ML20203P227
| ML20203P227 | |
| Person / Time | |
|---|---|
| Site: | Farley  |
| Issue date: | 07/31/1986 |
| From: | NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20203P223 | List: |
| References | |
| 50-348-86-14, 50-364-86-14, NUDOCS 8610270255 | |
| Download: ML20203P227 (33) | |
See also: IR 05000348/1986014
Text
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ENCLOSURE
SALP BOARD REPORT
U.S. NUCLEAR REGULATORY COMMISSION
REGION II
SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE
INSPECTION _
REPORT NUMBERS
50-348/86-14, 50-364/86-14
Alabama Power Company
Joseph M. Farley Units 1 and 2
January 1, 1985 through July 31, 1986
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8610270255 861022
PDR ADOCK 05000348
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I. INTRODUCTION
The Systematic Assessment of Licensee Performance (SALP) program is an'
integrated NRC staff effort to collect available observations and data on a
periodic basis and to evaluate licensee performance based upon this
information. SALP is supplemental to normal regulatory processes used to
ensure compliance with NRC rules and regulations. SALP is intended to be
sufficiently diagnostic to provide a rational basis for allocating NRC
resources and to provide meaningful guidance to the licensee's management to
promote the quality and safety of plant construction and operation.
An NRC SALP Board, composed of the staff members listed below, met on
October 2,1986, to review the collection of performance observations and
data to assess the licensee performance in accordance with the guidance in
NRC Manual- Chapter 0516, " Systematic Assessment of Licensee Performance." A
summary of the guidance and evaluation criteria is provided in Section II of ~
this report.
This report is the SALP Board's assessment of the licensee's safety
performance for the J. M. Farley facility for the period January 1,1985
through July 31, 1986.
SALP Board for the J. M. Farley facility:
L. A. Reyes, Deputy Director, Division of Reactor Projects (DRP), Region II
(RII) (Chairman)
A. F. Gibson, Director, Division of Reactor Safety, RII
J. P. Stohr, Director, Division of Radiation Safety and Safeguards, RII
D. M. Verrelli, Chief, Projects Branch 2, DRP, RII
L. S. Rubenstein, Director, Directorate 2, PWR-A Division, NRR
W. H. Bradford, Senior Resident Inspector, Farley, DRP, RII
E. A. Reeves, Project Manager, Directorate 2, PWR-A Division, NRR
Attendees at SALP Board Meeting:
K. D. Landis, Chief, Technical Support Staff (TSS), DRP, RII
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H. C. Dance, Chief, Project Section 2B, DRP, RII
L. P. Modenos, Project Engineer, Project Section 28, DRP, RII
B. R. Bonser, Resident Inspector, Farley, DRP, RII
II. CRITERIA
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Licensee performance is assessed in certain functional areas depending upon
whether _ the facility has been in the construction, preoperational, or
operating phase. Each functional area normally represents an area which is
significant to nuclear safety and the environment, and which is a normal
- programmatic area. Some functional areas may not be assessed because of
little or no licensee activities or lack of meaningful observations.
Special areas may be added to highlight significant observations.
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One or more of the following evaluation criteria were used to assess each
functional area; however, the SALP Board is not limited to these criteria
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and others may have been used where appropriate.
A. Management involvement in assuring quality
B. Approach to the resolution of technical issues from a safety standpoint !
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C. Responsiveness to NRC initiatives
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D. Enforcement history
E. Operational and construction events (including response to, analysis
of, and corrective actions for)
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F. Staffing (including management)
l G. Training and qualification effectiveness
! Based upon the SALP Board assessment, each functional area evaluated is
classified into one of the three performance categories. The definitions of
- these performance categories ar'e
} Category 1: Reduced NRC attention may be appropriate. Licensee
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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 or
l construction is being achiev'ed.
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! Category 2: NRC attention should be maintained at normal levels.
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Licensee management attention and involvement are evident and are
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concerned with nuclear safety; licensee resources are adequate and are
reasonably effective so that satisfactory performance with resoect to ,
j- operational safety or construction is being achieved. I
Category 3: Both NRC and licensee attention should be increased.
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Licensee management attention or involvement is acceptable and
i considers nuclear safety, but weaknesses are evident; licensee
resources appear to be strained or not effectively used so that
minimally satisfactory performance with respect to operational safety
or construction is being achieved.
