ML20214X387
| ML20214X387 | |
| Person / Time | |
|---|---|
| Site: | Haddam Neck File:Connecticut Yankee Atomic Power Co icon.png |
| Issue date: | 05/14/1987 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML20214X369 | List: |
| References | |
| 50-213-86-99, NUDOCS 8706160451 | |
| Download: ML20214X387 (65) | |
See also: IR 05000213/1986099
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U.S. NUCLEAR REGULATORY COMMISSION
REGION I
SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE
INSPECTION REPORT 86-99
CONNECTICUT YANKEE ATOMIC POWER COMPANY
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HADDAM NECK NUCLEAR POWER PLANT
ASSESSMENT PERIOD: MARCH 1, 1986 - MARCH 31, 1987
BOARD MEETING DATE: MAY 14, 1987
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TABLE OF CONTENTS
PAGE
I. INTRODUCTION......................................................... 1
II. CRITERIA............................................................. 3
III. SUMMARY OF RESULTS................................................... 4
Overall Facility
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A. Evaluation..................................... 4
B. Background...................................................... 5
C.. _ Facility Performance Analysis Summary........................... 6
D. Plant Trips and Unplanned Shutdowns................... ......... 7
IV. PERFORMANCE ANALYSIS................................................. 10
. A. Plant 0perations................................................ 10
B. Maintenance and Modifications................................... 14
C. Surveillance.................................................... 17
D. Fire Protection................................................. 20
E. Engineering Support............................................. 23
F. Licensing Activities............................................ 26
G. Refueling and Outage Management................................. 29
'H. Radiological Controls........................................... 31
I. Emergency Preparedness.......................................... 35
J. Security and Safeguards......................................... '38
K. Training and Qualification Effectiveness........................ 41
L. A s s u ra n ce o f Q ua l i ty. . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . . 43
V. SUPPORTING DATA AND SUMMARIES........................................ 46
A. ' Investigation and Allegation Review............................. 46
B. Escalated Enforcement Action.................................... 46
C. Management Conferences.......................................... 46
D. Review of LERs.................................................. 46
E. Summary'of Licensing Activities................................. 47
F. Description of Plant Activities................................. 49
TABLES
Table 1 - Inspection Activities
Table 2 - Inspection Hours Summary
Table 3 - Enforcement Activity
Table 4 - Licensee Event Reports
Table 5 - SALP History
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I. INTRODUCTION
A. Purpose and Overview
The Systematic Assessment of Licensee Performance (SALP) is an integrated
NRC staff effort to collect information periodically and evaluate licen-
see performance. SALP supplements the normal regulatory processes that
ensure compliance with NRC regulations. It is intended to be sufficiently
diagnostic for allocating NRC resources for providing meaningful guidance
to licensee management to promote the quality and safety of plant opera-
tion.
An NRC SALP Board met on May 14, 1987 to assess licensee performance in
accordance with NRC Manual Chapter 0516, " Systematic Assessment of Lic-
ensee Performance." A summary of the SALP guidance and evaluation cri-
teria is provided in Section II of this report.
This report assesses performance at the Haddam Neck Plant during the
13-month period from March 1, 1986 through March 31, 1987.
The SALP Board was composed of the following:
Chairman:
W. F. Kane, Director, Division of Reactor Projects (DRP) (part-time)
Members:
T. T. Martin, Director, Division of Radiation Safety and Safeguards,
(DRSS)
W. V. Johnston, Director, Division of Reactor Safety (DRS)
5. J. Collins, Deputy Director, DRP (part-time)
E. C. Wenzinger, Chief, Projects Branch 3, DRP (part-time)
E. C. McCabe, Chief, Reactor Projects Section 3B, DRP
P. 9 Swetland, Senior Pesident Inspector, Haddam Neck
C. O. Thomas, Director, Integrated Safety Assessment Project Directorate,
Office of Nuclear Reactor Regulation (NRR)
F. M. Akstulewicz, Licensing Project Manager, NRR
Other Attendees (Non Voting)
A. A. Asars, Resident Inspector, Haddam Neck
R. R. Bellamy, Chief, Emergancy Preparedness & Radiological Protection
Branch, DRSS (part-time)
D. L. Caphton, Senior Technical Reviewer, DRS
M. C. Kray, Reactor Engineer, DRP
W. J. Lazarus, Chief, Emergency Preparedness Section, DRSS (part-time)
W. J. Pasciak, Chief, Effluents Radiation Protection Section, DRSS
(part-time)
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M. M. Shanbaky, Chief,. Facilities Radiation Protection Section, DRSS.
3(part-time)
S..F. Shankman, Project Manager, Operator Licensing (part-time)
5. S.. Sherbini, Senior Radiation Specialist ~, DRSS (part-time)
A. B. Sidpara, Reactor Engineer, DRP (part-time)
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II. CRITERIA
Licensee performance is assessed in selected functional areas. These areas
are significant to nuclear safety and the environment, and are normal pro-
grammatic areas. The following criteria were used as appropriate to assess
each area.
1. Management involvement and control in assuring quality.
2. Approach to resolution of' technical issues from a safety standpoint.
3. Responsiveness to NRC initiatives.
4. Enforcement history.
5. Reporting and analysis of reportable events.
6. Staffing (including management).
7. Training effectiveness and qualification.
Based upon the SALP Board assessment, each functional area is classified into
one of three performance categories. These are:
Category 1. Reduced NRC attention may be appropriate. Licensee management
attention and involvement are aggressive and oriented toward nuclear safety;
licensee resources are ample and effectively used so that a high level of
performance with respect to operational safety is being achieved.
Category 2. NRC attention should be maintained at normal levels. Licensee
management attention and involvement are evident and concerned with nuclear
safety; licensee resources are adequate and reasonable effective such that
satisfactory performance with respect to operational safety is being achieved.
Category 3. Both NRC and licensee attention should be increased. Licensee
management attention or involvement is acceptable and considers nuclear safety,
but weaknesses are evident; licensee resources appear strained or not effec-
tively used such that minimally satisfactory performance with respect to
operational safety is being achieved.
The SALP Board has also categorized the performance trend over the course of
the SALP assessment period. The SALP trend categories are:
Improving: Licensee performance was determined to be improving near the close
of the assessment period.
Declining: Licensee performance was determined to be declining near the close
of the assessment period.
A trend is assigned only when a definite trend of performance is discernible
and the SALP Board believes that continuation of the trend may result in a
change of performance level.
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III. SUMMARY OF RESULTS
A. Overall Facility Evaluation
This SALP confirmed continued safe operation although overall performance
continued to decrease. In 1982, all nine functional areas were rated
as Category 1. In 1986, there were two Category 1 and eight Category
2 areas. This SALP found one Category 1 area (Security), ten Category
2 areas, and one Category 3 area (Fire Protection). Actions to improve
performance have been initiated by the new station superintendent (as-
signed in December 1986), but the effectiveness of these actions has not
yet been demonstrated.
Notable improvements were evident in operator requalification, onsite
safety review committee performance, and control room utilization. The
licensee continued to exhibit a strong safety perspective and aggres-
sively resolved those matters of immediate safety significance. In ad-
dition, risk reduction was actively pursued. When vulnerabilities were
identified, interim corrective actions were taken, and modifications were
proposed and implemented. Performance in the security area has remained
excellent. In this area, there was solid individual and supervisory
performance, aggressive identification and resolution of problem areas,
and a good interface with the corporate staff.
There were seven plant trips during this 13-month SALP period. Three
of these were attributed to personnel error. None of the five trips
during the preceding 12-month SALP period were due to personnel error.
These trips illustrate the need for improvement of both personnel and
equipment performance. Also, ALARA continues to be an area where im-
provement is needed. Radiation exposures remain exceptionally high and
there was an overexposure that resulted in a civil penalty. These mat-
ters and procedure inadequacy and adherence problems marred a generally
sound radiation protection program.
Site and corporate interface deficiencies appear to be a major contribu-
tor to performance weaknesses in a number of areas including fire pro-
tection, engineering support, radiological controls, licensing, and out-
age management. In addition, other long-standing problems continue to
affect performance. These include containment leak rate testing, the
large open items backlog, numerous facility modifications, converting
to standard technical specifications, and upgrading Safety Analysis
documentation of design data.
In summary, while safe operation continues, there are numerous perform-
ance weaknesses and an overall downward trend. Recent initiatives are
indicative of potential improvements, but tangible results to support
a conclusion of improving performance were not available during this SALP
period.
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B. Background
Licensee Activities
On March 1, 1986, at the start of the SALP period, the licensee was com-
pleting a refueling and maintenance outage. The plant was restarted on
May 6, 1986, after resolving an emergency core cooling deficiency which
prevented long term core cooling for a small break in the charging system
piping. Also, the licensee identified an unplugged, degraded steam
generator U-tube and was granted temporary, one-cycle relief for opera-
tion with the tube not plugged. Full power was achieved on May 22.
During the SALP period, seven plant trips occurred because of personnel
errors, equipment, and other maintenance related failures. Numerous
turbine load runback actuations resulted from nuclear instrumentation
system (NIS) fluctuations caused by aging components within the NIS.
The plant shut down for 20 days in July 1986 to plug steam generator U-
tubes in which defects were identified in the tubesheet area. The ap-
parent cause of the defects was the tube rolling process during fabrica-
tion and subsequent primary side stress corrosion cracking. During this
outage, a steam generator worker exceeded his quarterly radiation expo-
sure limit.
In December 1986, the licensee identified another emergency core cooling
system design discrepancy affecting long term core cooling in the sump
recirculation mode. Interim corrective actions involved a plant shutdown
on December 19 to perform a special flow test to verify the acceptability
of the proposed throttling of injection flow. The plant returned to
power on December 25, 1986. Overall, the plant achieved a capacity fac-
tor of 63% during this SALP period. The plant's lifetime capacity factor
is 76%.
A more detailed description of plant activities is provided in Section
V.F of this report.
Inspection Activities
Two NRC resident inspectors were assigned to the site during the assess-
ment period. The NRC inspections are summarized in Table 1 and represent
an inspection effort of 3590 hours0.0416 days <br />0.997 hours <br />0.00594 weeks <br />0.00137 months <br /> (3314 hours0.0384 days <br />0.921 hours <br />0.00548 weeks <br />0.00126 months <br /> per year), distributed
as shown in Table 2.
Special team inspections were made of licensee ALARA practices (April
7-11,1986); the annual site emergency exercise (April 25, 1986); licen-
see implementation of Appendix R fire protection requirements (June 16-20,
1986); and overall facility operation (November 14-21,1986).
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Twenty-one enforcement actions, including a Severity Level IIILviolation
and Civil Penalty for the radiation overexposure, were issued. Enforce-
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ment actions are tabulated in Table 3.
C. Facility Performance Tabulation
CATEGORY CATEGORY
LAST THIS
FUNCTIONAL AREA PERIOD * PERIOD ** TREND
1. Plant Operations 1 2 --
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2. ' Maintenance & Modifications 2 2 --
3. Surveillance 2 2 --
4. Fire Protection # 3 --
5. Engineering Support #- 2 --
6. ; Licensing Activities 2 2 --
7. Refueling / Outage Hanagacnt 2 2 --
' 8. Radiological Controls' 2 2 --
9. Emergency Preparedness 2 _2
--
10. Security & Safeguards 1 1 --
11. '. Training and Qualification 2 2 --
Effectiveness-
12. Assurance of Quality 2 2 --
- March 1, 1985 to February 28, 1986
- March 1, 1986 to March 31, 1987
- Not addressed as a separate area.
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D. Unplanned Trips and Shutdowns
-POWER ROOT FUNCTIONAL
DATE LEVEL DESCRIPTION CAUSE AREA
15/8/86 110% Automatic trip due to load Personnel error - Operations <
increase past 10% power (P7) operator exceeded
during main turbine testing P7 with an MSIV
with the #2 main steam iso- closed.
lation valve (MSIV) closed.
6/4/86 100% Anticipatory manual trip due Random equipment --
to prompt (and proper) opera- -failure - LCV
tor response to feedwater operator-to-stem
system fluctuations. Failure failure.
of a heater drain tank level
control valve (LCV) caused
rapidly decreasing steam
generator levels.
' 6/17/86 100% Anticipatory manual trip Inadequate failure Maintenance
similar to the trip on analysis.
6/4/86. The same LCV oper-
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ator separated again. The
cause of the previous LCV
failure was not adequately
identified therefore approp-
riate actions to prevent re-
currence were not implemented.
6/19/86 0% Automatic trip while trouble- Personnel error- Maintenance
shooting a blown fuse in a inadequate control
nuclear instrument power of maintenance.
supply. The fuse was re-
placed without determining
the initial cause of failure
or assessing the potential
effects of reenergizing a
faulted component. The shorted
power supply overloaded the
vital instrument bus and un-
blocked the existing closed
HSIV reactor trip signal.
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POWER ROOT FUNCTIONAL
DATE LEVEL DESCRIPTION CAUSE AREA
6/22/86 10% Automatic trip during 3-loop Inadequate control Maintenance
operation. Ceiling debris of cleanliness
lodged in a reactor protec- during maintenance.
tion system (RPS) relay after
the control room ceiling was
replaced. This created an un-
annunciated low flow half-trip
signal. A second half-signal
was actuated by the 3-loop
operating condition. The plant
tripped when the low flow trip
signal was automatically un-
blocked as turbine load ex-
ceeded 10% power.
7/11/86 100% Shutdown to mode 4 to re- Recurrent plant Maintenance
place a sheared charging shaft failure due
pump shaft. The licensee was to untimely failure
unable to affect repairs analysis and correc-
within the 72 hour8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> grace tive action in re-
period and placed the unit sponse to an earlier
in hot shutdown. shaft failure.
The plant remained shutdown
until August 5, 1986 for steam
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generator repairs described
below.
