ML17264A845

Insp Rept 50-244/97-01 on 970105-0223.Violations Noted.Major Areas Inspected:Operations,Maint,Engineering & Plant Support

ML17264A845
Person / Time
Site:	Ginna
Issue date:	03/25/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I)
To:
Shared Package
ML17264A843	List: IR 05000244/1994029 ML17264A842 ML17264A845
References
50-244-97-01, 50-244-97-1, NUDOCS 9704020021
Download: ML17264A845 (50)

See also: IR 05000244/1997001

Text

O

U.S. NUCLEAR.REGULATORY COMMISSION

REGION

I

License Nos.

DPR-1 8

Report No.

50-244/97-01

Docket No..

50-244

Licensee:

Facility Name:

Location:

Rochester

Gas and Electric Corporation (RGSE)

R. E, Ginna Nuclear Power Plant

1503 Lake Road

Ontario, New York 14519

Inspection Period:

January 5, 1997 through February 23, 1997

Inspectors:

Approved by:

P. D. Drysdale, Senior Resident Inspector

C. C. Osterholtz, Resident Inspector

T. A. Moslak, Project Engineer,

DRP

G. S. Vissing, Project Manager,

NRR

L. T. Doerflein, Chief

Projects Branch

1

.

Division of Reactor Projects

970402002i

970325

PDR

ADQCK 05000244

8

PDR

0

lk

EXECUTIVE SUMMARY

R. E. Ginna Nuclear Power Plant

NRC Inspection Report 50-244/97-01

This integrated inspection included aspects of licensee operations,

engineering,

maintenance,

and plant support.

The report covers

a 7-week period of resident inspection.

In addition, it includes the results of announced

inspections by a regional project engineer

and the NRR Project Manager for the Ginna.Station.

~Oerations

Classroom training on Improved Technical Specifications

(ITS) implementation was

appropriate

in that it focused on an area that had been identified as a potential weakness

with operations

personnel.

The performarice of mock simulator examinations

at the

beginning of the training week was c'onsidered

a good practice in identifying specific

training concerns.

Also, the programs in place to ensure that all license'd personnel

receive

a biennial physical examination was considered

adequate.

The practice of extending

a

normal shift to meet programmatic requirements for operator proficiency has been

forwarded to NRR for further evaluation.

This item is unresolved

(URI 50-244/97-01-01).

The implementation of Priority "D" ACTION Reports to capture.low threshold incidents that

reflect possible negative performance trends was considered

a good initiative;

Nuclear Safety Audit and Review'oard (NSARB) issues were discussed'in

depth and all

.

members who participated demonstrated

a questioning attitude.

The members outside

RGRE nuclear operations were particularly active, utilizing their experiences

at other plants

to positively contribute to the discussions.

The QA/QC subcommittee

meeting provided good oversight of plant activities and audit

results.

The subcommittee

provided self-critical assessments,

focusing management

attention on the underlying root causes.

Positive measur'es

were incorporated to

improve'ata,

presentation.

The licensee had made an improvement in th'e documentation

of safety reviews, and a

heighten'ed

awareness

existed throughout'the

Ginna operation staff of the 10 CFR 50.59

process.

With one. exception, the safety review's adequately substantiated

that no

unreviewed safety question requiring a safety evaluation existed.

'aintenance

Overall, maintenance

and surveillance activities were performed effectively and were well

coordinated with all site organizations.

Maintenance technicians demonstrated

good

technical abilities and performed their work in accordance

with procedure requirements.

Maintenance

on the A-safety injection pump breaker was performed effectively. The

maintenance

technician. demonstrated

good technical abiljty arid performed the work in

accordance

with the procedure requirements.

The licensee's efforts to augment their

Executive Summary (cont'd)

inspection of Westinghouse

breaker contacts during routine breaker maintenance

were

appropriate.

The licensee appropriately responded

to plant fire alarms and effectively secured the

Control Rod Drive B-Motor Generator

(B-MG) Set.

The licensee also appeared

to identify

the cause of the MG set failure. However, the generator bearings may not h'ave received

adequate

lubrication, and the maintenance

procedure may not have provided sufficient

guidance to ensure that bearings

are adequately

lubricated.

The licensee stated that they intend to resolve excessive

vibrations in recirculation line

instrument tubing in the safety'injection system during pump operations.

The inspectors

considered this appropriate.

The licensee i'nitiated appropriate action in response

to the identified discrepancy between

the Improved Technical Specifications and the Updated Final Safety Analysis Report

regarding the operability of the auxiliary feedwater recirculation line..

Given the discrepancy between the acceptance

criteria basis assumptions,

and the ITS and

Updated Final Safety Analysis Report, the licensee's intention to re-evaluate the

acce'ptance

criteria for diesel generator compressor

check valve leakage was appropriate.

A number of discrepancies

were noted in the Training System Description.

However, the

licensee stated that the noted discrepancies

would be resolved.

h

The licensee stated that they intend to revise the emergency diesel generator periodic tests

to ensure that the proper service water pump line-up would be selected after testing was

completed.

The inspectors considered'his

appropriate.

. The licensee responded

expeditiously to secure unlocked QA storage areas.

The inspector

considered this condition t'o be an inadvertent oversight, and follow-up actions to evaluate

the processes

for modifying QA area boundaries

were appropriate.

~En ineerin

The license'e adequately

responded

to a noted increase

in reactor coolant system activity

levels and determined that the increases

were still well below the ITS limits. Continuous

monitoring of these levels for trends was appropriate.

The licensee's

aetio'ns to eliminate unnecessary

feedwater flow oscillations were effective.

No excessive oscillations occurred in.the A-train following adjustments

in its control

system.

The licensee's intention to incorporate the same change to the B-train channel

was appropriate.

The licensee took appropriate actions to correct the low temperature

conditions in the

screenhouse

bay, and to evaluate the consequences

for design basis conditions.

The

reanalysis for lower inlet temperatures

effectively demonstrated

that the consequences

of lower temperatures

would have a minimal impact under accident conditions.

A

Executive Summary (cont'd)

re-evaluation of the operability criteria for screenhouse

minimum water level and

temperature

monitor inaccuracies

at the minimum and maximum lake water inlet

temperatures

was appropriate.

Plant Su

ort

The failure to perform whole body counts on at least three in'dividuals terminated from

Ginna had minor safety significance and.represented

a failure of the licensee to meet an

administrative requirement.

The licensee had initiated adequate

corrective action.

Accordingly, this constituted

a violation of minor significance and is being treated as a non-

cited violation.

The licensee's actions to implement effective radiological work practices, to maintain

proper contamination boundary controls, and to monitor. to assure

radiological work

practices adequately prevent the spread of'ontamination have not been effective over

~

recent months.

The actions taken to identify and correct poor radiological work practices,

and to prevent the spread of potential and actual contamination have also not been

effective.

These conditions represent a'violation of 10 CFR 50, Appendix B, Criterion XVI,

"Corrective Actions," (NOV 50-244/97-01-02).

Ig

TABLE OF CONTENTS

EXECUTIVE SUMMARY

TABLE OF CONTENTS ..... ~.................., ~.....

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v

I. Operations

01

05

07

08

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

01.1

General Comments

01.2

Summary of Plant Status

Operator Training and Qualification ..'. ~... ~......

05.1

Licensed Operator Training Observations

and Qualification

Review .....

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Quality'ssurance

in Operations

.

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07.1

Problem Identification and Resolution

07.2

Nuclear Safety Audit and Review Board (NSARB) Meeting ..

07.3

Self- Assessment

Activities

07 4

Review of 10 CFR 50.59 Activities ..................

Miscellaneous Operations Issues.........,......'; .:.....

~

08.1

(Closed) LER 96-013: Circuit Breakers Closed While in Mode

Due to Personnel

Error, Resulted in Condition Prohibited by

'echnical Specifications;

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6

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~ I. Maintenance

M1

M2

~ M3

M7

Conduct of Maintenance

M1 ~ 1

General Comments on Maintenance Activities

M1.2

General Comments on Surveillance Activities ......

~

~

~ ..

Maintenance

and Material Condition of Facilities and Equipment

M2.1

A-Safety Injection (A-Sl) Pump Breaker Maintenance

M2.2

Control Rod Drive B-Motor Generator (B-MG) Set Failure ...

M2.3

C-Safety Injection (C-Sl) Pump Quarterly Test

Maintenance

Procedures

and Documentation ....

~ . ~........

~

M3.1

B-Auxiliary Feedwater

(B-AFW) Pump Periodic Test .......

M3.2

Diesel Generator Starting Air Compressor

Discharge Check

Valve Closure Test

. ~............ ~....,........