The functional area being evaluated may have some attributes that would
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place the evaluation in Category 1, and others that would place it in either
Category 2 or 3. The final rating for each functional area is a composite ,
of the attributes tempered with the judgement of NRC management as to the
j significance of individual items.
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The SALP Board may also include an appraisal of the performance trend of a
functional area. This performance trend will only be used when both a
definite trend of performance within the evaluation period is discernable
and the Board believes that continuation of the trend may result in a change
of performance level. The trend, if used, is defined as:
Improving: Licensee performance was determined to be improving near the
close of the assessment period.
Declining: Licensee performance was determined to be declining near the
close of the assessment period.
No trends were noted by the Board for this period.
III. SUMMARY OF RESULTS
A. Overall Facility Performance
The Farley facility is well managed by qualified and experienced
personnel. Senior plant managers hold active senior reactor operator
licenses and the site is supported by a corporate organization that is
composed of personnel who have extensive backgrounds in nuclear plant
management and operations. The licensee remains responsive to NRC
concerns and the organization is safety criented. Strengths were
identified in the areas of plant operations, surveillance, radiological
controls, maintenance, fire protection, outages, and licensing
activities.
The Farley Nuclear Plant was effectively managed and continues to
achieve a satisfactory level of operational safety. The licensee has
strong programs in all aspects of plant operation. However, the
weakness noted in the last SALP evaluation of procedure adherence is an
area requiring continuing management attention. The licensee has
initiated corrective action which appears to be effective. This is
evidenced by a decrease in procedure violations in the surveillance
area. However, violation of failure to follow procedure led to a Level
III violation that rendered a train of the Residual Heat Removal System
incapable of fulfilling its design function. Even though this
condition was indicated on the main control board, it was not detected
for 96 hours0.00111 days <br />0.0267 hours <br />1.587302e-4 weeks <br />3.6528e-5 months <br />.
During the SALP pariod the Farley plant had high availability, fewer ,
than average number of reactor trips, few inadvertent ESF actuations,
efficient operational and hardware response to the events that have
occurred, prompt and thorough reporting of events when required, and
low occupational radiation exposures. The licensee recognized the
potential plant-specific and generic consequences of the tendon failure
problem and acted responsibly in reporting and resolving the event.
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The licensee has set high standards for cleanliness, radioactive waste.
control, general plant operations, and maintenance. The licensee is
dedicated to long run time and short refueling outages. Unit l's
longest run time was 321 days. This occurred during Cycle 6 from
April 25, 1984 to March 13, 1985. Nucleonics Week 1985 ranking of
commercial reactors b'ased on cumulative capacity factor ranked Farley
Unit 2 as #1 in the nation and #15 in world ranking.
B The performance categories fnr the current and previous SALP period in
each functional area are as follows:
August 1, 1983 - January 1, 1985-
Functional Area December 31, 1984 July 31, 1986
Plant Operations 1 1
Radiological Controls 1 1
Maintenance 1 1
Surveillance 2 1
Fire Protection 1 1
Security _
2 2
Outages-(includes refueling) 1 1
Quality Programs and 2 2
Administrative Controls
Affecting Quality
Licensing Activities 1 1
Training 1 2
IV. Performance Analysis
A. Plant Operations
1. Analysis
During this assessment period, inspections were performed by the
resident and regional inspection staffs. The licensee had a positive
-nuclear safety attitude and exhibited no significant administrative,
management control or material problems. The licensee's supervisory
staff was knowledgeable and proficient in day-to-day plant operations.
Major operational decisions were made at a management level adequate to
assure appropriate supervisory involvement. Plant operations were
generally conducted in a conservative manner to ensure plant safety.
Overall control of plant operations was satisfactory and was well
planned with established and realistic priorities. The licensee was
quick to take corrective action when problems or violations were
identified by NRC. The licensee has demonstrated responsiveness for
items identified by the internal audit group. Corrective actions in
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these areas were prompt. The licensee has demonstrated a thorough
knowledge of regulations, guides, standards and generic issues and
interpretations of these documents and associated issues were
conservative.
