IR 05000413/1989032
| ML19354E775 | |
| Person / Time | |
|---|---|
| Site: | Catawba  |
| Issue date: | 01/23/1990 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML19354E770 | List: |
| References | |
| 50-413-89-32, 50-414-89-32, NUDOCS 9002010321 | |
| Download: ML19354E775 (27) | |
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ENCLOSURE INTERIM SALP BOARD REPORT
U. S. NUCLEAR REGULATORY COMMISSION
REGION II
SYSTEMNTICASSESSMENT.OFLICENSEEPERFORMANCE INSPECTION REPORT NUMBER 50-413, 414/89-32 DUKE POWER COMPANY CATAWBA UNITS 1 AND 2 AUGUST 1, 1988 - OCTOBER 31, 1989-I l
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TABLE OF. SUMMARY OF RESULTS 1.
Overall Facility Evaluation During this SALP assessment' period, the ' Catawba-facility was effectively managed and achieved a satisfactory level of operational
safety, corporate leadership, direction,. and ' support.
The station
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staff is technically competent, well qualified.~and dedicated.
Management involvement in problem resolution as? well > as routine facility operations is focused on-safety.. A. comparison of the ratings of the last period and this period indicates no change in any
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area other than Emergency Preparedness which improved.
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The operating history for Catawba improved during ' this < assessment'
. period. Unit 1 continued to be a steady performer experiencing two
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period. Unit 2 showed markedzimprovement experiencing five reactor trips during this assessment period compared to ten during the last.
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period. This improvement is due in. part. to' plant: modifications to -
steam generator level instrumentation.
Some improvement was noted with respect to procedural adequacy and adherence but ambiguities and inadequacies led to misinterpretations and. inappropriate operator
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actions. Although the overall per_formance in this area was adequate, continued attention in the areas of procedural' adequacy, procedure adherence and personnel errors is warranted.
f The radiological controls prog' ram was-adequate to protect the workers and the general public.
The licensee : improved the personnel l
contamination. monitoring program and collectivei radiation doses L
remained low.
The radiation protection. staff is ' experienced and staffing levels were. adequate and remained steady. -Weaknesses in
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door' maintenance-and securing procedures'were identified as a result
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of several examples of unsecured high radiation areas.
Significant.
improvements were.noted in count room activities. Liquid and gaseous
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effluents were within regulatory limits.
The level of performance Lin the Maintenance / Surveillance area was.
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adequate, however, a number-of. problems were; detected = relative to i
procedure adequacy, and adherence.
The quality: of post' maintenance
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testing declined with several instances of degraded equipment being-
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returned to service after maintenance. -The Maintenance Engineering i
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Services group has matured over the assessment ~ period to become-a vital ~part of the statio'ns problem resolution program.
Improved performance in the Emergency Preparedness area was noted.
There was a high degree of management commitment towards improving
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emergency facilities.
Several strengths' were noted including-
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comprehensive, detailed independent audits; a well trained onsite emergency, organization and an effective tracking system to ensure prompt corrective action on identified problems. The annual exercise.
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was fully successful with no weaknesses identified.
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Performance in the Security and Safeguards area = declined over the assessment period. Multiple problems with the protected and vital areas access control program were experienced. Multiple examples of-
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failure of personnel to comply with access control and safeguards information-handling procedures and the several examples.of failure'
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to monitor and report.the training and qualification status of
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security personnel raised concerns -regarding the' effectiveness -of-
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these programs.
Licensee physical security, contingency and guard training and qualification plan change submittals_were normally well
coordinated and technically sound.
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Overall performance in the Engineering and. Technical. Support area has
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been adequate. Strengths were noted in the operability determination.
process and engineering support.
There were, however, several examples of engineering errors in operability determinations.
Effective engineering support was demonstrated on several issues, i
Design change development and implementation 'were' generally acceptable, however, an inadequate post modification test caused:
undetected inoperability of a safety system component and resulted in escalated enforcement. Drawing' deficiencies indicated a weakness in engineering support for design change implementation.
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Safety Assessment and Quality Verification-performance has been good.
Effective management involvement. in site activities' associated with the -licensing areas was evident and management is. well aware of."
generic and plant specific safety issues. The licensee has _ improved i
.in the area of self assessment and appears to have a broad. based and-usually aggressive program, i
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2.
Facility Performance Summary
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Rating Last-Rating This Functional Area period Period
. Trend Plant Operations
2 I-(Operations & Fire Protection)
Radiological Controls
2 Maintenance / Surveillance
'2 2.
1 Security
2 D'
Engineering / Technical. Support
.2 (Engineering, Training & Outages)
Safety Assessment /
2 Quality Verification (Quality Programs & Licensing)
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I - Improving D - DeciWng
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III. CRITERIA
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Licensee performance is assessed in the functional areas shown above.
L Functional areas normally represent areas significant to nuclear safety-and the environment. Special areas may be added to highlight significant-
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observations.
- l The evaluation criteria which were used, a; applicable, to assess each functional area are described _in detail in NRC Manual Chapter 0516. This
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chapter is in the Public Document Room files. Therefore, these criteria
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are not repeated here, but will be presented in detail at the public
meeting to be held with licensee mana0ement on January 31, 1990. However, the-NRC is not limited to.these criteria and others may have been used where appropriate, l
e On the basis 'of the NRC assessment, each functional area evaluated is_
t rated according to three performance categories. The definitions of these-performance categories are shown here only because of some changes in the
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NRC Manual Chapter noted above.
These new performance categories are
defined as follows:
1.
Category 1.