~ .

M3.3

A-Diesel Generator Perio'dic Test

. ~...: ~.............

Quality Assurance

in Mainte'nance Activities

M7.1

Unsecured

QA Material Storage Areas................

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.II. Engineering

E2

E8

Engineering Support of Facilities and Equipment .'..............

E2.1

RCS Total Gas Activity Increase

E2.2

Feedwater Flow Oscillations

E2.3

Temperature

Monitoring of Lake Ontario Inlet Water ~......

Miscellaneous

Engineering

Issues

E8.1

(Closed) LER 96-015: Based on Review of NRg Generic Letter 96-06, a Thermally Induced Overpressure

Transient Could

0 ccure

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IV. Plant Support ..................

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R3

Radiological Protection and Chemistry (RPSC) Procedures

and

Documentation

R3.1

Whole Body Counting for Terminated Employees

R7

Quality Assurance

in Radiological Protection and Chemistry

Activities

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R7.1

Poor Radiological Work Practices and Inadequate

Contamination Area Boundary Controls

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V. Management

Meetings ..

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X1

Exit Meeting Sum'mary............

L2

Review of UFSAR Commitments

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2.4

ATTACHMENT

Attachment

1 - Partial List of Persons

Contacted

- Inspection Procedures

Used

- Items Opened, Closed, and Discussed

- List of Acronym Used

I. 0 erations

01

Conduct of

Operations'1.1

General Comments

Ins ection Procedure

IP 71707

The inspectors observed

plant operation on a daily basis to verify that the facility

was operated

safely and in accordance

with licensee procedures

and regulatory

requirements.

This review included tours of the accessible

areas of the facility,

verification of engineered

safeguards

features

(ESF) system operability, verification

of proper control room and shift staffing, verification that the plant was operated

in

conformance with the Improved Technical Specifications

(ITS) and appropriate

action statements

for out-of-service equipment were implemented,

and verification

that logs and,records

accurately identified equipment status or deficiencies.

01.2

Summar

of Plant Status

The plant operated at full power (approximately 100%) for the majority of the

inspection period.

On'ebruary

11, 1997, power was reduced to approximately

96% for about three hours and returned to full power.

Plant power was again

reduced.to approximately 96% on February 12 for about five hours, and then

returned to full power.

These temporary powe'r reductions'ere

performed to

maintain minimum Improved Technical Specification (ITS) temperature

limits for

service water inlet temperature without exceeding the New York State requirements

for maximum differential temperature for lake water inlet and outlet temperature

(see section E2.3). Additionally, a failure of the control rod driv'e 8-motor generator

(B-MG) set occurred on'February

18, 1997.

Plant personnel

responded

appropriately and no plant transient resulted from the failure (see section M2.2).

05

Operator Training and.Qualification

05.'1

Licensed 0 erator Trainin

Observations

.and Qualification Review

a.

Ins ection Sco

e (71707)

The inspectors observed

a classroom training session

on Improved Technical

Specifications

(ITS) implementation and simulator training on a loss-of-heat-sink

and

a small break loss-of-coolant accident (LOCA). The. inspectors

also reviewed.

records pertaining to licensed operator medical and proficiency requirements.

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

reactor inspection report outline.

Individual reports are not expected to address

all outline

topics.

Observations

and Findin s

The inspectors attended

a classroom training session

on ITS implementation on

January 31, 1997.

ITS implementation by operations

personnel

had previously

been identified by the NRC as a potential weakness

(see Inspection reports (IRs)

50-244/96-11

and 50-244/96-05)

~ The training included analyses of the ITS bases

and was conducted interactively with operations

personnel

~

During the training

session, the instructor also addressed

questions from the operators concerning

specific plant scenarios

and applicable ITS requirements.

The inspectors

also observed simulator training for a licensed operating crew on a

loss-of-heat-sink

and a small break LOCA on February 3, 1997, the first day of the

training week.

The trainers initially administered

a practice examination on the crew

to identify any deficiencies that could be addressed

later in the week.

The

inspectors noted that all operator aid tags on the simulator were consistent with

tags in the plant's control room, and that the performance observations

recorded by

the licensee's training personnel were consistent with those identified by the

inspectors.

The inspectors reviewed medical records for all licensed personnel

and verified that

they all had received

a physical examination within the last two years as required by

10 CFR 55.53, "Condition of Licenses."

The inspectors

also reviewed the 1996

~ operations records for a sample of licensed operators to verify that they had

maintained their licenses active by standing

a minimum number of watches

as

specified in 10 CFR 55.53.

10 CFR 55.53(e) states,

in part, that, "To maintain

active status, the licensee shall actively perform the functions of an operator or

senior operator on a minimum of seven 8-hour or five 12-hour shifts per calendar

quarter."

The review indicated that all operators

had met the minimum number'of

total hours (56) required to maintain proficiency.

However, one instance was noted

during the second

calen'dar quarter'of 1996 where an operator stood one'regular"8-

hour shift, three regular 12-hour shifts, and then extended

a regular 8-hour shift to

12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> by relieving the watch 4 hours4.62963e-5 days <br />0.00111 hours <br />6.613757e-6 weeks <br />1.522e-6 months <br /> early.

His quarterly watches met the

. minimum of 56 hours6.481481e-4 days <br />0.0156 hours <br />9.259259e-5 weeks <br />2.1308e-5 months <br /> over five shifts, but it was not apparent whether the

regulatory requirements were being adequately satisfied under these circumstances.

Therefore, the inspectors forWarded theissue to The Office of Nuclear Reactor

Regulation

(NRR) for evaluation.

Conclusions

The inspectors concluded that tPe classroom training on implementing the ITS was

appropriate

in that it focused on an area that had been identified as a

potential'eakness

with operations

personnel.

The inspectors considered the performance of

mock simulator examinations at the beginning of the training week to be a good

practice in identifying specific training concerns.

Also, the programs in place to

ensure that all licensed personnel receive a biennial physical examination was

considered

adequate

in that no instance was found where the programmatic

requirements

were not met.

However, the practice of extending

a normal shift to

meet programmatic requirements for operator proficiency has been forwarded to

NRR for further ev'aluation.

Pending the results of the evaluation, this item remains

unresolved

(URI 50-244/97-01-01).

07

Quality Assurance in Operations

The inspectors

assessed

various licensee controls for identifying and correcting

problems.

This included a review of Revision 6 to IP-CAP-1, the problem reporting

procedure,

and attending

a Nuclear Safety Audit and Review Board (NSARB) QA/QC

subcommittee

meeting.

07.1

Problem Identification and Resolution

a.

Ins ection Sco

e (40500)

The inspector reviewed the licensee's

process for identifying, tracking, and

resolving problems.

b.

Observations

and Findin s

On November 11, 1996, IP-CAP-1, "Abnormal Condition Tracking Initiation or

Notification (ACTION) Report," was revised to include a lower thieshold of problem

reporting.

By management

request, ACTION Report initiation was desired for

. events or conditions that are of.very low risk, that do not require documented

investigation or corrective actions, but would provide useful precursor information if

tracked and trended for identification of repeat occurrence.

Such problems were to

be documented

with an ACTION Report and categorized

as Priority "D" by the

originator.

The PORC Chairman makes the -final determination on the appropriate

'ategory

during the daily report review.

Examples of items to be reported were

problems involving housekeeping,

administrative shortcomings,

and findings

identified duririg routine Shift Technical Advisor tours, the Supervisory Safety,

Housekeeping,

Administrative, Reliability, and Productivity (SHARP) Program tours,

and the Radiological Protection management

tours.

Management

has actively promoted use of the Priority "D" ACTION Reports through

an award/recognition

system.

The inspector examined the approximately three

dozen reports generated to date and determined that the problems described were

appropriately prioritized. The problems reflected minor lapses in the use,of 'personal

protective equipment, paperwork processing

discrepancies,

and abnormal plant.

equipment operating trends,.with no problem commonality evident.

The inspector considered the implementation of a Priority "D" ACTION Report to be

a good initiative to focus management

and'staff attention to low threshold, off

normal conditions

however, due to its-short history, the ariticipated benefits of this

initiative have not beeri realized in preventing, a more significant incident from

occurring.

c.

Conclusion

The inspector concluded that the implementation of Priority "D" ACTION Reports to

capture low threshold incidents that reflect possible negative performance trends

was a good initiative.

07.2

Nuclear Safet

Audit and Review Board

NSARB Meetin

a.

Ins

e tion Sco

e (40500)

The NRR Project Manager for the Ginna Station attended the NSARB quarterly

meeting held on January 29 and 30, 1997.

b.

