Text

Insp Repts 50-280/96-201 & 50-281/96-201 Conducted on 960122-0202.Violations Noted.Major Areas Inspected: Review of Insp Record & Performance History for 2 Yr Period & Overall Assessment Scope & Objectives
	ML18152A058
Person / Time
Site:	Surry
Issue date:	02/12/1996
From:	Mary Johnson, Koltay P, Norkin D; NRC (Affiliation Not Assigned)
To:	;
Shared Package
ML18152A059	List: IR 05000280/1996201; ML18152A058;
References
	50-280-96-201, 50-281-96-201, NUDOCS 9602200230
	Download: ML18152A058 (36)
	v • d • e

U.S. NUCLEAR REGULATORY COMMISSION

OFFICE OF NUCLEAR REACTOR REGULATION

NRC Inspection Report:

50-280/96-201 *

License No.:

DPR-32 and DPR-37

and 50-281/96-201

Docket No.:

50-280 and 50-281

Licensee: Virginia Electric and Power Company

Facility Name:

Surry Power Station, Units 1 and 2

Inspection at: Surry Power Station, Surry, Virginia

Inspection Conducted:

January 22 through February 2, 1996

Inspection Team:

Peter S. Koltay, Team Leader, Special Inspection Branch

Jeffrey B. Jacobson, Special Inspection Branch

Prepared by:

Reviewed by:

Approved by:

Paul P. Narbut, Special Inspection Branch

Edmund A. Kleeh, Special Inspection Branch

David L. Gamberoni, Inspection Program Branch

Steven P. Sanchez, Inspection Program Branch

James E. Wigginton, Emergency Preparedness and Radiation

Protection Branch

Lawrence K. Cohen, Emergency Preparedness and Radiation

Protection Branch

Robert B. Manili, Safeguards Branch

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Peter S. Koltay, Team Leader

Special Inspection Branch

Division of Inspection and Support Programs

Office of Nuclear Reactor Regulation

Donald P. Norkin, Section Chief

Special Inspection Branch

Division of Inspection and Support Programs

Office of Nuclear Reactor Regulation

Mic

el R. Johnson,

Special Inspection

anch

Division of Inspection and Support Programs

Office of Nuclear Reactor Regulation

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Date

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9602200230 960212

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Enclosure

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TABLE OF CONTENTS

EXECUTIVE SUMMARY .......... .

OVERALL ASSESSMENT SCOPE AND OBJECTIVES

ASSESSMENT METHODOLOGY

1.0

SAFETY ASSESSMENT AND CORRECTIVE ACTION

I.I

Problem Identification ....

1.2

Problem Analysis and Evaluation

1.3

Problem Resolution

2.0

OPERATIONS

2.1

Safety Focus

..... .

2.2

Problem Identification ind Resolution

2.3

Quality of Operations .

2.4

Programs and Procedures

3.0

ENGINEERING ........ .

3.1

Safety Focus ..... .

3.2

Problem Identification and Problem Resolution

3.3

Quality of Engineering

3.4

Programs and Procedures

4.0

MAINTENANCE ........ .

4.1

Safety Focus

. . . . . .

. .

.

4.2

Problem Identification and Problem Resolution

4.3

Equipment Performance and Material Condition

4.4

Quality of Maintenance Wor~

4.5

Programs and Procedures

5.0

PLANT SUPPORT ...

5.1

Safety Focus

5.1.1

Radiological Controls

5.1.2

Security ..... .

5.1.3

Emergency Planning

5.2

Problem Identification and Resolution .

5.2.1

Radiological Controls

5.2.2

Security .....

5.2.3

Emergency Planning

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5.3

Quality of Plant Support

5.3.1

5.3.2

5.3.3

Radiological Controls

Security .....

Emergency Planning

5.4

Programs and Procedures

5.4.1

5.4.2

5.4.3

Radiological

Security ....

Emergency Planning

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APPENDIX A - LIST OF REFERENCES. . . . . . . . . . . . . . . . . . . . . . A-1

APPENDIX 8 - PRELIMINARY PERFORMANCE ASSESSMENT/INSPECTION PLANNING TREE. 8-1

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EXECUTIVE SUMMARY

This report represents the results of the in-office review phase of the IPAP

for the Surry Power Station, Units 1 and 2.

The assessment was conducted by

the Special Inspection Branch of the U.S. Nuclear Regulatory Commission's

Office of Nuclear Reactor Regulation during the weeks of January 22 and 29,

1996.

The purpose of this assessment was to develop an integrated perspective

of performance strengths and weaknesses based upon an in-office review of

inspection reports, event reports, and other NRC and licensee generated

performance information.

The assessment covered a two year period from

January 1994 to December 1995.

A two week on-site assessment scheduled to

start on February 26, 1996, will be conducted to validate the observations

made during this in-office review.

The licensee's corrective action and performance assessment systems have been

effective at capturing equipment, program, and human performance deficiencies.

Assessments of program performance by the licensee's nuclear oversight

division were effective.

Problem analysis and evaluation for routine, low

level issues was generally good, but initial root cause analyses performed for

more complex equipment is;ues were sometimes ineffective.

Often, sufficient

root causes were not identified until after the problems resulted in plant

events.

Repeat failures were identified with the rod control system, the

auxiliary feedwater turbine-driven pump, esse,,tia~ service water pumps, and

component cooling water heat exchangers.

Trending of e4uipmenL ctnd human

performance was good as evidenced by quarterly trend reports which effectively

captured equipment and human performance issues, including specific

recommendations for management action.

Problem resolution was not effective. Corrective actions to longstanding

problems with the rod control system, hydroid growth in the essential service

water system, and the turbine-driven auxiliary feedwater pumps have sometimes

been delayed or have only partially been completed.

A review of recent

licensee corrective actions with regard to these issues will be conducted

during the team's on-site assessment.

In the area of operations, management involvement and safety focus were good;

however, instances were noted where management's decision making process was

non-conservative.

Management actions were conservative in interpreting

technical specifications, establishing additional supervisory oversight in the

control room, oversight of on-line maintenance, and in shutdown risk

reduction.

Convetsely, decisions associated with the sampling of charcoal

filters, pumping the containment sump, and assessing the operability of safety

related equipment potentially affected by common failure mode were .sometimes

inadequate.

Operators were knowledgeable and responded well to challenges

caused by the large number of equipment failures and reactor trips that were

often complicated by unexpected equipment responses.

Engineering management established a good safety perspective as demonstrated

by the progranvnatic controls for design related ~atters; however, there were

weaknesses in the implementation of programs.

For example, engineering safety

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evaluations and operability assessments provided good support, to maintenance

and operations.

However, the licensee did not maintain control of setpoints

when making changes to design basis calculations and their meth6dology.

In the area of maintenance, programs for problem identification, self-

assessments, and quality assurance department audits were well established;

however, problem resolution was sometimes ~1 ow and ineffective, resulting in

longstanding or recurring problems.