Licensee technical competence was well founded both in technical
matters and general plant operations. The plant staff responded to
plant trips and other operational events during this review period in a
professional and competent manner. Daily conduct of business in the
control room was performed in a professional manner. Access to the
control room was controlled and limited to personnel conducting
business. Radios and reading material not directly related to plant
operation are not allowed in the control room or plant. Housekeeping
throughout the plant was well maintained.
The licensee was well prepared at meetings with NRC. The licensee's
staff was able to make immediate commitments or state the utility's
position in a given area, especially in the enforcement conference held
at the Region on June 3, 1986.
The qualifications of ~ plant management exceeded NRC requirements.
Most senior plant managers hold senior reactor operator licenses.
Plant management was oriented towards safety and efficiency. This was
demonstrated by the close supervision of plant operations. The plant
was well managed with conscientious and capable personnel.
Licensee onsite evaluations were routinely performed to address, assess
and correct reportable events. An evaluation of the content and
quality of a representative sample of Licensee Event Reports (LERs) was
performed by the NRC using a refinement of the basic methodology
presented in NUREG/CR-4178. The results indicate that Farley has an
overall average LER score of 7.8 out of possible 10 points, compared to
a current industry average of 7.9. The principle weakness identified
in the LERs, in terms of safety significance, involve the requirements
to provide a safety assessment and to adequately identify failed
components in the text. A strong point for the Farley LERs is that the
requirement to provide the failure mode, mechanism, and effect of each
failed component was satisfied for all applicable LERs.
Seven reactor trips from power operation occurred on Unit 1 during the
assessment period. Six trips were caused by equipment. failures and one
by personnel error when a technician accidentally bumped a cable on a
main feed pump which broke a wire and caused the pump to trip. Unit 1
trip rate was about 0.72 trips per 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> of operation compared to
a national average of 1.47. Unit 2 had eight trips from power
operation and one trip during start-up at low power level . Five trips
were caused by equipment failure, one from personnel error, one by
lightning, and one at low reactor power level from low
electro-hydraulic fluid pressure at the steam generator feed pump.
Unit 2 trip rate was about 0.88 trips per 1000 hours0.0116 days <br />0.278 hours <br />0.00165 weeks <br />3.805e-4 months <br /> of operation.
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Reactor trips are described in Section V.J.
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Seven violations were identi fied. The plant staff is normally
observant of Limiting Conditions for Operation (LCOs) and was generally
conservative in its application of action statement requirements.
However, violations (a) and (b), listed below involved failure to
. follow procedures that resulted in improper system alignment.
Violations (c), (d), (e), (f) and (g) involved procedural inadequacies
and operator failure to comply with procedures. Procedural violations
indicates a lack of strict adherence in following procedures.
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The licensee has initiated strong corrective action to instill in all
personnel that plant procedures must be rigidly followed. The
elimination of personnel errors has beccme a pointed objective of
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supervision. This is being accomplished by corrective action directed
toward the employee as well as publication of personnel errors
committed by the various work groups in the plant. The personnel
errors are displayed on closed circuit TV monitors located throughout
- the plant and are tabulated and credited to the appropriate work
l groups. This process, though in the. early stages, has already shown
positive results.
a. Severity Level III violation without civil penalty for violating
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regulatory requirements in that procedures and technical
specifications (TS) were not adhered to which caused ECCS
subsystem "B" train of RHR to be incapable of transferring pump
suction to the containment sump during the recirculation phase of
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operation (348/86-10).
b. Severity Level IV violation for not placing an inoperable Unit 2
power range channel in the tripped condition within one hour as
required by TS (364/85-11).
c. Severity Level IV violation for failure to update control room
reference drawings to conform to as built status, failure to
! adhere to requirements of a p'rocedure, and inadequate procedure
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d. Severity Level IV violation for failure to have a continuous fire
watch posted when a fire door was blocked open by a rubber hose
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e. Severity Level V violation for failure to adhere to the
requirements of a procedure (348, 364/85-11).
i f. Severity Level V violation for failure to have an adequate
! procedure to set the flow rate for the control room chlorine '
detector (348/84-32).
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g. Severity Level V violation when a fire damper penetration was not
functional due to a telephone cord blocking the closure of the
damper (364/86-11).