7/15/86 0% Three week outage to plug 119 Construction --
steam generator tubes identi- defects and
tified as having flaws in the primary side stress
rolled region of the tubes. corrosion cracking.
8/30/86 100% Anticipatory manual trip due Random equipment --
to prompt (and proper) opera- failure - defects
tor response to a failed open in the FRV stem-to-
feedwater regulating valve plug weld.
(FRV).
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POWER ROOT FUNCTIONAL
DATE LEVEL DESCRIPTION CAUSE AREA
11/30/86 100% Automatic trip due to low Personnel error- Maintenance
steam generator level caused inadequate control
by inadvertent closure of of maintenance.
the #3 FRV. Technicians and
operations supervision did
not assure that the FRV
control system was deenergiz-
ed prior to work. FRV control
power supply was shorted out
causing the valve to fail
closed.
The plant remained shutdown for
4 days to repack a reactor
coolant loop isolation valve
which leaked following this
trip.
12/6/86 25% Shutdown to remove vibrating Design error. Engineering
moisture separator reheater Support
baffle plates which were in-
stalled during turbine over-
hauls. Thermal expansion
stresses broke away the
baffle plate mounts.
12/19/86 16% Shutdown to perform a special Design / analysis Engineering
flow test to verify the ade- error. Support
quacy of emergency core cool-
ing systems (ECCS) due to a
design deficiency affecting
the recirculation mode after
a medium sized LOCA.
NOTE: The root causes in this table are the SALP Board assessments based on in-
spector evaluations and may differ from the LERs.
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IV. PERFORMANCE ANALYSIS
A. Plant Operations (1270 hours0.0147 days <br />0.353 hours <br />0.0021 weeks <br />4.83235e-4 months <br />, 35%)
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1. Analysis
The last SALP rated this area as Category 1, consistent. No major
concerns were identified. Events involving personnel errors, the
quality of submittals to the onsite safety review committee (PORC),
and the effectiveness of root cause addressal were highlighted for
licensee evaluation. Corrective action effectiveness was shown
during the current SALP period by a reduction of items such as fire
barrier problems and late procedure reviews. As discussed in the
following text, personnel errors and quality of submittals to PORC
are still weaknesses.
This assessment is based on routine resident inspections throughout
the SALP period and a Region I Diagnostic Team Inspection conducted
in November 1986.
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A capacity factor of 63% (76% lifetime average) resulted from an
extended refueling outage, two mid-cycle shutdowns to address steam
generator and emergency core cooling system deficiencies, and seven
plant trips. One of these trips occurred due to operator error
during plant startup. The operators did not maintain power below
10% to assure that a bypassed main steam isolation trip signal was
not automatically reinstated. Two other trips occurred during
troubleshooting of instrumentation problems. Better coordination
between the operators and technicians could have prevented these
two trips.
Operating shift activities were conducted professionally. Shift
turnovers were thorough. Log keeping was adequate. Activities were
generally well understood, and were conducted with care and formality.
Operators responded effectively to equipment failures and plant
trips. Operator alertness was evident during dayshift and backshift
inspections. Die, tractions such as extraneous reading material were
not permitted in the control room. The licensee rearranged the
control room to remove unnecessary administrative materials and
provide more space away from the operator-controlled area for opera-
tor aides and for equipment tagging activities. Also, the licensee
initiated modifications to expand the control room to provide seg-
regated space for the shift supervisor.
The licensee has effectively minimized operator distractions due
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to illuminated main control board alarms. During routine operations
there are normally only one or two lighted alarms, and projects have
been initiated to correct these indications.
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Operator technical knowledge was good. NRC license examination re-
sults (14 of 15 passed) showed that the candidates were well trained
for initial license examination. Operator interviews by inspectors
and NRC participation in the operator requalification program demon-
strated that the operators were knowledgeable.
The NRC has been concerned about weaknesses in the operator requali-
fication program since 1984. Progress toward a new continuous re-
qualification program had been slow. During this SALP period, NRC
reviews of the requalification program found satisfactory training
effectiveness. However, some program administration details had
not been formalized. For instance, several draft lesson plans were
in use, and there was no method for documentation of removal of
operators from shift due to poor performance in requalification.
Also, the licensee's formal requalification program submittal did
not specify the attributes of an annual comprehensive written ex-
amination, and no formal program evaluation / audit was specified.
These factors indicate that more management attention to program
details is appropriate.
Management involvement in operations was evident in routine super-
intendent tours of the control room and plant spaces, in regular
plant material inspections by superintendents and department super-
visors, and in active and assertive participation in daily manage-
ment meetings. However, management was not effective in resolving
a problem involving several instances of manual containment isola-
tion valves being opened in violation of Technical Specification
(TS) requirements. While the safety significance of these viola-
tions was small, the failure to recognize the need for strict com-
pliance with TS requirements and propose prompt and comprehensive
corrective actions resulted in repeat violations.
The Plant Operations Review Committee (PORC) and the Nuclear Review
Board functioned well. Inspector observations characterized these
meetings as frank technical discussions with full management support
for the expression of dissenting views. PORC met more than once
a week to complete the large volume of procedure, operational event,
modification, and Technical Specification reviews. The licensee
made some progress in reducing the PORC workload by screening out
minor event reviews which are adequately handled by line management,
and by use of subcommittee pre reviews for large packages. Notwith-
standing, the quality of staff work prior to PORC review has re-
mained inconsistent. While the general quality of modification
packages has improved, several packages were returned for documen-
tation of the bases for acceptability. Also, the continuing large
number of procedure changes submitted to PORC indicated that the
previous review process has not been properly effective. The good
Nuclear Review Board (NRB) involvement was particularly evident in
the two safety analysis / emergency core cooling system design errors
identified during this period. Detailed NRB review and their in-
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quisitive approach to these problems contributed effectively to
satisfactory interim resolution. The NRB also noted inadequate
correction of the root cause in selected Licensee Event Reports and
initiated corrective actions.
NRC review of Licensee Event Report (LER) quality based on ten
selected LERs resulted in the same overall acceptable rating as in
the previous period. Improvement was noted in safety consequence
discussions and failure mode determinations. Weaknesses were noted
in identifying the dates and times of major occurrences, in failed
component information, and in cause and corrective action summaries
in the abstract section. During this SALP period, LERs improved and
were more consistent due to the licensee's formalization and track-
ing of LER development and commitment implementation. This action
resulted from NRC and licensee identified weaknesses in previous
LERs, and from licensee identification that an LER commitment to
prohibit 3-loop operation in the startup mode had not been imple-
mented. This resulted in plant operation outside this temporary
limitation. No similar problems occurred under the new tracking
system. The high proportion of personnel error-related events re-
mained about the same during this SALP period as it was before.
While LER improvement initiatives were evident, these error-related
events and other instances where LER corrective action was not ade-
quate indicate that further improvement in root cause analysis and
corrective action is warranted.
Plant housekeeping generally improved during this period. The lic-
ensee continued efforts to expand facilities for radioactive waste
storage and employee habitability. Early in the period, there was
an overall appearance of cleanliness but a lot of clutter due to
the uncontrolled storage of materials near equipment and atop cabi-
nets. An improvement was made in this aspect, but a formal program
for control of equipment storage was not implementea. Improved
cleanliness was also noted in the auxiliary feed pump area. In con-
trast, efforts to recover areas which had become dirty or contamin-
ated during the refueling outage did not extend into historically
unkempt areas such as the RHR pit and pipe trenches. Overall,
housekeeping was satisfactory.
In summary, improvements were noted in operator requalification,
PORC performance, and control room utilization. Steps were taken
to address correction of recurrent weaknesses. Further improvement
in the quality of input to the PORC and the effectiveness of root
cause analysis and corrective action for error-related events is
warranted. For the most part, plant down time was to correct iden-
tified equipment problems and reflected proper response to the
associated conditions. However, operator error caused one trip and
operator performance contributed to two other trips. Specific
licensee action to reduce the frequency of unnecessary trips is
needed. Overall, this SALP found satisfactory operating performance.
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. 2. Conclusion
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Category 2.
3. Board Recommendation
Licensee: Correct the causes of operator contribution to unnecessary
trips.
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NRC: None,
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B. Maintenance and Modifications (425 hours0.00492 days <br />0.118 hours <br />7.027116e-4 weeks <br />1.617125e-4 months <br />, 12%)
1. Analysis
The previous SALP rated Maintenance as Category 2, consistent. Con-
cerns included problems in the implementation of plant modifications,
poor documentation of maintenance, and the maintenance backlog.
Improvements were noted in these areas during this period, but those
problems are still a concern.
The licensee has a strong preventive maintenance program. Activi-
ties are tracked and scheduled through the computerized Production
Maintenance Management System (PMMS), and preventive maintenance
items on both safety-related and balance of plant components are
routinely completed. In one case, the licensee was not sufficiently
sensitive to the time consumed by maintenance. Inadequate pre-
planning prior to removal from service contributed to the diesel
fire pump being out of service on three consecutive days for work
that is normally completed in one day. Overall, however, the ef-
fectiveness of preventive maintenance was reflected in good equip-
ment reliability. No plant outages or safety-related equipment
inoperabilities were attributed to preventive maintenance weaknesses
during this SALP period.
Corrective maintenance was generally well-controlled. Identified
equipment discrepancies were entered into the PMMS system and auto-
matically converted to work control documents. In October 1986,
, the licensee initiated a new program to develop automated PMMS
tracking of equipment spare parts listings, bills of materials, and
recommended parts stock levels. Daily planning meetings at manage-
ment and supervisor levels contributed positively to the coordina-
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tion of repair activities. The licensee reduced the backlog of
, corrective maintenance actions by about 17 percent, but over 300
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outstanding work items remain open. Many of these await spare parts,
engineering review, or a plant shutdown.
The maintenance staff has effectively managed action items, assuring
that safety significant work is completed. During outages, an
inter plant maintenance force (IMF) is available to augment the
licensee's staff. However, staffing levels appeared strained by
increased commitments to support new technical training programs
implenented late in this SALP period and to provide maintenance
assistance at other plants. The licensee should take care to assure
the reduction in backlog continues despite training and IMF commit-
ments.
NRC observations of maintenance and discussions with workers and
supervisors identified competent and knowledgeable personnel. Job
supervisors were routinely present in the field and were observed
to provide valuable assessment of technical problems. However, job
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oversight was noted to be lacking in radiation exposure control
(see Functional Area H of this report) and there was inadequate
documentation of maintenance. For example, a violation was cited
for inadequate documentation of work scope and material control for
replacement of electrical leads for an environmentally qualified
motor-operated valve. Licensee quality assurance audits have iden-
tified similar problems in attention to detail and administrative
procedure compliance. Licensee management has emphasized procedural
compliance and careful documentation in departmental meetings and
written guidance to the staff. So far, observations have not found
the results to be effective overall.
The licensee has implemented expanded technical training programs
for maintenance personnel. These programs also include maintenance
team training for maintenance, health physics, quality control and
supervisory personnel. The teams repaired full-sized mock-ups such
as reactor coolant pump seals to improve the repair procedures,
worker effectiveness, and job exposure controls. Inspector discus-
sion of these training activities with workers revealed highly
positive evaluations of the quality of this program. The programs
are relatively new, however, and were not effective in preventing
the maintenance-related errors discussed below. In addition, in-
creased training commitments contributed to the backlog of out-
standing work items.
Two plant trips during this SALP period occurred during instrumen-
tation troubleshooting. In one case, the technician failed to de-
energize the circuit on which he was soldering. The plant tripped
when the_ soldering iron shorted the circuit and closed a main feed-
water regulating valve. In another case, a failed nuclear instru-
ment drawer was re-energized without assessing the potential con-
sequences, and the resultant surge on the vital bus unblocked an
existing plant trip signal. Four other plant trips involved equip-
ment problems or failure to correct the root cause of a previous
component failure. Two of these were assessed as due to random
equipment failure. Another occurred when the heater drain tank
level control valve failed closed, causing a loss of feedwater and
low steam generator levels. The valve operator-to stem coupling
had disengaged. This problem had occurred several days before (one
of the random failures), but repair was not effective in preventing
the second trip. After further repair (addition of set screws to
hold the coupling in place) the problem did not recur. Also, a trip
occurred because debris from the control room ceiling, which was
replaced during the 1986 outage, became lodged in a reactor protec-
tion system low flow trip relay. Maintenance controls during the
ceiling replacement and subsequent Halon discharge tests did not
protect the reactor protection system from falling debris. The
debris held a relay contact open and created an unannunciated half-
trip. When the plant was placed in 3-loop operation, a second half-
trip was generated and the plant tripped. These events point out
a need for better maintenance control.
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Fifteen equipment failure related events were reported during this
SALP period. Eight of these involved design and aging problems in
the low temperature overpressure protection and nuclear instrumen-
tation systems, respectively. The licensee has projects ongoing
to correct these problems by 1989.
A recurrent charging pump shaft failure also occurred'during this
period. NRC review of this repair and previous failure data con-
cluded that the failure mechanism was not apparent during the first
repair and that vendor assistance did not promptly identify effec-
tive corrective actions. Vendor recommended corrective actions to
lower axial clearances were subsequently incorporated.into charging
pump maintenance procedures, but no action was taken to verify con-
formance of the initial repair to these new pump clearances. After
the second failure, measures recommended by the vendor to reduce
cyclic stresses were implemented. These measures have been effec-
tive to date.
In summary, the maintenance and maintenance training programs are
good. However, personnel error. , other instances of ineffective
maintenance, procedural and docimentation inadequacies, and the
maintenance backlog were assess rd to be factors which prevented
achievement of overall maintenaice excellence.
2. Conclusion
Category 2
3. Board Recommendations
Licensee: Consider measures to improve equipment and maintenance
personnel performance in order to prevent unnecessary
trip:.
NRC: None.
_ _ _ _ _ _ _ _ _ _ _ _
.
.