Licensee management attention. and involvement are
readily evident and place emphasis on superior performance of nuclear safety or safeguards activities. with the resulting performance substantially exceeding regulatory requirements.
Licensee resources are ample and effectively used so that a high level of plant and personnel performance is being achieved.
Reduced NRC attention may
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be appropriate.
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Category 2.
Licensee management attention -and involvement in the performance of nuclear safety or safeguards activities are good. The
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licensee has attained a level' of performance above that needed to
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meet regulatory. requirements.
Licensee resources are adequate and reasonably allocated so that good plant and personnel performance is being achieved. NRC attention may be maintained at normal levels.
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Category 3.
Licensee management'. attention to and involvement in the performance of nuclear safety or safeguards activities are not sufficient. The licensee's performance does not significantly exceed that needed to meet minimal regulatory requirements.- Licensee-resources appear to be strained or not effectively-used._ NRC --
attention should bc increased above normal levels.
The SALP Board may also include an appraisal of the performance trend of a-functional area. This performance trend will only be used when both a
- definite trend of performance within the evaluation period is discernible and the Board: believes that continuation of the trend may result in a change of performance level.
The trend, if used, is defined as:
Improving: Licensee performance was determined to be improving during the assessment period.
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l-Declining: Licensee performance was, determined to be declining during the assessment period and the licensee had not taken meaningful steps to address this pattern.
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IV. Performance Analysis
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Plant Operations
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Analysis j
i Plant management is involved in the daily operations of the plant. _ Morning status meetings are effective in summarizing
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plannad activities, accomplishments, and identifying problem areas.
Similarly. management in most instances effectively _
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participated in the pursuit of problem resolution and the L assure.nce of quality.
Management policies are adequately established and disseminated to personnel.
Catawba continued to maintain high standards of housekeeping-throughout the plant. Periodic. inspections by plant supervision and the trending of results are effective in maintaining the
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material condition' of the plant.
Labeling programs for'
.l equipment and doors have enhanced the ability of personnel to correctly locate and identify components. Weak areas remain in i
the labeling of instrument root valves and' adequate lightingL in
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certain areas of the auxiliary building.
One. violation was identified involving inadequate measures to assure cleanliness in containment.
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l Unit 1 continued its steady performance during this assessment period. The unit experienced two manual-reactor. trips and one-l automatic trip followed by 'a safetyL injection.-
Both manual J
reactor trips resulted from identical causes involving the-same feedwater regulating valve failing shut. -The root cause i
for the first event was insufficiently evaluated-and:the failure-l I
' repeated itself. Evaluation and corrective action-following the second trip were more extensive and effective. The automatic
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l trip was the result of an operator's inattention -to - detail n
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during an attempt to operate a valve from the control' board. : A
- l wrong button was pushed causing a main steam isolation valve _to l
shut. A reactor trip and safety injection resulted.
R Unit 2 experienced five reactor trips (three of L which were j
manually initiated) and two safety injections during the.
assessment period.
The following personnel -errors were H
contributing factors:
Operator error while controlling main feedwater pump speed
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caused a loss of feedwater.
Operator failure to monitor cooldown rata contributed to a -
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safety injection.
Maintenance technician error while troubleshooting ' caused a
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main steam isolation valve to shut.
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Maintenance technician error while painting caused a loss
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of main generator cooling.
One trip was due to a decision by management to'not. replace previously identified defective fuses used in non-safety related-applications.
The safety related fuses were replaced as required.
l Secondary problems also contributed to the reactor, trips. The
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licensee subsequently performed modifications to the steam dumps
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and feedwater regulating valves which resulted in.a marked y
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improvement in performance.-
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-l The previous SALP noted a significant number of reactor trips on
Unit 2.
Many of these.-trips either directly or indirectly'
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related to the narrow steam generator water level operating band--
associated with the Westinghouse _Model D-5 steam generatort.
During this assessment period, none of the trips on Unit 2 werel R
related to 'the narrow operating water level band.
This - is partially due to the licensee's efforts to train, operators l to; more effectively respond to water level-transients.
Furthermore, during the Unit 2 refueling outage.which began in'.
March 1989, modifications were made to the steam generator-level monitoring instrumentation to expand the operating band such that the D-5 steam generators behave similarly to the Unit:1,-
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D-3 steam generators.
Following-implementation of1 the modification, operators observed steam generator: level control to be significantly l improved. - Unit 2 - experienced no' reactor.-
trips for the duration. of_ the period.
Several ambiguities or inadequacies -in Operations. procedures l
caused misunderstandings or misinterpretations by - personnel'.
Some resulted in violations 'of Technical Specifications..
Examples _of these included. the. following:-
an inadequate procedure which contributed to a situation where the" residual'
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heat removal flow rate. was. reduced below Technical Specification
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required values; -induced instrument _ errors in nuclear
instrumentation were not-accounted for in a. plant shutdown j
L procedure; the procedure for plantJshutdown during chemical
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contamination of steam generators required a power. ; reduction
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faster than the unit could reliably endure; and operatorsiin one
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case performed evolutions 'not covered by procedures to : adjust
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boron concentration in the' cold leg. accumulator by ' draining
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below the indicating range.
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The Emergency Operating Procedures were determined to adequately-l
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cover the broad range. of accidents and e'quipment failures necessary for safe control of the plant. Only minor' problems i
were found. There were many inconsistencies between.Lthe E0P.'s -
and the Plant Specific Technical Guidance (PSTG) where there.
should be none. The licensee committed to. provide sufficient resources to the E0P technical verification process.