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2. Conclusion
Category: 1
3. Board Recommendations
No changes in the NRC's reduced inspection resources are recommended.
B. Radiological Controls
1. Analysis
During this assessment period, inspections were performed by resident
and regional inspection staffs. This included confirmatory
measurements using Region II mobile laboratory.
The licensee's health physics and chemistry staffing levels were
appropriate and compared well to other utilities having a facility of
similar size. An adequate number of ANSI qualified licensee and
contract health physics technicians and of qualified chemistry
technicians were available to support routine and outage operations.
Key positions in the radwaste management program and environmental
surveillance programs were filled with qualified staff.
Two strengths of the health physics program were the quality of the
health physics technicians and the experience level of the site health
physics staff. The staff has a low turnover rate and an effective
training program.
.The performance of the health physics and chemistry staff in support of
routine operations and outages was good. No substantive issues were
identified in this area.
Management support and involvement in matters related to radiation
protection, radwaste control and chemistry was adequate. Health
physics management was involved sufficiently early in outage
preparations to permit adequate planning. The station health physicist
and plant chemist received the support of other plant managers in
implementing the radiation protection and chemical control programs.
Resolution of technical issues by the health physics and chemistry
staff was generally adequate and responses to NRC initiatives were
conducted in an effective and acceptable manner.
Audits performed by the corporate staff of the health physics,
radwaste, environmental and chemistry programs were of sufficient scope.
and depth to identify problems and adverse trends. Appropriate
corrective actions were taken and documented. Audits performed by the
site audit organization are discussed in Section IV.I.
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- The licensee's radiation work permit and respiratory protection
programs were found to be satisfactory. Control of contamination and
radioactive materials within the facility was generally adequate. From
January 1985 to July 1986, the amount of contaminated area decreased
from approximately 24,398 to 23,626 square feet which represents 20%
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percent of the radiologically controlled area of the plant. In 1985,
! there was a 33 percent decrease in the number of clothing and skin
contamination incidents when compared to 1984.
During 1985, the licensee's cumulative exposure was 400 man-rem per
unit. This compares favorable to the national average exposure of 425
man-rem per unit observed at similar PWR facilities. This lower than
average collective dose results from the aggressive exposure control
program established and implemented by the licensee.
During 1985, the licensee disposed of 8,730 cubic feet of solid
i radioactive waste per unit containing 410 curies. This is less than
the national average of 11,650 cubic feet per unit shipped by other
utilities with similar facilities. This low amount is due primarily to
a dedicated solid waste reduction program. A covered radioactive waste
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transfer facility is under construction to upgrade the outside bulk
l loading area and reduce exposure of waste handling personnel.
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In the area of radiological confirmatory measurements, the licensee
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participated in the NRC's spiked sample program. .The licensee has had
consistent problems with the FE-55 analysis for the past three years,
indicative of a weakness in the radiological measurements program;
however, the 1986 spiked sample results showed agreement with known
concentration of Fe-55, but only after reanalysis by the licensee.
The licensee submitted the required radiological effluent and
environmental reports during the evaluation period. Radioactive
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gaseous effluents for 1985, for Units 1 and 2 combined, were 2,368
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curies of noble gases, 6.0E-4 curies of halogens (including I-131),
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5.0E-5 curies of mixed fission product and irradiation product
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particulates, and 470 curies of tritium. Alpha-emitting particulates
! aerosols were-not detected. The 1985 Region II averages for a two unit
site (based on 21 operating PWRs) were 10,360 curies of noble gases,
0.12 curies of halogens and 190 curies of tritium. Radioactive liquid
i effluents for the two unit site totalled 0.071 curies of mixed fission
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and irradiation products, 1,105 curies of tritium and 2.0E-4 curies of
long-lived alpha emitters released in 2.1 E7 gallons of liquia plant
effluents. The 1985 Region II averages for a two unit site were 2.7
. curies of mixed fission products, 840 curies of tritium and 2.2E-4
curies of alpha emitters. The licensee's releases were less than the
average annual releases reported by 21 Region II plants of similar size
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and type for 1985, with the exception of tritium.