17
C. Surveillance (460 hours0.00532 days <br />0.128 hours <br />7.60582e-4 weeks <br />1.7503e-4 months <br />, 13%)
1. Analysis
Surveillance was rated Category 2 during the previous SALP. Con-
cerns included surveillance procedure adequacy, missed surveillances,
QA adequacy, and correction of long standing containment leak rate
testing issues. During the current period, licensee progress in
these areas was noted. However, problems identified indicated a
need for further improvement.
This assessment is based on six region-based NRC inspections and
frequent resident inspector observations of surveillance.
An effective and generally well controlled surveillance program is
in place using computerized scheduling and tracking programs. The
program is divided into surveillances (SURs) for Technical Specifi-
cation (TS) required testing and preventive maintenance (PMPs) for
other equipment. Both the SUR tracking system and Production Main-
tenance Management System (PMMS) effectively scheduled surveillance
requirements on a weekly basis. No missed surveillances were iden-
tified. There was, however, one NRC-identified instance where new
TS requirements for radiological chemistry surveillances were not
properly incorporated in the SUR tracking system. Also, these tests
were not completed. The licensee is evaluating changes necessary
to improve commitment documentation and tracking.
! The licensee has completed a 2 year effort to upgrade surveillance.
During this SALP period, the primary thrust of this program has been
l the upgrading of procedures. NRC inspectors noted improved quality
in many surveillance procedures, particularly those for inservice
'
testing of pumps and valves. Acceptance criteria were included in
those procedures and operators performing the tests were thereby
alerted to test failures. The NRC also noted inconsistencies in
procedure format within departments. These indicate that the sur-
veillance upgrade program was not fully successful. The licensee
recently initiated a program to standardize the format and improve
procedure details for technicians in the field.
No plant trips were caused by surveillance testing errors. There
were two minor NRC violations for using an alternate data measure-
ment technique and an uncalibrated test gauge. The licensee promptly
corrected these problems. Other surveillance program deficiencies
including delayed test result reviews, informal documentation, in-
adequate independent verification of a jumper / bypass, and an un-
approved procedure change to correct test connection locations for
a valve interlock test. The number and variety of these problems
indicate that formality of implementation and attention to procedure
details, particularly at supervisory levels, needs improvement.
_ _ _ _ - _ - _ _ _ _ _ _ - _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
-
i
,
.
18
During.the 1986 refueling outage, the licensee performed eddy cur-
rent testing of almost all of the steam generator U-tubes. The
eddy current test program was improved through the use of better
equipment and technology. This was evident in the licensee's iden-
tification of U-tube flaws in the rolled area of the tubes. These
defects were related to construction process deficiencies and pri-
mary side stress corrosion cracking at the stressed locations. Upon
identification of the problem after the outage, the licensee shut
down the plant, proposed acceptable defect acceptance criteria for
NRC approval, and plugged 119 tubes which exceeded the criteria.
There was one problem related to eddy current testing (ECT) during
this period. Poor communication and verification of ECT dc)ta re-
sulted in one tube with a 55% through wall defect not being plugged
during the refueling outage. The licensee identified this problem
during a subsequent review of the steam generator inspection report.
After justification was developed, the NRC approved the licensee's
request for leaving that tube unplugged during the current operating
cycle.
The post-outage startup physics testing program included precritical
tests, zero power physics tests, and power ascension tests. NRC
inspectors observed effective management involvement. Reactor en-
gineering staffing was ample to assure safety. Reactor engineers
were found knowledgeable in their assigned areas. New trending /
testing computer programs were developed. For instance, monitoring
rod drop measurements by computer permitted acquisition of data from
all rods in a bank at one time versus a single rod drop using the
old method. The reactor engineers were found to interface effec-
tively to accomplish their tasks. For example, they provided the
operating staff with proper updated graphs and figures and obtained
input from the chemistry group for fuel integrity evaluations.
Containment leak rate testing (CLRT) was assessed based on NRC ob-
servation of both local and integrated leak rate tests and problems
in resolving outstanding CLRT licensing issues. An integrated CLRT
was performed during this SALP period. The test procedure and
methodology were much improved since the last test, yet several
procedure weaknesses su@ as nonspecification of the start time and
stabilization period for 6he verification test were noted. Manage-
ment involvement was considered to be deficient, in that the test
director was generally uni #0rmed about the progress of integrated
CLRT preparation and was not '<nowledgeable of the location of the
verification test flow meter. Iirensee personnel conuucting the
test were found to be knowledgecble and competent. Staffing levels
were adequate during the CLRT.
As found containment leakage exceeded allowed limits for the second
consecutive integrated test and the third consecutive set of local
tests. Both failures were due to excessive penetration leakage.
The licensee plans to modify several penetrations during the 1987
,
.
.
19
refueling outage, including the most recurrent leaking penetrations.
NRC noted improved performance in the conduct and frequency of local
CLRTs. Procedures were more detailed, particularly regarding test
acceptance criteria and prompt notification of test results. This
> had been a weakness in previous CLRTs. The licensee also modified
! the component cooling water system by adding a slipstream filter
to reduce the impurities which had previously clogged the contain-
ment isolation valves in this system. Subsequent tests indicated
improved leak rate performance, but the success of this change will
depend on the as-found leakage following the current operating cycle.
Several outstanding CLRT Ifcensing issues remain open. Overall,
CLRT failures indicate that additional licensee attention to main-
tenance, modifications, and/or more frequent testing is needed to
i assure acceptable containment leakage throughout operating periods.
Quality assurance (QA) involvement in the surveillance program was
evident during this period. Expanded QA observations of field ac-
tivities included coverage of CLRT and startup physics testing.
Licensee findings related to procedure inadequacy / noncompliance
- supported the NRC conclusion that further management attention to
these areas is necessary. The NRC also noted that quality assurance
group reviews of TS surveillance documents failed to identify tardy
supervisory reviews and procedural deviations. That indicates
weakness in the quality of these QA reviews.
In summary, the surveillance program was technically sound and pro-
perly scheduled. Weaknesses regarding procedure adequacy, minor
noncompliances, and attention to detail in documentation have been
identified by the licensee and NRC. These indicate that previous
upgrade programs were not fully effective. As-found leak rate
testing results indicate that acceptable containment leakage has
not been assured throughout operating cycles. Corrective action
programs related to these problems are ongoing, but their effec-
l tiveness has not yet been evident. Quality assurance involvement
was improving, although the quality of document reviews was weak.
!
2. Conclusion
Category 2
3. Board Recommendations
Licensee: Assure that acceptable containment leakage is maintained
throughout operating cycles.
NRC: None
!
l
!
l
-
-.
.
20
D. Fire Protection (276 hours0.00319 days <br />0.0767 hours <br />4.563492e-4 weeks <br />1.05018e-4 months <br />, 8%)
1. Analysis
Fire protection was part of the Plant Operations area in the pre-
vious SALP. The frequency of reported fire protection system de-
ficiencies was highlighted for licensee attention. This concern
was addressed by the licensee through personnel indoctrinatiol and
training, and was not a concern during the current SALP period.
This area includes fire prevention, fire detection, the ability to
respond to fires with onsite and offsite forces, engineering and
hardware features to limit fire spread, and the ability to achieve
safe shutdown in the event of fire. An NRC team inspected the im-
plementation of NRC fire protection requirements (10 CFR 50, Appen-
dix R) during this SALP. Licensea performance in this area was
routinely observed by the resident inspectors.
The onsite fire brigade was well equipped and trained, with regular
' drills and classroom training for each brigade member. During this
SALP period, the fire brigade responded effectively to several small
fires within the owner-controlled area. Fire fighting equipment
was generally maintained in good working order through inclusion
inplantlsurveillanceandpreventivemaintenanceprograms. In one
.
case, however, the NRC identified omission of monthly checks on a
few portable fire extinguishers. Also, the licensee identified an
instance where a fire suppression system was rendered inoperable
when the actuation system failed due to exposure to the outside
environment. The actuation gircuit was rebuilt but the environ-
mental cause was not addressed. Offsite committee review of the
Licensee Event Report on this failure initiated further action to
protect the system from the outside environment.
NRC review identified weaknesses in site awareness of fire protec-
tion ~prcgram commitments and in program management of the Appendix
R upgrade ef fort. For instance, the safe shutdown analysis gene-
rated by the licensee lacked thoroughness and accuracy. Systems
necdssary for safe shutdown (such as component cooling) were not
includtd in the Fire Protection Evaluation Report. Critical calcu-
lations for a cooldown analysis contained errors.because the calcu-
lations failed to take into consideration actual as-built conditioas.
,
One deviation and four violations were idertified.
Plant staff,and engineering personnel in the company's headquarters ,
did not appen to communicate effectively. As a result, commitments
made to the NRC regarding electrical breaker control were not fully
implemented and procedures lacked critical elements such as accept-
ance criteria and an applicable setpoint change. In addition, the
licensee's interim procedures for shutting down following a control
room fire and evacuation did not provide for monitoring some im-
\
/
.
21
portant process parameters such as steam generator pressure and
reactor coolant loop temperatures. Also, the NRC identified weak-
nesses in emergency lighting. The licensee subsequently upgraded
the instrumentation and lighting for these evolutions.
A lack of thoroughness also appeared in some modification work.
The results of the control room Halon suppression system discharge
test were not adequately analyzed and dispositioned. The NRC found
that the test revealed that under certain plant conditions the Halon
system would not provide the appropriate level of fire protection.
After NRC identification, the licensee took action to prevent those
conditions which degraded Halon system performance. Another problem
was that plant modifications were made without taking into consi-
deration the effect of these changes on other safety-related equip-
ment. A gas bottle for remote operation of containment isolation
valves was lashed to a conduit in the auxiliary building without
evaluating the consequences of this action. Also, a wheeled breath-
ing air cart (approximately 1500 lbs) was located in the control
room without installing appropriate seismic restraints. The licen-
see began implementation of a comprehensive equipment control and
storage program.
Discrepancies were identified in the previous installation of fire
barriers in the control room and switchgear room. In response, the
licensee committed to review commitments from previous NRC fire
protection evaluations and verify continued maintenance and control
of these items. This project is ongoing.
Although the licensee was aware of the hands-on training expected
for fire watches (these requirements were identified at an inspec-
tion of another of the licensee's facilities), the licensee did not
provide this training at Haddam Neck until it became an issue during
an NRC inspection.
With regard to general fire protection concerns, the licensee had
no designated person dedicated to fire protection management at the
plant. As a result, some fire protection issues did not get the
attention that would be expected if a dedicated person were present
onsite. For instance, local storage of lube oil in unauthorized
containers throughout the plant was not corrected until NRC identi-
fled this problem. During this SALP period, however, the licensee
initiated weekly site inspections by a corporate fire protection
engineer in company with site supervisors. These inspections pro-
vided needed fire protection training onsite, as well as timely
identification of deficiencies in this area. In addition, a per-
manent fire protection engineer position was authorized in January
1986 and is to be filled prior to the July 1987 refueling outage.
- ___ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _
O.
.-
22
In summary, fire brigades were well trained and equipped to fight
on site fires. Significant fire protection program weaknesses were
identified in safe shutdown analyses, in equipment for shutdown
outside the control room, in emergency lighting, in fire watch
training, and in evaluation of plant modifications. 'These problems.
reflect a lack of management involvement in this area and inadequate
coordination of fire protection activities between the site and
corporate staffs. However, fire protection inadequacies did not I
cause inadequate fire. safety. !
,
2. Conclusion
Category 3.
3. Board Recommendation
Licensee: Take appropriate action to improve' management involvement
in the fire protection area and enhance the interface
between the site and corporate staffs.
NRC: None
i
i
_ _ _ _ _ - - - _ _
-
.
23
E. Engineering Support
1. Analysis
This is a new functional area. It is based on observations made
during resident and specialist inspection of technical activities
outside those provided by the operations, maintenance and instru-
mentation and controls (I&C) departments.
The licensee maintains a site engineering group of about 20 engi-
neers and technicians. Additional engineering support is available
from a large corporate support organization (NUSCO) which services
4 nuclear units and several fossil units. The site engineers are
relatively inexperienced and have had a high turnover (9) during
this period. They are responsible for site implementation of design
changes, and for engineering evaluation of operational problems.
More detailed engineering, design and analysis work is requested
from the NUSCO organization. Site engineering also includes reactor
engineering personnel to monitor reactor performance during opera-
tion, to coordinate refueling activities, and to conduct startup
physics tests. Inservice inspection and testing personnel are also
assigned to the site engineering staff.
In most technical disciplines, the large NUSCO support organization
has the engineering knowledge and expertise needed to support the
site. NUSCO supplies design and project management functions for
plant modifications, performs engineering evaluations of site prob-
lems as requested, and manages reactor fuel design and plant safety
analyses. NUSCO also has specialty groups which conduct containment
leak rate testing, monitor plant equipment reliability, develop
environmental qualification (EQ) programs, and provido probabilistic
risk assessment (PRA).
NUSCO PRA support has contributed positively to safety during the
current SALP period. Loss of offsite power was found to have a
potential for resulting in loss of the emergency diesels due to
cooling water valves failing shut. Changes were implemented to
assure that the valves are open to support diesel operation. Also,
as a result of PRA review, additional electrical separation measures
are being incorporated in the design of the new switchgear room
which is scheduled for completion in 1989. In addition, PRA was
instrumental in identifying two emergency core cooling system design
problems. During certain postulated loss of coolant accidents
(LOCAs) in the charging system or core deluge piping, long term core
cooling in the recirculation mode was not assured. Upon identifi-
cation of these problems, the engineering support organization an-
alyzed the safety consequences and promptly proposed interim actions
which were approved by NRC and implemented at the plant. One aspect
of these changes involved throttling a valve in the residual heat
removal system to prevent loss of safety injection flow through a
-- - - - . - - -- . -- -- . --
,-
'
.