Plant material condition has improved

steadily, but problems with material condition in the balance of plant and

with maintenance and surveillance personnel errors have resulted in numerous

equipment failures and plant perturbations.

In the plant support areas of Security, Emergency Preparedness and Health

Physics, overall strong performance was demonstrated.

However, some

weaknesses were noted in the procedure compliance area for health physics .

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OVERALL ASSESSMENT SCOPE AND OBJECTIVES

This Integrated Performance Assessment of both units of the Surry Power

Station Units 1 and 2 is being performed in accordance with NRC Inspection

Procedure 93808 "Integrated Performance Assessment Process." The assessment

is divided into: an in-office review performed at NRC headquarters; an on-site

assessment to validate the observations f~n~ the in-office review; and a final

analysis of the results of the assessments and developmen~ of inspection

reconvnendations.

The assessment is being conducted by the Special Inspection

Branch of the Office of Nuclear Reactor Regulation.

The in-office review was

performed during the weeks of January 22 and January 29, 1996.

The on-site

assessment is scheduled to be performed during a two week period starting

February 26, 1996.

The assessment objectives are to develop an integrated perspective of licensee

performance and arrive at reconvnendations for future inspection focus in the

areas of Safety Assessment/Corrective Action, Operations, Engineering,

Maintenance, and Plant Support.

The in-office review. covers NRC inspection

reports, licensee event reports (LERs), enforcement history, regional

assessments, and licensee internal and external assessments.

The results of

the in-office review are included in this preliminary assessment report.

The

references contained in the report a,~ listed in Appendix A.

The preliminary

results are presented on the Performance Assessment/Inspection Planning Tree

in Appendix 8.

Following the issuance of this report, the team will validate its observations

via a performance based, on-site assessment.

The results of the on-site

assessment and in-office review will be used during the final analysis and

development of inspection recommendations, and will be documented in a final

report to be issued after the conclusion of the on-site assessment.

The final

assessment report will include recommendations on where to focus future NRC

inspection effort, and these recommendations will be depicted on a final

Performance Assessment/Inspection Planning Tree.

ASSESSMENT METHODOLOGY

During the in-office review, the team evaluated the Surry Power Station

inspection record and performance history for a two year period spanning

January 1994 to December 1995.

Available licensee quality assurance (QA)

audit reports and other self-assessment documents were reviewed.

The review

results were utilized to assign performance ratings of either decreased,

normal, or increased inspection to the individual elements in each assessment

area.

Where the team's review of inspection data and licensee information was

inconclusive, or where sufficient information was not available to come to

meaningful conclusions, individual elements were rated as being indeterminate.

Ratings for the overall performance* in the areas of Safety

Assessments/Corrective Action, Operations, Engineering, Maintenance, and Plant

Support were not addressed during the in-office review phase.

The results obtained from the in-office review will be used by the assessment

team to develop individual on-site assessment plans for each of the assessment

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areas.

During the on-site review, the team will focus on those areas rated as

indeterminate and those where the inspection or performance data record

indicated potential performance weaknesses.

The team will also validate the

elements that were assigned decreased or normal inspection ratings.

Following

the on-site phase of assessment, the team will issue a final assessment

report.

1.0

SAFETY ASSESSMENT AND CORRECTIVE ACTION

1.1

Problem Identification

The licensee's corrective action and performance assessment systems have been

effective at capturing equipment, human performance, and program deficiencies.

The threshold for initiating corrective action documents appears to be

sufficiently low and the licensee has apparently been able to avoid a large

backlog of open corrective action documents.

A recent audit of the corrective

action process by the licensee's nuclear oversight division confirmed that the

corrective action system has been effective in the area of problem

identification (ref. 225).

Assessments of program performance by the licensee's nuclear oversight

division were noted in several inspection reports as being effective (refs. 3,

18, 26, and 29).

These assessments were seen as oeing insightful, in-depth,

and as having identified significant issues which could be used by the line

organizations for making meaningful improvements.

The effectiveness of line

organization follow-up actions to address assessment findings will be reviewed

during the team's on-site assessment.

The quarterly performance annunciator window program overseen by Station

Nuclear Safety group effectively communicated personnel, equipment, and

programmatic performance.

Numerous performance indicators with pre-determined

criteria have been established for each organization.

Data for individual

annunciator windows are provided by the line organizations.

The Station

Nuclear Safety group compiles the data and in conjunction with senior licensee

management, assigns overall performance ratings for personnel, equipment, and

program performance.

Areas receiving a red (significant weakness) or yellow

(improvement needed) rating require a line management response.

The

effectiveness of the annunciator window program in the area of operations

appeared to be indeterminate due to a lack of internal assessment data.

Of

the 16 operations areas rated by the program, 7 appeared to be primarily based

on NRC and the Institute of Nuclear power Operations (INPO) findings,

including items such as tagging, operations status, and configuration control.

A review of the actions taken by line management in response to the

annunciator window program will be conducted during the team's on-site

assessment.

The licensee is currently developing a more extensive self assessment program.

Other than the annunciator window program described above, formal line

organization self assessments have not been performed routinely.

In the

m~intenance area, self assessment repo~ts reviewed by the IPAP team appear to

have been effective in the identification and resolution of problems,

including the fact that certain relief valves and risk significant check

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valves were not being periodically tested. Engineerlng self assessments were

stated in inspection reports as being good {ref. 26), but it appears the self

assessments were limited to a review of the annunciator windows.

Likewise,

the team was not provided with any internal asse.ssments of the Operations

Division performance.

Normal inspection effort is recommended for th-h area.

1.2

Problem Analysis and Evaluation

Overall, performance at Surry appears to have been hindered by a large number

of challenges caused by equipment failures and human performance weaknesses.

Several plant trips and inoperable safety systems can be attributed to

uncorrected or partially corrected problems that had been previously

.

identified by the licensee.

For example, instances of inoperable or degraded

emergency service water (ESW) pumps (refs. 38, 42, and 45), control room

annunciators (ref. 43 and 45), station batteries (ref. 28), component cooling

water heat exchangers (CCWHXs) (ref. 45), and charging pump lube oil

temperature controllers (ref. 25) have b~P1, attributed to causes that were not

fully resolved by previous corrective action attempts.

Repeated challenges

were also caused by deficiencies in the turbine driven auxiliary feedwater

(AFW) pump (ref 14, 21, and 32) and in the rod control system (ref. 37).

More extensive root causes analyses were generally performed only after plant

events had occurred and the initial opportunities to prevent problem

escalation had not been successful. These root cause analyses provided the

necessary detail for the technical assessment, but did not fully evaluate the

contributing causes related to human performance, the corrective action system

itself, and management.

For example, the root cause analysis (ref. 94)

associated with the failure of the 2A station battery did not address why the

system engineer was unable to recognize that the batteries had become

inoperable or why plant procedures were not followed.