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Many _
examples of?
human factor and-technical
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. deficiencies were also identified-which the. licensee committed.
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to review and correct as necessary. Some examples of inadequate i
support equipment, such as ladders and emergency: lighting:for -
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local valve' operation required ~ in - emergency ' procedures, were.
noted.
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Weaknesses were identified in the removal of. equipment from-U service without complete. evaluation. of the consequences. One-
' J violation identified 'the failure - to' perform a 10 CFR 50.59
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evaluation prior to disabling a flow control valve in a Control Room area ' ventilation chiller.
A second violation was the failure ' to consider long term ' operability requirements off a diesel generator prior to removing _ both starting air compressors from -service.
Another weakness ' identified: the_ failure to j
implement compensatory measures during a
period. where-D modifications rendered. nuclear service' water system' strainer
differential pressure annunciators inoperable..
. Management's understanding and response to technical; issues from-
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a safety perspective was-apparent and generally conservative.-
This is supported by a plant shutdown when steam' generator tube j
leakage approached 100gpd;.although a higher value is. allowed by,
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Technical Specifications. Management elected _ to ~ suspend outage _.
critical path activities _= when one residual < heat Lremoval pump
was lost. The outage was extended to: repair; the pump. _If; the repair had been performed later in the outage it may_ not have H
i delayed critical path < activities.
This. demonstrated a i
sensitivity towards the potential for.losingt decay heat removal
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capabilities.
Also on an'other t occasion !with a - pressurizer
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oressure instrument line compression' fitting leak the plant was
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shutdown even though Technical Specifications did not require a shutdown.
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'i One example in which management decisions were not considered to:
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be conservative involved-actions taken following an overspeed l
event of the turbine. driven. auxiliary :feedwater' pump.
Ineffective repair activities, conducted outside the scope.of the maintenance program, allowed: the pump ' to be: returned to
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service without the ; root.cause being identified; and adequate.
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corrective actions taken.. Management relied upon the results of a surveillance test to justify operability although the pump was not adequately tested, in-- the '.'as found" condition.
The pump failed-a surveillance test one week-later.- An engineering evaluation was not performed ts justify operability nor were
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compensatory measures imposed; while scheduled ' maintenance proceeded on redundant equipment.
In another event, an operability evaluation, performed for an-auxiliary feedwater valve which had failed. to close under _dp conditions, concluded - the valve was operable pending de' sign engineering review.
In a subsequent analysis the event was determined to be the precursor of a generic operability issue.
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Management initiated programs to improve ;the: reliability of -
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annunciators and other equipment. A goal of achieving a " dark
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board" Control Room annunciator status was established.
Significant resources were required.to repair equipment.and to modify annunciator logic.
The status of outstanding
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annunciators is presented to management on a weekly basis. Dark.
boards have been achieved on.two occasions for brief-periods of
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time on Unit 1.
Management has taken -steps 'to minimize the length of time that
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Technical Specification required equipment is out of. service.-
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Personnel responsible - for removing equipment from ' services are held accountable to provide status updates at daily meetings.
These efforts have been effective and sensitivity levels have
been elevated in this area.
Operators maintained a high degree of professionalism in the :
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Control Room, which included, adequate access controls,-
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effective, shift turnovers, good Control.. Room decorum, thorough auxiliary operator rounds and professional-attitudes.
The previous SALP evaluation : identified a significant number of Engineered Safety. Features -(ESF) actuations. with;50% resulting from personnel errors. Management efforts to. reverse this trend i
focused on requiring more involvement ' by -Senior Reactor.
- j Operators in routine plant operations, and the encouragement. of
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all operators to work as a team and' consult: with each other.
Improved performance in this-area has been noted.- However, as previously mentioned personnel errors contributed :to.-several a
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reactor trips and safety injections, during this. period.
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Additionally the failure of-operators. to maintain visual-
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observation and communications during a refueling cavity-filling-i
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evolution resulted in overfilling the cavity.and a 25000 gallon o
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Operators generally responded'well to abnormal plant _ conditions.
One event involved the: loss of a train of essential switch gear, while shutdown,-due to an installation deficiency in a normal.
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auxiliary power supply circuit breaker.. This. caused a loss of-j residual heat removal, component cooling, and. the charging flow.
control valve to fail open resulting in a reactor. coolant system:
pressure increase.
Operators promptly diagnosed the.-transient and restored residual heat removal flow, component cooling and'
pressure control within.two minutes. Another event,:however, involved. the failure by an -operator ' to promptly respond to-indications of inoperable auxiliary feedwater trains and a demonstrated lack of understanding of. the safety significance of certain features of the system.
Two replacement examinations were administered with 5 of 6 R0s and 8 of 11 SR0s passing. One requalification examination: was'
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administered.
The operator training program was rated satisfactory.
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A problem was identified by operators involving' excessive noise levels in the~ Control Room following a safety injection due to the automatic startup of standby ventilation fans.
This presents the-potential for, communication difficulties during an emergency. The licensee has initiated a design study.to, conduct tests ar.d identify 'a resolution.
The last SALP identified problems with maintaining valves in a locked - condition., Licensee efforts to improve in this area -
included modifications to valve handwheels -to allow for easier locking techniques and additional operator training. However,
- two violations were identified involving valves found ' unlocked and several examples were noted where valves were not-properly.
j locked. Weaknesses were also~ identified involving inappropriate-
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use of the Operator Aid Computer for verifying valve position.
Inadequate-measures to-identify backseated valves: contributed to a valve being mispositioned and a resultant spill and per6onnel contamination.
