L failed to perform safety evaluations prior to making changes

to the facility as required by 10 CFR 50.59.

The following examples were

identified:

1) blocking open doors for rooms with high energy line break protective

features,

2) increasing the normal 250 Vdc system float voltage, 3) installing

temporary test equipment on the emergency diesel generators,

and 4) cross

connecting the normal and backup fire protection systems.

This was identified as a

violation of NRC requirements.

Plant Su

ort

The licensee met the requirements of its security plan with respect to vital area door

access.

The licensee's

surveillance activities were carefully and well performed.

Some aspects of general employee training could be improved to make the

operation of the door alarm system clearer to plant employees.

No violations of

NRC requirements

were identified.

On May 5, 1997, the licensee cross connected

the normal and backup fire

protection systems

and the systems remained in the cross connected

condition at

the end of this report period.

This alignment constitutes

a change to the normal fire

protection system that is described

in the FSAR and TS 3/4 7.6.

This change was

not preceded

by an evaluation to determine if an unreviewed safety question would

result from the cross tie of the two fire protection systems

and is example number

(4) of the 10 CFR 50.59 violation.

TABLE OF CONTENTS

I ~ Operations

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Conduct. of Operations...................

01.1

Unit 2 Refueling Activities

01.2

Unit 2 Turbine Building System Particulate

(SPING) Refueling Activities

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Iodine Noble Gas

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Operational Status of Facilities and Equipment .......

02.1

Control Room Annunciators Operability ..

02.2

Review of Licensee Condition Reports ...

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Miscellaneous Operations Issues .. ~.........

08.1

Review of Licensee Event Reports......

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I. Maintenance

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Conduct of Maintenance

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M1.1

Planned Maintenance Activity Review .....

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M1.2

Reactor Feed Pump Repair ActivityReview

M1.3

Surveillance Test ActivitySample Reviews

M1A Emergency Diesel Generator Monthly Surveillance Test ..

M1.5

Unit 1 Core Spray System Quarterly Surveillance Test

M1.6

Maintenance Activities that Resulted in Potential Industrial

Safety Situations ........

M1.7

Maintenance Activities in Support of Refueling Activities

M1.8

Maintenance Activities Resulting in a Plant Transient

M1.9

Maintenance Activities Under the Unit 2 Reactor Vessel

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M1.10 Emergent Work - Minor Maintenance

Maintenance

and Material Condition of Facilities and Equipment

M2.1

Material Condition of Plant Equipment and Systems ..

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M2.2

Unit 2 Containment Closeout Inspection

IVI2.3 Unit 2 Suppression

Pool Cleaning

Maintenance

Procedures

and Documentation ..... ~... ~....

M3.1

Maintenance

Rule Implementation - Back Draft Isolation

ampers ..............................

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M7

Quality Assurance

in Maintenance Activities

M7.1

Review of Post-Maintenance Testing....................

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E2

E8

Engineering Support of Facilities and Equipment............. ~...

E2.1

Hydraulic Compatibility of Atrium 10 Fuel in the Current Unit 2

Core Configuration

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E2.2

Engineering Support of 4160 Volt Circuit Breaker Operability...

Miscellaneous

Engineering

Issues

E8.1

Review of FSAR Commitments ........................

E8.2

(Closed) URI 50-387, 388/96-08-03: Open HELB Room Doors ..

E8.3

(Closed) IFI 50-387, 388/96-04-01: Battery Charger Setpoints

E8.4

(Closed) URI 50-387,388/95-24-01:

Temporary Monitoring

Equipment

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III. Engineering

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IV. Plant Support

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Conduct of Security and Safeguards

Activities ~..........

S1.1

Access Practices on Vital Doors

FS

Miscellaneous

Fire Protection Issues

F8.1

Review of UFSAR Commitments ................

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V. Management Meetings.....,...

X1.

Exit Meeting Summery

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34

Re ort Details

Summar

of Plant Status

Unit 1 operated throughout this inspection period at 100 percent power.

The Unit 2 eighth

refueling and inspection outage

(8RIO) lasted 56 days and ended

on May 5, 1997.

On

May 10, 1997, Unit 2 closed its generator output breakers ending the refueling outage and

reached 100% power on May 16, 1997.

Unit 2 operated

at 100% power throughout the

remainder of the inspection period.

I. 0 erations

01

Conduct of

Operations'1.1

Unit 2 Refuelin

Activities

a.

Ins ection Sco

e 71707

Unit 2 8RIO refueling activities were inspected to determine proper movement and

placement of fuel assemblies.

In addition, several refueling related issues,

and two

events were reviewed for adequate

licensee corrective action.

The two events

involved a single blade guide making contact with the reactor vessel flange cover,

and retaining a clip on the refueling bridge which failed and dropped into the fuel

pool near new and used fuel.

b.

Observations

and Findin s

The inspector reviewed a number of procedures,

drawings, reports and corrective

action documents

including:

GE Drawing A-17599-D, Refueling Platform

QS Surveillances97-035, 97-036

OP-ORF-005, Refueling Operations

Significant Operating Occurrence

Report (SOOR)93-347

Condition Reports (CRs) 97-0809, 1182, 1293,.1 222, 1271 and 1275

related to human performance

CRs 97-1175, and 1182 related to refueling bridge condition

'Topical headings such as 01, M8, etc., are used in accordance with the NRC standardized

reactor inspection report outline.

Individual reports are not expected to address

all outline

topics.

CRs 97-1222 and 97-1256 related to impacting the reactor pressure

vessel

flange protector with a single blade guide

Sin le Blade Guide Movement

SSES Significant Operating Occurrence

Report (SOOR)93-347 addressed

a double

blade guide handle that was bent during a 1993 transfer activity. The double blade

guide was bent during movement above the reactor vessel flange.

During the 1993

movement of the double. blade guide, an operator on the bridge noticed that there

was not enough clearance to move over the reactor vessel flange.

The operator

attempted to stop the movement of the bridge but was unable to prevent impact

between the double blade guide and the reactor vessel.

The root cause of the event

was determined to be that the governing procedure,

RE-081-032, Refueling

Operations,

did not require the mast to be in the full up position prior to entering the

transfer canal.

The governing procedure was modified to include the requirement to

have the mast in the full up position prior to movement in certain areas of the

reactor vessel and/or spent fuel'pool.

Similar events occurred on four occasions

prior to the 1993 SOOR.

Following replacement of the aforementioned

bridge, the operating procedures for

the bridge were modified.

The procedures

were modified to account for the

flexibilitythat the new bridge offered during fuel movement (movement in more

than one direction at a time) ~

On April 11, 1997, the licensee was conducting

a transfer of a single blade guide to

reactor core location 49-34, in the automatic mode of the refueling bridge.

The

single blade guide impacted the reactor pressure

vessel flange protector during the

move.

The refueling bridge platform had started to slow down per design near its

target position.

Concurrently, the grapple started to lower the blade guide into the

core.

Because of the combination of these motions, the location in the core that

the single blade guide was to be placed, and the light mass of the single blade

guide, which allowed it to glide while being moved, its placement was not vertical

during the lowering process.

The operator recognized the situation (in a manner

similar to the 1993 event) and stopped lowering the mast but was unable to

prevent the impact.

The operators entered offnormal procedure, ON-081-002,

Refueling Platform Operation Anomaly, and performed inspections of the grapple

and flange protector.

The inspector determined that OP-ORF-005, revision December 9, 1996, Refueling

Operations, contained precautions that disagreed with the body of the procedure.

The precautions

stated that the operators were not to use semiautomatic

or

automatic mode when moving double blade guides, while the body of the procedure

stated that the operators

are not to use these modes when moving double blade

guides or single blade guides.

This inconsistency

in the procedure resulted in a

misunderstanding

by refueling operators who thought that it was acceptable to

operate the refueling bridge in automatic and/or semiautomatic while moving a

single blade guide.

Technical Specification (TS) 6.8.1 requires that written procedures

be established

and implemented for applicable procedures

recommended

in Appendix 'A'f

Regulatory Guide 1.33 Revision 2,.February 1978.

Regulatory Guide 1.33 Appendix

'A'tem 1.1requires procedures for refueling operations.

Contrary to the above,

OP-

ORF-005, Refueling Operations, was inadequate

in that it did not clearly control the

movement of single blade guides and as a result, a single blade guide impacted the

reactor pressure

vessel flange cover.

This is example

(a) of VIO 387,388/97-03-01.

Human Performance

Durin

Refuelin

Activities

During the Unit 2 BRIO activities, a number of events were identified by the licensee

in condition reports.

Among these condition reports, the inspector identified a trend

related to human performance.

The condition reports that relate to human

performance

are listed in paragraph

1.1.6 of this report.

The inspector identified

this issue in parallel with the SSES second line refueling supervisor.

The SSES

supervisor documented

the human performance

issue in CR 97-1293 and took very

aggressive

corrective actions.

Refuelin

Brid e Cab Entr

Chain Cli

On April 9, 1997, while moving spent fuel from the rector vessel to a storage

location in the spent fuel pool, an operator leaned on the safety chain for the

refueling bridge cab entry.

The clip broke and a portion of the clip landed in the fuel

pool near a new fuel assembly (Atrium 10), The licensee stopped fuel movement

activities and developed

a plan to locate the missing portion of the clip. The

licensee subsequently

located the missing portion of the clip and determined that

there was no impact on the new fuel near where it fell. The inspector determined

that the actions of the licensee were responsive

and very conservative.

The SSES

second

line refueling supervisor took charge of the issue and implemented

a number

of very conservative actions that resulted in the retrieval of the item from the pool

and overall improvement of the condition of the refueling bridge.