The root cause analysis

associated with the inoperable turbine driven auxiliary feedwater pump (ref.

1178) failed to address why previous corrective action attempts were not

sufficient and why the plant was allowed to restart before ensuring that the

problem was completely resolved.

A recent audit (ref. 225) conducted by the

licensee of the corrective action program also raised concerns regarding the

effectiveness of the root cause analysis program.

The team also reviewed the third quarter 1995, quarterly trend report, issued

by Station Nuclear Safety.

The trend report of deviations effectively

captured station issues involving both equipment and human performance.

The

trend report provided a detailed summary of the issues and included a synopsis

of the actions planned or already implemented.

The report also included six

additional recommendations for management action which were gleaned from a

review of recent deviation reports.

The recommendations are required to be

tracked by the licensee's commitment action tracking system.

The team will

review the license's response to the recommendations during the on-site phase

of the assessment.

Overall performance in this area was rated as being indeterminate.

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1.3

Problem Resolution

The licensee has done an adequate job of addressing the majority of lower

level hardware concerns that resulted in challenges to the corrective action

system.

For example, corrective actions taken for a failed shunt trip relay

(ref. 6), low head safety injection check valves failing t~ seat (ref. 22),

unsuccessful attempts to load main.contra~ .*oom chillers (ref. 22), and

clogged suction strainers for the main control room chillers (ref. 26), were

stated as being adequate to resolve the identified problems.

However, the licensee's efforts to resolve several recurring major issues have

not been effective. Corrective actions were sometimes delayed or only

partially completed.

Examples include hydroid growth in the ESW and component

cooling water {CCW) systems, problems with the performance of the rod control

system, and governor problems with the turbine driven auxiliary feedwater

pumps

Also reviews to ensure the effectiveness of the corrective actions

taken have not always been sufficient to prevent reoccurrence.

Some of these

issues have finally been resolved, but follow-up during the on-site phase of

the assessment will be necessary to fu11y assess this area.

Based on the team's review of the licen~P~*s r2sponses to NRC violations,

actions taken in response to externally identified issues appeared to be very

good.

Corrective actions addressed both the specific prob1ems cited a,,d the

more general programmatic concerns as appropriate.

Tracking of corrective actions via the licensee's commitment action tracking

system will be reviewed during the on-site phase of the assessment.

Overall performance in this area was rated as being indeterminate.

2.0

OPERATIONS

2.1

Safety Focus

Management generally demonstrated conservative safety focus.

For example, the

licensee declared the condensate storcge tank inoperable even though only tank

level indication was lost (ref. 9).

The Station Nuclear Safety and Operating

Committee {SNSOC) assessments were effective and focused on safety.

For

example, to assess readiness for startup following outages, each department

presented a review of work status, and an action plan for remaining items to

the SNSOC {ref. 29).

Managers regularly interacted with the control room

crews to address issues.

For example, management was closely involved in

establishing appropriate and timely compensatory measures when Unit 1

annunciators were lost (ref. 43). Operations managers gave effective pre-

shift briefings to crews before complex evolutions (ref. 29).

However, several examples were noted where management decisions were not

conservative.

For example, an enforcement conference was held dealing with

the failure to promptly identify and implement corrective action when the

licensee did not sample both auxiliary ventilation exhaust filter trains

following a chemical release, a recognized potential common failure mechanism.

Upon delayed sampling, both charcoal banks failed to meet technical

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specification requirements (ref. 21).

Likewise, a nonconservative decision

was made to pump the containment sump when one of the sump valves, a

containment isolation valve, had already failed to close during testing and

the licensee was in an action statement to close and remove power from the

second valve.

When the second valve was later administratively reopened it

failed to close due to debris, resulting in entl'y into a one hour LCO (ref.

27).

Another example occurred on May 30, 1995 when ESW pump C failed its

operability test due to low flow.

Technical Sp~cificatiors required the plant

to shutdown if more than one ESW pump was inoperable.

The redundant ESW pump

A was not invnediately tested, even though it was reasonable to assume that the

A pump was also fouled.

The A pump had been operating in the alert range

since May 16, 1995 when it exhibited low flow during testing.

On June I, 1995

after ESW pump C was cleaned and operable, ESW pump A was cleaned and

subsequently tested satisfactorily (ref. 38). Similarly, LER 95-10 (ref. 85)

describes the inoperable condition of all four CCWHXs due to fouling.

Technical specifications required the plant to shutdown if more than one heat

exchanger was inoperable.

In this case no shutdown occurred since the

licensee cleaned one heat exchanger first and declared it operable, before

testing the other three.

The licensee implemented a comprehe~sive*program to reduce shutdown risk

(ref. 35).

However, an example of poor risk management was identified when

turbine building flooding occurred due to leaking canal damming devices during

outage work.

Despite the fact that turbine building flooding is a high risk

core melt sequence, no flood watches had been posted and the process of

installing the damming devices was not described in a procedure (ref. 46) .

Increased inspection in this area is recommended.

2.2

Problem Identification and Resolution

Overall, inspection reports indicated that problem identification was

adequate.

Operators had improved in their performance in writing problem

reports.

For instance, operators identified that the protection channels for

pressurizer pressure were indicating lower than the control channels and wrote

a deviation report.

This eventually led to the identification of the fact

that all three protection channels were inoperable (ref. 4).

Although there

were examples where operators did not initiate deviation reports when

appropriate, inspection reports state that operators have a generally low

threshold for reporting problems (ref. 1).

Some of the licensee problem reporting mechanism's appear to identify

important trends to management.

For example, the Third Quarter 1995 Station

Deviation Trend Report addressed configuration control issues as a degrading

trend area.

Another of the licensee's mechanisms for highlighting operations performance

for management attention, the Performance Annunciator Panel program, often

only uses NRC data, such as violations issued, to assess performance.

Seven

of the sixteen windows applicable to operations use only NRC data for success

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criteria. Examples include the windows for Operations Status and

Configuration Control; Operations Drawings, Documents and Procedures; Tagging;

and Labeling.

There were examples of a lack of aggressive management problem resolution

which resulted in repeated challenges to operators.

For example, there was a

repeated problem with unexpected changes in reactor vessel water level during

reduced inventory operations (ref. 1), reµeated problems with dropped rods

(refs. 36 and 37), and biofouling of CCWHXs (ref. 45).

Normal inspection is recorrvnended in the area of problem identification.

Performance in the area of problem resolution is indeterminate.

2.3

Quality of Operations

Operating crews responded promptly and effectively to operational events.

For

example, when challenged by a turbine run-back, operators prevented a plant

trip through detailed plant knowledge and skillful equipment operation (ref.

l); they performed a unit shutdown that required difficult manual control of

steam generator levels (ref. l); and operator crew responses to plant trips

was considered a strength ( ref. 9;.

On one occasion the operators prevented

an automatic trip by prompt response to a feedwater regulating valve failing

closed (ref. 21).