24 -
,
. core deluge. system break. Th'e licensee's proposal to throttle the
valve.without a confirmatory flow. test was not accepted because of
the sensitivity.of. flow rate to-valve position. The NRC required
a flow test to be conducted after setting the valve' position. Not- ~l
withstanding this testing weakness,'PRA-related engineering s'upport
has been an excellent contributor to assuring and improving-facility
safety.
The licensee is committed to maintaining the Haddam Neck PRA up-t'o- ,
- date in regard to new plant _ modifications. PRA review of a modifi-
cation to provide a nitrogen blanket on the demineralized water
-
storage tank (DWST) identified that.the modification -increased risk
because the probability of DWST. collapse;due to breather valve -
failure was high. The licensee disabled the nitrogen blanket system '
'
before-it was to be placed in. service. While'the' identification-
'
of this problem shows good performance and diversity within the
. licensee's organization,-it also. indicates a. potential. weakness in
engineering this modification without sufficiently addressing the
- reliability of the breather valve design.
. NUSCO developed the environmental = qualification program forlelec-
- : trical equipment. During the-1986 refueling outage, many component
deviations.from'the test report configurations.were identified.
These. discrepancies apparently resulted from inadequate field veri-
-
fication of various component configurations. NUSCO provided ef-
fective engineering services 1to identify and correct these problems-
prior to plant startup. This occurrence indicated a lack of co-
ordination between the site.and corporate organizations. Similar
coordination problems were noted in the implementation of fire pro _-
- tection program upgrades (see the Fire Protection Functional Area).
Engi_neering support for-plant modifications was> inconsistent during
this period. -The technical design work and safety reviews were-
genera 11y' good. Many modification packages from the 1986 refueling
outage were completed satisfactcrily. In contrast, inadequacies
in post-modification testing for the Appendix R shutdown indication
panel and a containment isolation valve replacement, and material
control and quality certification problems in the main steam isola-
tion valve closure system modification indicate that site engineer
training, supervision, and performance should be improved. The
licensee has been developing engineer training programs,.but imple-
mentation has not been accomplished. (The licensee plans to conduct-
job-specific training for site engineers prior to the July 1987 re-
fueling outage.)
In summary, the engineering support organization was competent and
staffed to support facility needs. PRA support was excellent.
Overall, appropriate attention to plant safety was assured by ex-
tensive reviews by diverse organizations. However, poor coordina-
tion between site and corporate staffs was evident in the EQ and
g
3 Y
,.
.
'25
fire protection programs. Also, problems with implementation of
modifications indicated' inadequate engineer training and/or ex-
perience.
2.- Conclusion-
Category 2
3. Board Recommendations
_
Licensee: Improve communication and coordination between the site'
and corporate staffs.
NRC: None.
.~
y; :
- . -
26
-
F. Licensing Activities
1. ' Analysis.
During the previous assessment period,'.this area:was rated Category
. 2. Further,--a declining ~ trend was noted, based upon an increase
(in the number of incomplete and/or untimely submittals and the.
appearance of a decrease in overall management oversight in assuring
quality submittals to the NRC. Better planning and attention to
regulatory deadlines-~was recommended.
l
The. basis'of the current appraisal was the licensee's support of
"
licensing actions that were either completed or active during the
rating period. These activities consisted of: amendment' requests,
exemption requests, responses to generic letters, TMI Action Plan
l" items, Systematic. Evaluation Program and Integrated Safety Assess-
ment Program (ISAP) topics, and.their related licensing actions.
Licensing activity during the current SALP period has remained very l
high. Forty-six (46) licensing actions were completed during this-
13-month rating period compared to 50 licensing actions' completed
-during the previous'12 month rating period. .In addition to routine
actions, major. activities that have been completed or are currently
ongoing include a fuel reload (Cycle 14), a steam generator tube
sleeving program, initiation of.the' Integrated Safety Assessment
Program (ISAP) pilot program, environmental qualification modifica-
tions, a schedular extension for implementing fire' protection re--
quirements (10 CFR 50.48), and modifications to the Emergency Core >
Cooling systems to resolve recently discovered deficiencies in.long
term decay heat removal. Although 46 actions were completed during
the rating period, 45.new' actions.were added. At the completion
of the rating period, 53. licensing actions-remained active.
During this SALP-period, the licensee has been heavily involved in
p providing the results of the;ISAP analyses for the Haddam Neck-Plant.
The licensee submitted both' the Haddam Neck probabilistic safety
study (PSS) and their proposed final ranking of licensing issues
in the ISAP. The PSS was particularly valuable because it identi-
b fled several areas of plant vulnerability (such as Motor Control
Center No. 5 system interdependencies and certain small break loss-
of-coolant accident scenarios) which were not previously analyzed.
Based upon these vulnerabilities, the licensee took prompt interim ~
action to modify plant systems and procedures. The licerisce also
proposed future modifications to reduce these vulnerabilities.
These actions demonstrate a strong management commitment to plant
safety and a comprehensive effort to continue to upgrade safety at
the Haddam Neck Plant.
_ _.,.___.,.__._.-----------------__u- -
-
.: ,
- i ,
- A ~
'
y
-27
-
.
During the'last rating period,_several licensee submittals were
-untimely and/or-inadequate for. review. -It was concluded that man-
' '
agement oversight in' assuring quality submittals did not appear to ,
be at the level.'of previous: periods. During the first six months. '
of the current rating period, the NRC observed an overalltimprove-
. ment in the quality and timeliness of.the licensee's submittals.
- However, the licensee did not-sustain the improved performance dur-
ing the remainder of the rating perio~d. During-that period, the
licensee requested seven' waivers of compliance-and/or emergency ,
license amendments. .The most notable subject area was-the need to- -
open manual containment isolation valves for routine plant opera-
tions.' Manual isolation valve TS problems had been identified
. earlier but-the licensee's submittals=were not timely or comprehen-
lsive and addressed each necessary operation one at a. time. .Overall,
. this. reflected.a lack of comprehensive planning and timeliness in
. these submittals.
The licensee. has usually exhibited an understanc") f licensing
issues and has generally employed a thorough and < mservative ap-
proach to address potential safety concerns.' .An exampla was'the
technical approach to resolution of ' issues concerning' long term corei
cooling following a postulated break in-the. charging'line. Another
example was steam ~ generator. tube end cracking in allifour steami
generators. These were' examples of.the licensee's clear-understand-
ing of the-potential safety significance.of each issue and'consci-
entious effort:to comply with the regulations using conservative
approaches. However, the approach to resolution-.of an additional
emergency core cooling problem with a break in the core dekge sys-
~
. tem was not up to previous standards.' The_ licensee proposed re-
solution by setting a flow control' valve based'upon calculation
rather than an. actual flow' test, even though-the flow rate wes very
sensitive to. valve position and the margin for error was very small.
The staff could not conclude that the licensee: employed a conserva-
-tive approach in resolving this issue,~and required testing.to' prove -
the acceptability of the valve positioning.
The licensee's response to NRC initiatives has generally been tech-
nically thorough and usually timely. When delays in providing in-
- formation have occurred, adequate' justifications were usually pro- ,
vided. However, the-licensee still has not.been able to resolve ~
several long-standing NRC concerns such as the NUREG-0737 technical.
specification upgrades and degraded grid voltage technical specifi-
cations.
In summary, licensee management and staff have demonstrated dedi-
cated and competent involvement in safety at the Haddam Neck plant.
,
The licensee showed some improvement in the quality and timeliness
! of submittals but this was not sustained throughout the period.
. ;
4
W
- . ,
>
.c_ , . _ _ _ . - _ _ _ _ _ . . _ ._ . _ , _ , _ . _ ___. . _ _ .. _ _ _ _ , _ _ , ,, _ _ _ ,- ._ - _ - -
_
.- - _ . _ _ _ . . . _ . . .
, .
~
...
~
- 28
-
'
-The submitta1 of information was generally timely and technically
sound. However,.too many submittals continue to require the staff
'to request significant amounts of supporting technical information.
,
-
!
, 1
2. Conclusion
Category 2.- >
j
!
3. Board Recommendations
. Licensee: Develop an overall approach for planning submittals-to
assure they are fully supported and submitted in a timely
fashion.-
'
NRC: None.
.
f
.
I
1
4
4 .
4
.
Y
t
4'.
+
F
'".' g; 5
. - -, , . . , , - , , . ._.-~m , - , . . . . - .-, - , - , - - _ _ . , . , , . . . , , , , . . - . . , -
. . . . . . , . , , , . - ~ , - - - . . -
. - _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ - _ - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ - _ - _ _ _ _ _
.
.
29
G. Refueling and Outage Management (240 hours0.00278 days <br />0.0667 hours <br />3.968254e-4 weeks <br />9.132e-5 months <br />, 7%)
1. Analysis
During the previous SALP, this area was rated Category 2. NRC con-
cerns centered on inadequate preplanning and preparation of modifi-
cations and the reactive mode of management caused by the need to
simultaneously plan, review, and implement these changes.
At the beginning of the current assessment period, the 1986 outage
was still in progress. It was completed two months later. Inade- l
quat' preplanning of jobs for this outage reflected an inadequate '
corporate / site interface and challenged the licensee's plant man-
agement and supervisory capabilities. Extraordinary plant staff
efforts generally overcame the negative impact of inadequate pre-
planning. Refueling outage coordination employed computer-assisted
critical path scheduling and management coordinators on each shift.
Project schedules received wide distribution and were reviewed twice
each day at plant status meetings. An indication of the adverse
effect of the high outage workload was the identification, after
plant start-up, that a degraded steam generator U-tube had not been
plugged.
In 1987, licensee management initiated efforts to improve preplan-
, ning of the refueling outage scheduled to begin in July 1987. Re-
(- gular outage planning meetings were widely attended by organizations
involved in outage planning, scheduling, and implementation. NRC
observers particularly noted the development of pre-outage activity
schedules which track the development and planning of outage jobs
! using the same detail and critical path tracking used during outages.
An unplanned shutdown and 15 day maintenance outage occurred in July
l 1986 because of a charging pump failure and the subsequent plugging
of previously identified defects in many steam generator U-tubes.
This identification was due to the application of improved surveil-
lance capabilities. The licensee implemented work activities in
accordance with a preplanned shutdown work list which was supple-
mented by the steam generator repair activities schedule.
1
The main inspection effort during this SALP period occurred during
the last two months of the 1986 refueling outage. The assessment
of that outage, as described in the previous SALP, is unchanged.
While positive steps have been made to improve this area, results
of these efforts will not be demonstrated until they are challenged
during the next refueling outage.
_ _ - _ _ _ _ _ - _ _ _ _ .
'
.y
,
.
30
2. Conclusion
Category 2.
3. Board Recommendations
None.
.
- _
, , . . _ . _ .-__
.
.- t
31
H. Radiological Controls (507 hours0.00587 days <br />0.141 hours <br />8.382936e-4 weeks <br />1.929135e-4 months <br />, 14%)
1. Analysis
The last SALP rated this area as Category 2. Concerns included:
recurrence of minor, self-identified violations; weak job planning,
faulty procedure adherence and corrective actions, and use of poor
judgement; and tracking of exposures without effective control.
A radioactive drum was compacted in violation of procedures, with
unnecessary worker exposure and work area contamination. Also, a
worker exceeded his assigned exposure by about a factor of two while
staying unnecessarily in a high radiation area-to resolve a problem
with mismatched couplings. Exposures were consistently higher than
licensee projections. Proposed corrective actions were not effec-
tively supported by management. Comprehensive licensee review and
upgrade of the ALARA program was recommended.
The radiological controls program and organization are acceptable.
Lines of authority are clear. The staff is professionally capable.
During this SALP period, external radiation control, dosimetry, and
respiratory protection were generally good. The radiological con-
trols staff provided many suggestions for improvement. Management
was generally responsive. A noteworthy example was implementation
of a zone concept for control of radiation work areas, with each
zone supervised by a health physics foreman, improving supervisory
oversight.
Escalated enforcement action was taken for the exposure of a steam
generator worker to a quarterly total of 3.3 rem (3 rem federal
limit). Station health physics supervision had not been overseeing
or monitoring this very high radiation field work. Station manage-
ment had not observed and evaluated the activity. This event re-
flected unacceptable program implementation and management. NRC
review also concluded that the licensee had treated health physics
events as individual failures to follow procedures without address-
ing the related supervisory and program implementation inadequacies.
Evaluation of the effectiveness of the licensee's corrective actions
awaits observation of refueling outage activities.
Radiation protection procedure inadequacies remain an NRC concern.
Survey and radioactivity counting procedures are ambiguous. Many
procedures are difficult to understand. Key concepts are omitted
or poorly defined. There are multiple technical and typographical
errors. A typical inadequacy was the specifying of undefined acti-
vities (e.g., intermittent surveys, continuous coverage) in proce-
dures. These were not well understood and contributed to health
physics events including the overexposure identified in the preced-
ing paragraph. This demonstrates lack of attention to detail in
procedure development and inadequate initial and periodic procedure
review.
, . x
_
32
Y
Deviation-from p'rocedures also remains an NRC concern. .An effective
-means of assuring technician: familiarity with procedures affecting
-
.*. their work areas,:and with associated changes, has not been evident.
The ineffectiveness of the existing briefings.by supervisors and
individual certifications of knowledge and understanding.was shewn
by inadequate control of significant radiological operations'and
of high radiation area entries.
Despite the' management support evident in the raising of the value
of preventing a man-rem to $20,000,.ALARA; weaknesses-continue.. The
p
"
ALARA staff consists of a coordinator and an assistant. Though
qualified, they were overwhelmed by'the number and scope of needed
ALARA initi_atives. Computers were available, but a lack of software
. necessitated use of less efficient manual methods. For example, .