The inspector identified that the applicable GE Drawing A-17599-D, Refueling

Platform, indicated a specific part number for the correct clip. It was determined

that the clip that broke and possibly the chain on which it was connected

did not

correspond to the part number identified in the GE drawing.

Because of the

responsive

corrective action by the licensee, the conservative

action of the second

line refueling manager,

and the determination that there was no impact on the new

fuel, no violation of NRC requirements were determined to occur.

Conclusions

Unit 2 BRIO refueling activities were inspected to determine proper movement and

placement of fuel assemblies.

The fuel movement activities involved several

refueling related issues that were documented

by the licensee with condition

reports, and two events.

The two events were:

1) an impact between

a single

blade guide and a reactor vessel flange cover and 2) the failure of a safety chain clip

resulting in a portion of the clip being dropped into the fuel pool.

Each event was

responded

to aggressively

by the licensee.

The licensee also identified and resolved

generic issues with human performance related to refueling activities.

Despite the

two events and the refueling related issues, refueling'activities were well supervised

and were conducted

in a safe and conservative

manner.

Second line management

supervision of refueling activities and refueling bridge physical condition were

considered

a strength.

01.2

Unit 2 Turbine Buildin

S stem Particulate Iodine Noble Gas

SPING

Refuelin

Activities

a.

Ins ection Sco

e 71707

On May 2, 1997, with Unit 2 in cold shutdown,

a Stack Monitoring System Alarm

was received.

The inspector reviewed/inspected

this event to determine if the

operators responded

appropriately, plant procedures

and practices were adequate,

and if operator response

was similar to previously inspected plant events.

b.

Observations

and Findin s

In order to evaluate the event that occurred on May 2, 1997, it was necessary to

examine three aspects of the event.

The adequacy of the Alarm Response

Procedure

(AR) in addressing

the condition

The adequacy of the Shift Supervisor's

response

in resolving

divergent procedural guidance

The determination of whether the event represented

a real release

with potential impact on the public.

The Ade uac

of the Alarm Res

onse Procedure

On May 2, 1997, with Unit 2 in cold shutdown,

a Stack Monitoring System Alarm

was received.

A Plant Control Operator (PCO) responded

to the alarm, using alarm

response

(AR) AR-015-D4, Stack Monitoring System Alarm (OC630), Hi Hi

Radiation and initially determined that the cause of the alarm was the Unit 2 turbine

building Iodine above 3.59 E-8 micro Ci/cc. AR-015-D4 step 2.2 requires the

operator to perform a number of actions including substep 2.2.1b - Notify chemistry

to confirm alarm validity and to take appropriate actions.

The PCO implemented the

proper procedural steps including substep 2.2.1b.

Substep 2.2.1c states that for

valid alarms, evaluate data for entry condition into EO-100-105, Reactivity Release

Control.

Section 5.0 of the SSES Emergency

Plan states that an Unusual Event should be

declared

as soon as it has been indicated and verified. However, it sets the

parameters

of the time expected to verify the need for an Unusual Event by stating

that all reasonable

efforts are implemented to make this verification within fifteen

minutes of the initial indication of the event.

Because the AR procedure limits the

operator to a validation process which could take up to two hours before directing

him to the Emergency Plan, the AR does not adequately support the implementation

of the Emergency Plan.

Technical Specification (TS) 6.8.1 requires that written procedures

be established

and implemented for applicable procedures

recommended

in Appendix 'A'f

Regulatory Guide 1.33 Revision 2, February 1978.

Regulatory Guide 1.33 Appendix

'A'tem 5 requires procedures

for emergencies

and item 6 requires procedures for

abnormal, offnormal or alarm condition.

Item 6 further states that the procedures

for abnormal conditions should include immediate operator action.

Contrary to the requirement of TS 6.8.1, AR-015-D4, Stack Monitoring System Hi

Hi Radiation, was inadequate

in that substep 2.2.1b requires the operator to notify

chemistry to confirm the validity of a SPING alarm, effectively eliminating a

procedural route to the Emergency

Plan, prior to the completion of the validation by

chemistry.

Because the validation process implemented by chemistry can take up

to two hours, compliance with the AR inhibits compliance with SSES Emergency

Plan which would have all reasonable

efforts implemented to complete the

verification within fifteen minutes of the initial indication of the event.

This is

example

(b) of VIO 387,388/97-03-01.

This issue is similar to a violation identified in NRC inspection report 387,388/97-

01, which also involved the adequacy of operator response to an AR. The events

share

a common thread in that there appears to be an SSES tendency to delay

operator action in certain instances until a validation of a control room alarm is

- performed.

Although.this is not an incorrect approach overall, in the two specific

cases cited, this tendency was a contributor to the noted weaknesses.

Following a discussion between the SSES Supervisor of Operations, the inspector

determined that the licensee had implemented

a number of initial corrective actions

including a procedural change to the AR (PCAF 1-97-0348) which describes

a

different method of determining the alarm validity. The initial corrective actions

were determined to be very good.

The long term corrective actions are as yet

incomplete and will be addressed

by the licensee's

response to the violation

indicated in the previous paragraph of this report.

The Ade uac

of the Shift Su ervisor Res

onse

The Shift Supervisor (SS) determined that it would take up to two hours for

chemistry to validate the alarm and that the alarm data, as read, indicated that the

limit defining an Unusual Event in EP-PS-100-6,

EAL 15.1 release rate at 1.41

E 5

micro Ci/min was being exceeded

(note: this is a converted quantity that agrees

in

scale with the AR scale).

He further realized that section 5.0 of the SSES

Emergency Plan required that all reasonable

efforts be implemented to make a

verification of an Unusual Event within fifteen minutes of the initial indication of the

event.

The SS chose not to enter the emergency plan and did not declare an Unusual

Event.

Based on the inspector's discussion with the SS and his supervisor, the SS

chose to declare the SPING alarm an invalid alarm based

on data other than'that

called for in the AR. His determination was based on:

The immediate previous operating history of the unit (greater than 30

days shutdown and completing a refueling period) did not support

such

a high SPING indication.

Area monitors near the SPING monitors were not alarming and were

in fact reading normally.

The mechanical vacuum pump had been removed from service and

isolated four hours prior to the alarm and, therefore,

a release path

was isolated.

Work activities in progress

did not include activities that would

produce the SPING monitor alarm.

His choice was determined by the inspector to be a conservative

one which was

later verified to be technically correct.

However, the fact that the SS was forced

into such a position indicated a procedural weakness that is being addressed

as a

violation.

Im act of the Event on the Public

The inspector determined that there was no actual release,

that, by not declaring

an Unusual Event, the Shift Supervisor made a conservative decision, and that there

was no impact on the public.

C.

Conclusions

The operators responded

appropriately, the SS made conservative decisions which

ensured the safety of the unit and the public.

The inspector determined that one of

the plant alarm response

procedures

was inadequate.

02

Operational Status of Facilities and Equipment

02.1

Control Room Annunciators 0 erabilit

ao

Ins ection Sco

e 71707

The inspector selected

a number of normally alarmed annunciators

in the control

room and reviewed the impact that they had on the control room operators, the

licensee response to the lighted conditions, the impact on Technical Specification

operability requirements,

compliance with NRC requirements

and compliance with

selected

SSES design standards.

b.

Observations

and Findin s

The control room annunciators

selected for this review each affect SSES ventilation

systems.

There are a total of seven annunciators

in the lighted condition.

They

include:

AR-106-C16 "1C276 Delta Press Swings

AR-106-D16 "Delta Press Swings"

AR-106-E16

"RW Building HVAC Panel OC377 System Trouble"

AR-106-F16

"TB Supply Filter Hi DP"

AR-206-C16

"Circ Space Hi/Lo DP"

AR-206-D16 "Area DP Swing"

AR-206-F16

"HVACTurbine Building Panel 2C175 Trouble"

The inspector determined that the licensee currently has an engineering project

scheduled to eliminate these lighted annunciators.

The project includes initial

completion dates between June and August 1997.

The licensee has had similar

projects that date back to approximately 1988, which had similar goals and

objectives.

The inspector determined by interviews that the lighted annunciators

did not

distract the operators.

Occasionally the corresponding

alarms intermittently

annunciated.

In those situations, the operators were more directly affected.

The

inspector also determined that the licensee's current response to the lighted

conditions was adequate

and that it had a corrective action plan in place.

Previous

action plans were determined to be ineffective in that there was very little progress

made to the present to correct these specific alarmed annunciators.

The previous

plans were eventually abandoned

when milestones were routinely missed.

No

present impact on Technical Specification operability requirements,

compliance with

NRC requirements

or compliance with selected

SSES design standards

was

identified.

C.

Conclusions

Approximately seven lighted control room annunciators

correspond to chronic

conditions in SSES ventilation systems.

The licensee currently has a corrective

action plan in place to correct these conditions.

Relative to these specific alarms,

previous action plans were determined to be ineffective over a period of years.

There is no present impact on Technical Specification operability requirements,

compliance with NRC requirements

or compliance with selected

SSES design

standards.

02.2

Review of Licensee Condition Re orts

a.

Ins ection Sco

e 71707

The inspector reviewed approximately 800 CRs written during this report period

and/or associated

with the Unit 2 8RIO. The CRs were reviewed for initial licensee

response,

impact on Technical Specification operability requirements,

compliance

with NRC requirements

and compliance with selected

SSES design standards.

b.

Observations

and Findin s

A cursory review of approximately 800 CRs was performed by the inspector during

this inspection period.

Of these, approximately 100 involved level 2 or

1

conditions.