Observations

and Findin s

h

The NSARB meeting was conducted

in accordance

with the Updated Final Safety

Analysis Report (UFSAR). The membership

included four individuals from outside

organizations,

as well as personnel from RGRE management.

Particular attention

was addressed. on the review of the draft letter in response to the.10 CFR 50.54(f)

letter of October 1996.

Other discussions

focused on p)ant performance,

Licensee

Event Reports (LERs), NRC inspections; the 10 CFR 50.59 process, the. proposed

modification to the spent fuel storage pool, the proposed

exemption to the

regulations relating to the low temperature

over-pressure

(LTOP) requirements,

and

the recent submittal to the NRC regarding the Probabilistic Safety Assessment.

The.

NSARB Chairman was particularly interested

in the feedback provided by outside

board'members.

Those items on the agenda

not discussed

were rescheduled

for

subsequent

meetings.

C.

Conclusions

The Project Manager concluded that the issues were discussed

in depth and all

members who participated demonstrated

a questioning attitude.

The members

outside RGKE nuclear. operations

appeared

particularly. active, utilizing their

experiences

at other plants to positively contribute to the discussions.

07.3

Self- Assessment

Activities

P

Ins ection Sco

e (40500)

The inspector assessed

the QA/QC subcommittee to the NSARB in evaluating audit

results and recommending

corrective actions.

b.

Observations

and Find n s

On January

16, 1997, the inspector attended

a meeting of the QA/QC

subcommittee to the NSARB. The meeting addressed

the status of open items from

past meetings,

a review of recently completed audit reports regarding maintenance

personnel training/qualification, the Cooperative Management Audit Program

(CMAP), and ACTION Report trending data.

The subcommittee

also addressed

the

results of a Human Performance

Enhancement

System

(HPES) evaluation relating to

an operator inappropriately closing certain safety-related

breakers

as a result of poor

communications

between the control room foreman and the operator.

The subcommittee's

discussions

were properly focused on safety and areas for

management/supervisory

improvement.

New methods of presenting ACTION

Report data to better direct attention on repetitive problems by Equipment

Identification Number/Cause

Code and by Organization/Cause

Code were discussed.

Additionally, management

directed that criteria be improved for selecting the "Top

Ten" ACTION Reports to better focus management

attention on the underlying

causes.

C.

Conclusion

The QA/QC Subcommittee

meeting provided good oversight of plant activities and

.audit results.

The subcommittee

provided self-critical assessments,

focusing

management

attention on underlying root causes.

Positive measures

were

incorporated to improve data presentation.

07.4

. Review of 10 CFR 50.59 Activities

a ~

Ins ection Sco

e (37001)

\\

The NRR Project Manager for the Ginna Station performed

a sample review of the

licensee's

10 CFR 50.59 activities.

b.

Observations

and Findin s

The Project Manager reviewed a sample of Safety Reviews of modifications

performed by the licensee

in accordance

w'ith IP-SEV-1, "Preparation,

Review and

Approval of Safety Reviews."

Additionally, since the licensee was in the process of

revising IP-SEV-1, the latest draft of this procedure was also reviewed.

It was

noted that several improvements were made, particularly in the appendices that

provided guidance

in. answering questions'that

determined if an unreviewed safety

question existed..

The Project Manager also reviewed

a selection of safety evaluations included in the

Annual Report of Facility Changes,

Tests, and Experiments Conducted Without Prio'r

Commission Approval, dated December 16, 1996.

These included:

SEV-1020, "Steam Generator Rigging and Handling"

SEV-1057, "18-Month Fuel Cycle"

SEV-1062, "Revision

1 to the.Technical Requirements

Manual"

SEV-1064, "MOV-4616 50% Open Throttling"

SEV-1020, SEV-1057, and SEV-1062 all appeared

to be comprehensive,

addressing

each of the three criteria for determining if an unreviewed safety question existed.

The Project Manager was concerned that SEV-1064 did not address the possibility

of degradation

occurring in MOV-4616 by operating it throttled at 50%.

However,

the licensee indicated that they had intended this modification to be temporary, and

the actuator for the valve was changed after approximately three months to allow

for full closure within its required time limit.

The licensee had performed an audit to assess

the adequacy of its 10 CFR 50.59

process.

The audit identified strengths

and weaknesses

and made

recommendations

accordingly.

Although the audit noted that a strength was in the

training of preparers

and reviewers, it recommended

related refresher training and

training to PORC and NSARB members regarding their specific unreviewed safety

question review responsibilities.

Other recommendations

included revising training

materials to be consistent with 10 CFR 50.59 requirements

and NRC guidance,

establishing

a process for self assessment,

and reviewing current practices

regarding the designation of inconsequential

changes.

C.

Conclusions

Based on previous inspections conducted

over the past year and interviews of

licensee personnel during this visit, the Project Manager concluded that the licensee

had made an improvement in the documentation

of safety review's.

A heightened

awareness

also existed throughout the Ginna operation staff of the 10 CFR 50.59

.process.

With the exception of SEV-1064, the safety reviews adequately

substantiated

that no unreviewed safety question requiring a safety evaluation

existed.

The Project Manager considered

one weakness

in SEV-1064 was in not

identifying the modification as temporary.

08

Miscellaneous Operations Issues (92700)

08.1

Closed

LER 96-01'3: Circuit Breakers Closed'While in Mode 3

Due to Personnel

Error Resulted in'ondition Prohibited b

Technical 'S ecifications.

LER 96-013 was submitted on November 27, 1996, as a result of an improper

closure of breakers for safety injection (Sl) pump discharge valves prior to entry into

MODE 4 during a reactor plant cooldown (see IR 50-244/96-11).

The breakers

were shut for a short period of time with the plant about 3'F above the maximum

average

RCS temperature

(Tave) allowed by the ITS for entry into MODE,.4

'<350~F).

However, this resulted in two trains of the emergency core cooling

system

(ECCS) being considered

ino'perable, which required entry into ITS limiting

condition for operation

(LCQ) 3.0.3.

'he

LER attributed the underlying cause of 'this event to personnel

error, resulting

from less than adequate

verbal communications

between the control room and in-

plant operators,

and the use of unclear supervisory methods.

The licensee

performed a Human Performance

Enhancement

System

(HPES) evaluation in.

response to this event.

The evaluation reinforced the causes

identified in the LER,

and recorr)mended that procedural, steps performed by operators

outs(de the control

room be more formally controlled.

The HPES recommended

that labels'be attached

to the Sl pump discharge

breakers to indicate the conditions under which they may

be operated.

Administrative procedure A-52.1, "Shift Organization and Responsibilities," was

revised to indicate that local operations affecting reactivity or manipulation of

safeguards

equipment shall be communicated to the control room just prior to

performance.

The licensee also indicated that condition labels would be attached to

the Sl pump discharge breakers.'he

inspectors considered that the LER and HPES

adequately described this event,.and

appropriately addressed

the root causes

and

corrective actions.

This LER is closed.

II. Maintenance

M1

Conduct of Maintenance

M1.1

General Comments on Maintenance Activities

a 0

Ins ection Sco

e 62707

The inspectors observed portions of plant maintenance

activities to verify that the

correct parts and tools were utilized, the applicable industry codes and,technical

specification reqUirements were satisfied, adequate

measures

were in place to

ensure personnel safety and prevent damage to plant structures, systems,

and

components,

and to ensure that equipment operability was verified upon completion

of post-maintenance

testing.

b.

Observations

and Findin s

the inspectors observed portions of the following maintenance

activities:

A-Safety Injection (A-Sl) pump breaker maintenance;

observed

on February

4, 1997 (see section M2.1)

Control rod drive B-motor generator

(B-'MG) set maintenance

performed in

response

to equipment failure~ observed from February 18 - February 21,

1997 (see section M2.2)

c.

Conclusions

Overall, the inspectors concluded that the observed

maintenance

activities were

performed effectively and were well coordinated between the applicable site

organizations,

Maintenance technicians demonstr'ated

good technical abilities and

performed their work in accordance

with procedure requirements.

4

8

M1.2

General Comments on Surveillance Activities

a.

Ins ection Sco

e 61726

The inspectors observed

selected surveillance tests to determine whether approved

procedures

were in use, details were adequate,

test instrumentation was properly

calibrated and used, technical specifications limiting conditions for operation

(LCOs)

were satisfied, testing was performed by qualified personnel, test results satisfied

acceptance

criteria or were properly'dispositioned,

and correct post-test system

restoration was performed.

b.