,
ALARA. personnel were observed to be manually totalling exposure data
to the detriment of their'other duties.. Radiological controls per-
sonnel appeared ineffective in limiting access, stay times, and-
-
nonproductive entries. Job completion was often delayed, with im-
proper fit of components a typical problem. Packages for onsite-
ALARA' review were typically incomplete and so late that there was
.
insufficient review time. Exposure reduction-initiatives focused
on shielding,1 decontamination, job cancellation, and other technical
aspects. Weaknesses in access'and stay time. control were. generally
not addressed. ALARA goals established by the corporate staff were
not based on specific job analyses, and.were therefore regularly
exceeded. -The:1700 man-rem exposure for 1986 is considered to be
unnecessarily and. extremely high. .While the ALARA goal for steam
generator work in July 1986 was achieved and exposure goals are
normally met for non-outage work, the problems identified show that
ALARA controls.have not been properly effective.
'A December 1986 licensee ALARA appraisal attributed unnecessary
radiation exposure to several of the ab'ove weaknesses. Corrective
-actions were not formulated during the SALP period. Improved pre-
parations for'.the 1987 refueling outage have, however, been evident.
Regular planning sessions have included ALARA personnel. -Several
iterations of ALARA goals for outage projects have been evident.
Daily management planning meetings have incorporated increased at-
tention to radiation exposure goals. Long-standing program defi-
ciencies have been addressed by clear assignment of responsibility
for exposure controls ~and by more stringent management insistence
upon procedure adherence. The onsite health physicist-position
was filled by an individual who appears to be changing the job
function from a consulting role to one of active participation in
ongoing activities. Also, a new staff position was added under the
Station Services Superintendent. That position, although it does
not directly involve the radiation protection staff, should free
the station services superintendent for more attention to health
physics matters. These steps show management support, but overall
assessment of their effectiveness must await evaluation of the
i forthcoming refueling outage.
- y ,
,
Q-
- o
'
c.
-
- 1 '
33
.
-The corporate. office provided the site with standardized-health
. physics ' procedures. These, though different from the ones used on
. site, .were generally very well written and provided valuable tech-
- nical. references. Monthly corporate audits of site activities were
performed. The NRC concluded.that these audits were not of'suffi-
cient depth and. breadth to effect significant' program improvements.
Overall, corporate radiation protection staff input was-assessed
as not having a significant impact on site radiation protection.
INPO accreditation of health physics-technician training is'sched-
uled for 1987. This.. training has been upgraded in scope and depth.
Lesson plans have been improved. More time _is spent in training.
Effectiveness is better measured by increased use of exams. These
are-substantial program' improvements, but newness prevented assess-
ment of their effectiveness.
Facilities and equipment were generally adequate. Concerns included-
irregular timing of' quality control (QC) checks on instrumants and
.QC. checks being performed without a clear understanding of the sig-
'nificance of the results. For' example, control charts-can be'used
to plot results of successive tests, with the acceptance band
clearly denoted. Chart examination illustrates both the current
performance and the trend. Such control charts we're not used at'
Haddam Neck, and NRC inspection found a lack of technician awareness
of the associated performance characteristics. Also, although
access to computers was available,. software development has lagged.-
-The resultant need to perform functions manually was assessed as
prone to error and an inefficient use of licensee staff time.
Effluent controls are the responsibility of the chemistry eeganiza--
tion. Extensive laboratory. training facilities-including plant
specific, state of the art equipment have been completed. Based
on participation in the new chemistry program and inspector discus-
sions with technicians, the NRC concluded that:the chemistry staff
was acceptably qualified. INP0 accreditation of chemistry techni-
cians is scheduled for 1987.
The offsite dose calculation manual was generally implemented as
required by.the Radiological Effluent Technical Specifications
(RETS). However, the licensee did not perform a pre-implementation
audit of the RETS. Onsite licensee QA review of RETS surveillance
implementation was scheduled for six months after program implemen-
tation. Four months after implementation, the chemistry organiza-
tion identified two surveillances which were not addressed in pro-
cedures and initiated procedure coverage. No additional licensee
review was implemented. Later, the NRC identified two violations
of the RETS. One involved the procedures not prescribing all of
the sampling and analysis frequencies for continuous liquid releases.
The second involved daily grab samples not being collected and
weekly analyses not being performed for continuous releases from
. .
- _ - _____ __ - _ _ _
.
.
34
steam generator blowdown and service water effluent. The NRC also
identified record management weaknesses. These included undated
offsite dose calculations which did not demonstrate compliance with
the 31-day frequency requirement and unavailability of a gamma re-
lease record for a batch liquid release. Overall, RETS implementa-
tion reflected a need for better site and corporate interaction and
for pre-implementation validation of new programs and procedures.
Radwaste was generally well controlled. Processing systems were
well maintained. Outstanding maintenance actions were less than
two months old. Licensee personnel showed good understanding of
system capabilities and alarm responses. Minor weaknesses were ;
evident in violations related to characterization and documentation
of radionuclides in radwaste shipments. Previous NRC appraisals
had highlighted inadequacies in QA for waste packaging and shipment;
corrective actions had not prevented subsequent violations. Cor-
rective actions now appear effective, in that no subsequent problems
have been identified. Also, after the SALP period, NRC inspection
,
found the programmatic deficiencies to be corrected.
In summary, the basically sound radiation controls performance has
been marred by an above limit personnel exposure, by weaknesses in
exposure controls, by procedure inadequacies, by failures to follow
procedures, by ALARA weaknesses, by a lack of effective corporate
radiation protection staff input, and by inadequacies in RETS im-
plementation. Recent site management attention to problems has been
obvious, but the results will not be tangible until the refueling
outage occurs.
l 2. Conclusion:
Category 2
i
3. Board Recommendation
Licensee: Assure effective site and corporate coordination and pre-
planning of radiological controls activities.
NRC: Perform a special inspection of refueling outage radiation
controls.
>
_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ _ _
_
.
.
35
I. Emergency Preparedness (201 hours0.00233 days <br />0.0558 hours <br />3.323413e-4 weeks <br />7.64805e-5 months <br />, 6%)
1. Analysis
During the previous assessment period the licensee was rated Cate-
gory 2 in this area, principally because of weaknesses noted in
communications between emergency response facilities and the number
of weaknesses noted during the annual exercise.
During the current assessment period there were two region-based
inspections. One inspection included NRC observation of the licen-
see's April 25-26, 1986 annual full participation emergency pre-
paredness exercise. Another inspection covered changes to the
Emergency Preparedness (EP) Program; shift staffing and augmentation;
knowledge and performance of duties; dose calculation and assessment;
and licensee audits.
During the 1986 emergency exercise the NRC observation team identi-
fied several strengths and weaknesses in the licensee's EP program.
The overall assessment wa: that performance provided reasonable
assurance that the public could be protected. Specific strengths
were noted in the areas of accident mitigation, communications with
the state, classification of plant conditions, protective action
recommendations, rumor control, press releases, and strong technical
management in the corporate Emergency Operations Center (E0C).
Weaknesses identified by the previous exercise as requiring correc-
tive action were adequately demonstrated and closed. However,
several new items were identified as potential weaknesses. The more
significant of these include:
--
Control room personnel interchanged ALERT and SITE AREA
EMERGENCY terminology during communications;
--
Site evacuation was ordered without consideration of the tor-
nado watch which was in effect;
--
Director of Station Emergency Operations (DSE0) did not inform
the Manager of Control Room Operations that he was assuming
DSE0 duties; and
--
Communication between the site and corporate EOCs was weak,
in that the corporate E0C was not clearly informed of the basis
for the SITE AREA EMERGENCY declaration.
Communication of information between remote emergency operating
facilities has been a long standing weakness. The above noted ex-
ercise observations indicated that communication problems were still
evident. Also, NRC observed slow implementation of control room
data transmission functions during a March 1987 emergency staff
augmentation drill.
-
..
s
jl~
"
. 36
NRC review of EP implementing procedures (EPIPs) revealed'that there
was no recognition of an interface between the licensee's DSE0 and
an NRC Director of Site Operations' (D50)/ Site Team Leader, _.if an
- NRC' response. team was on-site. Specifically,.the need to keep the
NRC DSO informed of any notifications / protective' action recommenda-
tions being transmitted to the State was not specified. The licen-
see was responsive.to this' concern and agreed to make necessary
procedure changes and stress this issue during training.- Another
NRC concern was the effectiveness of the emergency staff augmenta-
-tion plan. The licensee's. Emergency Plan extends the. allowable
staff. augmentation time to one hour rather than the recommended 30
minutes. This. capability had not been demonstrated by unannounced
.
call-in. exercises. This concern was reemphasized on March 24, 1987
when-the licensee conducted an unannounced. drill and three responders
did not reach the site within one hour. Strong management response
was noted in emphasizing both personal and supervisor accountability
for the performance of scheduled responders. The' licensee plans
to continue to randomly exercise the augmentation plan.
The licensee's emergency response facilities are generally well-
.
designed and complete with the exception of the installation of a
safety parameter display system (SPDS) and_the monitoring of certain
process variables. The system improvements await. installation of
a new main frame computer which is scheduled for the July 1987 out-
age. Final implementation.and training for SDPS is-scheduled for
the spring of 1988. This schedule ha's been accepted by NRC. Upon
completion, the licensee facilities will be fully assessed in an
NRC Emergency Response Facilities Appraisal.
The licensee's corporate. organization performs audits and appraisals
of_the emergency plan, EPIPs and corporate emergency procedures.
These audits cover all the major elements of the EP program-at least
once every_12 months. The audits are thorough and detailed, and
provide assurance that weaknesses or potential weaknesses will be
identified. The audit review and distribution process.is extensive
and adequate administrative controls exist for documentation and
follow-up of corrective actions.
~The Emergency Preparedness Staff at Haddam Neck is adequate, con-
sisting of an onsite Emergency Preparedness Coordinator and Assist-
ant Emergency Preparedness Coordinator. Additional assistance is
available from the Supervisor Emergency Preparedness at.the Corpor-
ate Headquarters in Berlin, Connecticut. The licensee continues
to maintain an excellent-working relationship with the State of
Connecticut and local governmental agencies as evidenced by the
continuing cooperation demonstrated during exercises.
- _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _
_ _ . _
.
- .
37
Overall, the licensee maintains an EP program capable of providing
adequate protective measures in the event of an emergency. Corpor- l
ate involvemeni,in EP is evident and slow but steady improvements
have been observed. The good working relationship with the State
of. Connecticut is considered a strength. Exercise performance is
satisfactory, but better communications between E0Cs and the control
room, improved emergency staff augmentation performance, and com-
pletion of outstanding response facility hardware upgrades will
-increase the overall effectiveness of the licensee's emergency
response functions.
2. Conclusion
Category 2.
3. Board Recommendation
None
!
_ _ _ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ - - _ _ _
< l
l
.
38
J. Security and Safeguards (188 hours0.00218 days <br />0.0522 hours <br />3.108466e-4 weeks <br />7.1534e-5 months <br />, 5%) '
1. Analysis
During the previous SALP period, no regulatory concerns were iden-
tified and the licensee's performance was assessed as Category 1.
The NRC has not cited any violations of its requirements during four
consecutive rating periods (1980 to present).
The licensee has maintained this high level of security program
effectiveness for a period of almost seven years. NRC attributes
this to:
--
strong management involvement and support for the program in-
cluding effective licensee oversight of the contract security
force;
--
a comprehensive security program based on NRC performance ob-
jectives rather than minimum regulatory requirements, and up-
grading of that program to meet changing nec-ds on a regular
basis;
--
utilizing state-of-the-art systems and equipment and maintain-
ing systems and equipment in good working order by providing
the necessary technical oversight;
--
establishing and maintaining a well trained, aggressive and
highly motivated security organization that emphasizes personal
accountability; and
--
compensation for a high security force turnover rate by estab-
lishing and maintaining a highly effective training program
that includes feedback derived from personnel performance,
audits and surveillances, and program upgrades.
Corporate support for the program was demonstrated by the licensee's
active participation in various nuclear industry professional groups
and associations and by providing the necessary resources to upgrade
the program on a regular basis. An example of this support was the
licensee's development of a read and sign indoctrination training
program for NRC inspectors. This program significantly improved
the licensee's capability to provide prompt, unfettered NRC access.
To ensure the continuing effectiveness of the security program, the
licensee utilizes comprehensive corporate auditing and site-based
self-assessment programs that are carried out by personnel with
nuclear plant security expertise. The corporate audit program is
conducted on an unannounced basis and includes periodic reviews of
selected aspects of the security program throughout the year. All
aspects of the program are reviewed by the end of the audit year.
.
...
39
Audit reports were distributed to responsible senior management per-
sonnel for review. Corrective actions on identified deficiencies
were prompt and effective in all cases and followup was conducted
to ensure that corrective actions were effective. The site security
organization also performed continuing self-assessments of the pro-
gram using NRC inspection criteria and program performance objec-
tives. These assessments were conducted independently by propri-
etary supervisors or, in some cases, by contract security force
supervisors with proprietary oversight. Using these techniques,
the licensee was able to identify and correct weaknesses and poten-
tial violations before they adversely affected the program. Ex-
amples are listed below.
Four security event reports were submitted in accordance with 10
CFR 73.71. One event involved a brief system failure as a result
of a component malfunction. The remaining events involved the lic-
ensee's identification and repair of two vital area barrier weak-
nesses and of an isolated personnel error which resulted in an un-
secured vital area door. The licensee's response to these events
was prompt and thorough. In the case of the barrier weaknesses,
those problems were identified during scheduled security plan self-
assessment efforts and follow-up on an NRC Information Notice. Each
event was appropriately handled and compensatory measures were
promptly initiated, when required. Event reports were clear, con-
cise and submitted to NRC in a timely fashion. The small number
of reportable events is further evidence of the quality of the lic-
ensee's program.