About half of the CRs address

equipment failures (400) and half of

those (200) were resolved prior to the end of this inspection period.

The CR

generation rate has dropped since the last inspection period but remains high based

on historical data.

The licensee has directed considerable efforts toward the resolution of CRs and the

inspector observed

PP5L management

levels up to and including the Vice President

Nuclear Operations

and the General Manager Nuclear Engineering routinely

addressing

the resolution of individual CRs.

The licensee demonstrated

during this

inspection report period that they were capable of addressing

and resolving CRs

adequately

in the short term.

c. 'onclusions

E

The licensee demonstrated

during this inspection period that they were capable of

addressing

and resolving CRs adequately

in the short term.

08

Miscellaneous Operations Issues (92700)

08.1

Review of Licensee Event Re orts

a.

Ins ection Sco

e 90712

The inspector reviewed Licensee Event Reports (LERs) submitted to the NRC to

verify that the details of the event were clearly reported, including the accuracy of

the event description, cause and corrective action.

The inspector evaluated

whether further information was required from the licensee, whether generic

implications were involved, and whether the event warranted onsite followup.

b. 'bservations

and Findin s

The following LERs were reviewed and closed during this inspection period:

Closed

LER 50-387 97-008-00: Instrument Response

Time Testing

0

On March 26, 1997, with Unit 1 at 100% power and Unit 2 in refueling, PPRL

determined that the requirements

of TS surveillances 4.3.1, 4.3.2, and 4.3.3 for

Response

Time Testing were not fulfilled. The failure to satisfy TS surveillance

requirements for response

time testing resulted in numerous instruments

and

systems being declared inoperable and required operators to enter TS 3.0.3 at 8:25

p.m.

Enforcement discretion was granted by.the NRC.allowing Unit 1,to exit TS 3.0.3 at 9:00 p.m. This event is discussed

in NRC Inspection Report

50-387/97-02.

The licensee's failure to adequately perform the TS surveillances

is a violation, This

licensee-identified

and corrected violation is being treated as a Non-Cited Violation,

consistent with Section VII.B.1 of the NRC Enforcement Policy.

Closed

LER 50-387 97-009-00: Fire Watch Rounds Not Completed On Time

On April 2, 1997, PP5L line management

identified that on two separate

occasions,

roving fire watch personnel

did not survey areas as required by TS Action 3.7.7.a.

PPSL determined that these events were caused

by ineffective on-the-job fire watch

training and qualification, and that there was no provision for timely feedback of

problems during fire watch rounds.

Short term corrective actions implemented by'PPSL included surveys of the missed

areas, initiation of refresher training for fire 'watch personnel,

and increased

supervisory oversight.

Long term actions described

in the LER were completion of

the refresher training, implementation of a formal training and qualification process,

training for fire watch supervisors,

and evaluation of feedback methods for fire

watch patrols.

The inspector found that a number of long term corrective actions described

in the

LER have expected completion dates in September

1997.

The inspector discussed

the completed short term actions with the responsible

PPSL manager and concluded

that the short term actions for this LER are acceptable.

However, to provide

assurance

that future fire watch rounds will not be missed in other areas of the

plant, more comprehensive

action will be needed.

This licensee-identified

and corrected violation is being treated as a Non-cited

Violation, consistent with Section VII.B.1 of the NRC Enforcement Policy.

Conclusions

The events reported. by PPRL in the LERs reviewed during this period were

appropriately reported, and provided an accurate description of the causes

and

corrective actions.

The inspector determined that for the LERs discussed

in brief,

the corrective actions were reasonable,

and that these events require no additional

onsite followup.

10

II. IVlaintenance

M1

Conduct of Maintenance

M1.1

Planned Maintenance Activit Review

a.

Ins ection Sco

e 62707

A variety of maintenance

activities were reviewed on the basis of their complexity,

safety (or risk) significance, or other considerations.

A sample of work permits,

equipment tagouts, procedures,

drawings, and vendor technical manuals associated

with these maintenance

activities were reviewed as part of the inspection.

Through

observation of the maintenance

activities, review of appropriate documentation

and

interviews with maintenance

personnel, the inspector sought to verify that the

activities were performed in accordance

with procedures

and regulatory

requirements, that personnel were appropriately trained and qualified, and that

appropriate radiological controls were followed.

b.

Observations

and Findin s

The following maintenance

activities were reviewed through direct observation

and/or review of the completed work packages:

WA S71662

Spent Fuel Pool Temperature

Instrument

WA V71171

Spent Fuel Pool Temperature Instrument

MT-GE-005

Circuit Breaker Inspection and Maintenance of 5 and 15 Kv

Breakers

WA V70005

LPRM Maintenance

c.

Conclusions

With respect to the selection of maintenance

activities indicated in this section, the

work activities were adequately controlled and observed portions were performed in

accordance

w'ith station procedures.

In the case of the spent fuel pool temperature

monitor maintenance

activity the work was well performed and controlled.

A

M1.2

Reactor Feed Pum

Re air Activit Review

a.

Ins ection Sco

e 62707

Maintenance activities associated

with the repair of the Unit 2 'C'eactor Feed

Pump (RFP) were reviewed on the basis of their potential for catastrophic failure of

the pump bearing during inprocess testing and/or normal plant operation.

11

b.

Observations

and Findin s

The following data were reviewed/inspected

during this inspection activity:

WA P61666

Reactor Feedwater

Pump (RFP) Disassemble/Inspect

MFP-QA-2309

Design Change Package/Engineering

Change Order

Preparation

NDPA-QA-206

NDAP-QA-502

Replacement

Item Evaluations

(RIE)

Work Authorization (WA)

MT-048-001

RFP Disassembly Inspection and Reassembly

TP-245-004

Overspeed

Trip Test of RFP

Two aspects of this maintenance

activity indicated weaknesses

in the licensee's

control of balance of plant related work. These aspects were; 1) the acceptance

of

an out of tolerance bearing and 2) the performance of a test without clear

communication to the Unit Supervisor of the out of tolerance condition and/or

acceptance

criteria of the performance test.

Acce tance of an Out of Clearance

Bearin

WA P61666 was written to disassemble

and inspect the Unit 2 'C'FP.

During the

performance of the WA, the low pressure

(LP) bearing was replaced.

The as-found

clearance of the old bearing met the MT-048-001 calculated clearance limits of 12

to 20 mils. The new replacement

bearing did not meet the clearance limits when

installed.

The clearance of the new bearing was 6 mils. The new bearing was

accepted

by the System Engineer as out-of-tolerance

and the procedural clearance

limits were relieved unilaterally by the System Engineer.

The inspector reviewed several plant processes

to determine the level of review and

approval, design control, and test control that are normally provided at SSES in

cases such as the one encountered

by the RFP System Engineer.

These processes

are discussed

in the paragraphs

below.

Section 6.4.8 of the SSES MFP-QA-2309 allows a process referred to as a minimod

for "mundane and simplistic changes" to plant equipment.

Minimods permit

implementation of a minor change

as a maintenance

activity using the normal work

authorization proces's

and post modification configuration control.

Minimods are

required to meet NDAP-QA-1202, section 6.2.5, and must be authorized by a

modification group lead and the supervisor site modification design.

The indicated

review and approval assures that the change fully satisfies the design basis of the

equipment.

12

Section 6.6.6 of NDAP-QA-0502 describes the process of WA field changes

used

by the licensee.

The procedure states that field changes

cannot result in a change

in plant design or configuration.

Section 1.0 of NDAP-QA-0206 describes the process for evaluating non-identical

replacement

items for use at SSES, including the identification of installation

requirements,

and the maintenance

of configuration control.

Section 3.5 describes

the exemptions allowed for use of a non-identical item outside of the approved

process.

The inspector determined that the use of the out-of-tolerance

bearing was a defacto

design change because

it resulted in a change to the bearing design specifications

as established

by the SSES procedure.

The acceptance

of the out-of-tolerance

bearing by the System Engineer did not meet the intent of any of the SSES

procedures

discussed

above.

Following the failure of the post maintenance

test, the

RFP was brought into compliance with the acceptance

criteria and a subsequent

RFP acceptance

test was performed.

The subsequent

acceptable

performance of a

post-maintenance

acceptance

test indicated that there was little potential for impact

on the safe operation of the unit.

Performance of a Post-Maintenance

Acce tance Test

Following the acceptance

of the out-of-tolerance bearing, the System Engineer

requested

and performed (with the assistance

of Operations) post-maintenance

test

TP-245-004.

The inspector reviewed the interaction/communication

between the

Unit Supervisor and the System Engineer prior to the performance of this test.

The

inspector understands

that the System Engineer did not make it clear to the Unit

Supervisor that the bearings did not meet the required design tolerances

and that

the test was intended to run the bearings in. Both the Unit Supervisor (US) and the

Plant Control Operator stated that no temperature criteria had been established

by

the System Engineer and that the US and PCO set an interim temperature

limit

based on their expectations

of the post-maintenance

test.

The inspector concluded

,that the US was not provided adequate

information by the System Engineer in order

to make knowledgeable

decisions concerning the testing of plant equipment.

This is

considered

a weakness.

Potential Re viator

and or Safet

Im act of the Maintenance Activit

Failure of the reactor feed pump would result in bounded plant transients

and the

reactor feed pump is not safety related.

The acceptable

performance of a post

maintenance

acceptance

test conducted after the initial test failure ensured that

there was little potential for impact on the safe operation of the unit from this

problem.

Therefore, no violations of NRC requirements

occurred.