However, there is one example where an acting control room

supervisor lost command and control of ongoing plant evolutions (ref. 102).

The event represents a significant but isolated case.

A negative trend in operator performance is indicated by recurring personnel

errors (ref. 45).

For example, operators made configuration control errors

such as failure to open the hot leg stop valve within two hours of filling the

loop as required by technical specifications (ref. 6), and failure to lock a

makeup water isolation valve as required by technical specifications (ref. *

14).

Five equipment lineup deficiencies occurred in one month.

Also, a power

loss to emergency buses was caused by an operator opening the wrong fuse

drawer due to inadequate self checking (ref. 46).

Likewise, operations had

inappropriately released work on the seal table and pressurizer relief valves

when the reactor coolant system was still pressurized (ref. 49).

Equipment failures continued to challenge the operators during startups,

shutdowns and during normal operation.

For example, in separate instances in

July and August 1995, a partial failure of the control room annunciators

occurred (refs. 43 and 45).

Additional examples were losses of nuclear

instrument power and the generator hydrogen seal oil pressure (ref. 37).

Also, biofouling made the CCWHXs inoperable 1-4 times per week (ref. 45).

Other hardware problems included battery cell problems (ref. 29), control rod

dropping problems (refs. 36 and 37), and chiller unit trips (ref. 29).

Normal inspection in this area is recommended.

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2.4

Programs and Procedures

Procedures were generally followed.

However, three violations were identified

for failure to follow procedures involving a reactor coolant system inventory

reduction (ref. 49).

Additionally, there were other isolated examples of

lack of procedure compliance.

An operator performing a quarterly pump test

entered a procedure at the wrong place (ref. 24), operators exceeded the

pressurizer maximum heatup rate (ref. 32), and: ;censee contractors, performed

fuel manipulations without following the procedure (ref. 34).

Isolated instances of inadequate procedures were identified.

For example, the

procedure for opening the loop stop valves was not detailed and did not point

out that the cold leg valve must be shut to open the hot leg valve.

Consequently, the valve was not opened within the technical specification

limit of two hours after loop fill (ref. 6). Similarly, plant cooldown was

accomplished using main steam bypass valves which was a method not addressed

by the procedure (ref. 27).

The procedure for maintaining containment

integrity during refueling was inadequate.

Some valves needed for integrity

were kept open and would only be closed by operators if an event occurred.

The licensee consequently changed the ryrocedure (ref 34).

Normal inspection in this area is recommended.

3.0

ENGINEERING

3.1

Safety Focus

Generally, a conservative safety focus was exhibited by engineering during the

review period.

This conservative safety focus was evident during

engineering's attempts to improve emergency diesel generator (EOG) reliability

(ref. 22) and eliminate power oscillations caused by degrading steam

generators (ref. 13).

Conversely, the lack of a formal setpoint control

program led to an overpower event resulting from the licensee's failure to

update the calorimetric computer program in accordance with the latest revised

base calculation (ref. 8). A licensee root cause evaluation (RCE) of this

event identified potential problems with other station-instrument settings and

test/calibration procedures, that were also attributed to a lack of formal

setpoint control program (ref. 113).

The engineering work load of open design change packages (DCPs) and drawing

revisions has been maintained below management established goals (ref. 22).

Operability and saf~ty evaluations were generally comprehensive and provided a

sound basis for conclusions regarding safety impact.

Examples included an

evaluation of power oscillations due to steam generator tube-sheet blockages

(ref. 25) and the by-passing (jumpering) out of two cells of battery 2A (ref.

25).

The engineering safety evaluations associated with deviation reports

also were generally adequate to support continued operation (ref.22).

However, there were two instances where the safety evaluation process failed

to provide acceptable results.

In one case, a safety evaluation to allow for

th~ administrative control of a manual valve for the low pressure carbon

~;oxide fire suppression system was performed only upon NRC request (ref.4);

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in the other case, improper safety analysis of a rev1s1on to a test procedure

resulted in a "water hammer" in the reactor coolant system (RCS) and safety

injection (SI) piping systems (ref. 6).

Normal inspection effort in this area is recommended.

3.2

Problem Identification and Problem RP~olution

The licensee generally provided prompt i~sessments and resolution to issues

emanating from events.

For example, engineering provided good assessment of

issues pertaining to the cleaning of steam generator tube sheets to eliminate

a power oscillation (ref. 13), open circuited charging station batteries

(ref. 46), the erosion of the protective lining for the ESW piping (ref. 4),

and the marine-fouling of ESW pumps and CCWHXs (ref. 45).

However, there were

some notable examples where the licensee failed to recognize significant

deterioration of equipment that caused delays in the corrective actions and

allowed continued plant operation with degraded equipment.

For example,

engineering failed to*identify that Battery 2A was not operable (ref. 28) and

that rod control circuit cards and other vital relays had a limited design

life (ref. 6 and 98).

Engineering self assessments consist of monthly and qu~rterly audits,

primarily of design functions, along with specific audits in targeted areas.

Overall engineering assessments are monitored by the Engineering Program

Performance Annunciator Panel Report.

QA audits, engineering self

assessments, and performance monitoring by management were positive

indications of management's efforts to improve overall engineering performance

(ref. 26).

The engineering assessments and audits were effective at

discovering and resolving problems (ref. 26).

The engineering department uses station deviation reports to document, track,

and resolve long standing problems.

The reports reviewed indicate that the

engineering staff adequately responded to these deviation reports with

evaluations and recommendations that appeared to resolve the problem (ref.

22).

The ability of the licensee to promptly resolve problems was

demonstrated during the replacement of aluminum bronze ESW valves (ref. 4),

the elimination of power oscillations by steam generator tube sheet cleaning

(ref. 25), the modification to reduce pressurizer safety valve seat leakage

(ref. 29), during the actions taken to improve the reliability of station

batteries (ref. 94), and during plant modifications to improve EOG reliability

and availability (ref. 22).

To the contrary, it is not clear whether engineering has been able to resolve

some lingering long term problems.

Examples of these problems include the

loss of reactor water level indication (ref. 1), AFW pump-turbine governor

valve problems that resulted in numerous trips for both units (ref. 14, 21,

and 117), problems with the charging pump lube oil temperature control systems

(ref. 25), several problems i~ the rod control system which have caused rod

drops and manual trips (ref. 6, 37, and 98), and the repeated problems with

the macro fouling of the CCWHXs and SW pumps (ref. 22, 85, and 79).

Also root

cause evaluations were not always effective in p, eventing recurring equipment

problems like control rod system failures, Kaman radiation monitors spurious

8

alarms, and individual rod position indication erroneous readings (ref. 21).

In stime of these examples engineering failed to determine whether the problems

were attributed to a materials issue, a degraded environment, or a combination

of both.

In some cases, corrective actions were initiated to alleviate a

condition, but only the symptoms were addressed allowing the problem to

reoccur.