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Both liquid and gaseous effluents were within regulatory limits for
quantities of radioactive material released and for dose to the
maximally exposed individual. For 1985 releases, the maximum
, calculated total body dose to a member of the public was 0.03 mrem from
liquid releases and 0.13 mrem from gaseous effluents. These calculated
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doses represented 0.12 percent and 0.52 percent of-the 40 CFR 190 limit
of 25 mrem / year. There were two unplanned gaseous releases and one
unplanned liquid release during the evaluation period. The liquid
release was the result of leakage from the Component Cooling Water
System into the Service Water System. The gaseous releases were caused
by inadvertent venting of the Hydrogen Recombiner System into the
Auxiliary Building. The design that vented the RHR Sump Vent into the
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Component Cooling Water Heat Exchanger Room was corrected. The total
activity for unplanned releases was 0.006 curies for liquid and 11.5
- curies for gas. Unit 2 had no unplanned releases during this
assessment period.
In the area of plant ' chemistry the steam generators had, in prior years
of operation, accumulated significant amounts of iron-copper oxide
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< sludge as well as potentially corrosive species (e.g, chloride,
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sulfate) that were present as " hideout return." Consequently, several
days were required during startup after each lengthy outage to achieve
the desired level of chemi stry control . During the last two fuel
cycles of each unit the licensee had achieved stable plant operation
and a high level of chemistry control while making progress in removing
- both sludge and reducing the effects of hideout from the steam
generators. In an effort to eliminate the detrimental effect of copper
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as a corroding element, the licensee had replaced all copper heat
exchanger tubes in the condensate /feedwater train. In addition,
inleakage of air condenser cooling water through the condenser had been
effectively eliminated. All elements of the chemistry program had been
upgraded to implement the recommendations of the Steam Generator Owners
Group.
! Two violations were identified for failure to assure that radioactive
material shipped for burial was without free standing liquid.
a. Severity Level IV violation for failure to assure that radioactive
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material shipments for burial were without free standing liquids
(348,364/85-34).
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b. Severity Level IV violation for failure to have adequate
l procedures ' to preclude shipping radioactive material for burial
- with free standing liquids (348, 364/85-34).
2. Conclusion
Category 1
- 3. Board Recommendations
No change in the NRC's reduced inspection resources are recommended.
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C. Maintenance
1. Analysis
During this assessment period, inspections were conducted by the
resident and regional inspection staffs. The maintenance program
continued to be strong. Management involvement in maintenance planning
and practices were evident. First line supervisors and maintenance
personnel indicated a high awareness for procedural adherence. This is
indicative of the positive nuclear safety attitude in the preventive
and corrective maintenance programs. The licensee's approach to the
resolution of technical issues continues to be sound.
The licensee's maintenance program was well controlled by specific
procedures. The personnel participating in activities affecting
equipment on the Q-list were aware of the quality assurance (QA)
controls. The craft personnel performing maintenance and surveillances
were knowledgeable of maintenance procedures and plant equipment.
Maintenance Work Request (MWR) packages had the required reviews and
approvals prior to the start of the work. The MWR indicates the proper
Q-list classification, work was completed and inspected as required,
and post-maintenance testing was conducted.
Use of the Nuclear Plant Reliability Data System (NPRDS) has increased
the licensee's awareness of potential plant problems. Upgrades in the
Computer Historical and Maintenance Program System (CHAMPS) and
implementation of data verification has improved the data base used for
maintenance planning and scheduling. Staffing increases added
maintenance planners who provided better scheduling and coordination of
the activities of each maintenance discipline.
During a regional maintenance inspection, instances of breakdowns in
the corrective maintenance process occurred. One event included the
licensee's failure to properly conduct corrective maintenance
activities involving wiring errors associated with the feedwater flow
control valves. A second particular concern was the licensee's
inadequate processing of the maintenance work request on a failed
electrical penetration which had caused a forced outage. These
instances indicate that under some circumstances, a less than
meticulous attention to detail has been directed towards corrective
maintenance activities.
A special inspection conducted to evaluate the licensee's actions in
response to Generic Letter 83-28 revealed that maintenance activity and
post-maintenance testing were adequate to ensure reactor trip system
reliability.
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Once identified, equipment and-components received adequate preventive
maintenance. However, inadequacies were identified in the scope of the
preventive maintenance program .which included the failure to incor-
porate compressed air system pressure switches and inverter circuit
breakers.