However, the

inspector was concerned

because

safety related activities are subject to the same

maintenance,

engineering

and communication restraints as were described

in this

case.

13

c.

Conclusions

During the performance of maintenance

on the Unit 2 'C'eactor feed pump, the LP

bearing was replaced.

The new replacement

bearing did not meet the clearance

limits when installed.

The new bearing was accepted

by the System Engineer who

unilaterally revised the procedural clearance limits. The System Engineer did not

effectively communicate to the Unit Supervisor that he had not met the bearing

clearance

limits when he requested

and operators performing a post-maintenance

acceptance

test.

The test was terminated by the US on high:bearing temperature.

The bearing was brought into tolerance and a successful post maintenance

acceptance

test was performed.

The reactor feed pump is not safety related and its

failure would result in bounded plant transients.

Although these maintenance

and

communication activities were weak, there was no impact on the safe operation of

the unit. Therefore, no violations of NRC requirements

occurred.

M1.3

Surveillance Test Activit Sam

Ie Reviews

ao

Ins ection Sco

e 61726

The inspectors observed portions of selected surveillance tests involving different

technical disciplines for safety-significant systems.

b.

Observations

and Findin s

Through observation

and review of records, the inspectors verified that the test

activities were properly released for performance, that the test instrumentation was

within its current calibration cycle, and that it was being performed by qualified

personnel

in accordance

with approved test procedures.

The inspectors also

verified that the tests conform to TS requirements

and that applicable limiting

condition for operations

{LCOs) were taken.

The following activities were reviewed

during this period:

TP-149-060

SR-200-008

OP-261-002

OP-249-002

. SO-024-001

SO-200-006

SE-1 70-01

1

Unit 1 Shutdown Cooling Suction Flush

Shutdown Margin Demonstration

Precoating Reactor Water Cleanup Filter

Shutdown Cooling Flush

E Emergency Diesel Generator

Shift Surveillance Log

18 Month Secondary Containment lnleakage Test

c.

Conclusions

The routine surveillance activities observed during this inspection period were

adequately performed.

14'mer

enc

Diesel Generator Monthl

Surveillance Test

Ins ection Sco

e 61726

The inspectors reviewed the conduct of the monthly surveillance test for the

'C'mergency

diesel generator

(EDG).

Observations

and Findin s

On April 22, 1997, the inspector observed

selected portions of the monthly

surveillance test for the 'C'DG. The inspector found that an operator and

technicians were actively monitoring the EDG, and the test was performed using

approved procedures.

The inspector noted that a "standpipe level high" alarm was lit, which indicated that

a high level existed in the jacket water cooling system standpipe.

The inspector

discussed

the alarm with the operator, who pointed out that the standpipe

level

indication oscillated significantly, causing repeated

alarms.

The operator had

verified that the average

level was not increasing; an increasing level would indicate

a possible jacket water heat exchanger tube leak.

The operator understood

the

cause of the alarm, and was aware of the alarm response

procedure.

He also

recognized the oscillating standpipe

level as a known,'xpected

minor deficiency.

Yet, the inspector did not observe any deficiency tags or other documentation

indicating that this was a known condition.

The inspector noted that step 2.3 of the local alarm response

procedure,. LA-0521-

003, specified that the operator notify chemistry following EDG shutdown to

sample the jacket water.

However, the operator stated that he would not notify

chemistry because the average standpipe

level did not increase.

The inspector observed that the minor equipment condition associated

with the

oscillating standpipe

level was an accepted

condition that led the operator to believe

that compliance with the alarm response

procedure was not necessary.

Although a

minor issue, this represented

an example of failure to follow alarm response

procedures.

This failure constitutes

a violation of minor significance and is being

treated as a Non-Cited Violation, consistent with Section IV of the NRC

Enforcement. Policy.

Conclusions

The monthly surveillance test for the 'C'mergency diesel generator

(EDG) was

generally performed according to approved surveillance test procedures.

However,

the inspector observed

a minor failure to comply with alarm response

procedures for

a known equipment condition associated

with an oscillating jacket water standpipe

level indication.

This is being treated as a Non-Cited Violation consistent with

Section IV of the NRC Enforcement Policy.

15

Unit 1 Core S

ra

S stem Quarterl

Surveillance Test

Ins ection Sco

e 61726

The inspector observed portions of the Unit 1 core spray system quarterly

surveillance test, SO-151-A02, conducted

on April 24, 1997.

Observations

and Findin s

The inspector observed

an operator performing the preparations,

valve positioning,

and other actions specified by the test procedure.

The test was generally well-

controlled.

During the observations,

the inspector identified a weakness

in the methodology for

performing independent

verifications of valve positions as part of the test.

Specifically, the inspector found that isolation valves for the core spray pump

suction pressure

gages were operated multiple times during the test without

independent

verification of each manipulation, as required by the procedure.

Although the independent

verifications were done by a second operator at the

conclusion of the test, this did assure that the interim valve operations were

performed as specified.

The operator performing the test had questioned

the unit

supervisor prior to the test on how the independent

verifications were to be

performed, but due to communications

weaknesses

or incomplete review of the

'rocedure,

the decision was made to perform the verifications only at the

conclusion of the test.

The inspector brought this issue to the attention of plant

management.

Operations staff determined that the independent

verification steps

were not performed according to expectations.

Operations also determined that the

interim independent

verification steps were not necessary

and initiated a procedure

change to remove them.

Following the start of the 'C'ore spray pump, the inspector observed that the

discharge check valve indicator did not indicate fully open, as expected.

The

inspector brought this to the attention of the operator, and condition report 97-

1475 was initiated. The licensee's operability determination concluded that this

was an indication discrepancy only and, because

system flow rates were within

specification, there were no operability concerns.

During the test, the inspector observed

an action that resulted in the pumps being

tested in a condition that was altered from their as-found condition.

The inspector

noted that the surveillance procedure, SO-151-A02, required the operator to vent

the core spray pumps prior to starting them.

The purpose of the venting steps

apparently was to remove any trapped air that could lead to air binding of the

pumps.

The inspector discussed

this observation with the operator, who stated that he

knew of no instances

in which air was actually vented from the pumps during

surveillance testing.

He stated that the steps provided additional assurance

that no

air was in the pumps.

The inspector also brought this issue to the attention of plant

16

management,

who stated that venting was considered

by operations management

to be a good operational practice, but was not required.

Based on this information,

the inspector considered that there were no pump operability issues.

NRC

Information Notice (IN) 97-16, issued April 4, 1997, discussed

several examples of

unacceptable

preconditioning actions that licensees

have performed before

Technical Specification surveillance testing.

One of the examples cited was a

practice of venting residual heat removal pumps immediately prior to conducting

surveillance testing.

IN 97-16 further notes that equipment should be tested in the

as-found condition, and any disturbance

or alteration to equipment would be

expected to be limited to the minimum necessary to perform the test and prevent

damage to the equipment.

The inspector concluded that the practice of venting the core spray pumps

immediately prior to starting them for their quarterly surveillance test was an

unnecessary

preconditioning action.

This poor surveillance practice resulted in the

pumps being tested in a condition that was different'from the as-found condition

an'd thus made questionable

the validity of the surveillance test results.

The failure

to perform the core spray surveillance test under suitably controlled conditions is

considered

a violation of 10 CFR 50 Appendix B, Criterion XI, Test Control.

(VIO

50-387/97-03-02)

C.

Conclusions

The performance of the quarterly surveillance test for, the Division I core spray

system was generally well-controlled.

The inspector observed that the methodology

for performing independent

verifications within the test procedure was weak and did

not meet licensee expectations.

The inspector concluded that the practice of

venting the core spray pumps immediately prior to starting them was an

unnecessary

preconditioning action.

The failure to perform the core spray

surveillance test under suitably controlled conditions is considered

a violation.

M1.6

Mainte'nance Activities that Resulted in Potential Industrial Safet

Situations

a 0

Ins ection Sco

e 62707

Two maintenance

activities associated

with the restoration of Unit 2 8RIO

conditions were inspected.

b.

Observations

and Findin s

During a Unit 2 reactor building tour, the inspector identified personnel working

below and to the side of a suspended

drywell hatch.

The hatch, which weighed in

excess of 1000 pounds, was supported

by a lifting rig. When the inspector

questioned

the attending first line supervisor,

he agreed that the position of the

hatch was not safe and stated that it would be repositioned.

Upon returning to the

work area, the inspector found that the hatch had been moved, but was still in a

position to affect workers, if the rig from which it was suspended

failed. The

17

inspector notified the SSES safety organization and the SSES safety organization

worked with maintenance

line management

to resolve the issue adequately.

On April 24, 1997, a nylon sling, which was being used to support

a tool box,

separated.

The sling was suspended

from the Unit 1 Reactor Building Crane

auxiliary hoist.

One end of the box dropped approximately eight feet striking the

edge of a stored Unit 2 reactor cavity shield plug.

The tool box weighed

approximately 4000 pounds.

The inspector determined,

in parallel with the

licensee, that one of the root causes of this event involved weaknesses

in the

routine testing of nylon slings.

SSES safety and maintenance

departments

adequately

responded to the event.

Because

no personnel injuries occurred and there was no impact on safety related

equipment,

no violations of NRC requirements

occurred.

However, these events

constitute

a weakness

in the industrial safety practices at the site.

C.

Conclusions

Two maintenance

activities associated

with the restoration of Unit 2 BRIO

conditions were inspected.

Each of the activities had the potential for personnel

injury, although no injury occurred.

Both issues were adequately

resolved by the

licensee.

No personnel injury occurred, there was no impact on safety related

equipment,

and no violations of NRC requirements

occurred.