The licensee's security organization was adequately staffed with
well qualified personnel. Both proprietary and contractor super-
visors were well trained, professional and provided effective super-
vision. A noteworthy strength in the organization was the licen-
see's policy of requiring security force members to qualify by ex-
amination in all positions, up to Alarm Station Operator, before
being eligible for promotion to the rank of Sergeant. This policy
ensured a detailed knowledge of security program requirements, sys-
tems and equipment prior to consideration for a supervisory position.
This appeared to significantly enhance program implementation and
quality. All members of the force were well trained, knowledgeable
of the program, and highly motivated. This was consistently demon-
strated in interviews and records reviews conducted during NRC in-
spections and again demonstrates the licensee's commitment to a
quality program.
Facilities were clean and well maintained. Sufficient space was
allocated for the administrative and operational needs of the pro-
gram. Records were also well maintained and readily retrievable.
Record repositories were found to be in accordance with NRC require-
ments and properly secured.
. _
,
.y
s .
y
40
.
Training facil'ities are well designed and provide for the efficient
administration of the training program. The training was admini-
, - -stered by qualified, full-time instructors. Lesson plans and in-
'structional aids were maintained current' utilizing feedback from
day-to-day ~ operations and the audit and.'self-assessment. programs.
.The training instructors also played a key' role in evaluating the
effectiveness of.the security-force drill program and.in assessing
individual and team proficiency levels.during these drills. A-total
of 196 drills and 75 individua1' performance evaluations were con-
' ducted during'1986. These drills served to reinforce the training
and qualification program and increase the performance level and
- proficiency.of the-security force.
.Two revisions-to the Security Plan were submitted under the provi-
sions of;10 CFR 50.54(p) and:the licensee also responded to the
Amendment to 10 CFR 73.55, codified by NRC on August 4, 1986. The
licensee's staff responsible for these actions had a thoroughL
knowledge and understanding of NRC security program objectives
and ensured that plans were current and that changes'were-properly-
coordinated, when required. Security. Plan changes were ' coordinated
.with: regional safeguards licensing personnel to ensure that a clear-
understanding of each change existed. These changes were usually'.
of~high quality, indicating appropriate staffing and in-depth man-
.
agement review. -With respect to the revision currently under NRC
review, some changes were not-adequately reflected in the. text of
the plan.
'
' '
In summary,.the licensee continued to maintain a highly' effective
and proactive security program. This achievement reflects favorably
upon the-managers and supervisors at both corporate and site levels,
and on members of the security force who have aggressively pursued.
-excellence in the discharge of-their responsibilities. Management
attention to and support of the program were clearly evident from
the high degree of success which was achieved.
2. Conclusion
Category 1.
3. Board Recommendation
None.
_
..- ;
l
.
41
K. Training and Qualification Effectiveness
l
1. Analysis
Training and Qualification Effectiveness is an evaluation criterion
for each functional area. During this SALP, it also is being
separately addressed as a synopsis of the assessments in the other
areas. Training effectiveness has been measured primarily by the
observed performance of licensee personnel and, to a lesser degree,
through program review.
Training effectiveness was rated Category 2 during the previous SALP
period. NRC concerns included recurrent weaknesses in ALARA, modi-
fication control and fire protection. During this SALP period,
training contributed to the reduction in fire barrier related re-
portable events. Weaknesses in exposure control and modifications
were again apparent, however.
Management commitment to quality training was evident from the ex-
cellent facilities and staff which have been developed. During this
SALP period, the operator license programs received INPO accredita-
tion and a new, plant specific simulator was put into operation.
In addition, senior operations personnel were promoted to the
training staff, improving training background in operations. The
training staff includes more than 20 instructors, most of whom hold
NRC licenses or instructor certifications. However, many of these
instructors are not licensee employees. The licensee has made an
effort to reduce the number of contractors in the training organi-
zation, but the large number remaining provides a high potential
for an unanticipated reduction of plant specific expertise. However,
no associated impact on operator performance or training effective-
ness was identified during this SALP period.
Initial operator training continued to be a program strength with
14 of 15 successful license candidates during this SALP period.
Significant improvement was noted in the operator requalification
program, although administrative controls were not completely im-
plemented.
Technical staff training was also improved in most functional areas
during this period. This training includes academic instruction
in areas such as mathematics and science as necessary for job-
related tasks, and practical applications in the icboratory and
jobsite locations. While the training organization is located off-
site, sufficient training staff works at the site to provide on-the-
job training and operational feedback to the training programs.
The technical training programs are scheduled for INP0 accreditation
in 1987. Personnel interviews have indicated generally positive
evaluations of the quality and applicability of the new training
initiatives. On the otter hand, operator and maintenance errors
_-
r
- .
42
caused three plant trips during this SALP period and there continued
to be a high proportion of reported events involving personnel error.
In addition, weaknesses-in chemistry department training related
to implementation of the Radiological Effluent Technical Specifica-
tions contributed to'several NRC and licensee identified problems
in that area. These factors and the recurrent instances of proce-
dural compliance and documentation inadequacies noted in the main-
tenance,. surveillance, and radiation protection areas indicate-that
training in these areas has not yet been properly effective.
The licensee has not yet implemented formal training programs ad-
dressing concerns about modification control and testing. This lack
of training, particularly for relatively inexperienced plant staff
members, has contributed to recurrent problems such as inadequate
post modification testing of the Appendix R shutdown instrumentation
and of a replacement containment isolation valve.
The General employee training (GET) program adequately addressed
orientation, radiation protection, security, emergency planning,
safety and assurance of quality. GET content is directed by a
steering committee made up of station managers who determine program
emphasis based on station performance goals. During this SALP
period, GET emphasized the importance of fire barrier control. No
further events occurred in this area and NRC noted appropriate con-
cern for fire barriers during discussions with personnel in the
field.
The licensee's extensive training program for the plant security
staff was judged to be a significant factor in the continuing ex-
cellence of the security organization.
In summary, the licensee's significant commitment to training is
evident in the development of extensive training facilities and
competent management and staff. Operator license training programs
were effective and the security staff training was excellent. With
the exception of engineer training, improvement was noted in the
technical training programs. However, training did not prevent
recurrent problems with personnel errors and radiation exposure
control.
2. Conclusion
Category 2
3. Board Recommendation
None
l
.
..
43
L. Ass'urance of Quality
1. Analysis
Management involvement in assuring quality is an evaluation cri-
terion for each functional area. Quality assurance (QA) also is
an. integral part of each' functional area. This Assurance of Quality
~ area.is a synopsis of the applicable aspects of other areas, in-
cluding worker and supervisor. performance, management oversight,
and safety review committee activities. The last SALP rated this
area as Category 2, with the major concern being for the effective-
ness of self-evaluation and resolution of problems. It was recom-
mended that the associated licensee programs be reevaluated.
During the current SALP period, licensee personnel were assessed
as technically competent and as focused on doing things properly
the first time. Most activities were carefully conducted in ac-
cordance with relevant directives. The necessity for careful at-
tention to detail was acknowledged. In contrast, a willingness to-
deviate from procedural requirements was evident from the analyses
of the preceding functional areas. Also, the total of 20 violations
and 17 personnel error related events (including three plant trips)
showed multiple inadequacies in adherence to requirements. The
number and nature of these occurrences show substantial room for
improvement in performance and supervisory overview.
Breakdowns in the overall assurance of quality were evident in the
above limit exposure of a steam generator worker and in the rest
of the violations of NRC requirements. For the overexposure, it
was particularly relevant that neither senior health physics nor
senior station managers had provided evaluation / overview of the
activity, which had been ongoing for several weeks when the over-
exposure occurred.
First line supervisors generally provided oversight of activities
and were knowledgeable of associated design and administrative re-
quirements. They were often observed to be providing guidance at
work sites. Operating shift supervisors exhibited thorough knowl-
edge of plant activities. Daily first line supervisor meetings were
held to coordinate work activities, and this new initiative was
found to be an effective means of controlling mutual interference.
Department supervisors also were observed to be in the plant fre-
quently. They were found knowledgeable of their departments' acti-
vities and of significant plant problems. Daily meetings to discuss
planned activities, review problems, and assign corrective action
responsibilities were considered effective. Licensee tracking sys-
tems were found effective in closing items, except that the large
number of NRC items open for a prolonged period indicates a lack
of aggressive licensee follow-up of such items.
.
e
44
The unit and plant superintendents made frequ'ent control room and
plant tours. Formal weekly inspections by the superintendents and
department supervisors were made, and improved housekeeping was one
result.
Senior plant staff members acted as management duty officers during
operations and outages. In addition, management representatives
were assigned for full time coverage of outages. These measures
and the daily staff meetings were generally effective in assur_ing
proper. activity control. Station management was noted to be ac-
tively involved in the daily meetings, with action items and due
dates being generated as a result.
Rapid identification of problems was achieved by the Plant Incident
Report (PIR) system. There was a low threshold for PIR initiation,
and over 200 PIRs were written in 1986. The PIRs were an excellent
tool for site managers, who maintained awareness of the associated
root causes and corrective actions.
QA routinely provided inputs to plant managers on the effectiveness
of activities. There were frequent quality control (QC) inspections
of the proper completion of safety-significant jobs. QA/QC person-
nel were appropriately qualified and certified.
QA surveillance of in process activities was increased, with about
70 separate evolutions observed. Findings were provided to depart-
ment supervisors, and corrective actions were verified during sub-
sequent surveillances.
The QA audit system functioned acceptably. Schedules were followed.
Checklists were well organized and comprehensive. Corrective action
timeliness and thoroughness remains a concern, however. Procedural
deficiencies were the subject of audit findings in 1981, 1983, 1985,
and 1986, clearly indicating ineffective correction of this generic
problem.
The Plant Operations Review Committee (PORC) and the Nuclear Review
Board (offsite committee) provide quality oversight of safety-
related activities. NRC observed frank, open, and knowledgeable
discussions of issues and a sound approach to safety was clearly
demonstrated. The contribution of these committees was shown both
in the routine review of plant documents and changes, and in assess-
ing the safety significance of two safety system design errors which
were identified during this SALP period. The committees were in-
strumental in development of corrective actions with appropriate
attention to continued plant safety.
Nuclear Safety Engineering (NSE), the independent safety engineering
group which is part of the corporate staff, was active in its
coverage at Haddam Neck. This onsite group had ready access to the
_-
e
4
45
plant staff, equipment, and records. NSE assessed plant safety
programs.and evaluated plant operating experiences through reviews
of. procedures and data including independent reviews of the resolu-
tion of some Licensee Event Reports (LERs) and PIRs. In addition,
NSE performed special reviews and evaluations as requested by plant
management. An example is an ongoing NSE review of industry pro-
cedures.
In summary, assurance of quality was generally satisfactory. Good
performance was evident in most aspects of performance. But there
was unacceptable control over the high radiation work involved in
steam generator entries, an ineffective ALARA program, a high number
of plant trips, a high number of violations of NRC requirements,
and a high number of personnel errors resulting in plant transients
and events. Sound recent management initiatives were evident in
outage planning and in initiation of action items and schedular
tracking on problems identified during daily activities. Overall,
it appears that recent positive initiatives have, so far, had less
affect on performance than long-standing practices and attitudes.
2. Conclusion
Category 2.
3. Board Recommendation
None
"
3,
- .
46
1/.' . SUPPORTING DATA AND SUMMARIES
A. Investigation and Allegation Review
None
B. Escalated Enforcement Actions
1. Civil Penalties
A $50,000 civil penalty was issued on December 10, 1986 for an
occupational exposure exceeding 10 CFR 20 quarterly limits.
2. Orders
None
'
3. Confirmatory Action Letters
None
C. Management Conferences
On June 18, 1986, an enforcement conference was held at the NRC
'
- 1.
Region I office to discuss violations and program weaknesses related
to the radioactive waste transportation area.
2. On April 7, 1986, a management meeting was held at the Northeast
Utilities Corporate Office to discuss the results of the independent
review of the design change control program required by NRC Order
dated December 13, 1984.
3. On September 3, 1986, an enforcement conference was held at the NRC
Region I office to discuss an above linit occupational whole body
exposure and several violations of NRC fire protection requirements.
D. Review of Licensee Event Reports (LERs)
1. LERs Reviewed
LER No. 86-10 to 87-03
2. Tabular Listing
a. Personnel Errors 17
-
b. Design / Man./Const./ Install 7
c. External Cause 0
d. Defective Procedure 0
e. Component Failure 15
<
x. Other 3
Total 42
.__ _ . _ _ _ _ _ _ . ____ _ _ _ __ ~ .____ _ . _ _ _ ._.
- - _ .
.
.
47
'A tabulation of LERs by functional area and an LER synopsis ~are
attached as Table 4.
3. Causal Analysis (Review Period 3/1/84 - 3/31/87)
Four sets of common mode events were identfied:
a. LERs 86-21, 22, 23, 24, 30, 32 and 87-02 reported age-related
nuclear instrumentation anomalies. Five of these resulted in
turbine load runback actuations.
b. LERs 85-17, 85-29, 86-13, and 86-48 reported engineered safe-
guards systems design / analysis errors.
c. LERs 84-10, 86-04, 86-33 and 86-46 reported problems with
operability of the low temperature overpressure protection
system.
d. LERs 86-43 and 87-03 reported problems with the operability
of the wide range plant stack monitor.
Other notable trends were the continued high percentage of personnel
errors and component failures.
E. Summary of Licensing Activities
1. Schedular Extensions Granted
August 25, 1986 Schedular Extension for 10 CFR 50.48,
Switchgear Room Modifications
December 15, 1986 Schedular Extension to Commitments for
Halon System Mcdifications in the Switch-
gear Room
2. Reliefs Granted
May 12, 1986 Relief from Inservice Inspection Require-
ments from the 1980 Edition of the ASME
Code
3. Exemptions Granted
April 28, 1986 Exemption from General Design Criteria 35 - Small Break LOCA Single Failure
Requirements
-
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4. License Amendments Is' sued' ; ' '
. ~ y
s Amendment -73 , Primary Syst$n Leakage
April 14,p;h,986
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.