The licensee utilizes a predictive maintenance analysis which includes
oil and vibration analysis on mechanical equipment and infrared
analysis on electrical equipment. These techniques have enabled the
licensee to predict degrading trends in equipment performance and
effect repairs before equipment failure occurs.
The licensee was responsive to NRC concerns and conducted evaluations
to identify and correct, if required, activities related to maintenance
which appeared to be contrary to the prescribed function of equipment.
This is exemplified by the licensee's action in the investigation and
repair of Unit 2 containment building post-tensioning system and
modification to the Atwood Morrill Company main steam isolation valves
on Unit ? in which the valve shafts were modified. Unit I will be
modified during the next refueling outage.
Six. violations were identified. Five of these resulted from not
following existing procedures and drawings as noted below. One
involved independent in process inspections.
a. Severity Level IV violation for not performing independent
in process inspection for Class 1 and 2 pipe welds (348/85-33).
b. Severity Level IV. violation for failure to wire feedwater control
valve in accordance with the work request and complete maintenance
prescribed procedures or drawings (343/85-40).
c. Severity Level V violation for failure to follow prescribed
procedures for Class 2 pipe support spring hangers in that the
support settings were not recorded and verified (348/85-33).
d. Severity Level V violation for failure to install spacers between
the cells of the service water batteries and the uninterruptible
power supply batteries as required by sendor drawings (348,
364/85-20).
e. Severity Level V violation for having combustible liquids
unattended in the hot machine shop (348/85-24).
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f. Severity Level V violation for having combustible liquids
unattended in the hot machine shop (348, 364/86-11).
2. Conclusion
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3. Board Reccmmendations
Due to the findings identified during. a team inspection, the Board
recommends that the routine inspection program be conducted in the
maintenance area.
D. Surveillance
1. Analysis
During the assessment period, inspections were performed by the
resident and regional inspection staffs. These included activities
related to inservice inspection and testing: surveillance, containment
building tendons, and containment intergrated leak rate testing.
Routine plant surveillance related activities appeared to be planned
and well defined. The licensee has continuously upgraded the
surveillance program. Review of surveillance activities was performed
by prescribed licensee reviewers who were qualified to perform these
activities. Review of surveillance records revealed that they were
readily available, complete, and adequately maintained. Onsite
evaluations were routinely performed to address, assess and correct
surveillance concerns. The licensee's nnsite corporate QA organization
was heavily involved in the surveillance program.
Licensee response to NRC initiatives was timely and there were few
long-standing regulato ry issues attributable to the licensee. ,
Understanding of technical issues was apparent with timely resolution. '
Viable, sound and thorough responses were offered.
Licensee management involvement in Inservice Inspection and Inservice
Testing activities was adequate. Decision-making was usually at a
level that assured adequate review. Corporate management was involved
in site activities, and reviews were timely, thorough and technically
sound. Records were complete, well maintained, and readily available.
The surveillance procedures reviewed, tests that were witnessed, and
examinations of test results, revealed that the licensee's surveillance
procedures were technically adequate and satisfactorily executed.
Inspection of the snubber surveillance - program-identified a problem
with the visual inspection procedure. The licensee agreed to . revise
their procedure to eliminate any confusion with the TS requirements.
The snubber surveillance records were complete, well maintained,
legible, and retrievable.
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- The four violations identified are not considered to indicate
significant programmatic deficiencies. Violations (a) and (d) involved
procedure violations. Violations (b) and (c) concerned control of
measuring and test equipment.
a. Severity Level IV violation for not conducting a review of a
completed surveillance test procedure within the time frame
specified by an administrative procedure (364/85-44).
b. Severity Level V violation for changing the work sequence from
that which was specified on the Maintenance Work Request without
appropriate review and approval (348/85-17).
c. Severity Level V violation for not establishing suitable
environmental conditions for calibration of measuring and test
equipment (348,364/85-25).
d. Severity Level V violation for not establishing adequate measures
to assure that measuring and testing devices are calibrated with
sufficient frequency to assure accuracy (348, 364/85-25).
2. Conclusion
Category 1
3. Board Recommendations
No change in the NRC's reduced inspection resources are recommended.