Review of-the fire protection and preventive program y
implementing procedures,- surveillance inspection ~and test results, fire brigade staffing and training, and the status of
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fire fighting equipment and fire detection systems demonstrated
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that plant fire protection. features are inservice and functional.
Fire brigade performance was observed during an i
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unannounced drill. The inspector concluded that the performance
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of the fire brigade leader and the fire brigade : was
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The performance of the fire protection program meets NRC requirements.
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Performance Rating Category: 2 Improving
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Recommendations Improvement has been noted. in the Operations area.
Reactor i
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trips were reduced and. personnel errors were down. Deficiencies in operating i procedures and attention to detail by operators contributed to operating errors. These areas require additional i,
management attention.
B.
Radiological Controls
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Analysis Overall. the. licensee's radiological controls program was adequate to protect the workers and the general public.
The licensee improved the personnel contamination monitoring program and collective radiation doses remained lo.
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The' licensee's radiationL protection staff was experienced, staffing levels were adequate and remained steady and=approriate
~a support by contractor personnel-was available during outages.
Several examples of unsecured.high radiation areas were noted by i
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the-licentee and the NRC inspectors.. There were no unexpected q
doses, however, weaknesses in door maintenance and securing.
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procedures were identified, l
During' the previous SALP per"od che NRC identified failures to; adhere to radiological control-procedures. for personnel
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contamination monitoring resulting in. ' violations.- Initially-i during this period the licensee failed to take effective corrective ' action resulting 'in ' a repeated violation this.
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assessment period.
An additional corrective action plan to
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aodress the problem-areas. was required and implemented by the licensee.
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I The addition of new personnel ' contamination monitors: has.
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significantly improved the. licensee's. ability lto-detect'
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New whole body' contamination monitors were installed in early 1989.
This is.one area whereL corrective'=
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l actions taken have improved performance.
As of ' October 31, 1989, the licensee had already recorded 306 personnel skin and clothing contamination events for:1989. This isthigher than the l
total documented in the previous two years in which the/ licensee i
documented 220 and 281 personnel; contaminations respectively.
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capabilities of their new monitoring equipment -
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The licensee formed a crew dedicated to decontamination eff' ort, The contaminated area. decreased - from 16,000; square feet to-the u
current area of about -10,000 square feet or' 6 percent: of-the radiologically controlled areas of.the plant. Further reduction:
was planned. The licensee's goal for-1989 was4to treduce the total area contaminated in non-outage. periods.to 5 percent.
i The station's 1989 collective dose goal was: established at 474
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person-rem. By the end of the assessment -period, the licensee
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had accumulated 318 person-rem of the 1989: dose goal
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licensee.was below the projected 1989 collective -dose total for
the period and on schedule to meet-the annual goal. This can be
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attributed to-section supervisors increased involvement with.the J
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facility's ALARA program through'ALARA Committee'assignmentst
1 Significant improvements were noted concerning: count room
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quality assurance as compared to. the previous period.
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Comparisons of counting results were in agreement with NRC
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results with one exception. Disagreements were due to' problems in: splitting ' a-gaseous sample and were 'not. attributed to
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instrument calibrations or laboratory quality-assurance.
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Liquid and gaseous radioactive effluents were within-the-Technical Specification limits and'in compliance with 40 CFR 190 limits for radiation dose.= One unplanned gaseous release was reported during the SALP per_iod.
The release: occurred during April 1989 when approximately 63 cubic feet' of radioactive gas -
leaked from the Shutdown Waste Gas Decay Tank into the Auxiliary Building. The release was a small fraction of the -Technical Specification limits.
There have been problems with radiation monitors.
Steam-generator blowdown monitors inoperable for several months. The monitor for each unit is located on a~ sampling _line off the steam generator blowdown common header. The-licensee determined that the system has problems with flow control and plugged'11nes to the monitor.
A station modification had been initiated to install 4 new-EMF monitors (one for each steam generator) to i
monitor individual ' generator blowdown' and is scheduled to be-implemented during 1990.
The Unit 2 turbine building sump '(TBS) monitor. was-. inoperable from March to August, 1989 due.to an inoperable flow switch and low flow rate out of the pump. One weakness involved a failure-
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to conduct analyses of the TBS liquid on" a 24 hour: basis when-
the TBS monitor was inoperable.
l The licensee has ' effectively maintained _' primary chemistry-well l
within Technical Specification - requirements and secondary
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chemistry ' well within the limits recommended by the. Steam Generators Owners Group..Biofouling problems = in the service water systems have received increased attention.
The licensee has increased maintenance inspections and, cleaning of - the systems susceptible to this type of corrosion.
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Performance Rating
Category:
d 3.
Recommendations j
The Board notes that the licensee has now aggressively pursued improved ~ personnel contamination monitoring.
The-Board.
encourages the licensee to continue proactive efforts to' correct-those areas or activities-that contribute to personnel contaminations.
l C.
Maintenance / Surveillance
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Analysis
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The maintenance organization is well staffed and adequately-trained. The licensee has organized many of the mechanical
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i maintenance crews according'. to component specialization. The'
licensee has observed many advantages in specialized crews such I
as increased experience and expertise,, an. increased sense of component ownership and increased opportunity - for innovative techni. ques and tool development.
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Maintenance-Engineering Services, which was organized -to I
assemble 42 component. experts under a Senior Reactor Operator i
licensed engineer, has matured to.become a vital component in:
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the licensee's program for solving persistent problems and-i
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maintaining plant reliability.
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Technical: Support programs are' being developed.by each engineer j
in their respective field to document the design basis,
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L regulatory requirements, and predictive maintenance practices to;
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enhance establishing a-proactive approach to maintenance.