Observations

and Findin s

The inspectors observed

portions of the following surveillance tests:

~

PT-16Q-B, "AFW Pump B - Quarterly," observed

on January 31, 1997 (see

section M-3.1)

PT-2.1-0, "Safety.injection System Quarterly Test," was performed on the

C-Sl pump and observed

on February 5, 1997 (see section )VI-2.3)

PT-12.7, "Diesel Generator Starting Air Compressor

Discharge Check Valve

Closure Test," observed

on February 11, 1997 (see section M-3.2)

~

PT-12.1, "Diesel Generator A," observed

on February 19, 1997 (see section

M-3.3)

w

~

PT-12.2, "Diesel Generator B," observed

on January 27, 1997.

t

c.

Conclusions

Overall, the inspectors concluded that the observed

surveillance activities were well

controlled and coordinated with other site organizations.

Testing was performed in

accordance

with procedure requirements

and technicians demonstrated

a good

understanding

of the functional'requirements

of the equipment being tested.

M2

. Maintenance and Material Condition of Facilities and Equipment

I

l'

M2.1

A-Safet

In'ection

A-Sl Pum

Breaker Maintenance

a.

Ins ection Sco

e (62707)

The inspectors observed

and reviewed routine maintenance

associated

with the

A-Sl pump breaker.

b.

Observations

and Findin s

On February 4, 1997, the inspectors observed

routine maintenance

on the A-Sl

pump breaker in accordance with procedur'es

M-32.1.50, "DB-50 Circuit Breaker

Maintenance,"

and M-32.8, "Installation and Testing of Amptector Overcurrent

Devices for DB-25, DB-50, and'DB-75 Westinghouse Breakers."'he

technician,

performing the maintenance

noticed that two of the breaker's secondary contacts

were slightly deformed, and exhibited some slight bowing.

Additionally, these two

contacts

had much less tension (more play) when manually pressed.

The technician

replaced the two contacts with new ones.

Breaker contact deformity had recently

occurred

in other Westinghouse

breakers

in the plant (see IRs 50-224/96-11

and

50-244/96-12), but the technician indicated to the inspectors that it had not been

a

problem prior to those instances.

Other maintenance

personriel present indicated

that they were working o'n a method to verify breaker contact operability by

comparing exi'sting contacts to a predetermined

standard.

C.

Conclusions

The inspectors concluded that the breaker maintenance

was performed effectively.

The maintenance

technician performing the work demonstrated

good technical

ability in identifying the beaker contact deformity and performed the work in

accordance

with the procedure requirements.

Given the recent failures in these

components,

the inspectors

also concluded that the licensee's efforts to augment

their inspection of Westinghouse

breaker contacts during routine breaker

maintenance

were appropriate.

M2.2

a 0

Control Rod Drive B-Motor Generator

B-MG Set Failure

1

'Ins ection Sco

e (62707)

'he inspectors observed

arid reviewed corrective maintenance

associated

with a

.failure of..the control rod drive B-MG set.

b.

Observations

and Findin s

At 1:45 a.m. on February 18, 1997, the control rod drive B-MG set,was secured

due to smoke from the generator of the B-MG setting off area fire alarms.

The fire

brigade responded,

the apparent fire was extinguished,

and'the equipment secured

in approximately 10 minutes.

The licensee initially attributed the problem to bearing

failure in the generator since intermittent abnormal noises had been reported by

plant personnel

during the preceding two weeks.

After the failure, the MG was

.

dismantled and sent to an offsite repair facility the following day.

The generator

bearings were replaced and the MG was returned to the site on February 21, 1997.

At the end of the report period, the licens'ee indicated that the MG should be

returned to service early the'next week.

The inspectors reviewed Maintenance

Procedure M-1015, "Three Month Lubrication

and Maintenance

Inspection of Rotating Equipment," which provided the following

guidance for generator bearing lubrication:

"Grease Generator while running, if possible...NOTE:

DO NOT overgrease,

use discretion."

Other equipment lubricated as part of M-015 included the MG set motor bearings,

the component cooling water (CCW) pump bearings, the service water (SW) pump

'motor bearings,

and the instrument air (IA) compressor motor bearings.

The

licensee indicated that they intended to do a thorough root cause analysis in

response

to this failure.

C.

Conclusions

The inspectors concluded that the licensee appropriately responded

to the fire

alarms and effectively secured the MG set, as evidenced

by the absence

of a plant

transient, and the short period of time between fire alarm initiation and the

equipment being secured.

The licensee also appeared to identify the component

failure in the MG set.

However, the inspectors were concerned that the generator

bearings may not have received adequate

lubrication, and that the maintenance

procedure to lubricate bearings

in rotating equipment may not provide sufficient

guidance to ensure that bearings are adequately lubricated.

This was of particular

concern because

M-1015 provided similar guidance for lubrication of engineered

safety features

(ESF) equipment.

M2.3

C-Safet

In'ection

C-Sl

Pum

Quarterl

Test

a.

Ins ection Sco

e (61726)

The inspectors observed

and reviewed PT-2.1-Q, "Safety Injectioh System Quarterly

Test," for the C-Sl pump.

b.

Observations

and Findin s

\\

This periodic test was performed on February 5, 1997 to demonstrate

operability of

the C-Sl pump.

The test included verifications of pump differential pressure

and

discharge pressure,

discharge check valve operability, Sl recirculation line

operability, and pump vibration limits. During performance of the PT while the

pump was running, the inspector identified that tubing to pressure indicator Pl-915

(Sl pump recirculation flow) was vibrating and the indication on the gauge was

oscillating significantly. The inspector informed licensee test personnel that the

.vibration appeared

excessive.

Design engineering

personnel subsequently

indicated

that they intended to analyze the problem for resolutioh.

11

Conclusions

The inspectors concluded that the surveillance was performed in accordance

with

procedure

requirements

and that technicians demonstrated

a good understanding

of

the functional requirements of the equipment being tested.

The Jicensee's

intention

to resolve excessive vibrations in recirculation line instrument tubing during pump

operations was appropriate.

M3

Maintenance Procedures

and Documentation

M3.1

B-Auxiliai

Feedwater

B-AFW Pum

Periodic Test

Ins ection Sco

e (62716)

The inspectors observed

and reviewed portions of PT-16Q-B, "AFW Pump B-

Quarterly."

b,

I

'I

PT-16Q-B, "AFW Pump,B - Quarterly" was performed on January 28, 1997, for the

inservice test requirements

under the ASME Code,Section XI. Pump performance

was satisfactorily tested.

However, the pump's recirculation valve, AOV-4310,

failed to fully open when the pump was throttled to 40 gpm output flow per local

indication as required. by the PT. Therefore, on January 31, 1997, the PT was re-

performed in part to calibrate AOV-4310., The inspectors questioned the licensee's

practice of performing instrument calibrations as part of a periodic pump test.

However, the licensee indicated that the AFW recirculation line was not required for

pump operability in accordance

with the Improved Technical Specifications (ITS).

The inspectors reviewed the Updated Final Safety Analysis Report (UFSAR) and"

noted that the function of the pump's recirculation flow was inconsistent with

the'TS

operability requirements.

Section 10.5-8 of the UFSAR stated that "The motor-

driven and turbine-driven pumps each have an automatically controlled minimum

flow recirculation system sized and periodically tested.to ensure that minimum flow

will be provided under all accident and'normal operating conditions'to prevent pump

damage from overheating."

The ITS did not have this requirement.

The licensee

generated

an ACTION Report (97-0284) to address the inconsistency.

C.

Conclusions

The inspectors concluded that the surveillance was performed in accordance

with

plant procedures

and that the technicians demonstr'ated

a good understanding

of

the functional requirements

of the equipment being tested.

The licensee initiated

appropriate action in response to the identified discrepancy between the ITS and the

UF,SAR.

0

M3.2

12

Diesel Generator Startin

Air Com ressor Dischar

e Check Valve Closure Test

a 0

Ins ection Sco

e (61726)

b.

The inspectors observed

and reviewed the performance of PT-12.7, "Diesel

Generator Starting Air Compres'sor

Discharge Check Valve'Closure Test."

Observations

and Findin

s

PT-12.7 was performed on February 11, 1997, to verify the diesel air start receivers

would not excessively depressurize

if the air compressor

separated

from its

discharge

line during a seismic event.

A check valve is installed in the air line

between the receivers and the compressor to prevent receiver depressurization.

The PT's acceptance

criteria for check valve leakage was 5 psig/min, which was

based

on an assumed

normal receiver operating pressure of 250 psig to provide

.

5 minutes for operator action to manually start the EDG before the air pressure

reached the minimum 225 psig required for. operability.

During performance" of the

PT, actual compressor check'valve leakage for both diesel generators

w'as 0 psig

over a test period of 3 minutes.