M1.7

Maintenance Activities in Su

ort of Refuelin

Activities

a.

Ins ection Sco

e 62707

During fuel movement activities, a fuel assembly was suspended

(less than one

foot) above its lower fuel support piece without the ability to raise or lower it

through normal means.

The inspector observed portions of the refueling activities

and the ensuing maintenance

support activities.

b.

Observations

and Findin s

On April 15, 1997, a fuel assembly was suspended

(less than one foot) above the

lower reactor vessel fuel support piece without the ability to raise or lower it

through normal means.

The inspector observed portions of the refueling activities

and the ensuing maintenance

support activities.

It was determined from the control

room that the inability to move the fuel assembly was the result of a rod movement

interlock from a lost position on rod 22-47.

The Unit Supervisor initiated a maintenance

activity (Work Authorization V71079)

to resolve the interlock. The inspector observed portions of the WA and reviewed

the following documents:

18

WA V71079, 22-47 Rod Out Interlock

MT-AD-509, Control of Minor Maintenance

CR 97-1303, Emergency Work Authorization to Clear Refueling

Platform Interlocks

NDAP-QA-0500, Conduct of Maintenance

NDAP-QA-'0502, Work Authorization System

MI-PS-001, Work Package Standard

Section 6.6.4 of NDAP-QA-502 addresses

emergency work authorizations.

It states

that the Shift Supervisor assumes

responsibility during offnormal hours for all

'groups represented

in the Work Authorization procedure.

No other allowance is

given by NDAP-QA-502 concerning the conduct of work, and there is no relief given

for the processes

that actually control work in the field under NDAP-QA-0502.

During the performance of work under WA V71079, the inspector noted that the

technician was using an "information only" SSES Training Department drawing to

guide his activities,

In addition, the technician was not documenting

his activities

on a Status Control sheet, NDAP-QA-502-5, nor was he documenting

his activities

on an Actions Taken form NDAP-QA-502-3.

The training drawing used by the technician to support WA 71079 was not

'prescribed by the licensee for work at SSES.

Because the licensee was able to

determine that none of the activities performed impacted on the operability of the

reactor protection system, this issue being treated as a Non-Cited Violation

consistent with Section IV of the NRC Enforcement Policy.

c.

Conclusions

During fuel movement activities a fuel assembly was suspended

(less than one foot)

above its reactor vessel fuel support piece without the ability to raise or lower it

through normal means.

Maintenance activities were initiated on the Unit 2 reactor

protection system to resolve this condition.

Maintenance activities performed under

Unit 2 work authorization WA 71079 on the reactor protection system were

performed using an "information only" SSES Training Department drawing which

was not authorized for use.

This issue being treated as a non-cited violation.

M1.8

Maintenance Activities Resultin

in a Plant Transient

ao

Ins ection.Sco

e 62707

On February 25, 1997 maintenance

activities resulted in a loss of Unit 1 condenser

vacuum.

The inspector reviewed an SSES Event Review Team (ERT) Report and

additional information supplied by the licensee to evaluate the event.

b.

Observations

and Findin s

NRC Inspection Report 387,388/97-02 section M21.b.1 discussed the event.

In

that section it states that the concrete boring work was directly above the Unit 1

19

hydrogen analyzer cabinet.

Following a review of the ERT and WA C63269, the

inspector determined that a more accurate description of the location of the work

with respect to the hydrogen analyzer would be approximately 15 feet above and

offset approximately 8 feet to the north west.

The previous report incorrectly

discusses'the

event.

The report should have stated that the 1997 event was similar

to a 1996 transient which occurred in anticipation of a loss of condenser

vacuum,

and that the event was the result of weak circulating water pump maintenance.

The weak maintenance

included:

the failure to remove a buildup of corona

discharge material inside a connector box; the use of sealants

on the motor

connection box; and the failure of preventive maintenance

activities to identify the

connector cable damage, the corona discharge material or a contaminated standoff

insulator prior to failure.

c.

Conclusions

The conclusions of NRC Inspection Report 387,388/97-02 section M21.b.1 are

unchanged.

M1.9

Maintenance Activities Under the Unit 2 Reactor Vessel

a.

Ins ection Sco

e 62707

The inspector reviewed and observed

(through a video link) maintenance

activities

conducted

under the Unit 2 reactor vessel.

b.

Observations

and Findin s

C.

As a result of weaknesses

identified in the under-vessel

maintenance

activities

during the Unit 1 refueling outage, and CRs written by the licensee, the inspector

observed

and reviewed under-vessel

activities during the Unit 2 refueling outage.

The licensee issued

CRs 97-0941, 97-0947, and 97-0936 which addressed

split

cables identified on Unit 2.

Unit 1 had experienced

split cables and the licensee's

corrective actions included the placement of cable protectors to prevent

maintenance

related damage.

The inspector determined that the licensee's

corrective actions were adequate

and that the inherent tight quarters under the

vessel made maintenance

activities very difficult.

e

Conclusions

As a result of weaknesses

identified with under-vessel

maintenance

activities during

the 1996 Unit 1 refueling outage,

and condition reports written by the licensee, the

inspector observed/reviewed

under vessel activities during the Unit 2 refueling

outage.

The licensee issued and resolved

a number of condition reports and took

adequate

corrective actions.

No violations of NRC requirements were identified.

20

M1.10 Emer ent Work - Minor Maintenance

Previous NRC inspection observations

and non-cited violations identified that the,

documentation

of actions taken and the "as left" condition following minor

maintenance

on safety related equipment was weak.

On May 6, 1997, the

inspector observed minor corrective maintenance

on a suppression

pool temperature

indicator, Tl-15751, located on the remote shutdown panel in Unit 1. The work

was performed by the emergent work action crew (EWAC) in accordance

with

maintenance

procedure MT-AD-509, Minor Maintenance,

under work authorization

S71 572.

The inspector concluded that the scope and complexity of the work observed was

similar to the examples provided in the Minor Maintenance procedure.

The

inspector observed

appropriate documentation

of the actions taken, appropriate post

maintenance testing, and good communication of the as left configuration.

During

the activity, good supervisory interaction and communication between the work

crew and the control room were noted.

M2

Maintenance and Material Condition of Facilities and Equipment

M2.1

Material Condition of Plant E ui ment and S stems

a 0

Ins ection Sco

e 62707

During routine observations

of plant operations, the general condition of equipment

was examined to determine the effectiveness of licensee controls for identification

and resolution of maintenance

related problems.

b.

Observations

and Findin s

The general condition of the facilities was discussed

routinely with SSES operators

and system engineers

and was inspected

in the field. Five issues were identified

that required varying degrees of inspector review.

These five issues were:

Seven annunciated

plant conditions associated

with Unit 1 and Unit 2

ventilation

This issue is discussed

in section 2.1 of this report

Unit 2 'D'esidual heat removal (RHR) pump oil leak and Division 2 RHR

Swing Bus Motor Generator Set slinging oil onto the wall while being

internally contaminated

Each of these issues had been previously identified by the licensee and had

corrective action plans (CR 97-1568, 97-1535) in place and analyses to

indicate that there was no impact on the safe operation of the involved unit.

21

'E'mergency diesel fuel line abutting a support during operation.

This issue was identified during a routine resident inspector tour of the diesel

building. A CR and a WA were written to resolve this issue.

No immediate

impact on diesel operability was identified by the inspector or the licensee.

Bonnet leak on HV10606B

On a plant tour, the inspector identified that the leakage from the Unit 1

Reactor Feed Pump discharge check was impacting plant cabling.

The

licensee determined that the affected conduit tray was approximately 170'F,

there was no immediate impact, and is completing

a long-term evaluation.

c.

Conclusions

With the exception of the ventilation annunciator issue, the licensee has initiated

adequate

corrective actions.

M2.2

Unit 2 Containment Closeout Ins ection

a.

Ins ection Sco

e 71707

Prior to final closure of the Unit 2 containment, the inspector performed

a

walkdown of the drywell to assess

PP&L's effectiveness

in removal of foreign

material, restoration of pipe insulation, cable raceway covers, electrical connections,

flanges, piping, and supports.

b.

Observations

and Findin s

In general, the cleanliness of the containment was very good.

The inspector

performed

a containment walkdown in parallel with PPRL management's

final tour.

A small number of items were identified and removed during this tour. These items

included plastic tie wraps, small pieces of wire, and duct tape.

The inspector identified three items that did not appear consistent with the

expected equipment configuration during the containment tour:

~

Screens protecting the air inlet for containment coolers 2V421A and

2V415A were missing.

~

A number of hold-down clips were missing from a section of floor grating on

elevation 738.

~

A bundle of small gauge wire was installed with tie wraps on existing

conduit and structures.

22

PP&L initiated a CR for each of these items and documented

an, operability

determination.

In all three cases,

PP&L determined that the items were acceptable

for operation and would not impact operability.

Conclusions

PP&L's efforts to remove foreign materials from the Unit 2 containment following

the refueling outage were very good.

The final containment walkdown and

inspection by the Operations

and Maintenance department managers

were viewed

as strength.

Based on the areas reviewed during the inspector's containment

walkdown, PP&L was effective in restoring equipment (hatches,

insulation, hangars,

etc.).to the condition required for plant operation.

Unit 2 Su

ression

Pool Cleanin

Ins ection Sco

e 62707

The inspector reviewed PP&L's suppression

pool cleaning activities during the

Unit 2 refueling outage and confirmed implementation of the compensatory

actions

requested

by the NRC in a letter dated February 19, 1997.