Scheduling of periodic surveillance activities was implemented effectively with good use of' plant computers to track ~ and plan i
i testing. Surve111ances were integrated with planned maintenance t
l 1tems to optimize overall system availability. and 'to_ ensure the
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removal from service of one train at a time. Daily planning activities were Lobserved to be efficient including worksite inspection and meetings to prioritize'the work list.-
Several violaticns and weaknesses in the ; performance' _of.
i maintenance and Technical Specification required surveillances were caused by procedure deficiencies.
Surveillances on the Containment?. Hydrogen Monitors and the
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Reactor Vessel Level Instrumentation System were: determined to
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be inadequate in'that -acceptance: criteria.was either lacking or
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insufficient to verify operability.
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Errors in surveillance and maintenance procedures. led to
personnel confusion on a number of occasions ~ and eventual
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violations of-Technical Specifications, i
a Methods employed t'o designate procedural steps which were not applicable to the task beinc performed swere ? inconsistent, contributed to-confusion. and. in one': case. had a _ bearing'-on' the -
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failure to restore the Reactor Vessel Level Instrumentation System after maintenance.
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Other-examples of. test
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weaknesses-included, procedure:
insufficient measures to prevent operating 'a containment. spray -
pump with flows below-the minimum value and an: incomplete method-
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to fill and vent a piping section following a check valve leak.
rate teat.
The method used to full flow test residual ' heat
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removal system check valves was identified to be deficient ire that flow rates of check valves -in_ parallel' were checked,-. as'
opposed to individual valve flow rates. Additionally, a;
maintenance procedure inadequacy for flushing service water
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j lines contributed to an event where auxiliary feedwater valve--
l switches were mispositioned rendering the system inoperable, n
Examples were identified involving an excessive number of pen y
and ink changes making readability difficult and requiring procedure retypes.
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The licensee has recognized the procedure problems' and has taken
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corrective actions. - - Actions -include.- procedure -improvements,
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development of section specific guidance relative to: procedure l
adherence, additional training and - increased ' emphasis : on-
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in-depth root: cause evaluation for errors involving procedure.
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Improvement and strength's :were observed in many1of the:
licensee's preventive. maintenance-practices. Rotating machinery vibration is routinely trended and analyzed in detail.
Diesel
fuel oil and machinery lubricating oil;is routinely sampled and analyzed.for. contaminants and indications or wear.
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Corrective maintenance programs on some equipment, identified to be problem areas in the previous SALP, 'showed ; improvement
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particularly on diesel air systems and: residual heat removal i
isolation valves. Minor improvement was noted in main feedwater regulating valve performance as failures conHnued to produce
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safety system challenges and caused three-_ reactor trips.
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Chronic seat leakage on the-steam-admission. valves to.the
turbine driven auxiliary feedwater pump contributed; to several t;
problems with the pump.
The licensee's performance in the area of testing systems 'and '
components following maintenance activities clearly declined.-
In several instances 1the post maintenance testing program was
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inadequate to demonstrate-the ability; of components to perform
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satisfactorily in service. ^ This. became. apparent ' when ; the
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components subsequently failed.
The circuit breaker for a
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Containment Hydrogen Skimmer Fan was-replaced and returned to j
service without the requirement.to run the fan for checkout.
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The first start attempt, approximately one month _later, failed
"I when the circuit breeker tripped on overcurrent.
. Testing-following a turbine _ driven auxiliary feedwater pump overspeeding event failed to consider that the governor ~ valve = linkages had-I recently - been manipulated _and, therefore, would not detect binding problems.
Operability determination-was based on -the recommendation. of.a: maintenance engineer following' local
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i exercising of the linkage and'one successful start.
Personnel errors were associated with the failure to_ perform H
post maintenance testing on some components prior. to returning
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them to service.
Examples of-these included a1 containment
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isolatior. ' valve,. a main' steam isolation valve and the Safe Shutdown Facility batteries.
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IS Due to procedural inadequacies and personnel errors a Reactor Vessel Level Instrumentation System post maintenance test failed to detect an improperly restored system.
In response to the above m nts the licensee has been developing a retest manual, however, progress with this effort' has been slow.
Additional guidance is also being developed for functional verification and retests.
The previous'SALP identified a high rate of personnel errors by Instrument 'and' Electrical (IAE) maintenance personnel.
An improving trend was evident during this assessment period as IAE personnel errors were notably reduced. Management attention in
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this ' area has been effective.
Isolated cases of errors did occur and one reactor trip was attributed to inadequate
troubleshooting activities.
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The Inservice Inspection' (ISI) program was observed to. be adequate. ISI personnel were knowledgeable, well trained, and
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qualified to perform inspections-within their respective areas of certification.
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Licensee activities for detecting and correcting a steam-
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gererator tube leak on Unit I were observed. The licensee has a strong program for monitoring SG tube leakage, tube inspection, and taking corrective action when leakage is. identified.
Plant management as well as corpocate personnel were actively involved in analyzing the leak, inspection of the SGs, and repair of the leaking tube, j
2, performance Rating
Category:
3.
Recommendations The areas of procedural adequacy and compliance, and post maintenance testing continue to result in operational problems and require coatinued management ettention. Actions t.sken in both areas appear to be effective and should continue to be pursued aggressively.
D.
Analysis
The emergency preparedness program.has received; strong management support to ensure that the licensee maintained the
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basic elements needt.d to promptly identify, correctly classify,
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adequately staff, and to implement the key elements of the
l Radiological Emergency Plan (REP) and, procedures-in-response to emergency events.