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April (14,11986 .. Amendment 74 Cycle 14 Reload. , , f.
, .:
3
April 29, 1986 Amendment 75
~
' Firs * Detection and' Spray
'
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Systems 3
'
c, ,
-
.
~
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- May 14, 1986
-i s .
' Amendment 76 - Steam Generator Tube- -
,
Plugging
- July 11, 1986 . '
Amendment 77 - High Pressure Recirculation
1 for Emergency' Core Cooling
,
'
',- Systems
., . .,
July 14, 1986 Amendment 78 - Steam Generator Tube
Sleeving
August 6,.1986 Amendment 79 - Administrative and Reporting
Requirements _
c August 7,'1966 z '#-Asendment 80.- Minimum Shift Crew Composi-
L tion
August 18, 1986 ' Amendment 81 - Control Woom '
Fire Detection-
Systsm
, . .
.
(4,
- Se'ptember 3, 1986 Amendment 82 - Inservice ,In,spection of
- '
Steam Generator Tubes-
E \>)
September 9, 1986 Amendment 83'- Fire Protection Audits
-September 18, 1986 Amendment 84 - Quadrant Power
_q sa Tilt Ratio
- September 29,.1986 Amendment 85 - Monthly Operating Reports
'
- Cctober 30, 1986 i Amendment 86-ManualC'AntainmentIdolation
Valves i
o. >
, p,
November.12, 1986 *
Amendment 87 * Inservice Inspection of s ,
s .", Reactor Coolant Pump
"
Flywheels
"
r ,, f) , . <-
December 24,'1986 Amendment 88 - RHR Flow Control Valve 796
in Emergency Core Cooling
'
,
' . ..
' *
. System A>
s ,
February 9,1987 . Amendment 89 - Charging Pump fechnical
'
, . ,
- Specification's
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E,
.
.
49
- March 11, 1987 Amendment 90 - Manual Containment Isolation
Valves (Neutron Shield Tank)
- Waiver of Technical Specifications compliance granted.
5. Emergency Technical Specifications Issued
May 14,-1986 Amendment 76 - Steam Generator Tube
Plugging
December 24, 1986 Amendment 88 - RHR Flow Control Valve
796 in Emergency Core
Cooling System
6. Orders Issued
o July 2, 1986 Revision to TMI Confirmatory Order
Related to SPDS/DCRDR Schedules
'
F. Description of Plant Activities (3/1/86 - 3/31/87)
At the beginning of this SALP period (March 1, 1986), the facility had
been shutdown for a refueling and maintenance outage since January 4,.
1986. Outage activities included the tore XIV reload, steam generator
eddy current testing and tube plugging, containment leak rate testing, .
backfit modifications and plant equipment upgrades, and secondary system
overhaul / repair work. On March 2, the licensee safely recovered the fuel
element which was dropped onto the core February 26, 1986, during reactor
disassembly for fuel offload. On April 1,1986, the licensee reported
an error in the existing plant safety analysis for small break loss of
coolant accidents (LOCA). Certain small breaks at one reactor coolant
system location could preclude the proper operation of emergency core
cooling systems (ECCS) in the containment recirculation mode. The lic-
ensee developed a set of administrative and procedural controls to mini-
mize this event pending ECCS modifications scheduled for a future outage.
NRC approved the licensee's program including a temporary exemption from
the single failure criterion on April 28, 1986.
Plant heatup began on April 27, 1986, and low power physics testing was
conducted satisfactorily on May 6-8, 1986. On May-7, the licensee dis-
covered that one defective steam generator (SG) U-tube in SG #2 had not
been plugged as required by Technical Specifications (TS). The licensee
isolated SG #2 and reactor coolant system (RCS) loop #2 and requested
TS relief to allow cycle XIV operation with the 55% through-wall defect
unp' lugged. NRC Licensing granted this relief on May 8, 1986. i.thile
still-in 3-loop operation on May 8, the plant automatically tripped from
10 percent power. The blocked main steam isolation valve (MSIV) trip
signal, which existed because of the closed MSIV on SG #2, automatically
unblocked when power was raised above the 10% trip reset point during
a main turbine balancing evolution. The trip was due to operator in-
_ _ - _ _ _ _ _ - _ _ _ - _ - _ _ _ _ _ - _ _ _ _ _ _ _ _
.
.
50
_.
o
attention to the existing blocked trip signal and its automatic unblock
setpoint. While the plant was shutdown, SG #2 and RCS loop #2 were un-
isolated since TS relief had been granted to allow operation of SG #2.
The reactor was restarted on May 8 and the unit was phased to the grid
on May 10. Full power was reached on May 22, 1986.
During this assessment period, spurious fluctuations in nuclear instru-
mentation (NIS) channels caused several inadvertent turbine load runback
actuations. The licensee determined that component aging within the NIS
was the primary cause of these fluctuations. In addition to repairing
individual failed components, a project to replace the NIS with state
of the art equipment has been initiated.
On June 4, plant operators manually tripped the plant at full power due
4 to a loss of main feedwater (rapidly decreasing SG levels) caused by a
--'
\ failed closed heater drain tank level control valve (LCV). Following
repairs to the LCV on June 5, the plant operated at full power until June
17, when the same LCV again failed shut and operators again manually
tripped the plant in anticipation of an automatic trip. This time the
LCV was modified to prevent the separation of the valve stem and operator,
which caused both the June 4 and 17 trips. Main Steam Isolation Valve
(MSIV) stroke testing following the June 17 trip revealed that the MSIV
air operating accumulator did not have adequate capacity to close all
four MSIVs without non-safety-related make-up air. The reactor was re-
started, but power operation was delayed until MSIV air system modifica-
tions were coapleted on June 22. Durirg this period, an automatic reac-
tor trip from zero power occurred on June 19 while troubleshooting a
failed NIS power ranga drawer. Operators and technicians failed to re-
cognize the consequences of re-energizing the shorted NIS drawer. A
second trip occurred during plant startup on June 22. The plant had been
placed in a 3-loop operating conriition to measure RCS flow in this
operating mode. A reactor protection system (RPS) loop low flow relay
contact had been obstructed by a small piece of ceiling tile, creating
an unnannunciated reactor loop low flow trip signal. Two low flow trip
signals are required to trip the reactor between 10% and 74% power. A
second trip signal was actuated and annunciated by the idle loop in the
3-loop operating mode. When the low flow trip signal was automatically
ublocked by design at 10% power, the reactor tripped. The ceiling tile
most likely entered the RPS relay as a result of control room work during
the 1986 refueling outage. Upon identification and correction of the
cause of the trip, plant operation resumed on June 24. Full power was
reached on June 26 following a controlled shutdown on June 25 to recover
the idle RCS locp.
The plant operated at full power until July 11, when a load reduction
was initiated due to failure of the "A" charging pump. The pump, which
failed on July 8, could not be repaired prior to expiration of the TS
limiting condition for operation. Therefore, the unit was placed in a
hot shutdown condition. On July 15, the licensee placed the unit in cold
shutdown to accomplish steam generator (SG) U-tube repairs. This 3-week
, _ . .. - . ~ , . - - - - - . . ~ . . . . _, . . - , , .
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J unplanned'-. --shutdown resulted from an evaluation of-SG eddy-current test-
M
3
data (from the 1986 refueling / maintenance outage), which indicated pre-
{} :viously unidentified SG tube cracks.in the rolled area of the U-tubes.
J ~
=.During the-mini-outage; on July 23,'a contractor SG worker was exposed
' -
.;to 110% of his-allowable quarterly radiation exposure limit. Following.
E^ , ,
.the mini-outage, full power was reached on August 6. On August 30,-
operator 31 manually tripped the plant after a failed-open feedwater regu-
.
lating valve (FRV) resulted in rapidly increasing SG levels. : Power
, , _ . operation resumed on' September 3,1986, upon completion of FRV repairs.
Full power operation continued (except for brief load reductions for. *
routine tests and maintenance) until November 30, 1986, when an automatic-
.
Ltrip occurred when a technician shorted the SG feedwater control ' system
while repairing an energized circuit. Increased RCS leakage'inside con-
Ltainment (identified during post-trip reviews) resulted-in a partial
plant cooldown-for repairs from December 1-5. The plantfoperated briefly
,
-
on December 6 until further secondary. system equipment failures (moisture -
separator. reheater baffles. dislodged) were identified. :.The plant was
~
-
shut down for repairs until' December 11. After plant startup on December .
.12, power. was- held at 16 percent because the -licensee identified a prob-- ,
lem that could affect the operability.of the ECCS. systems in the'high '
head recirculation mode following certain medium- sized-break LOCAs. On
. December 19, the plant shut down to ' perform a special flow test on the.
ECCS system to verify the adequacy of ~ the proposed problem resolution.
, Following. successful completion of the test on December 20, and NRC
approval of-the. licensee's corrective actions, the plant was restarted
on December 23 and reached full power on December 25, 1986. With the
>
exception ~ofibrief power reductions for routine testing and maintenance,
l' full- power operation continued through the end of the SALP period (March
31, 1987). ,
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- 9-
TABLE 1
INSPECTION REPORT ACTIVITIES-
REPORT / DATES INSPECTOR ' HOURS AREAS INSPECTED
213/86-03 RESIDENT 469 ROUTINE RESIDENT
2/11-4/15/86
213/86-04 SPECIALIST '35 RADWASTE TRANSPORTATION
'3/10-14/86
213/86-06 RESIDENT 138 ROUTINE RESIDENT
14/15-5/27/86
.213/86-07. SPECIALIST 20 PDCR MANAG'iENT MEETING-
3/26/86
213/86-08 SPECIALIST 38 CONTAINMENT INTEGRATED LEAK RATE TEST
3/10-14/86 (CILRT) PREPARATIONS
213/86-09 SPECIALIST 94 CILRT AND RESULTS
3/31-4/16/86
213/86-10 CANCELLED
213/86-11 SPECIALIST 120 ALARA PROGRAM APPRAISAL
4/7-11/86
213/86-12 SPECIALIST 150 1986 EMERGENCY EXERCISE
4/25-26/86~
213/86-13 SPECIALIST 22 INSERVICE TEST PROGRAM REVIEW
4/29-5/2/86
213/86-14 SPECIALIST 32 OPERATOR REQUALIFICATION
6/16-17/86
213/86-15 SPECIALIST 74 RADI0 ACTIVE EFFLUENTS
'5/19-22/86
213/86-16~ RESIDENT 156 ROUTINE RESIDENT
5/28-7/8/86
213/86-17 SPECIALIST 170 APPENDIX R IMPLEMENTATION
6/16-20/86
213/86-18 CANCELLED -
-
T1-1
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'
-
s .
REPORT / DATES INSPECTOR HOURS AREAS INSPECTED
213/86-19 SPECIALIST 38 MAINTENANCE PROGRAM
'7/7/86-11/86
213/86-20 RESIDENT 142 ROUTINE RESIDENT
7/8-8/14/86
213/86-21 CANCELLED-
213/86-22 SPECIALIST 73 FOLLOWUP DN PERSONNEL'0VEREXPOSURE
-7/22-25/86
~ 213/86-23 SPECIALIST 37 STARTUP TESTING
7/28-8/1/86=
213/86-24 RESIDENT 137 ROUTINE RESIDENT
'
8/15-9/30/86
213/86-25 SPECIALIST 100 MASONRY WALL DESIGN
10/21-24/86
213/86-26 SPECIALIST *
OPERATOR LICENSE EXAMS
11/3-7/86
213/86-27 RESIDENT 261 ROUTINE RESIDENT
10/1-11/17/86
213/86-28 SPECIALIST' 41 NON-LICENSED STAFF TRAINING
11/3-7/86
213/86-29 SPECIALIST 509 EVALUATIVE TEAM INSPECTION
11/14-21/86
213/86-30 RESIDENT 158 ROUTINE RESIDENT
11/18-12/17/86
213/87-01 SPECIALIST 44 CALIBRATION
1/6-9/87
213/87-02 RESIDENT 229 ROUTINE RESIDENT
12/18/86-2/9/87
213/87-03' SPECIALIST 37 EMERGENCY PREPAREDNESS
2/2-05/87
213/87-04- SPECIALIST 6 PHYSICAL SECURITY
1/27-28/87
T1-2
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- - . . _. .. . --
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>
>
REPORT / DATES INSPECTOR- HOURS . AREAS INSPECTED
213/87-05 SPECIALIST -35 FIRE PROTECTION FOLLOWUP
1/12-2/17/87-
- 213/87-06 : RESIDENT- 225 ROUTINE RESIDENT.
'2/10-3/18/87-
'* This report documents an operator-licensing examination. There were no-
inspection related hours.
T1-3
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TABLE 2'
- 1
INSPECTION HOUR SUMMARY
. AREA' HOURS HR/YR % OF. TIME
OPERATIONS 1270- 1172- 35.4
MAINTENANCE 426 393 11.9 '
SURVEILLANCE. 460 425 12.8.
FIRE. PROTECTION. :276 -255 7.7
LICENSING -- --
0.0
OUTAGES 240 222 6. 7
'
RAD PROTECTION 507 468 14.1
EMERGENCY PREP. 201 186 5.6
-SEC/SAFEGUAROS 188 173 5.2
. TRAINING -- --
0.0
= ASSURANCE OF QUALITY 22 20 0.6
TOTALS: 3590 3314 100.0
T2-1
- ~ .
7; _
- e.
TABLE 3
. ENFORCEMENT SUMMARY
'
SEVERITY LEVEL
AREA 1 2 3 4 5 DEV TOTAL
OPERATIONS 3 1 4
MAINTENANCE- 2 1 3
5URVEILLANCE 3 1 '4
FIRE PROTECTION 2 3 1 6
LICENSING
OUTAGES
RAD PROTECTION 1 2 1 4
EMERGENCY PREP.