Observations

and Findin s

The Unit 2 suppression

pool cleaning was performed by divers and included filtering

the water and vacuuming the floor, structures,

and reactor pedestal openings.

The pool water was filtered/vacuumed

using 1.0 micron filters'and resulted in the

collection of approximately 1100 pounds of wet sludge and 71 pounds of rust

particles.

Debris collected during this evolution included rust particles, small pieces

of tape, tie-wraps, small pieces of paper/plastic,

a 25 foot hose, tags, strips of

metal, a glove, a boot, small pieces of wire, small pieces of hose and rope, a hard

hat, a soda can, a piece of weld guard (approximate

1 square foot), a 4" wire

brush, nuts and washers.

PP&L's final inspection of the pool floor found no signs of debris and no signs of silt

accumulation.

The cleaning resulted iri improved visibility (to approximately

11 feet

below the water surface) with a fine particulate still suspended

in the water.

Visibilityat the bottom of the pool was reported to be 2 to 4 feet.

All 87 downcomers were inspected

and 5 were found to contain floating debris.

The debris consisted of a piece of rope, 2 rubber'boots,

a paper tag, and small

pieces of paper/plastic.

PP&L documented

the items discovered during the inspection pool cleaning in the

CR process.

In all cases,

PP&L determined that the items would not have prevented

the emergency core cooling systems taking suction on the pool from performing

their intended function.

23

The inspector observed

PPRL's final inspection in the drywell to verify the proper

configuration and condition of insulation.

This activity also included a walkdown to

verify that all foreign material had been removed from the drywell. These activities

were noted to be through and are discussed

in more detail in Section M2.2 of this

report.

The Inspector discussed

the Unit 2 suppression

pool cleanout and inspection

results, and the implications for Unit 1, with the cognizant system engineer.

PPRL

considers the sludge removed from Unit 2 to be bounded

by the operability

evaluation (dated November 15, 1995) performed in response to NRC Bulletin 95-02.

Based on review of this evaluation, the inspector determined that the Unit 2

suppression

pool cleanout results did not invalidate PPKLs assumptions

or

conclusions.

C.

Conclusions

PP5L's efforts to reduce foreign debris in the Unit 2 containment and suppression

pool during the Spring 1997 refueling outage were through.

Management

involvement in the final inspection of containment was viewed as a strength.

The

compensatory

actions requested

by the NRC in conjunction with deferral of the final

resolution of Bulletin 96-03 were implemented by PPtkL. The Unit 2 suppression

pool cleanout results were consistent with the assumptions

contained

in PPRL's

existing operability evaluation for the suction strainer clogging issue communicated

in NRC Bulletin 95-02.

No information was identified that would invalidate PP&L's

conclusion regarding operability of either SSES Units'uppression

pool strainers.

M3

Maintenance Procedures

and Documentation

M3.1

Maintenance

Rule Im lementation - Back Draft Isolation Dam ers

Ins ection Sco

e 62707

b.

The inspector reviewed PPSL's Design Guide for System Scoping for Maintenance

Rule Applicability, GDS-18, to determine if the program for implementation of the

Maintenance

Rule (10 CFR 50.65) identified the safety function of back draft

isolation dampers

(BDIDs) in the reactor building ventilation system.

II

Observations

and Findin s

Piping systems whose failure might generate

hazardous

environmental conditions

are located in rooms which are capable of being isolated from required safety

systems.

Isolation of these rooms is provided, in part, by automatic BDIDs that

actuate on differential pressure between the room and the general reactor building.

The inspector was concerned that although pressure switch testing is periodically

conducted, there was no evidence that BDIDs were exercised to demonstrate

functionality.

24

.10 CFR 50.65(b) states that the scope of the monitoring program required by the

rule is to include safety related structures, systems, or components that are relied

upon to remain functional during and following design basis events.

FSAR Section

3.6.1.1 describes the Susquehanna

design basis for a postulated

pipe break outside

the containment.

Piping systems whose failure might generate

hazardous

environmental conditions are located in compartments which are capable of being

isolated from required safety systems.

The isolation of those compartments

is, in

part, accomplished

by BDIDs in the reactor building ventilation system.

The inspector's review of GDS-18, Revision 3, System Scoping for Maintenance

Rule Applicability, dated July 15, 1996, found that the BDIDs were not identified by

PPSL as a maintenance

rule function of the reactor building ventilation system.

The

reactor building ventilation system is identified as being within the scope of

maintenance

rule and fourteen separate

maintenance

rule functions of the system

are identified. The inspector discussed

this finding with the cognizant nuclear

system engineering

(NSE) supervisor,

and the supervisor acknowledged the need for

these dampers to be covered by PPSL's program.

In response to this issue,

CR 97-

1648 was initiated by PPSL to address the omission of the BDID function from the

maintenance

rule program scope and the fact that no testing has been performed

that confirms the dampers

are capable of closing.

As part of the corrective actions

for CR 97-1648 the licensee prepared

an interim operability determination that

concluded that the BDIDs were operable because

failure of the equipment was not

expected

based on the testing of the solenoids and the pressure switches.

The

inspector determined that the operability determination was weak in that it did not

support why the BDIDs were expected to function mechanically when called upon.

At the end of the inspection, the following information was needed to determine

the operability of the BDIDs and their status within the Maintenance

Rule program.

Industry data regarding the functionality of unexercised

dampers is needed.

A determination by the licensee of whether or not the BDIDs have ever

stroked on demand.

A determination whether future testing of the dampers

is needed.

A determination by the licensee whether the BDIDs should be included in the

scope of the reactor Building Ventilation system under GDS-18 criteria, and

whether a BDID failure is risk significant.

This issue willremain unresolved

pending completion of the CR 97-1648 corrective

actions (URI 50-387, 388/97-03-04).

Conclusions

The back draft isolation dampers

are safety related components within the non-

safety related reactor building ventilation system.

Although the reactor building

ventilation system is addressed

by the maintenance

rule program at SSES, the

25

function of the BDIDs was not included in the licensee's evaluation of the reactor

building ventilation system.

There was no performance history to indicate that the

BDIDs would function on demand.. A determination of the operability and

maintenance

rule status of the BDIDs in the reactor building ventilation system will

be tracked as an unresolved item.

M7

Quality Assurance in Maintenance Activities

M7.1

Review of Post-Maintenance

Testin

80

Ins ection Sco

e 62707

The inspector reviewed the most recent Nuclear Assessment

Services (NAS) audit

of the test control program at SSES, dated September

11, 1995.

Specifically,

PPRL's review of post-maintenance

testing (PMT) was evaluated against the SSES

Operational Quality Assurance

Manual policy for Control of Inspection and Testing

(OPS-14).

b.

Observations

and Findin s

NAS performs an audit of the test control program every two years as discussed

in

FSAR Chapter 13.4.

The 1995 audit reviewed a sample of fifteen completed work

authorizations

(WAs), out of approximately 22,000 activities in 1994/1995, to

determine the adequacy of post-maintenance

test activities.

The inspector found

that the sample included only WAs that had gone through the work planning

process

and did not appear to have sampled WAs for minor corrective maintenance

performed under the licensee's

Maintenance Investigation Instruction (superseded

by the Minor Maintenance

process).

The 1995 audit found that the post-maintenance

functional test requirements,

required by NDAP-QA-0482 were defined by the work group and that all WAs

reviewed contained test requirements.

In addition, the PPS.L sample found that test

results were properly documented,

analyzed,

and evaluated against test acceptance

criteria to verify completeness

and achievement of test objectives.

The inspector concluded that the 1995 NAS audit sampled too few safety related

maintenance, activities (15 out of 22,000) to provide a representative

sample.

In

addition, the omission of unplanned

maintenance

(ie., minor maintenance)

from the

sample was considered

a program weakness.

C.

Conclusions

The Nuclear Assessment

Services audit of the Test Control Program (Audit No.95-059) provided an adequate

review of post-maintenance

testing as required by

the Operational Quality Assurance

Manual.

However, the inspector considered the

audit sample size to be small relative to the number of safety related work

authorizations

processed

in a two year period and is considered

a potential

weakness.

The lack of an NAS audit in the minor maintenance

area was also

0

26

considered

a weakness

in testing program oversight.

Based on these weaknesses,

the inspector could not determine whether the 1995 NAS audit assessment

provided PPRL management

a representative

assessment

of all types of post-

maintenance

testing.

III. En ineerin

E2

Engineering Support of Facilities and Equipment

E2.1

H draulic Com atibilit of Atrium 10 Fuel in the Current Unit 2 Core Confi uration

a.

Ins ection Sco

e 37551

The inspector reviewed the cycle 9 Unit 2 core configuration for hydraulic

compatibility and stability. This review was performed separate from the core

thermal analysis that was conducted for the most recent fuel-related TS

amendment.

b.

Observations

and Findin s

The following proprietary documentation

was made available by the licensee for

NRC review:

Thermal Hydraulic Characteristics of the Atrium 10 Fuel Design for

Susquehanna

Susquehanna

SES Unit 2 Cycle 9 Hydraulic Compatibility Evaluation

Susquehanna

2 SQB-8 Design Report Mechanical and Thermal

Hydraulic Design for SPC Atrium 10 Fuel Assemblies

The review was conducted to determine if there were conditions that resulted in

flow instabilities, reversals or other anomalies (separate from thermohydraulic

conditions reviewed under a recent TS change).

No such conditions were identified.

The inspector did not retain any of the proprietary documentation.

C.

Conclusions

The hydraulic performance

(separate from thermohydraulic performance) of the

Atrium 10 fuel was reviewed and determined to be adequately

bounded by analysis.