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- There were several program strengths:
Independent audits were detailed and comprehensive; a strong management commitment to i
the emergency response program; an effective tracking system was j
maintained for ensuring that prompt and adequate corrective action was taken on items identified during drills and exercises; a well staffed and trained onsite emergency i
organization; and two emergency response equipment improvements
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discussed below were noteworthy.
In 1988, sirens for the H
Catawba Alert and Notification System (ANS) had an availability i
of 96.27 percent based on the results of silent, growl, and full
cycle tests.
In 1989, the full cycle test indicated a
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reliability of 94.4 percent.
The overall availability during i
the SALP cycle has been good. Duke Power is in the process of l
upgrading the Catawba ANS to include a computerized control and
feedback system for periodic system interrogation and polling.
The licensee determined that the ENS phone equipment was powered from a non-dedicated, non-safeguards circuit which would have lost power in the event of a loss of offsite power. The need to-have a more reliable power source was immediately recognized and j
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a station modification was written. The modification design:has been completed but not yet implemented.
One contributor to the successful implementation of the -
J emergency program was the strength,of the training program which j
included drills conducted approximately monthly.
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The emergency exercise demonstrated that the licensee could effectively implement the Emergency Plan and procedures.
The-
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emergency exercise was considered fully successful with no exercise weaknesses identified.
Changes incorporated as Revision 11 to the Catawba Emergency
Plan were consistent with the requirements of 10 CFR 50.47(b)
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and Appendix E to 10 CFR 50.
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The licensee took action to improve the emergency response j
facilities.
The Operations Support Center (OSC) was moved to a
cable spreading room where it is maintained in an operational mode.
Several improvements were made in the Technical Support
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Center (TSC) including improvements in the work areas for key management positions.
The licensee's overall performance indicated that the emergency
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preparedness program is being implemented in a manner ~ adequate
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to protect the public health and safety.
2.
Performance Rating
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Category:
3.
Recommendations
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E.
Security and Safeguards
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1.
Analysis j
Security Force staffing levels and training appeared to be adequate, although the deficiencies identified with regard to access control and protection of safeguards information involved
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failures of plant and security personnel to comply with I
established procedures and licensee policy.
In all cases, the i
personnel involved had been trained in tha correct procedures regarding access control and the handling of safeguards information, but failed to perform in accordance with those procedures.
Several examples were also identifieo in which
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guards were permitted to perform duties for which they were' not trained or qualifieo. This appeared to result from inadequacies
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in monitoring training expiration dates by the security training j
staff and communicating that information to the security shift supervisor.
The licensee experienced multiple problems concerning the
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protected and vital area access control program during this
assessment period. These problems, which were identified by the i
licensee, included several instances in which security badges
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and keycards were incorrectly issued, vital area access
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improperly authorized, and security badges not removed from the i
badge issue area (although the individual's access authorization had been terminated).
In one case, a deceased individual remained on the validated access list for four months, although-the list was reviewed, updated, and approved by responsible plant supervisory personnel on a monthly-basis.
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l The apparent failure of personnel to comply with access control
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and safeguards information handling procedures, along with the i
failure to monitor and report the training and qualification
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status of assigned security personnel,' raised questions l
concerning the effectiveness of.the licensee's programs in these
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areas.
An enforcement conference was held in August 1989 to l
discuss this subject.
At that conference, the licensee described actions undertaken or planned to correct identified i
problems.
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Physical security, contingency, and guard training and
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qualification plan change submittals made ender the provision of
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10 CFR 50.54(p) were normally well-coordinated and technically
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sound.
However, two of the twelve plan changes submitted required additional clarification or other action by the
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licensee before they were acceptable.
An NRC Regulatory Effectiveness Review of the physical security program conducted in March 1989, identified certain weaknesses
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in the implemented program.
Corrections for the majority of these weakness have been undertaken, completed, or scheduled.
2.
performance Rating Category:
2 Declining 3.
Recommendations Management is encouraged to continue aggressive self-assessment
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activities in this area.
Increased management oversite is needed to assure adequate corrective actions are implemented.
F.
Engineering /Techincal Support 1.
Analysis Overall performance in this functional area has been adequate with certain exceptions demonstrated by examples of engineering errors in operability determination evaluations and drawing control deficiencies.
Design change development and implementation were generally acceptable. However, an inadequate post modification test for a Hydrogen Sktmmer System modification failed to detect inoperability of the safety related system resulting in escalated enforcement this assessment period. Licensee actions to improve post modification testing included revision of Design Engineering (DE) procedures to emphasize post modification testing and acceptance criteria and related training for DE engineers.
Additionally, the licensee has assigned responsibility for post modification test. development ' and verification to the system experts within the on-site performance Engineering organization.
Numerous drawing control deficiencies identified by the NRC were indicative of a programatic weakness in the engineering design change implementation program. Critical drawings in the Control Room, Technical Support Center, and Crisis Management Center did not consistently reflect modified plant configurations.
Similarly drawing update failures were identified for. partially completed modifications in which the system had been returned'to service and temporary modifications were not. consistently incorporated into drawings. In response to drawing control deficiencies, all critical drawings were reviewed and either replaced or updated with applicable interim attachments.
Drawing control procedures were revised to ensure incorporation of modifications prior to return to operation of the system, i
In November 1988, DE was, reorganized, providing dedicated engineering resources to each Duke Power nuclear station.
Additionally, a small onsite DE group was maintained to improve the interface between the site and DE.
Engineering resources i
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were dedicated to evaluate deficient _ perfomance in post
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modification testing weaknesses and recommend corective actions.