Normal inspection effort in the areas problem iJentificati0n and problem

resolution is recommended.

3.3

Quality of Engineering

The quality of modification and design change packages was generally good.

The licensee had a strong program for the review, prioritization, and

scheduling of plant modifications which appropriately emphasized nuclear and

personnel safety in lieu of operational improvements (ref. 22).

Based on the

design change packages reviewed by previous inspectors, there were no

significant safety related deficiencies identified with reviews, walkdowns, or

installation instructions.

The only weaknesses identified concerned pre-

installation modification package development (ref. 26).

The number of outstanding drawing revisions was not excessive and the controls

for updating and maintaining critical drawings were effective (ref. 22).

The

support of operation and maintenance by engineering was considered good.

Normal inspection is recommended for this area.

3.4

Programs and Procedures

The inspection reports reviewed indicated that surveillance activities were

appropriately performed and that implementing procedures were being followed

(ref. 18).

Controls were adequate to ensure that effective updates of

procedures were performed for design changes (ref. 26); however, as previously

stated, concern was raised over the control of set-points.

Reviews of

licensee procedures and the witness of surveillance tests indicated that the

surveillance procedures were adequate to support safe operation of the plant.

Weaknesses were identified with the ma,ntenance procedure for testing of

individual battery cells {ref. 28), with the design change control process for

updating the computer calorimetric program (ref. 8) and with a revision of a

test procedure that deleted important 'caution' statements {ref. 6).

A high percentage of the engineering staff, including system engineers, had

either full senior reactor operator {SRO) training or certifications. System

engineers had a strong knowledge of assigned systems {ref. 22} and were

actively involved in supporting plant operation and maintenance {ref. 26).

The licensee instituted programs and conducted several assessments that were

effective in evaluating and maintaining plant systems and components.

Examples of these programs include: the {ISI) program which had well written

procedures as demonstrated by the high quality reactor vessel examinations and

the evaluation of ultrasonic data (ref. 2 and 33); the flow accelerated

9

.

-~

. y

corrosion program to maintain high energy carbon steel pipe within acceptable

wall limits (ref. 33); and the motor operated valve (MOV) compliance program

in which a high number of valves were tested (ref. 15).

Reduced inspection in this area is reconvnended.

4.0

MAINTENANCE

4.1

Safety Focus

Maintenance management's focus on safety resulted in effective planning and

scheduling, supervisory oversight of complex jobs (ref. 32), and basic use of

risk-informed decision making (refs. 35).

One example of effectively managing

risk during a shutdown was the rescheduling of an EOG surveillance during the

1995 Unit 2 refueling outage (ref. 35).

Procedures for on-line maintenance practices requiring voluntary entries into

technical specification limiting conditions for operation (LCO) action

statements were determined to meet NRC guidance in this area (refs. 27 and

35).

Prior to the 1995 Unit l shutdown for refueling an increase was noted in

the licensee's practice of this type of maintenance (ref. 45).

The most

significant was work on an EOG that was scheduled for 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br />, but took 128

hours.

This represented a significant delay in the availability of the

emergency diesel generator.

Pre-activity briefings were thorough (ref. 32).

The most recent outages for

both units were completed ahead of schedule, due in part to good outage

planning (ref. 35).

The maintenance backlog was well managed (ref. 35).

Normal inspection in this area is recommended.

4.2

Problem Identification and Problem Resolution

Maintenance self-assessments and quality assurance (QA) department audits of

maintenance were generally effective in identifying problems.

For example,

the 1994 Check Valve Program Annual Assessment (ref. 151) noted that the

failure rate of check valves reportable to the Nuclear Plant Reliability Data

System (NPRDS) is currently below industry average; however, some of the check

valves listed as risk significant in the individual plant examination (IPE)

were either not part of the Check Valve Program or have never been tested to

ensure operational reliability. The 1994 Safety and Relief Valve Program

Assessment (ref. 150) noted that the licensee continues to experience failures

of small safety and relief valve during testing.

Additionally, a QA audit of the measuring and test equipment (M&TE) program,

(ref. 169) concluded that it does not meet regulatory requirements and is not

being effectively implemented.

Specific examples included programmatic

weaknesses with the program's control of M&TE, the use of calibrated standards

and M&TE, the recording of usage data, the eval~ations for retesting, the

storage and identification of M&TE, and the trending of M&TE-related

aeficiencies.

10

. -

Category 1 root cause evaluations identified root causes for most of the

significant plant issues. Several long standing and recurring equipment

problems were corrected including replacement of pressurizer safety valves and

component cooling water heat exchangers (ref. 51).

In addition, condenser

outlet expansion joints that are considered to be a major contributor to IPE

internal flooding scenarios, were replaced.

The licensee was slow to take effective cn~~ective action on a low voltage

condition with station battery 2A, cell 52 (refs. 28 and ~9).

The licensee

concluded that this event was caused by inadequate post-maintenance testing

and a personnel error.

Normal inspection in the problem identification area is recommended.

The overall performance in problem resolution area was indeterminate.

4.3

Equipment Performance and Material Condition

Material conditions of selected safety systems such as the low head safety

injection, safety injecti0n, EOG air start, and AFW systems were found to be

acceptable (refs. 9, 14, and 36).

However, during this assessment period,

approximately 16 out of 18 plant perturbations (reactor trips, turbine

runbacks, and power reductions or shutdowns) were the result of either plant

material conditiorr (primarily in the balance of plant) or maintenar1ce and

surveillance errors.

For example, a reactor coolant pump motor failure (refs.

82); a main transformer differential lockout (refs. 38 and 80); and a trip of

a main feedwater pump due to a lubricating oil fitting failure (refs. 29, 67,

and 69) caused reactor trips.

In addition, other equipment failed during the post trip recoveries.

For example, following Unit 2 reactor trips the main steam reheater control

system initially would not reset, a condensate polishing building bypass valve

would not close, one of the main steam dump valves did not automatically open,

an individual rod position indication light was delayed (ref.82), and a main

steam dump valve opened unexpectedly and remained open longer than expected

(refs. 38 and 80).

Following a Unit 1 reactor trip the turbine-driven AFW

pump tripped on overspeed.

Less significant discrepancies were also noted

during the post-trip response including; two reactor coolant pump annunciators

alarmed, a feedwater pump recirculation valve position indicator light did not

illuminate, and an individual rod position indication light was delayed (refs.

38 and 80).

Other unexpected equipment failures also challenged site personnel, such as a

Unit 1 turbine runback that occurred following a failure of the K-2 control

rod position indication (ref. 29), and a hole in the service water outlet

piping for the recirculating spray heat exchanger represented a potential

pathway for radioactivity to leak outside containment.

Increased inspection in this area is recommended.

11

4.4

Quality of Maintenance Work

The quality of maintenance and surveillance activities was generally good.