E2.2

En ineerin

Su

ort of 4160 Volt Circuit Breaker 0 erabilit

a.

Ins ection Sco

e 62707

On April 18, 1997, a PCO attempted to start the Unit 2 'A'HR pump and received

no response.

The inspectors reviewed the root cause of this event (CR 97-1363),

e

27

the impact of the failure on the operability of other safety related equipment, the

licensee's

corrective actions, and the generic implications of the failure.

b.

Observations

and Findin s

The subject breaker contains

a personnel protection device designed to not allow

the breaker to close if it is fully or partially racked out.

This protection device is

referred to as a tripper lever and was the subject of a previous NRC Information

Notice, (IN) 96-50, September 4, 1996.

Upon investigation of the failure, the

licensee determined that this tripper lever was not in the fully down and level

position, causing the breaker not to close when called on to perform by the PCO.

The licensee responded

very conservatively to the previous issue discussed

in the

IN. However, the previous failure mode had been isolated to cases where the

breakers had been racked-in just prior to a test.

In the SSES RHR case, the breaker

had been cycled successfully prior to its failure and had not been racked-out/in just

prior to the test.

The licensee took immediate actions to ensure the operability of

the other safety related breakers on the operating unit and took generic action on all

other safety related breakers.

Included in the corrective actions was the

performance of a number of dimensional m'easurements

of the breakers.

Several

minor indications were identified through visual inspections

and were corrected by

the licensee (examples

CR 97-1547, and 97-6125).

The inspector reviewed the availability of the Unit 2 'A'HR pump and determined

.that no TS violations occurred.

The appropriate technical information was transferred to the SSES NRR Licensing

Project Manager for generic review and resolution.

C.

Conclusions

Following the failure of a safety related 4160 Vac breaker to perform when

requested,

the licensee identified a possible new failure mode involving a personnel

protection device referred to as a tripper lever. Although the failure mode was

different, this same device that was the subject of IN 96-50.

The licensee's

response to the IN was very conservative

and aggressive.

The involvement of first

and second

line engineering management

in this issue was laudable.

The generic

aspects of the issue have been forwarded to NRR for review.

E8

Miscellaneous Engineering Issues (92902)

E8.1

Review of FSAR Commitments

A recent discovery of a licensee operating its facility in a manner contrary to the

Updated Final Safety Analysis Report (UFSAR) description highlighted the need for a

special focused review that compares plant practices, procedures

and/or parameters

to the UFSAR description.

The inspectors reviewed the applicable portions of the

UFSAR and SSES emergency operating procedures that relate to post-accident

28

alternative water supplies to the core in order to determine if the licensee

maintained the capability of accessing

these alternative water supplies.

b.

Observations

and Findin s

The inspectors reviewed the following SSES operating and emergency operating

procedures:

ES-150/250-002,

Boron Injection via Reactor Core Isolation Cooling

ES-013-001,

Fire Protection System Cross Tie to RHR Service Water

SO-013-001, Monthly Hose House Inspection

SO-253-004, Quarterly SBLC Flow Verification

The licensee maintained and documented

adequate

access to the emergency

alternate water sources identified in the SSES FSAR.

C.

Conclusions

The licensee maintained the capability to utilize emergency alternate water sources

identified in the SSES Emergency

Plan and'discussed

in the FSAR.

No violations of

NRC requirements were identified.

E8.2

Closed

URI 50-387

388 96-08-03: Open HELB Room Doors

a.

Ins ection Sco

e 37551

On:three occasions,

doors for personnel

access to rooms equipped with high energy

line break (HELB) protective features were blocked open during on-line maintenance

activities.

The inspector opened the unresolved item pending additional information

from PPSL that was necessary to determine if this condition was adequately

bounded

by existing analysis.

b.

Observations

and Findin s

The plant design includes the ability to sustain

a high energy pipe break accident

coincident with a single active failure and retain the capability for safe cold

shutdown (reference

NUREG 0776).

The plant was designed

in accordance

with

Branch Technical Position (BTP) ASB 3-1 "Protection Against Postulated

Piping

'ailures in Fluid Systems Outside Containment" and PPKL used separation

as the

primary means of protection.

At the time the open doors were identified, PPSL was using a TS interpretation (No.

1-92-006) in place to provide operational restrictions for opening doors, hatches,

and plugs.

This interpretation permitted the opening or removal of HELB room

boundaries

in support of work activities but prohibited their being left open

indefinitely. The inspector's review of PPSL's basis for the interpretation (EWR

M10103) found that it did not address

generic assumptions

regarding the open

room boundary dimensions,

or the potential impact on environmental qualification

29

and divisional separation..PPSL's

immediate actions (after the third door was

identified) included removal of the subject TS interpretation pending

a review of its

basis.

In response to this issue, PPKL performed operability determinations for the three

subject plant areas where doors were blocked open in support of

maintenance'reference

CRs 96-2191

and 96-0748)

~ The evaluations addressed

seventeen

aspects of the SSES design basis, including those originally questioned

by the

inspector.

In each evaluation,

PPS.L concluded that there was no impact on the

operability of the subject design features.

The inspector's review of these

operability determinations

did not identify any problems with PPSL's assessment.

Due to the complexities of these assessments,

PP&L has determined that a generic

assessment

to cover all plant areas and combinations of open. doors, floor plugs,

hatches, etc., is not practical.

PPtkL management

stated that, in the future, a

safety evaluation would be performed to support specific work activities that would

block open a room boundary that could affect the HELB protective features.

The

inspector noted that this type of evaluation is now required by NDAP-QA-0409,

which was approved

on February 27, 1997.

Although PPRL's analysis after the events determined that the plant had been

operated within its design basis, the failure to perform safety evaluations before

blocking open the HELB room doors is a violation of 10 CFR 50.59.

This violation is

example (a) of VIO 97-03-04.

C.

Conclusions

E8.3

aO

PPSL's failure to perform safety evaluations prior to blocking open doors for rooms

with HELB protective features is a violation of 10 CFR 50.59.

Subsequent'PthL

evaluations determined that blocking open the HELB room doors created no adverse

impact on operability or conditions outside the plant's design basis.

Closed

IFI 50-387

388 96-04-01: Battery Charger Setpoints

Ins ection Sco

e 62703

This inspection followup item was opened pending NRC review of PPSL's formal

evaluation of increased 250 Vdc system float voltages.

b.

Observations

and Findin s

In 1994, the 60 month performance discharge test for 125 Vdc battery 2D620,

found the battery had 83.5% capacity.

Although this met the TS required minimum

capacity of 80%, this result was unexpected for a 5 year old battery.

PPRL

determined that the cause of the capacity degradation was sulfation resulting from

low float voltage,

As corrective action for this problem, PPRL increased the float

voltage for both 125 and 250 Vdc batteries.

30

NRC Inspection Report (IR) 96-04 reviewed the change

in float voltages after

discrepancies

were identified in plant log acceptance

criteria.

Section E2.1 of IR

96-04 stated that PP&L calculation EC-088-0530, Revision 1, did not address the

acceptability of the 125 Vdc float voltage.

However, this was not correct;

Calculation EC-088-0530, Revision 1, did not address the acceptability of the 250

Vdc float voltage.

PP&L's review of this issue (CR 96-0475) found that NSE had authorized the

increase

in float voltage for the 250 Vdc battery (under WAs S51447 and V50816)

as a long-term corrective action for the 1994 failure. However, a request for

Systems Analysis to evaluate the effects of the increased float voltage on

equipment connected to the 250 Vdc battery was not processed.

Revision 2, to EC-088-0530, Attachment 6, evaluated the effects of raising the

Class 1E 250 Vdc battery charger float voltage from 265 Vdc to 268 Vdc (each

with a band of a3 volts) to assure that there would be no adverse effects on the

safety function of connected

components.

PP&L concluded that all safety related

250 Vdc equipment can withstand continuous operation at the new float voltage

.without loss of life or adverse impact to nuclear safety.

The inspector reviewed this

revision of the calculation, discussed

sever'al questions with the cognizant PP&L

engineer,

and concluded that a technical basis exists for PP&L's conclusions.

PP&L's failure to perform a safety evaluation for the increase

in 250 Vdc float

voltage is a violation of 10 CFR 50.59.

This violation is example b of Vio 97-03-04.

C.

Conclusions

Revision 2 of PP&L's calculation EC-088-0530, documented

the evaluation of 250

Vdc battery 'float voltage and appropriately considered the potential for degradation

of connected

safety related equipment.

No degradation of the equipments ability to

perform its intended safety function was identified.

ff

E8.4

Closed

URI 50-387 388 95-24-01: Temporary Monitoring Equipment

a 0

Ins ection Sco

e 37551

This unresolved item was opened

pending PP&L's documentation

of a safety

evaluation for temporary test equipment (visicorder) used on the emergency diesel

generators.

The inspector found that the Bypass Program, which controlled

temporary monitoring equipment, did not require a formal evaluation for use on

safety related equipment, when the monitoring equipment was in place for less than

seven days.

b.

Observations

and Findin s

In response to the inspector's findings documented

in IR 50-387/95-24 and issues

raised by NRC Information Notice 95-13, PP&L took the following actions:

31

~

Administrative procedures

governing the Bypass Program (NDAP-QA-0484)

and the Work Authorization System (NDAP-QA-0502) were revised to

remove the exception that allowed temporary'monitoring equipment to be

installed for seven days without a safety evaluation.

~

A review was conducted to ensure that the Bypass Program required

evaluations for installation of temporary monitoring equipment would provide

sufficient barriers for concerns

raised in IR 95-25 and IN 95-13 (and IN 95-

13, Supplement

1).