Engineering errors were made in operability deteminations. In January 1989, the Control Room Ventilation system was determined
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operable based on an incorrect assumed rather than actual system I
configuration.
DE incorrectly determined the Containment Air i
Return System to be operable based on the use of inappropriate i
electrical drawings. This deficiency was identified by the $1te i
engineering organization.
In June 1989, a review by the NRC identified that the calculations utilized to support an j
operability determination of the Hydrogen Skimmer System did not
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substantiate the conclusion that the system was operable.
In determining the ability of the Annulus Ventilation System to j
meet secondary containment negative pressure requirements, the
licensee failed to utilize the acceptance criteria stated in the i
Technical Specifications.
The licensee identified - that the.
periodic test procedure which was utilized to verify the Control j
Room Area Ventilation. System operability was deficient in that i
it tested a system configuration cifferent than the system i
design base description. The system failed to meet requirements when tested in the required design base configuration. Initial i
evaluation of auxiliary feedwater valve closure failures did not
identify the failure cause and resulted in failure to identify
this failure mechanism in other plant valves.
Although engineering's eventual dedication of resources to this issue was I
extensive, initial action was weak.
An initial operability-determination for the valves was not technically supported and-
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initial evaluations did not identify the generic significance of I
the event although current industry infomation was available related to similar failures i.e.
Rotork actuator torque calculation deficiencies.
In response to deficient performance in operability determinations, DE developed comprehensive guidelines for performing and-documenting operability determinations.for identified non-conformances.
The program required research of the design basis and assessment of the non-confomance to meet the basis.
Effective engineering support has been demonstrated en numerous issues.
For example, a
steam generator level control instrumentation modification was developed and installed which resolved problems associated with steam generator level control ~
instrumentation and resultant reactor trips. Corrective actions resulting from an engineering study of previous diesel generator failures contributed to improved reliability of the diesel generators. Other plant issues receiving effective engineering support include, steam generator PORV modifications, Raw Water Systems evaluation, Service Water Pump Cooling System modifications, and the 600 VAC Essential Shared Power system.
Ths engineering and technical support staff demonstrated a high __
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level of competence in the review and evaluation of the First Ten-Year Interval Inservice Inspection Program Plan.
Another example of excellent engineering support involved activities associated with the removal of a stuck reactor vessel head stud. The planning and execution of the job resulted in successful stud removal with minimal complications and dote particularly considering that the equipment and techniques had never been used in U.S. plants.
I Licensee response to NRC. initiatives was evident in -technical support improvements this assessment period. Improvements were demonstrated in-temporary modification program activity due to increased management - focus and program controls following NRC identified weaknesses early in -the assessment period.
Resolution of identified Environmental Qualification- (EQ)
deficiencies was timely and an engineering design study of EQ equipment reference documentation contributed to improved performance in EQ related maintenance and design activity.
2.
performance Rt 'ino Category: 2 3.
Recommendations None G.
Safety Assessment / Quality Verification 1.
Analysis Overall corporate management leadership, direction and support is good.
Effective management involvement in site activities associated with licensing areas was evident through prior planning, assignment of priorities, and decision ma king processes.
Management is well aware of generic and plant-specific safety issues and the schedule for their resolution.
The licensee understands the technical issues and considers carefully the impact of various NRC requests and positions on the plant.
Conservatism is generally exhibited in the licensee's approach to the resolution of technical ise.as'from a safety perspective, and the approaches-are generally sound and -
thorough, Nevertheless, a failure to report a significant event in a - timely manner was identified as delineated in an enforcernent action dated May 19, 1989 on the inoperability of Unit 2 Containment Air Return and Hydrogen Skimmer fan damper.
Recently, when the licensee has found that reportability is questionable, a courtesy LER was sent. - The staff finds this practice fosters good communications with the NRC. For example, the licensee decided to send an LER on Catawba steel-containment j
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corrosion.
The LERs submitted adequately described the major aspects of the events, including components or system failures that contributed to the event, and the significant corrective actions taken or planned to prevent recurrence.
The reports were thorough, detailed, well written and easy to understand.
The licensee makes effective use of meetings with the NRC when
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appropriate to resolve licensing issues, The licensee is generally well prepared and provides ample support for its positions during such meetings.
This was exhibited during meetings on quality verification and response spectra.
Furthermore, the licensee supports regular interface meetings with the NRC on licensing activities.
The licensee has reduced staffing levels, however, this reduction did not significantly affect the performance of station or licensing personnel who were reorganized under a regulatory compliance group.
The licensing staff has good knowledge of the plant, of technical issues, and 3 - good historical knowledge of plant systems and program integration.
The licensee has taken effective measures to minimize dependency upon outside contractors, and this has increased the ability to provide more timely responses.
Duke is an active participant, and frequently assumes a leading role, in nuclear industry activities regarding matters. of generic concern.
The licensee has improved in the area of self assessment and appears to have a broad based and usually aggressive program for self assessment. The licensee's self assessment is performed mainly by three entities:
the. corporate QA Department, the onsite QA Section and the Catawba Safety Review Group.(CSRG).
The corporate QA group conducted 14 audits of various
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activities.
These audits were more-results oriented than had been the case previously, with more emphasis on the review of activities rather than a review of documentation. The corporate QA Department also conducted a Self Initiated. Technical. Audit (SITA), of the 600 V-AC Essential Auxiliary Power System. These audits were thorough and valuable.
The on-site QA section performed routine surveillance activities throughout the period.