E. Fire Protection
1. Analysis
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During this assessment period, inspections were performed by the
resident and regional inspection staffs. A special team inspection was
conducted of the licensee's fire protection / prevention program
reevaluation of 1985 with respect to compliance with 10 CFR - 50
Appendix R, Sections III.G., III.L., and III.O.
On November 19, 1985, the Commission granted exemptions to 10 CFR 50 -
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Appendix R at the licensee's request for 33 of 49 specific fire areas
of Unit 2 and areas shared with Unit 1. The remaining requests for
exemptions for 16 fire areas are under Commission review. Additional
justification have been provided by the license. In addition,
exemptions for 27 specific fire areas of Unit 1 were requested by the
licensee. The exemption resulted from the licensee's fire program
reevaluation noted above. As of July 31, 1986, Commission action
remains open for these 27 fire areas on Unit 1.
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.
The licensee has issued the appropriate alternative shutdown procedure
for a fire in the cable spreading room and control room. This
procedure was found to meet Appendix R,Section III.L. requirements.
With respect to Section III.G., Fire Protection of Safe Shutdown
Capabilities, the fixed fire protection extinguishing systems,
fire / smoke detection systems, one hour raceway fire barrier enclosures,
and fire area three hour fire barrier boundaries were found to be in
Service. In addition, these permanent plant fire protection features
were found to be adequate with respect to maintaining one train of
systems necessary to achieve and maintain hot standby free from fire
damage. The inspectors also found the reactor coolant pump oil
collection system design to meet the seismic and oil collection
requirements of Section III.O.
The licensee identified, analyzed and reported fire prevention events
and discrepancies as required by license condition or technical
specifications. These reports were reviewed and found to be
satisfactory.
In general, the management involvement and control in assuring quality
in the fire protection program is evident as demonstrated by the
completeness of the engineering analysis associated with the inple-
mentation of the Appendix R requirements. The licensee's apprcach to
resolution of technical fire protection issues indicates a clear
understanding of the issues. The responsiveness to NRC initiatives are
technically sound and thorough in almost all cases. Licensee
identified fire protection related events or discrepancies are properly
analyzed, promptly reported and effective corrective actions taken.
The previous SALP report refers to a well qualified staff and high
quality training program.
No violations of deviations were identified.
2. Conclusion
Category 1
3. Board Recommendations:
No changes in the NRC's reduced inspection resources are recommended.
1. Analysis
During the assessment period, inspections were performed by resident
and regional inspection staffs. These included observation of a small
scale emergency preparedness exercise in September 1985.
.
-
.
. ..
.
15
Inspections disclosed that the licensee had adequate emergency
preparedness organization and staffing at the plant and corporate
level. Corporate management appeared committed to an effective
emergency response program. Senior corporate officials were directly
involved in the annual emergency exercise, drills, and followup
critiques.
The following essential elements of emergency response were determined
to be acceptable: emergency detection and classification; protective
action decision making, except as discussed below; shift staffing and
augmentation; training, except as indicated below; dose calculation and
assessment; public information; annual quality assurance audits of
plant and corporate emergency preparedness programs; changes to
emergency plan and implementing procedures; coordination of offsite
agencies; identification of weaknesses during drills and exercises.
The exercise demonstrated that the emergency plan and respective
procedures could be implemented, although one violation involving
notification of emergencies within 15 minutes was noted.
Three of the four weaknesses identified were identified by the
licensee. The licensee failed to follow the format agreed upon between
the states and the licensee in making initial offsite notification.
During the exercisc, no protective actions were taken onsite during and
following the simulated plume passage. Onsite communications needed
improvement in that the Recovery Manager was not informed of the
Emergency Director's reclassification of the Site Area emergency to an
Alert until 16 minutes after the reclassification had been announced to
offsite officials. Also, one plant procedure did not clearly define
the use of plant personnel in the dose assessment group. The licensee
committed to resolve the above exercise weaknesses.
The violation noted is not indicative of a programmatic breakdown.
Severity Level IV violation for failure to provide the capability
and ' procedures for notification of offsite State and local
agencies within 15 minutes following emergency declarations (348,
364/85-37).
2. Conclusion
Category 2
3. Board Recommendations:
No change in the NRC's reduced inspection resources are recommended.