SEC/ SAFEGUARDS
TRAINING
ASSURANCE OF QUALITY
TOTALS: I 12 7 I YI
INSPECTION VIOL. FUNCTIONAL
'
REPORT /DATE REQUIREMENT LEVEL AREA VIOLATION
t
213/86-03 TS 6.8 5 MAINTENANCE LICENSEE FAILED TO IMPLEMENT
2/11-4/15/86 PROCEDURES FOR STEAM GENERATOR
WET LAYUP MODIFICATION AT SYSTEM
TURNOVER IN ACCORDANCE WITH
PROCEDURE 1.2-3.1
l- 213/86-04 10 CFR 50 4* RAD-CHEM RADWASTE PROCEDURES ACCEPTANCE
l 3/10-14/86 APP B, CR5 CRITERIA DID NOT PROVIDE FOR
) VERIFICATION OF RADIONUCLIDES
L-
IN SOLID WASTE SHIPMENTS
213/86-04 10 CFR 2.311B 4* RAD-CHEM RADWASTE SHIPMENT MANIFESTS DID
3/10-14/86 NOT IDENTIFY IRON 55 AS A CON-
TAINED RADIONUCLIDE IN VARIOUS
SHIPMENTS'
213/86-04 10 CFR 20.311 4* RAD-CHEM INADEQUATE CERTIFICATION OF RAD-
03/10-14/87 C WASTE SHIPMENT DESCRIPTION
213/86-04 10 CFR 71.5 4* RAD-CHEM SHIPPING PAPERS DID NOT INCLUDE
03/10-14/87 49 CFR 172 IRON 55 CONTAINED IN MANY SHIP-
MENTS
T3-1
_ - _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ - _
___ _-
,
o
INSPECTION VIOL. FUNCTIONAL
-REPORT /DATE . REQUIREMENT LEVEL AREA VIOLATION
213/86-08. TS 4.4 4 SURVEILLANCE LOCAL LEAK RATE TESTING USING
03/10-14/87 AN UNAUTHORIZED TEST METHOD
213/86-08 10 CFR 50.59 4 SURVEILLANCE FAILURE TO DOCUMENT A SAFETY
03/10-14/86 EVALUATION FOR CHANGES TO THE
CONTAINMENT B0UNDARIES
213/86-08 TS 3.11 4 OPERATIONS . BREACH OF CONTAINMENT INTEGRITY
03/10-14/86 ON 3 OCCASIONS
213/86-15 TS 8.1.1.1 -4 RAD-CHEM LICENSEE FAILED TO PERFORM
5/19-22/86 CHEMICAL ANALYSES AT THE PRE- l
l
SCRIBED FREQUENCY.
213/86-15 TS 6.8 5 RAD-CHEM CHEMISTRY PROCEDURES DID NOT PRE-
5/19-22/86 SCRIBE ALL REQUIRED ANALYSES FOR
LIQUID RELEASES
-213/86-17 10 CFR 50 4 FIRE PROT. CONTROL ROOM HALON SUPPRESSION
6/16-20/86 APP R SYSTEM TEST DID NOT MEET COM-
MITTED ACCEPTANCE CRITERIA.
'
213/86-17- 10 CFR 50 4 FIRE PROT. ALTERNATE SHUTDOWN /COOLDOWN
6/16-20/86 APP R ANALYSIS INADEQUATE BECAUSE OF
ERROR IN STEAM GENERATOR VENT
i
CAPACITY.
213/86-17 TS 6.8.1 5 FIRE PROT. INADEQUATE BREAKER C0 ORDINATION
6/16-20/86 SETTING PROCEDURES.
l
'
213/86-17 10 CFR 50 5 FIRE PROT. INADEQUATE EMERGENCY LIGHTING
6/16-20/86 APP R FOR THE B CHARGING PUMP.
213/86-17 CYAPCo LTR DEV FIRE PROT. BREAKERS FOR CERTAIN MOTOR OPER-
6/16-20/86 9/16/86 ATED VALVES WERE NOT LOCKED OPEN
AS COMMITTED BY THE LICENSEE.
213/86-19 TS 6.8.1 4 MAINTENANCE FAILURE TO PROPERLY DOCUMENT AND
7/7-11/86 CONTROL WORK SCOPE AND MATERIAL
ISSUE.
213/86-20 TS 3.11 4 OPERATIONS MANUAL CONTAINMENT ISOLATION
7/8-8/14/86 VALVES (SI-V-863A,B,C,D) WERE
OPENED COMPROMISING CONTAINMENT
INTEGRITY AS DEFINED IN TS 3.11.
T3-2
L
- - - - _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _
1
6
INSPECTION' VIOL. FUNCTIONAL
-REPORT /DATE REQUIREMENT LEVEL AREA VIOLATION
213/86-20 TS 6.8.1 5 FIRE PROT. CONTROL ROOM FIRE EXTINGUISHERS
7/8-8/14/86 WERE NOT INSPECTED MONTHLY AS
. REQUIRED BY PLANT PROCEDURES.
213/86-22 10 CFR 20.101 3# RAD-CHEM CUMULATIVE EXPOSURE OF STEAM
7/22-25/86 GENERATOR WORK IN EXCESS OF 3
REM /QTR.
213/86-22 TS 6.11 3# RAD-CHEM FAILURE TO CONTROL WORKER EX-
7/22-25/86 POSTURE IN ACCORDANCE WITH AD-
MINISTRATIVE CONTROL PROCEDURES.
213/86-22 10 CFR 20.201 3# RAD-CHEM FAILURE TO PERFORM ADEQUATE SUR-
l
7/22-25/86 VEYS TO EXTEND WORKER STAY TIME
IN A HIGH RADIATION AREA.
213/86-27 TS 3.11 4 OPERATIONS UNAUTHORIZED OPERATION OF CON-
10/1/-11/17/86 TAINMENT ISOLATION VALVE CC-V-884.
213/86-27 10 CFR 50 4 MAINTENANCE INADEQUATE POST-MAINTENANCE
10/1-11/17/86 APP. J TESTING FOR VALVE CC-CV-885.
213/86-27 . TS 6.8 5 OPERATIONS FAILURE TO FOLLOW SURVEILLANCE
10/1-11/17/86 PROCEDURE SUR-5.1.4 SUCH THAT
VALVE SI-V-865 WAS NOT RECLOSED
UPON TEST COMPLETION.
213/86-27 TS 6.8 5 SURVEILLANCE FAILURE TO FOLLOW SUR 5.7-19 SUCH-
10/1-11/17/86 THAT PRESSURE GAUGES WERE NOT
CALIBRATED AND DATA SHEETS NOT
ATTACHED FOR TEST.
213/87-01 10 CFR 50 4 SURVEILLANCE INADEQUATE ADMINISTRATIVE CON-
1/6-9/87 APP B TROLS OVER SUBSTITUTION / CHANGE
OF TEST METHODS AND EQUIPMENT.
- These four violations were cited as an aggregate Severity Level 4 violation.
- Three violations were cited as an aggregate Severity Level 3 violation.
T3-3
,. --. . . , . - . _ - .
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1
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y :5
TABLE 4
. LISTING OF LERs BY FUNCTIONAL AREA
- AREA. A B' - C, D E -X . TOTAL
-
OPERATIONS. .5 4- 9 2 20
-MAINTENANCE: 3 l' 4
SURVEILLANCE 3 2 4 9
FIRE PROTECTION 4 1 2 7
.--
LICENSING
-0UTAGES +
- RAD PROTECTION- 2 2 :
EMERGENCY-PREP.
SEC/ SAFEGUARDS
TRAINING-
- ASSURANCE OF QUALITY ~ - -
TOTALSi 17 7 - 15 3 42
Cause Codes *:
A - Personnel Error
B - Design, Nanufacturing,: Construction, or Installation Error
C - External Cause '
D - Defective: Procedures
E - Component failure
X - Other
- Cause Codes in this table a're based on inspector evaluations and may differ from
those specified~in the LER.
t
T4-1
r-
-C.
LER EVENT CAUSE
NUMBER DATE CODE DESCRIPTION
86-10 2/26/86 A UNAUTHORIZED ENTRANCE TO A HIGH RADIATION AREA.
86-11 2/28/86 A .IN0PERABLE FIRE SPRINKLER SYSTEM.
86-12 2/26/86 X DROPPED FUEL ELEMENT EVENT.
86-13 3/25/86 B INADEQUATE SMALL BREAK LOSS OF COOLANT ACCIDENT
ANALYSIS.
66-14~ 3/8/86 E IN0PERABLE FIRE WATER PUMPS
86-15 3/19/86 E CONTROL R00 WEAR AND CRACKING.
86-16 3/12/86 E INOPERABLE C02 FIRE PROTECTION SYSTEM.
86-17 3/17/86' A DEGRADED FIRE BARRIER SEALS.
86-18 4/9/86 B REACTOR COOLANT SYSTEM WIDE RANGE PRESSURE
UNCERTAINTY
86-19 5/6/86 A DEFECTIVE STEAM GENERATOR TUBE NOT PLUGGED
86-20 5/8/86 A REACTOR TRIP DURING TURBINE VIBRATION TESTING
86-21 5/24/86 E NUCLEAR INSTRUMENTATION DROPPED R00 SETPOINT
DRIFT
86-22 5/17/86 A TURBINE RUNBACK DURING NUCLEAR INSTRUMENTATION
ADJUSTMENT
86-23 5/18/86 E SPURIOUS LOAD RUNBACK ACTUATIONS
86-24 6/9/86 E SPURIOUS TURBINE LOAD RUNBACK
86-25 6/19/86 A REACTOR TRIP DURING NUCLEAR INSTRUMENTATION
TROUBLESHOOTING
86-26 6/22/86 X LOW FLOW TRIP CAUSED BY FOREIGN MATERIAL IN
86-27 6/4/86 E REACTOR TRIPS DUE TO LOSS OF HEATER DRAIN PUMP
FLOW
86-28 6/10/86 A INOPERABLE SWITCHGEAR HALON FIRE PROTECTION
SYSTEM
T4-2
1,
C-
LER EVENT CAUSE
NUMBER CATE CODE DESCRIPTION
86-29 6/17/86 B MhIN STEAM ISOLATION VALVE STROKE TEST FAILURE
86-30 6/27/86 E SPURIOUS LOAD RUNBACK
86-31 7/3/86 A HIGH PRESSURE SAFETY INJECTION RECIRCULATION
VALVE FAILURE
86-32 7/10/86 E DROPPED R0D PROTECTION SETPOINT DRIFT
86-33 7/15/86 B LOW TEMPERATURE OVERPRESSURIZATION PROTECTION
ISOLATION VALVE INTERLOCK FAILURE
86-34 7/11/86 B INADEQUATE 3-LOOP REACTOR COOLANT SYSTEM FLOW-
RATE
86-35 7/15/86 E
MULTIPLE STEAM GENERATOR TU8E DEFECTS (MINI-
OUTAGE)
86-36 7/8/86 X CHARGING PUMP SHAFT FAILURE
86-37 7/8/86 A MISOPERATION OF CONTAINMENT ISOLATION VALVES
86-38 8/6/86 A INADEQUATE RETEST OF NUCLEAR INSTRUMENTATION
86-39 7/23/86 A RADIATION WORKER OVEREXPOSURE
86-40 7/24/86 A LOW TEMPERATURE OVERPRESSURIZATION PROTECTION
RENDERED IN0PERABLE FOR TESTING
86-41 8/30/86 E REACTOR TRIP CAUSED BY FAILED FEED REGULATING
VALVE
86-42 10/16/86 A OPERATOR SURVEILLANCE MISSED IN0PERABLE LOCK
ON SAFETY INJECTION VALVE
86-43 11/8/86 E WIDE RANGE STACK MONITOR FAILED LOW
86-44 11/30/86 A AUTOMATIC REACTOR TRIP DUE TO PERSONNEL ERROR
DURING MAINTENANCE
86-45 11/14/86 B TURBINE SPRINKLER SYSTEM DISABLED TO PERFORM
APPENDIX R MODIFICATIONS
86-46 12/4/86 E LOW TEMPERATURE OVERPRESSURIZATION PROTECTION
RELIEF VALVE FAILED TO RESEAT FOLLOWING IN-
ADVERTENT RCS PRESSURE SPIKE
T4-3
_
._ _ _ _ - _ _ _ _ _ . _ _ _ . _ _ _ _ _ _ _ _ - _ _ - _ _ - _ _ - _ _ _ _ _ _ - _ _ _ - - _
__
.>'
- LER- EVENT CAUSE
NUMBER DATE CODE DESCRIPTION
86-47 12/8/87 A PLANT IN UNANALYZED CONDITION DUE TO ADMINI-
STRATIVE ERROR
86-48 12/12/86 B EMERGENCY CORE COOLING SYSTEM PUMP RUN0VT
DURING CORE DELUGE LINE BREAK
87-01 1/17/87 .A FIRE WATCH DEPARTED WITHOUT RELIEF.
87-02 .03/02/87 E SPURIOUS TURBINE LOAD RUNBACK.
87-03 03/06/87 _ E FAILURE OF THE WIDE RANGE STACK MONITOR.
L
!
T4-4
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,
.
. .
-a.
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TABLE 5
SALP HISTORY
Ratings for Period Ending
AREA ~ 6/81 8/82 8/83 2/85 -2/86
-~0perations 'l 1 1- 1 1
Radiological Controls 'l 1 1 2 2
Maintenance .1 1 1 1 2
. Surveillance 2 1 2 2 2
Fire Protection 1 .1 *
2 **
Emergency Preparedness 2 1 1 2 2
Security and Safeguards- 1 1 1 1 1
Refueling and Outage .1 1 1 1 2
Management
Assurance of Quality : 'l ** ** '2 2
Licensing ** 1 1 1. 2
Training & Qualifications 2 ** ** ** 2-
'Effecitvenss
.
- No basis available for a rating.
- Not rated as a separate area.
T5-1
t- i