The inspectors reviewed the design basis requirements

regarding the diesel

generator air start system, consulting the UFSAR, ITS, and the Standard

Review

Plan (SRP).

The 'UFSAR specified that the compressors

will automatically start at

230 psig and automatically stop at 250 psig to keep the receivers charged.

This

appeared to be inconsistent with the assumption of 250 psig normal operating

pressure

used by the li'censee in the acceptance

criteria basis for check valve

'leakage.

The licensee indicated that they intended to re-evaluate this acceptance

criteria since 250 psig is not.the normal operating pressure,

and 5 minutes would.

only represent the maximum time available for operator action before re'aching

225 psig.

Section 9..5.6.II.4.g of the SRP (General Design Criterion 17), specified that, "As a

minimum, the air starting system should be capable of cranking a cold diesel engine

five times without charging the receivers."

Section 9.5.6 of the UFSAR, "Diesel

Generator Starting System," stated that "Testing performed'during the 1992 MODE

6 (Refueling) outage demonstrated

that with an initial pressure of 225 psig the air

start system receiver tanks was sufficient to crank the diesels for a minimum of five

starts without recharging the receivers...."

Howev'er, Section 3.8.1 of the ITS

bases stated, "Since the DGs must start and begin loading within 10 seconds,

only

one air start must be available in the air receivers as assumed

in the accident

analysis."

The inspectors

also reviewed Training System Description (TSD) RGE-S, "Standby

Generation Diesel Generators,"

and noted'the following discrepancies:

~

The TSD incorrectly stated that, "The compressors

will automatically start

at 220 psig to charge the receivers."

The compressors

actually start at

230 psig.

13.

~

The TSD incorrectly indicated

a "...normal receiver pressure of 250."

Normal

receiver pressure

actually varies continuously between 230 psig'and 250

pslg.

~

The TSD incorrectly stated that, "A relief valve set at 265 psig provides

overpressure

protection."

This was inconsistent with the UFSAR

requirement and actual setpoint of 275 psig.

Although not controlled documents,

the TSDs are used as credible reference

documents for training, and are widely used by personnel throughout the plant, and

are regarded

as essentially accurate.

C.

Conclusions

~ the licensee stated that'the note

'3.3

A-Diesel Generator Periodic Test

The inspectors concluded that the surveillance was properly performed in

accordance

with plant procedures

and that the technicians demonstrated

a good

understanding

of the functional requirements of the equipment being tested.

Given

the discrepancy between the acceptance

criteria basis assumptions

and the ITS and

the UFSAR, the inspectors considered that the licensee's intention to re-evaluate

the acceptance

criteria for diesel generator, compressor check valve leakage was

appropriate.

The inspectors were concerned with the number of discrepancies

noted in the TSD, since these documents

are widely used at the station.

However,

d discrepancies

would be revised.

a.

Ins ection Sco

e (61726)

The inspectors observed

and reviewed the performance of PT-12.1, "Diesel

Generator A."

b.

Observations

and Findin s

PT-12.1, "Diesel Generator, A," was performed on February 19, 1997, to test the

A-Emergency Diesel Generator (A-EDG) for operability.

The test verified that the

'-EDG would start from the control room and that its associated

feeder breakers to

the plant's safeguard

buses would shut.

The inspector noted that step 6.4 of,the

test procedure directed the operators to start or verify that the service water (SW)

pump currently selected for automatic safeguards

start was running.

This was to

prevent an automatic pump start when the diesel-generator

supply breaker to Bus

~ 18 was closed.

The operators placed the safeguards

selector switch from the

'-'SW

pump to the C-SW pump, since the C-SW pump was currently running.

The

'T

was then continued and the diesel-ge'nerator

was tested satisfactorily.

However,

the PT gave no direction to the operators to return the safeguard

selector switch.to

its original line-up.

The desirable line-up for automatic actuation of SW pumps was to have the non-.

running pump selected.

This line-up was preferred due to the possibility that the

14

running pump could trip on overcurrent if an undervoltage

condition preceded

a loss

of offsite power.

This would prevent the pump from restarting on an engineered

safety feature (ESF) actuation.

After the PT concluded, operations

personnel

returned the SW ESF selector switches back to their original line-up without

procedural guidance.

The licensee indicated that they intended to revise procedures

PT-12.1

(A-EDG) and PT-12.2 (B-EDG) to direct the safeguard switches be placed in

their pre-test positions following EDG testing.

C.

Conclusions

The inspectors concluded that the surveillance was performed in accordance

with

plant procedures

and that the technicians demonstrated

a good understanding

of

the functional requirements of the equipment being tested,

The inspectors

considered the licensee's intention to revise the EDG PTs was appropriate to ensure

that the proper SW pump line-up would be'selected'after

testing was completed

M7

'Quality Assurance in Maintenance Activities

M7.1

Unsecured

QA Material Stora

e Areas

a 0

b.

Ins ection Sco

e (40500)

t

The inspector review'ed the licensee's follow-up response to two QA material

storage areas in the plant that were discovered unlocked.

I

Observations

and Findin s

On February 2, 1997 (Sunday), the inspector discovered

an unlocked access

door

to the plant's QA stockroom.

Upon entering'the area, no sign of a forced entry was

apparent

and no contents of the room appeared to be disturbed.

No person was

seen inside, nor did any stockroom personnel

appear to be on duty that day.

The

inspector notified an auxiliary operator and the shift supervisor, who dispatched

the control room foreman (CRF) to lock the door.

The CRF locked the door within

10 minutes, but also found a door ajar to an adjacent QA storage enclosure.

Operations

personnel

later performed

a walk-through in the. stockroom and agreed

that nothing had been disturbed.

The following day, stock room personnel

performed

a thorough walk-through of both aieas.

Most of the QA inventory is

stored in separately locked cabinets,

and none had been disturbed.

A stockroom

'mployee

apparently left the stockroom door unlocked.

The return spring o'n the

QA storage enclosure door was weak and not able to fully latch the door when it

closed on its own.

The doors became part of the stockroom when the floor plan was modified in

February 1996 to expand the total area.

This expansion

opened

an exterior wall to

allow for another window counter, and.created

additional office space.

The area

. that was newly incorporated

included the two doors that then became additional

.

ent/ances to the stockroom.

However, the latch on one of the new entrance doors

w'as not modified to make it like the other stockroom doors which could not be

15

permanently unlocked with a key.

The following day, the lock on the stockroom

door was upgraded to the desirable type, and.the spring on the storage enclosure

door was replaced.

'I

The licensee initiated an'ACTION Report to evaluate their process for modifying the

boundaries

of designated

controlled QA storage areas,

and to determine what QA

management

reviews are appropriate for such modifications.

Although the

stockroom is temperature

and humidity controlled, it is officiallycontrolled as a level

"B" storage area (i.e., temperature

only) ~ 'It was not apparent'that this was

adequately

considered

during the modification process to ensure that the proper

level of storage was maintained.

The licensee's evaluation was in progress at the

end of the inspection period.

C.

Conclusions

The licensee responded

expeditiously to secure the unlocked QA storage areas.

The inspector considered this condition to be an inadvertent oversight, and follow-

up actions to evaluate the licensee's

processes

for modifying QA area boundaries

were appropriate.

III. En ineerin

E2

Engineering Support of Facilities and Equipment

E2.1

a 0

b.

RCS Total Gas Activit Increase

Ins ection Sco

e (37551)

'I

The inspectors reviewed the licensee's

response to an increase

in reactor coolant

system

(RCS) total gas and Iodine-1,31

(1-131) gas activities.

Observations

and Findin s

4

On January'13,

1997, the licensee observed

a notable increase

in RCS total gas

activity from an average of 0.19 micro curies per gram (pCI/gm) to 0.424 pCI/gm.

Iodine-131 activity increased from 4.15 E-3 pCI/gm to 5.05 E-3 pCI/gm.

The

licensee initiated an ACTION Report (97-0042) in response to these increases.

Based on a computer analysis, the licensee determined that the inciease was most

likely caused

by a defect in one fuel pin: The ITS limits for activity were 1.0 yCi/gm

for l-131, as well as 100/EyCI/gm (=58 yCI/gm) for RCS gross specific activity

(corresponds

to 1% failed fuel). The licensee indicated that they intend to monitor

activity levels closely for changes

or trends.

C.

Conclusions

The inspectors concluded. that the licensee adequately

responded to the noted

increase

in RCS activity levels and that the increases

were still well below ITS

limits. Continuous monitoring of these levels for trends was appropriate.

EZZ

Feedwater Flow Oscillations

16

Ins ection Sco

e (37551)

The inspectors reviewed the licensee's

response to excessive flow oscillations in

the main feedwater system.