The activities indicated greater emphasis on field observations and included the areas of operations, station testing, station-security, maintenance and chemistry.
i The CSRG performed several In-plant-Reviews which were independent reviews of selected station activities.
These In-Plant Reviews included the areas of periodic surveillance, I
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plant equipment, operations; plant maintenance and facility status. The reviews were substantive and technically oriented.
A weakness was identified relative to the program for followup
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of NRC Information Notices. Untimely followup occurred relative to. a notice involving inadequate design and - testing of the Annulus Ventilation System.
2.
Performance Rating i
Category:
1 3.
Recommendations
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None l
V.
SUPPORTING DATA j
A.
Escalated Enforcement Actions 1.
Civil Penalties /No Civil Penalties I
Severity Level III violation, 50-413,414/88-38, $75000 Civil Penalty issued for failure to maintain both trains of the i
Containment Air Return and Hydrogen Skimmer System operable, j
Severity Level III violation, 50-413,414/88-38, no CP, issued i
for failure to submit a Licensee Event Report on the Hydrogen Skimmer issue within 30 days.
Severity Level III violation against Unit
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50-413,414/89-19, no CP, -issued for failure to maintain both
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channels of the Reactor Vessel Water Level Instrumentation operable,
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Orders Order imposing a $75000 civil penalty issued for Containment Air-1 Return and Hydrogen Skimmer System Severity leve3 III violation.
B.
Management Meetings August 10, 1988 -
Management Meeting held at Region II to discuss past performance.
October 21, 1988 - Management Meeting held at Catawba to present SALP report.
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December 9, 1988 - Management Meeting held-at Region II to discuss
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Duke Power Company's (DPC) reorganization.
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March 9, 1989 -
Enforcement Conference held in Region II to discuss Containment Air Return Damper.
May 25, 1989 -
Management Meeting held at Region II to discuss
DPC Nuclear Plant Design Basis Documentation Program.
June 21, 1989 -
DPC/NRC interface meeting held at McGuire site.
July 20, 1989 -
Enforcement Conference at Recinn II to discuss inoperability of Reactor Veivel Level Instrumentation System (RVLIS).
August 15, 1989 -
Continuation of RVLIS Enforcement Conference at Region II.
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August 29, 1989 -
Management Meeting at Region II to discuss
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licensed operator medical exam two year requirements.
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August 29, 1989 -
Enforcement Conference held in Region 11 to discuss potential safeguards violations.
October 12, 1989 - Enforcement Conference at Region II to discuss I
Auxiliary Feedwater System issues.
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C.
Review of Licensee Event Reports-(LER)
During the assessment period, a total of 56 LERs were analyzed (31 for Unit I and 25 for Unit 2) The distribution of these events by cause, as determined by the NRC staff, is as follows:
i Cause Unit 1 Unit 2 Total
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Component Failure
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Design
3
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Construction, Fabrication,
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or Installation Personnel Error
- Operating Activity
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- Maintenance Activity
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- Test / Calibration Activity
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- Other
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Other
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Total
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Note 1: With regard to the area of " Personnel," the NRC considers lack of procedures, inacequate procedures, and erroneous procedures to be classified as personnel error.
Note 2: The "Other" category is comprised of LERs where there was a spurious signal or a totally unknown cause, j
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Licensing Activities During the assessment period the staff completed approximately 80 i
licensing actions.
E.
Enforcement Activity
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No. of Deviations and Violations in Functional Each Severity Level (Unit 1/ Unit 2)
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Area Dev.
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Plant Operations 1/1 6/4 1/1
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Radiological Controls 1/1 4/4
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Maintenance / Surveillance 3/3 0/1 Emergency Preparedness j
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Security 3/3 Engineering / Technical 1/2 i
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Safety Assessment / Quality 3/4 1/1 Verification
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TOTAL 1/1 1/1 20/20 2/3
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F.
Rear +or Trips and Unplanned shutdowns 1.
Unit 1 Reactor Trips March 5, 1989, the unit tripped and a safety injection occurred on steam pressure rate when an operator inadvertently shut a main steam isolation valve instead of a steam ' generator PORV block valve.
Manual Reactor Trips
June 26, 1989, the unit was manually tripped due to a feedwater regulating valve failing closed.
August 24, 1989, the unit was manually tripped due to a feedwater regulating valve failing closed.
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Reactor Trips January 12, 1989, the unit tripped on steam generator low level following failure of a feedwater regulating valve due to a blown fuse.
Febreary 21, 1989, the unit tripped and a safety injection i
occurred when a personnel error caused a main steam isolation valve to shut.
l September 29, 1988, the unit was manually tripped following loss
of main generator cooling and erratic steam dump operation, i
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November 23, 1988, the unit was manually tripped due to a main
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feedwater isolation when a spurious high doghouse water level l
signal occurred.
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i January 21, 1989, the unit was manually tripped due to a loss of l
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stator cooling and turbine runback and personnel error causing :
loss of main feedwater pumps.
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Unscheduled Manual Shutdowns 1.
Unit 1 August 6,1988, the unit was shutdown due to a steam generator leak.
March 31, 1989, the unit was shutdown due to a technical specification required shutdown when a control room ventilation fan motor bearing-failed with the opposite train inoperable for r:Ntenance.
April 20, 1989, the unit was shutdown due to an instrument leak in the annulus.
June 16, 1989, the unit was shutdown to repair a reactor coolant pump motor oil leak.
Ocotber 24, 1989, the unit was shutdown to repair a main condenser tube leak.
2.
Unit 2 June 23, 1989, the unit was shutdown due to main feedwater flow orifice cavitation and check valve flutter.