NRC inspectors identified few maintenance personnel shortcomings during their

observations of maintenance work.

For example, maintenance support of the

Unit 2 core uprate generally enhanced plant safety (ref. 45).

Ho~ever,

personnel errors during maintenance and survei 11 ance activities resulted in

unnecessary challenges to equipment and personnel.

For exlmple, 50 percent of

the Unit 1 control room annunciators fail.ed due to maintenance error during

troubleshooting (refs. 43 and 45); a Unit 1 manual reactor trip was initiated

in response to a loss of a main feedwater pump due to accidental bumping of a

relay during a routine safeguards actuation logic test (refs. 9 and 61); and

welding activities on the Unit 1 primary system initiated a hydrogen burn

inside-the pressurizer (refs. 1 and 52).

Personnel errors also resulted in the loss of both Unit 1 source range nuclear

instruments for approximately one minute (refs. 48 and 86), and caused

all three pressurizer pressure protection transmitters to be out of

calibration (refs.* 35, 73, and 74).

Foreign material exclusion problems were identified as a 1ong standing and

recurring prob1em by both the NRC and the licensee (ref:;. 6, 14, and 24).

The

1icensee has initiated corrective actions and plans to audit this area pr'ior

to the 1996 Unit 2 refue1ing outage (ref. Surry Integrated Assessment

Schedule).

Normal inspection in this area is recommended.

4.5

Programs and Procedures

The quality of procedures steadily improved as a result of the licensee's

technical procedure upgrade program (ref. 51).

Procedural usage by

maintenance personnel was-usually consistent with licensee management

expectations (refs. 35, 43, 45, and 48).

The maintenance process for

troubleshooting and repairing a turbine-driven auxiliary feedwater pump was

ineffective in part because procedures were not used to perform the

maintenance (refs. 29 and 32).

Licensee assessments identified problems in the maintenance and test equipment

program and in the safety and relief valve program (refs. 150 and 169).

All

three Unit 2 pressurizer low pressure protection channels were inoperable due

to the use of an uncompensated test gage and weaknesses in the M&TE program

(refs. 35, 73, and 74).

The licensee's in-service inspection (ISi) program contained the necessary

procedures, which were well-written.

ISi examinations were performed

satisfactorily (ref. 33).

The licensee's snubber surveillance program was

inspected and it complied with technical specification requirements (ref. 44).

The licensee's preventive maintenance program will be evaluated further during

the on-site assessment to determine if it is effective in preventing material

condition deficiencies in aging plant equipment.

12

. - *

~

Normal inspection in this area is reconvnended.

5.0

PLANT SUPPORT

5.1

Safety Focus

5.1.1

Radiological Controls

The overall radiation protection safety focus was strong and well directed.

Strong corporate and station management support (along with active worker

involvement) for the ALARA program was instrumental in the program's success

(ref. 3).

The Five-Year Rad Reduction Program includes plans for radiation

source-term reduction along with direction and guidance for continued ALARA

program success (ref. 40).

The radiation protection organization was stable,

with no significant changes in lines of authority, and the number/level of

staffing adequate to support outage and normal operations (ref. 31).

The recent outages for both Units have been well planned, supported, and

effectively managed; the 1995 Unit 2 outage had the lowest ever worker

cumulative exposures (ref. 143).

Successful outages are clear indications of

good cooperation and communications between radiation protection and the

operations and crafts (refs. 4, 31, 34, 3:, 42, and 50).

Management consistently placed strong emphasis on improving and maintaining

the material conditions by actively reducing contaminated areas.

For example,

the auxiliary building restoration project significantly improved and eased

worker access (refs. 29 and 34).

Reduced inspection effort in this area is recommended.

5.1.2

Security

Management safety focus was evidenced by continued improvements in the

program.* An example was the recent implementation of the hand geometry based

access control system.

Management support for the physical security program

at the site ensured adequate level of staffing, training, and motivation for

the security force (ref. 518)

Reduced inspection in this area is recommended.

5.1.3

Emergency Planning

The satisfactory performance of the licensee during drills and exercises

demonstrated their ability to respond effectively to emergencies at the site.

During the 1995 exercise, the emergency facility was promptly staffed and

activated.

The on-site emergency organization was effective and had

sufficient staff to deal witn the simulated event.

The scenario was

challenging and fully exercised the licensee's on-site and off-site emergency

organization.

The licensee's ability to classify the simulated event was an

exercise ~trength.

Minor problems with the off-site notification system were

~uickly resolved (ref. 39)

13

' "'*

In support of the emergency planning program, emergency response facilities

continue to be maintained.

Additional communications capability has been

added to p_rovide additional communication channels for various emergency teams

(ref. 19). Training programs continued to be effectively implemented (ref.

51).

Interviews with off-site agencies revealed that the licensee has

developed and maintained very strong relationships with the state and local

support agencies.

Excellent critiques, detailed*audits, and tracking

corrective actions, strong management support indicated an overall excellent

program (refs. 19 and 24).

Normal inspection in this area is recommended.

5.2

Problem Identification and Resolution

5.2.l

Radio1ogica1 Contro1s

Qua1ity Assurance audits, survei11ance programs, and the radiation protection

se1f-assessment program are we11 organized and provide effective oversight of

the radio1ogica1 program.

These audits (or.sistently identified substantive

issues and problems, and tracked appropriate corrective actions.

The audits

have a low threshold for problem identification, as evidenced by the number

and type of findings.

Lessons learned and items for improvement are clearly

communicated (refs. 3, 11, 40, 42, and 50).

For example, QA found that forms

in use in the plant were not consistent with the current applicable procedure

(ref. 158).

The detailed radiological self-assessments were comprehensive and well

documented.

One excellent example of tracking and trending program

performance was in the area of radiation controls procedure compliance.

The

QA group had earlier made a finding in this area and decided to keep it open,

based on continued similar occurrences where workers were not following RWP

access requirements (ref. 3, 11, 40, and 50).

Some weakness was noted in the corrective action program in response to NRC

and licensee identified events.

Two instances where corrective actions to

prevent recurrence of NRC-identified problems only focused on the immediate

event, and did not address a broader programmatic view.

Both instances

involved the use of a forms that were not part of the approved governing

procedures (refs. 23 and 31).

Reduced inspection effort is recommended for problem identification.

Normal inspection effort in the area of corrective actions is recommended.

5.2.2

Security

The licensee effectively used the yearly Quality Assurance Audits and the

quarterly security self assessment program to identify weaknesses (refs. 154,

162 and 165).

The QA Reports identified a number uf deviations in the

f~~ness-for-duty program and procedures, communication equipment testing, and

the uninterruptable power supply (UPS) within the security plan.

The security

self assessment program provided an internal review of the security

14

~.

l'*7:.;;*

department's performance.

These quarterly repor~s examined each segment of

the security program from personnel, equipment, procedures, and other

supporting programs and conclusions were made {refs. 178 and 179).