Observations

and Findin s

Main feedwater flow oscillations had been occurring in the A-train of the feedwater

system since the reactor start up on November 12, 1996.

The A-train oscillations

occurred at irregular intervals an'd varied from between 200-300 Klbm/hr (normal

flow oscillations were between 60-80 Klbm/hr). The oscillations could be curtailed

by taking manual control of the feedwater regulating valve (FRV) and placing the

'RV

bypass valve in automatic.

The licensee indicated that the FRV bypass, valve

actuators were faster acting to feedwater system inputs, which allowed them to

respond more efficiently to system demands.

However, the licensee also indicated

that the FRV bypass valves did'not have redundant controller inputs in case of an

input failure, but that the FRVs did. Therefore, it was more desirable to analyze the

FRV actuators for possible adjustment. and.resolution.

The licensee's resolution was to reduce the gain on the A-FRV actuator by 6.5%.

~ This value was selected to reduce the sensitivity of the system to prevent the

swings from occurring, but to not reduce sensitivity to the point where the FRVs

would respond too slowly to a plant transient.

The adjustment was made on

January 25, 1997, and immediately reduced oscillations to 60-80 Klbm/hr.

No

excessive flow oscillations were noted throughout the rest of the report period.

The

licensee indicated that they intend 'to perform the same adjustment to the B-FRV in

the near future.

C.

Conclusions

. The inspectors concluded that the licensee's

actions to eliminate unnecessary

feedwater flow transients were effective in that no excessive

oscillations in the.

A-train feed flow had occurred since January 25, 1997.

The inspectors

also

concluded that the licensee's intention to incorporate the same change to the

B-train channel was appropriate

and would make both channels consistent with

each other.

E2.3

Tem erature Monitorin

of Lake Ontario Inlet Water

a 0

Iris ection Sco

e (37551)

The inspectors reviewed the licensee's

actions to evaluate discrepancies

between

the various lake inlet water temperature detectors,

and to ensure Compliance with

the ITS requirements for minimum screenhouse

inlet plenum temperature.

17

Observations

and Findin s

'he

licensee performed cold weather walkdowns on a weekly basis during the.

winter months to ensure that plant systems

are not subjected to freezing

.temperatures

and that the lake inlet water is free from ice that. could block flow'to

the circulating water (CW) and service water (SW) systems.

Administrative

procedure A-54.4.1, "Cold Weather Walkdown Procedure,"

directed plant operators

.,

to maintain a heightened

awareness

for the possible presence

of "frazil" ice in the

.

intake structure, arid to energize the intake heater voltage when lake temperature

is

less than 33

F.

ITS Basis 3.7.8 required that the screenhouse

bay temperature

be

maintained above 35'F in order to satisfy an accident analysis assumption that the

service water inlet temperature to the containment recirculation fan coolers (CRFCs)

be above the minimum temperature

used in the analysis.

Operators adjusted the

service water inlet temperature with warm recirculation flow from the circulating

water discharge that is mixed with the lake inlet water in the screenhouse

inlet

plenum.

Adjustments to the recirculation flow were made periodically to maintain a

desired service water inlet temperature

above 38

F.

During the week of January

13 -'17, 1997, the inspector noted that the plant

computer's daily recorded lake inlet temperatures

indicated betw'een 30 - 3,1

F;

however, no visual indication of ice was present on the lake or in'the'screenhouse

bay.

In addition, it was noted that the three installed instruments used for lake

temperature

monitoring did not display consistent indications, and varied over a 3-

4

F range.

The inspector questioned

the licensee about which instrument was the

most accurate, which was used for moriitoring potential icing conditions in the inlet

water, and which instrument is used to ensure that the screenhouse

bay

temperature

is maintained above 35

F.

Operators indicated that they used the one detector that inputs to the plant

computer for lake temperature

as an indication for when to regulate recirculation

flow, and they used the temperature detectors on the inlet to the condenser

waterboxes to keep the screenhouse

bay temperature

above 38'F.

A condenser

inlet temperature

above 38

F had been assumed to provide a sufficient margin to

ensure that service water inlet temperature

remained above 35'F.

The licensee

investigated the variation in temperature

indications from the three different

detectors

and determined that they had an accuracy of +2% of full calibration

range.

For the three detectors,'this tolerance could allow indications to vary from

30 to 34

F.

Based on these determinations,

it appeared that two of the

temperature detectors were reading lower than actual lake temperature.

However,

the licensee recalibrated all'three instruments to make them more consistent.

All

three of the temperature

detectors. are mounted on the north wall of the

screenhouse

close to the lake inlet flow stream before it mixes with the warm

recirculation flow. Consequently,

these detectors

read closer to the actual lake

temperature.

There is currently no temperature detector in the screenhouse

that

measures

the lake inlet mixed with the recirculation water that enters the CW and

SW inlet bays.

18

In order to obtain a more accurate indication of SW pump inlet temperature,

the SW

system engineer and operations personnel

began taking manual temperature

readings

in the SW bay using a high accuracy hand-held instrument.

On January

31, 1997, the licensee determined that some of the SW inlet water was below

35

F.

It appeared that a'"streaming" effect caused colder water in the middle of

the inlet bays to stream toward the SW pump suction.

Operators immediately

opened the recirculation gatq to increase the inlet pl'enum temperature,

and SW inlet

water was restored to above 35. F within one-half hour.

Operators later responded

to a similar occurrence

on February 11, 1997, and restored SW inlet temperature to

above 35

F within approximately 20 minutes.

Subsequent

evaluation by the licensee indicated that ITS surveillance requirement 3.7.8.1 had not been satisfied when the SW inlet temperature

dropped below 35

F,

and that represented

a condition requiring entry into ITS limiting condition for

operation

(LCO) 3.0.3.

Although a formal entry into LCO 3.0.3'was required on

both occasions,

the licensee's corrective actions were accomplished

in much less

time than the time limits permitted (6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br />) for these conditions.

However, the

licensee determined that this situation was reportable to the NRC under the

requirements of,10 CFR 50,73, and initiated a Licensee Event Report (LER). At the

end of the inspection,'the

licensee anticipated the LER would be submitted in early

March 1997.

In an effort to evaluate the consequences

of th'e lower SW'temperature

on accident

conditions, the licensee's.engineering

and licensing departments

consulted with

Westinghouse,

who subsequently

recalculated the peak containment pressure

and

temperature

assuming 30

F service water inlet to the CRFCs.

The results indicated

that the lower SW temperature

Would have a minimal impact.

Since the SW piping

design analysis and s>stem design minimum terriperature is 32

F, RGRE changed

the ITS basis to a minimum temperature

of 32'F.

The inspector reviewed the

licensee's safety evaluation (SEV-1090) that supported the ITS basis change

and

found it adequate

based

on the results of the recalculated accident analysis.

A

Safety evaluation was done to support

a change to the ITS Basis to permit the

operation of SW inlet water to drop to 32

F. Also, operations procedure 0-6.13,

"Daily Surveillance Logs," was changed to direct operators to take manual readings

of the screenhouse

bay to ensure that temperatures

are maintained at )32

F.

4

The inspector'also

noted that the ITS basis for SW pump operability stated that the

pumps are operable with only 5 feet indicated water level in the screenhouse.

However, the bottom of the pump suction is exactly at the indicated 5 foot level ~

When the inspector questioned this, the licensee agreed to evaluate the

appropriateness

of this operability requirement.

The, licensee also indicated that

they intend to evaluate the possibility that inaccuracies

allowed in the screenhouse

temperature

monitors could also represent

a potential problem for lake inlet

temperature

monitoring during the summer months'when the maximum lake inlet

temperature

is limited to,80

F by the current design basis analysis.

0

C.

Conclusions

The licensee took appropriate actions to correct the low temperature

conditions in

the screenhouse

bay, and to evaluate the consequences

for design basis conditions.

The reanalysis for lower inlet temperatures

effectively demonstrated

that the

consequences

of 32

F service water would have minimal impact under accident

conditions.

A reevaluation of the operability criteria for screenhouse

minimum

water level and temperature

monitor inaccuracies

at the minimum and maximum

lake water inlet temperatures

was appropriate.

ES

Miscellaneous Engineering Issues

E8.1

Closed

LER 96-015: Based on Review of NRC Generic Letter. 96-06

a Thermall

Induced Over ressure Transient Could Occur

'

LER 96-015 was submitted on January 22, 1997, in response to the licensee's

~ discovery that the possibility existed for a thermally induced overpressurization

that

could fail the charcoal filter dousing lines between the check valves in the cross-

connect piping for the A-.and B-containment spray (A- and B-CS) system headers.