~

Training was conducted for station engineering

personnel

regarding the

changes to the Bypass Program (deletion of the seven day allowance),

industry events,

IN 95-13, and the 10 CFR 50.59 evaluation requirements

for temporary changes.

~

Training was conducted for maintenance

production supervisors,

similar to

the training for engineers, to emphasis the procedural changes that require all

temporary monitoring instrumentation to be processed

as a bypass.

The inspector reviewed the actions taken in response to these issues, the safety

evaluation for the diesel generator temporary monitoring equipment and discussed

the Bypass Program review process with a cognizant engineer and NSE supervisor.

The inspector concluded that the reviews required by the Bypass Program provides

controls to ensure that the in-field configuration matches the approved

configuration, the reviews addressed

relevant design considerations,

and that there

is no impact on operation of equipment due to the installed monitoring

instrumentation.

PPS.L failed to perform a safety evaluation to support the installation of temporary

test equipment on the emergency diesel generators.

This failure constitutes

a

violation of 10 CFR 50.59, "Changes, Tests and Experiments."

This violation is

example (c) of VIO 97-03-04.

Conclusions

PPRL's failure to perform the safety evaluation required by 10 CFR 50.59 prior to

installing temporary test equipment on the emergency diesel generators

is a

violation. A subsequent

evaluation determined that there was no impact on

operability.

As part of the corrective action for this violation, changes to the

Bypass Program and Work Authorization System removed the inappropriate

exemption from performing safety evaluations that allowed the violation to occur.

0

32

IV. Plant Support

Conduct of Security and Safeguards Activities

Access Practices on Vital Doors

~Scc

e

Th'e inspector reviewed security surveillance activities intended to determine the

proper operation of site vital area access doors.

~Findin

e

Surveillance NS-SSP-004, Test Check and Inspection of Security Data System, and

NS-SO-004-1, Alarm Log, were reviewed in regards to a recent performance test.

PP&L decided to perform the test after questions were raised by the inspector

regarding the proper operation of doors and red indicating lights prior to personnel

access into vital areas.

The door alarm system was evaluated

by the inspector to perform as described

in

the SSES security plan.

The inspector determined that in some instances,

personnel

may be given the impression that they were inappropriately given access to vital

areas, based

on the illumination of red indicating door lights.

General employee

training requires

a practice of calling security after receiving two illuminated red

indicating lights following security door access attempts.

The practice of calling

security after receiving two red indicating door lights on sequential door key access

attempts may not be uniformly followed in the field. The inspector identified this

weakness in'the application of general employee training in the field. This

weakness

may have led to employees misunderstanding

the meaning of illuminated

red door indicating lights. However, this misunderstanding

does not bear on the

adequacy of the security system design nor the licensee's

implementation of the

security plan, both of which were determined to be adequate.

Conclusion

The licensee met the requirements of its security plan with respect to vital area door

access.

The licensee's

surveillance activities were carefully and well performed.

Some aspects of general employee training could be improved to make the

operation of the door alarm system clearer to plant employees.

No violations of

NRC requirements

were identified by the inspector.

Miscellaneous Fire Protection Issues

Review of UFSAR Commitments

A recent discovery of a licensee operating its facility in a manner contrary to the

UFSAR description highlighted the need for a special focused review that compares

plant practices, procedures

and/or parameters to the UFSAR description.

The

33

inspectors reviewed the applicable portions of the UFSAR and the SSES Fire

Protection Review Report (FPRR) that relate to the back up fire protection system.

Observations

and Findin s

The SSES FPRR, section 4.1, states that, in addition to the normal fire protection

system, the SSES site has a backup fire protection system which consists of a

2500 gpm diesel driven fire pump, a jockey pump and a dedicated water supply.

The 2500 gpm pump is not part of TS requirements

and is isolated from the main

yard loop.

The backup fire protection system and the normal fire protection system

can be cross tied.

The normal and backup fire protection systems

are isolated during routine standby

alignment.

On May 5, 1997, the licensee cross connected the systems and the

systems have remained in the cross connected

condition to the end of this

inspection report period (May 19, 1997).

The systems were cross connected to

prevent leakage through back flow preventer valve 022-528.

The valve had two

associated

deficiency tags attached to it when the inspectors traced parts of the

system in the field. The deficiencies (21690, 21045) were written on April 21,

1997 and May 4, 1997, respectively.

The licensee used operating procedure OP-013-003, step 3.4, to cross connect the

backup fire protection system with the normal fire protection system plant yard

loop.

The inspector walked down portions of the system and reviewed the

applicable procedures to determine if the backup system was operated

and

maintained in the same general condition as the normal fire protection system.

The

following procedures

were reviewed:

TP-013-026, 18 Month Function Test of Common Out Building

Sprinkler Systems

CL-013-031, Backup Fire Protection System Electrical

CL-013-032, Backup fire Protection System Mechanical

The inspector determined that the physical condition and procedural requirements of

the backup fire protection system were similar to those of the normal fire protection

system.

SSES calculation EC-013-0996 evaluated the basis and values for the backup fire

'rotection system jockey pump and diesel pump setpoints.

This calculation states

that the backup fire protection system shall operate as an integral part of the normal

fire protection system by acting as a supplemental

source of pressurized water.

The normal fire protection system provides sufficient water to satisfy design basis

requirements.

However, in an extreme case, the demand for pressurized water may

exceed design basis expectations.

Setpoints should be selected so that when the

demand exceeds

design basis expectations,

the cross-ties may be opened

and the

backup fire protection system brought on line as a supplemental

source of

pressurized

water.

This calculation does not address the cross connection of the

two systems for other than extreme conditions and does not address the routine

f

34

supply of keep-fill pressure to the backup fire protection system piping from the

normal fire protection system.

Based on the licensee description in the SSES FPRR and the three conditions used

to calculate system setpoints in EC-013-0996, the inspector determined that the

configuration, as described

in the FPRR and understood

by the NRC, is that the two

fire protection systems should normally be operated isolated from each other.

Further, according to the PPRL setpoint calculation, the cross-tied condition would

be reserved for extreme demand conditions - beyond design basis expectations.

The inspector determined that the normal and backup fire protection systems have

remained in the cross connected

condition without an adequate

safety evaluation.

This constitutes

a change to the normal fire protection system which is described

in

the FSAR and TS 3/4 7.6.

The change was not preceded

by an evaluation to

determine if an unreviewed safety question would result from the cross tie of the

two fire protection systems

and is a violation. This violation is example (d) of VIO

387/388/97-03-04.

c.

Conclusions

On May 5, 1997 the licensee cross connected the normal and backup fire protection

systems

and the systems have remained in the cross connected

condition at the

end of this report period.

This alignment constitutes

a change to the normal fire

protection system which is described

in the FSAR and TS 3/4 7.6.

The change was

not preceded

by an evaluation to determine if an unreviewed safety question would

result from the cross-tie of the two fire protection systems.

V. Mana ement Meetin s

Xl.

Exit Meeting Summery

The inspectors presented

the inspection findings for this report period to members of PP5L

management

at the conclusion of the inspection on May 20, 1997.

the licensee

acknowledged the findings presented,

with no exceptions taken.

No proprietary

information is included in this report.

~Oeoed

387,388/97-03-01

387,388/97-03-02

387,388/97-03-03

387,388/97-03-04

ITEMS OPENED, CLOSED, AND DISCUSSED

VIO

Two Examples of Inadequate

Procedures

(Refueling

Operations,

SPING Alarm Response

Procedure)

VIO

Failure to Perform Core Spray Surveillance Test Under

Controlled Conditions

URI

Omission of the Back Draft Isolation Dampers Function

in the SSES Maintenance

Rule Program

VIO

Four Examples of a Failure to Perform a Safety

Evaluation Prior to a Design Change

Closed

50-387/97-008-00

50-387/97-009-00

50-387,388/96-08-03

50-387,388/96-04-01

50-387,388/95-24-01

LER

Instrument Response

Time Testing

LER

Roving Fire Watch Rounds Not Completed On Time

URI

HELB Room Doors

IFI

Battery Charger Setpoints

URI

Temporary Monitoring Equipment

LIST OF ACRONYMS USED

AR

BDID

BTP

CFR

CR

EAL

ECCS

EDG

EP

ERT

EWAC

FPRR

FSAR

HELB

IN

LCO

LER

LP

NAS

NCV

NOV

NRC

NRR

NSE

PCO

PMT

QA

RFP

RHR

RIE

SALP

SOOR

SPING

SS

SSES

TS

UFSAR

US

WA

Alarm Response

Back Draft Isolation Dampers

Branch Technical Position

Code of Federal Regulations

Condition Report

Emergency Action Level

Emergency Core Cooling System

Emergency Diesel Generator

Emergency Preparedness

Event Review Team

Emergent Work Action Crew

Fire Protection Review Report

Final Safety Analysis Report

High Energy Line Break

Information Notice

Limiting Conditions for Operation

Licensee Event Report

Low Pressure

Nuclear Assessment

Services

Non-Cited Violation

Notice of Violation

Nuclear Regulatory Commission

Office of Nuclear Reactor Regulation

Nuclear System Engineer

Plant Control Operator

Post Maintenance Testing

Quality Assurance

Reactor Feed Pump

Residual Heat Removal

Replacement

Item Evaluations

Systematic Assessment

of Licensee Performance

Significant Operations Occurrence Report

System Particulate Iodine Noble Gas

Shift Supervisor

Susquehanna

Steam Electric Station

Technical Specification

Updated Final Safety Analysis Report

Unit Supervisor

Work Authorization