Based on the sample of problems- identified by audits, inspections, and event

logs, the review and analyses were appropriately assigned, analyzed and

prioritized, and corrective action was adequately performed in a timely

manner.

Any negative trends noted resulted in the licensee developing an

enhancement program to correct the situation (ref. 518).

Reduced inspection in this area is recommended.

5.2.3

Emergency Planning

The licensee conducted thorough and extensive annual audits of the emergency

preparedness program with the assistance of technical specialists from outside

the company.

The overall finding of the audits was that the emergency plan

satisfied regulatory requirements and that these requirements were effectively

implemented.

No major deficiencies were identified.

However, minor

procedural issues were identified during the audits concerning controlled

procedures and inventory of emergency kits (ref. 167).

The licensee critiques following exercise and drills were very detailed and

complete.

Follow-up was prompt and thorough.

The identified it~ms were

corrected in a timely manner (ref. 39).

Normal inspection in this area is recommended

5.3

Quality of Plant Support

5.3.1

Radiological Controls

The radiological support program provides good job coverage and technical

support to operations and crafts, during norma~ and outage conditions.

Especially noteworthy was the well planned support for the off-normal moisture

carryover testing {ref. 43).

The licensee effectively uses numerous ALARA

techniques, including temporary shielding, hot spot elimination, remote HRA

surveillance, mockup training, and teledosimetry {ref. 3 and 45).

The

controls established to control access and work in the incore sump rooms

(full-time designated VHRA), along with the tag-out of the moveable incores

guard against inadvertent, serious worker overexposure {refs. 3 and 23).

The

control and elimination of contamination is a program strength.

The major

revision to 10 CFR Part 20 was successfully implemented, with a notable

aggressive reduction in respirator use, while maintaining TEDE ALARA with

proper use of engineering controls {refs. 3, 23, and 40).

The 1994 health

physics technical continuing training program was successfully completed (ref.

31).

Three examples of failure to follow radiation c~ .. trols procedures governing

access to the incore sump rooms were identified by the NRC and resulted in a

severity level IV notice of violation {NOV) (refs. 23 and 31) .

15

Normal inspection effort in the occupational radiation program area is

reconvnended.

5.3.2

Security

The licensee effectively maintained radiologica1 controlled area boundaries

and the security perimeter (ref. 43).

Sec11rity personnel were knowledgeable

about fire control responsibilities in conjunction with ccmpensatory measures

during battery room maintenance (ref. 25).

Reduced inspection in this area is recommended.

5.3.3

Emergency Planning

Licensee's performance in the Emergency Planning area continues to be

excellent (ref. 51).

Response of the licensee during exercises and drills was

very good.

During the 1995 exercise, the emergency facility was promptly

staffed and activated.

The on-site emergency organization was effective and

had sufficient staff to deal with the simulated event.

The scenario was

challenging and fully exercised the licensee's on-site and off-site emerge~cy

organization.

The licensee's abi1iJy ~o classify the simulated event was an

exercise strength (ref. 39).

There is strong management support for the program.

Changes in organization

in 1995 did not affect the effectiveness of the program (ref. 19). Audits and

exercise critiques were found to be detailed and comprehensive.

Corrective

actions were timely and thorough (refs. 19 and 24).

Normal inspection in this area is recommended

5.4

Programs and Procedures

5.4.1

Radiological

The site has effective effluent and environmental controls programs, with well

trained, knowledgeable health physics technicians (HPTs).

Radwaste processing

and shipping was conducted in a competent, professional manner, while the

volume and number of radwaste shipments has remained relatively constant over

1991-1994.

An example of the quality of the radwaste program and the

commitment to occupational ALARA is the significant exposure reduction for

processing a high integrity container (refs. 11, 20, and 42).

A NCV was issued as a result of multiple licensee-identified examples of

workers failing to adhere to rad control procedures during outage work;

examples included failure to wear extremity dosimeters for S/G work, not

wearing teledose dosimetry, work in the seal table area without dosimetry

(DAD).

The invnediate corrective action was effective.

As a result, the

licensee requested independent corporate assessment and kept specific work

groups out of radiological controlled areas (RCA) until corrective actions

were implemented.

They also instituted dedicated health physics technician

16

(HPT) at access points to improve access compliance during the outage.

Previous licensee audit findings had identified problems with worker

compliance with HP procedures (ref. 50).

A notable. radwaste ~perational problem was the over pressurization of a

chemical radwaste tank, which resulted in personnel injury and the declaration

of a notification of .unusual event (NOUE).

The event details and the

licensee's corrective actions will be reviewed during the on-site IPAP team

inspection (refs. 27 and 34}.

Normal inspection is recommended in this area.

5.4.2

Security

The NRC verified that the security plan and procedures were in compliance with

the regulations.

The licensee effectively implemented the security plan and

applicable procedures for training and q11alification, contingency plans,

security patrols, inspection and testing, defensive positions, physical

security barriers, security alarm station operation, and security records

(refs. 5, 30, and 518).

The Fitnes5-For-Duty Program was thorough (ref. 12).

The licensee's Access Authorization ;,ograrn was revi~wed against the

requirements in 10 CFR 73.56.

The program and procedures appeared to be well

managed and thorough for all aspects of the program (ref. 41).

Reduced inspection in this area is recommended .

5.4.3

Emergency Planning

Minor problems with procedures involved the maintenance of controlled copies

in one of the emergency response facilities and controlled documents in the

emergency kits (ref. 167).

The licensee's system for making changes to the Surry Emergency Plan and EPIPs

was found to be effective.

Changes to EPIC and plans were approved,

documented, and distributed on a timely basis (ref. 19).

Normal inspection in this area is recommended

17

..:.-:..:-........

APPENDIX A

LIST OF REFERENCES

Inspection Reports 1993-~ 1994

Ref.

No.

0

93-16

SALP Report for Period, 4/5/92 to 7/3/93

l

2

3

4

5

6

7

8

9

10

11

12

13

14

15

16

17

Jg

94-02 Core Resident Inspection; Inservice Testing

94-04 Routine Inservice Inspection in regard to 2nd ten year interval

94-05 Occupational Radiation Exposure

94-06 Core Resident Inspection; Refueling

94-07 Physical Security

94-08 Core Resident Inspection; LER Fol,ow-up

94-09 Service Water System Operational Performance Inspection

94-11

Core Resident Inspection; On-site Engineering Review

94-12 Core Resident Inspection; On-site Engineering Review; LER

Follow-up

94-13* Low Head Safety Injection flow testing; Setpoint Validation

Program

94-14 Radwaste Mgt.; Effluent Reports; TS Chemistry Parameters

94-15 Fitness for Duty

94-16 Review of S/G tube pulse cleaning; FW system leaks; Flow Accel.