This failure could occur from a heatup of the water-filled line inside containment

following a design basis loss-of-coolant accident:

Such an overpressurization

could

render both trains of CS incapable of performing their safety function because

of

the excessive flow diversion from the CS system through the break.

The dousing

function was not credited in the Ginna Accident Analysis, and the licensee isolated

and vented the line on December 21, 1996.

On December 23, 1996, the licensee

completed calculations confirming that a thermally induced overpressurization

in the

dousing line was possible,

and reported the condition to the NRC under 10 CFR 50.72. (b)(2)(iii)(D) as a condition that could prevent a safety function from

~ mitigating the consequences

of an accident.

On January 3, 1997; the licensee installed a thermal relief valve in the containment

spray charcoal filter dousing lines to prevent a thermal over-pressurization

of the

dousing lines; The relief valve was set for a maximum relief cap'acity of 10 gpm at

500 psig to prevent pressure fram exceeding the rated maximum piping design

pressure.

Aftei the relief valve 'was installed, the dousing line was unisolated.

The

irispectors considered that the LER adequately

described this potential event, and

~ appropriately addressed

the root causes

and corrective actions.

This LER is closed.

IV: Plant Su

ort

R3

Radiological Protection and Chemistry (RP8cC) Procedures

and Documentation

R3.1

Whole Bod

Countin

for Terminated

Em lo ees

Ins ection Sco

e

(83750)

The inspector reviewed the licensee's implementation of site procedures

regarding

conducting whole body conducting for terminated employees.

20

b.

Observation

and Findin s

The inspector reviewed licensee activities related to conducting whole body counts

(WBC) for individuals terminating their activities at Ginna.

Interviews were

conducted with the Manager, Chemistry 5 Radiological Controls and the Manager,

Dosimetry, applicable procedures

were reviewed, and dosimetry records of recently

terminated individuals were examined.

Licensee procedure

"Dosimetry Program

Administrative Procedure",

RPA-DOS, defines the requirements for internal and

external dose processing

and recording.

Paragraph

5.3.7, "Whole Body Counting

Frequency" states in part,.

~ ~"TLD badged individuals shall receive a whole body

count under the following conditions:

~ Upon termination of work activities at Ginna."

Through review of dosimetry records for individuals terminated in 1996, the

inspector determined that WBC.were not conducted

in compliance with the

procedural requirement.

Specifically, at least three individuals terminated their work

activities in 1996 at Ginna but did not have a whole body count as required by the

procedure..

However, evidence exists, in the form of'radiation w'ork permit. entries,

airborne survey data, and dates when the individuals had their an'nual whole body

count, that these individuals did not obtain an uptake of radioactive material that

.would require WBC.

The RGRE RP5C staff acknowledged

th'e'procedure

shortcoming and is presently

evaluating procedure changes to better justify when an exit WBC is needed;

in lieu

of terminating work activities at Ginna.

C.

Conclusion

This issue had minor safety significance and represented

a failure of the licensee to

meet an administrative requirement.

The licensee had initiated adequate

corrective

actions.

Accordingly, this failure constituted

a violation of minor significance and is

~

being treated as a non-cited violation, consistent'with Section IV of the enforcement

policy.

R7

R7.1

Quality Assurance in Radiological Protection and Chemistry Activities

Poor Radiolo ical Work Practices and lnade

uate Contamination Area Boundar

Controls

a 0

Ins ection Sco

e (40500)

The inspector reviewed the licensee's corrective actions associated

with ongoing

incidents where poor radiological work practices and inadequate

contamination

area

boundary controls led to the uncontrolled release of potentially and actually

contaminated

material to an uncontaminated

area..

C

21

b.

Observations

and Findin s

In December 1994, the'RC identified inconsistent. contamination

area boundary

labels, and a'loss of some contamination

area boundary controls in the auxiliary

building.

Equipment crossing contamination boundaries

was not secured

in place,

and the demarkation between contaminated

and uncontaminated

portions could'not

be identified.

These conditions were discussed

with the licensee who.indicated that

.,

the conditions would be evaluated for the appropriate corrective actions (see IR

50-244/94-29). -The NRC identified a similar situation in October 1996, where two

'instances of poor radiological work practices and inadequate

contamination

boundary controls existed inside designated

contamination

areas were not adequate

to prevent a potential spread of contamination to uncontaminated

areas

(see

IR'0-244/96-11).

Specifically, significant amounts of debris, tools, and equipment

used for maintenance

work were allowed to cross over the contamination

demarcation

boundaries

around the refueling water stor'age'tank

and around an

MOV-857A work area, without having been surveyed by radiological protection (RP)

technicians for potential contamination.

After all materials. were surveyed, it was

determined that no actual spread of contamination. resulted from these incidents.

In

September

1996, the licensee identified 12 items in the maintenance

shops

contaminated

from. 1500 to 65,000 dpm/100cm'hat

had been removed from a

radiologically controlled area without a proper contamination survey, and without

placing the"items into a controlled contamination storage area.

This was reported

on RGSE ACTION Report No. 96-0902.

All of the above iristances did not meet RGSE's

RP program requirements

or

management

expectations for radiological work or the expected contamination

area

boundary controls designed to prevent the spread of contamination.

All material

inside a designated

contamination

area at the Ginna Station is consi'dered

contaminated

until cleared by an RP technician.

In October 1996, the licensee

initiated specific corrective actions, and also identified other contaminated work

areas where radiological work practices were not meeting

RP program and

management

expectations.

The inspectors followed up on these actions to evaluate

'heir'effectiveness,

and idehtified several actions taken that were designed.to

correct these conditions.

Corrective actions included:

1) upgrading contamination

boundary labels throughout the plant, 2) Increasing the level of RP staff monitoring

of radiological work activities, and 3) conducting additional training for qualified

radiological workers and RP technicians.

However, the licensee considered that no

programmatic changes

or improvements were necessary to correct, the inadequate

work practices and barrier controls.

On February 9, 1997, the inspector observed maintenance

repairs on, the A-'Sl pump

inside desi'gnated

contamination

areas around and on the pump and its pedestal.

The contamination boundaries

were small and the workers were required to reach

into the area from outside.

During the course of the work, the inspector observed

approximately 10 rags and

a wire brush straddling the contamination boundary on

the pedestal of the pump, an. area that had been previously surveyed and

determined to have loose smearable

surface contamination above the licensee's

'imit

of 500 dpm/100cm'.

Several other tools that had been used inside the

'

22

contamination boundary around the pump bearing were laying on the floor on an

uncontaminated

surface.

The general work area was not roped off around the

pump, and tools and equipment used inside the area were not surveyed prior to

being removed:

The inspector reported the condition to the RP technician covering

the job. The technician evaluated the situation and then surveyed the tools and the

pump pedestal.

The pump pedestal

area in contact with the rags and wire brush

contained'smearable

loose surface contamination of approximately 1600

dpm/100cm'.

The inspector considered that these conditions represented

a

fundamental. disregard for boundary 'controls and the basic purposes for observing

contamination boundaries.

However, none of the tools or floor areas outside the

boundary marker were contaminated.

A larger temporary area could have controlled

the entire floor around the pumps as contamination

areas,

and the tools could have

been bagged

and surveyed at the control point.

The inspector later jnterviewed one of the mechanics involved with this job. The

mechanic could not remember the levels of contamination detected

by the HP

technician before the job started.

Other qualified radiological workers in the plant

were also interviewed.

It was apparent to the in'spector that there was wide spread

confusion and disagreement

among radiological workers regardin'g contamination

boundaries

and whether or not reaching across

a contamination boundary to work

inside an area was an acceptable practice.

Confusion also'xisted over the meaning

of radiological contamination boundary markers, i.e., ropes, tape, and signs.

One

qualified radiation worker did'not know that yellow and magenta tape represented

a

contamination boundary at Ginna.. The inspector noted several examples in the

plant where the application of boundary markers was contradictory and confusing to

plant workers.

For example:

1) contamination area tape on the floor of the

contaminated

storage'building

(CSB) threshold of all outside doors, but the CSB

floors were clean and the boundary postings specifically designated

it as a restricted

~ area/radiation

area only; -2) on the auxiliary building intermediate level near the

entrance to the CVCS resin ion exchanger

room, barrier tape was on the floor

around piping penetrations,

but a nearby. contamination barrier sign stated that'the

contamination was on the overhead

horizontal piping which extended

beyond the

vertical plane described by the tape on the floor; 3) barrier tape was not on the

floor around all edges of the spent fuel pool, but the hand railing was posted

as a

contaminated

area; 4) the penetration fan coolers in the auxiliary building were,

surrounded

by a barrier rope and floor tape, and some workers did not know if the

"contamination barrier vertical, plane" extended to the ceiling of the auxiliary

building (any a~ca more than 8 feet above

a floor, requires contact with an RP

technician prior to entry); and 5)'the area around the sink inside the radiation

chemistry laboratory was a designated'contamination

area, but several pieces of lab

equipment,

a set of tongs, an absorbent

diaper, etc, were straddling the barrier

marker.

An RP technician considered

this to be an acceptable

situation; however,

an RP supervisor did not.

The inspector reviewed al( the ACTION Reports written in 1996 that concerned..

contamination deficiencies.

Four Reports involved contaminated. material found

outside

a radiologically controlled area; six Reports involved deficient boundary

controls; two involved a spread of contamination;

and four involved miscellaneous

2'3

incidents including personal contaminations

or failure to adhere to radiological work

permit requirements.

The inspector discussed

the level of training given to qualified radiological workers

at the Ginna Station with a training department instructor who stated that if a

worker was previously a "fullyqualified radiological worker'" at another nuclear

facility, and then employed at Ginna, no requalification training is given by RGB.E.

The individual must, however, perform satisfactorily in the normal General Employee

Training (GET) given annually to qualified radiological workers.

This was also the

current process for qualified RP technicians.

RP technicians who were previously

~

certified at another facility were granted

a certification upon entering employment at

Ginna.

There was currently no periodic recertification of RP technicians at Ginna,

but these individuals also needed to pass the annual GET training.

This situation

may be related do the notable variation in radiological work practices and control

standards

among the Ginna work force.

On February 17, 1997, the inspector observed

contaminated

water dripping'rom a

fitting on the differential pres'surd transmitter at flow indicator Fl-116 (seal injection

return. flow from the B-reactor coolant pump).

A small leak had previously

developed

on December 2, 1996, and.was causing

an accumulation of boron

crystals around the fitting, but it was not large enough to form drops that would fall

to the floor. On February 17, approximately 3 drops per minute were dripping onto

an absorbent towel (i.e., a "launderable decon rag") that was placed on the

uncontaminated

floor beneath.FI-116.

It appeared that the towel was placed on the

floor to collect the drippage since it was folded into a small 6" X 6" area and placed

directly below the leaking fitting. At the time of discovery, the towel was saturated

'with water and a small stream was flowing onto uncontaminated

floor

surfaces'way

to a lowpoint where it.collected into a small puddle.

Boric acid crystals

.

approximately, 1/8 - 1/4 inch in size had formed qn the surrounding floor due to

evaporation of the fluid. The inspector contacted the RP technician on duty, who

immediately placed

a bucket under the dripping fitting and then cleaned up the

water and boron crystals.

A subsequent

survey indicated that the floor area directly

below Fl-116 was contaminated

from 700 to 2700 dpm/100cm'.

A later survey of

50 ml of the leaking water. indicated an activity level of 2.75E-3 uCi/gm.

Both the

.

RP.technician

and the RP supervisor agreed that the absorbent tbwel was not an

appropriate method to contain the dripping water and that a catch or bucket should

have been used.

It appeared that the licensee did not currently have RP program

controls that required the use of catches or buckets under potential leaks of

contaminated

water that would prevent the spread of, contamination.

Conclusions

The licensee's

actions to maintain proper contamination boundary controls, and to

monitor to assure radiological 'work practices adequately prevent the spread of

contamination have not been adequate.

Actions to identify and correct poor

radiological work practices,

and to prevent the spread of potential and..actual

contamination,

also have not been effective.

These conditions represent

a violation

of 10 CFR 50, Appendix B, Criterion XVI, "Corrective Actions,"

(NOV 50-244/97-01-02).

V. Mana ement IVleetin s

X1

Exit Meeting Summary

The inspectors presented

the inspection results to members of licensee management

on February 25, 1997, after the <nspection was concluded.

The licensee

acknowledged

the findings presented.

The inspectors

asked the licensee whether any materials examined during the

inspection should be con'sidered

proprietary.

No proprietary information was

'dentified.

L2

Review of UFSAR Commitments

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

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

special focused revjew that compares plant practices, procedures

and/or parameters

to the UFSAR description.

While performing the inspections discussed. in this

report, the inspector reviewed the applicable portions of the UFSAR that related to

the areas inspected.

Except in the instances

noted below, the inspector verified

that the UFSAR wording was consistent with the observed plant practices,

procedure and/or parameters.

UFSAR Section 10.5-8; AFW pump recirculation line availability was not

consistent with ITS definition of pump operability (see section M3.1).

The EDG aii start system check valve test acceptance

criteria basis was not

consistent with UFSAR Section 3.8.1 (see section M3.2).

~

e ~

ATTACHMENT 1

PARTIAL LIST OF PERSONS CONTACTED

B. Flynn

C. Forkell

G. Graus

A. Harhay

J. Hotchkiss

G. Joss

R. Marchionda

R. Mecredy

R. Ploof

R. Smith

J. Smith

J. Widay

T. White

G; Wrobel

Primary Systems

Engineering

Manager

Electrical Systems

Engineering Manager

IRC/Electrical Maintenance

Manager

Chemistry 5 Radiological Protection Manager

Mechanical Maintenance

Manager

Results and Test, Supervisor

Production Superintendent

Vice President,

Nuclear Operations

Secdndary

Systems

Engineering

Manager

.

Senior Vice President,

Customer Operations

Maintenance

Superintendent

Plant Manager

Operations

Manager

Nuclear Safety 5 Licensing

Manager.'NSPECTION

PROCEBURES USED

IP 37001:

IP 37551:

IP 40500:

IP 61726:

IP 62707:

IP 71707:

IP 71750:

IP 83750:

IP 92700:

10 CFR 50.59 Safety Evaluation Program

Onsite En'gineering

Effectiveness of Licensee Controls in Identifying, Resolving, and Preventing

Problems

e

Surveillance Observation

Maintenance

Observation

Plant Operations

Plant Support

Occupational

Radiation Exposure

Onsite Follow-up of Written Reports of Nonroutine Events at Power Reactor

~ Facilities

ITEMS OPENED, CLOSED, AND DISCUSSED

~Oened

1) The practice of extending

a normal shift to meet programmatic requirements for

operator proficiency under 10 CFR 55.53 (URI 50-244/97-01-01).

2)

Inadequate

corrective actions for poor radiological work practices and inadequate

contamination

area 'boundary controls (NOV 50-244/97-01-02).

Closed

1)

Circuit Breakers Closed While in Mode 3, Due to Personnel

Error, Resulted in a

Condition Prohibited by Technical Specifications

(LER 96-013).

2)

Based on Review of NRC Generic Letter 96-06, a Thermally Induced Overpressure

Transient Could Occur (LER 96-015).

e

AFW

, ALARA

ASME

CCW

CW

CFR

d/p

dpm

ECCS

EDG

FME

HPES

IFI

IPE

IR

ISI

IST

ITS

LCO

LER

LOCA

MG

MSIV

NCV

NRC

NRR

NSARB

PCT

PORC

ppm

PSA

PT

pslg

PWR

QA

QC

.

RCA

. RGSE

RHR

RP

RP5C

RWP

SEV

Sl

'RV

SW

TSD

UFSAR

VIO

LIST OF ACRONYMS USED

Auxiliary Feedwater

As Low As Reasonably

Achievable

American Society of Mechanical Engineers

Component Cooling Water

Circulating Water

Code of Federal Regulations

differential pressure

disintegrations

per minute

Emergency Core Cooling Sy's tem

Emergency

Diesel Generator

Foreign Material Exclusion

Human Performance

Enhancerr}ent System

Inspection Follow-up Item

Individual Plant Evaluation

Inspection Report

Inservice Inspection

Inservice Test

Improved Technical Specifications

'imiting Condition for Operation

Licensee Event Report

Loss-of-Coolant Accident

Motor-Generator

Main Steam.Isolation Valve

Non-Cited Violation

Nuclear Regulatory Commission

Nuclear Reactor Regulation

Nuclear Safety Audit and Review Board

Peak Cladding Temperature

Plant Operations Review Committee

parts per million

Probabilistic Safety Assessment

~

Periodic Test

pounds per square inch gage

Pressurized

Water Reactor

Quality Assurance

Quality Control

Radiological Controlled Area

Rochester

Gas and Electric Corporation

Residual Heat Removal

Radiation Protection

Radiological Protection and Chemistry

Radiation Work Permit

Safety Evaluation

Safety Injection

Safety Relief Valve

Service Water

Training System Description

Updated Final S'afety Analys'is Report

Violation