Corrosion

94-17 Core Resident Inspection

94-18 Licensee Compli~nce with Generic Letter 89-10 for Safety-related

MOVs

94-19 Review of Station Blackout Commitments

94-20 Review of Requalification Program for ROs and SROs

94-21

Core Resident Inspecti*on; Safety Assessment & Quality

Verification; LER Followup; Engineering Tech Spec Review

A-1

Ref.

No.

1994 Inspection Reports Cont'd .

19

94-22 Emergency Planning

20

94-23 Org. of Chem.Dept.lRW Group;PW l SW Chemistry;REMP;CR Emer. Vent.

21

94-24 Core Resident Inspection; ~ER Followup

22

94-25 Engineering l Tech. Support (Org.; Training; Mgt.; Interfaces;

Controls)

23

94-26 Review Implementation of New JOCFR20

24*

94~27

Core Resident Inspection; On-site Engr. Review; Annual Emergency

Exercise; LER Followup

25

94-28 Core Resident Inspection; On-site Engr. Review; Plant Support;

LER

26

94-30

27

94-31

LER

28

94-32

Foll owup

Engineering & Tech. Su,pport {Procedu,*es; Personnel Interviews;

Activities in Progress)

Core Resident Inspection; On-site Engr. Review; Plant Support;

Followup

Special Inspection To Review Inoperability of Unit 2 Station

Battery

29

94-33

Core Resident Inspection; Plant Support; Safety Assessment &

Quality Verification

1995 Inspection Reports

30

95-01

Physical Security

31

32

33

34

35

95-02 Occupational Radiation Exposure

Core Resident Inspection; On-site Eng.

95-03

95-04

Inservice Inspection; Flow Accelerated

Replacement

95-05 Core Resident Inspection; On-site Eng.

Foll owup

95-06 Core Resident Inspection; On-site Eng.

A-2

Review; Plant Support

Corrosion; FW Heater

Review; Plant Support;

Review; Plant Support

LER

. *.

~ :

.. *-~** .. : :. -~ ...

.. ...

1995 Inspection Reports Cont'd .

36

95-07 Core Resident Inspection; Plant Support; LER ~ollowup

37

95~08

Core Resident Inspection; On-site Engineering Review; Plant

38

95-09 Core Resident Inspection; Plant Support; LER followup

39

95-10

Emergency Planning

40

95-11

Occupational Radiation Exposure

41

95-12 Access Authorization Program per lOCFR73.56

42

95-13 Org. of Chem. Dept. l RW Group; PW l SW Chemistry; REHP;

Transportation of RadwJste

43

95-14

Core Resident Inspection; On-site Eng:neering Review; Plant

Support; LER Followup

44

95-15 Surveillance of Snubbers; New Construction; Transport of Spent

Fuel Casts

45

95-16 Core Resident Inspection; On-site Engineering Review; Plant

Support

46

95-17 Core Resident Inspection; On-site Engineering Review; Plant

Support, Turbine Building Flooding; Discrepancy Reports Review

47

95-18 Review requalification program of ROs and SROs

48

95-19 Core Resident Inspection; Refueling; On-site Engineering Review;

Plant Support; Self Assessment; LER Followup

49

95-20 Special Inspection for Loss of Rx Inventory

50

. 95-21

Occupational Radiation Exposure

51

95-99

SALP Report for Period 7/14/93 to 1121/95

SIA

95-22 Core Resident Inspection; Reactor Trip

518

95-25 Physical Security

A-3

~

(.-

v'

-

Licensee Event Reports in 1994

Ref.

No.

52

94-001-00 (50-280)- Hydrogen in PZ 0 ~gnited by welding

53

94-001-00 (50-281)- Both trains aux. vent. filtered exhaust inoperable

94-002-00 (50-280)- Test of MSSVs indicated setpoint drift

54

55

56

57

58

59

60

61

62

63

64

65

66

94-002-00 (50-281)- Both trains aux. vent. filtered exhaust inoperable

94-003-00 (50-280)- Small hole in piping of RSHX B

94-003-00 (50-281)- Failure to close eves makeup valve following

dilution

94-004-00 (50-280)- Hot leg isol. valve not operated per T.S. time limit

94-004-00 (50-281)- Leo not timely for Inoperable Battery

94-005-00 (50-280)- Unit Output exceeds T.S. limit due to instrument

.error

94-006-00 (50-280)- Manual Rx trip due to MFW isolation causing S/G

levels to rise

94-007-00 (50-280)- Contrary to T.S.-Vent Stack Rad Monitor INOP

94-008-00 (50-280)- Both aux. vent. trains INOP due to a single event

94-008-01 (50-280)- Revised LER 94-008-00 to show violation of T.S.

because sampling of filters not performed

94-008-02 (50-280)- Inconsistency in verifying operability of new

charcoal beds thus LER revised

94-010-00 (50-280)- EDG Battery Surveillance Missed Due to Personal

Error

A-4

..

. -

.,_(

Licensee Event Reports in 1995

Ref.

No.

67

68

69

70

71

72

73

74

75

76

77

78

95-001-00 (50-280)- Auto Rx Trip duP to failure of MFP Coupling

95-001-00 (50-281)- PZR H/U exceeds T.S. (Procedure Control)

95-001-01 (50-280)- Revised_LER due to clarification for turbine driven

AFW

pump trip following the auto Rx trip

95-002-00 (50-280)- Smoke detectors not tested per T.S. Surv. Time limit

95-002-00 (50-281)- MS & PZR Safety valves out-of-tolerance

95-003-00 (50-280)- Auto. initiation of AFW@ Lo-Lo S/G level

95-003-00 (50-281)- PZR P.T.s out of calibration due to faulty gauges

95-003-01 (50-281)- PZR P.T.s out of calibration due to faulty gauges

95-004-00

95-005-00

(50-280)-

(50-281)-

(50-280)-

not being temperature compensated basis for LER

revision

Missed Battery Surveillance Due to Personal Error

Installation of Damaged Circuit Card Resulted in

Unit 2 Manual Trip

Error in calculation to convert T.S. NAOH Volume

Level

(50-281)- Manual Rx Trip due to control rods dropping into

core

to

79

95-006-00 (50-280)- ESW Pumps inoperable due to marine growth

80

95-006-00 (50-281)- Auto Rx trip due to main xfmr protective relay

actuation

81

95-007-00 (50-280)- Operation with non-isolable leak in PZR

instrumentation nozzles

82

95-007-00 (50-281)- Rx trip due to failed Rx coolant pump motor

83

95-008-00 (50-280)- PZR Safety valve as found setpoint out of tolerance

84

95-009-00 (50-280)- Loss of 4KV BUS then EOG s+3rt due to personal error

85

95-010-00 (50-280)- Four CCW heat exchangers inoperable due to

macrofouling

A-5

,.

' ....

-.... ..

Licensee Event Reports in 1995

86

95-011-00 (50-280)- Both source range nuclear instruments de-energized

due to personal error

ML18152A058

Text

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