.
e
._ . . - . . - - . - - . . . _ _ _ - - . - _ . _- . . . _ _ - - .. = - - _ . . .-
,
. ..
.
!.
16
. G. Security and Safeguards
1. Analysis
During the assessment period, inspections were conducted by the
i resident and regional inspection staffs. A Regulatory Effectiveness ;
Review (RER) was also conducted. Although the RER report was not ]
issued during the rating period, the licensee has taken actions to
compensate for a potential safeguards vulnerability identified during l
'
the RER. Further, the licensee has established a program for upgrade
,
and/o'r replacement of physical security hardware identified as
inadequate in the RER. Additionally, the licensee is using members of
the security force as compensatory measures for RER identified
safeguards inadequacies and concerns until completion of hardware
j: upgrade / replacement. Procedures have been placed in effect by the
j licensee to correct other procedural problems identified by the RER.
'
The licensee is working with the Division of Safeguards, NMSS, in order
to resolve other issues arising from the RER.
Authority and responsibilities associated with the security organi-
.! zation were cleariy delineated and appeared to be effective. The site
>
organization is adequately staffed and appropriately trained and
j- equipped. The facility guard Training and Qualification Plan is
i implemented on a continuing basis at all levels of the security
i
organization using the onsite training staff supplemented by corporate
specialists.
Changes to the licensee's Physical Security Plan were submitted on a
timely basis under the provisions of 10 CFR 50.54(p).
- The licensee's independent security program audit covers all aspects of
! the site security program and the program auditors seem well acquainted
with the prsgram.
t
.! Three violations were identified. The violations appear to have been
caused by a lack of effective program oversight rather than guard force
inadequacy,
a. Severity Level IV violation for failure to protect vital. equipment
with two physical barriers (348, 364/85-08).
b. Severity Level IV. violation for several protected area perimeter
- inadequacies
- several protected area gates were not protected by
- an intrusion detection system, the security fence was not secured
- at the bottom, and a portion of the microwave system was
inadequate (348, 364/85-08).
I
'
c. Severity Level V violation for inadequate security procedures
(348,364/85-08).
1
i,
{
f
-_ - - , - . - , . . . . . , . . . . - _ - _ _ . - - . _ _ - . - - - _ _ . _ - - _ _ - _ - - . - . - . . - -
_ _ . . . _ _ _ __ -.- _ . . _ _ - _ _ _ . _ ._
- _
i
- . ..
.
I
17
1
l 2. Conclusion-
4
Category 2.
3. Board Recommendations:
Management attention should be directed to the prompt resolution of the
RER findings. No changes to the NRC's inspection resources are
recommended.
H. Outages +
i
i Analysis
1.
During the assessment period, inspections were performed by the
resident and regional inspection staffs. The regional staff. also
! reviewed the design change program, inservice inspections of safety-
. related components and associated piping, supports, and snubbers;
j inservice testing of pumps and valves; welding and nondestructive
testing.
Unit I had one refueling outage from April 6,1985 to J - 4, 1985.
Two refueling outages were performed on Unit 2, January b, '985~to
l March 3, 1985 and April 5, 1986 to May 29, 1986. Major act. .ities
-
conducted during these refueling outage consisted of:
' ~
a. The Anti-vibration bars (AVB's) in Unit 1_ steam genera * ,r 1B, and
i Unit 2 steam generators 2A and 2B were replaced wi a modified
'
AYB's to reduce tube wear.
i b. An extensive inspection and repair program was completed on the
!- containment building tendons of_ Units 1 and 2. See also Section
IV.J. of this report.
I
- c. The high pressure turbine rotor, blade rings and nozzle blocks
were replaced.
i d. All feedwater heaters on both units have been replaced with
! heaters having stainless steel tube bundles.
I
i e. Eddy current testing of steam generator tubes and tube removal on
- Unit 2.
!
! f. Local leak rate testing and containment integrated leak rate
j testing on Unit 2.
i
l g. Unit 2 reactor vessel level monitoring system installation.
I
!
!
L
!
!.
L
. _ . - - - _ _ . - _ _ , . _ _ _ _ . . . . _ . _ _ _ _ - . _ _ _ _ _ _ - - _ -
_ _ _
, . . . . . - . . .-_ ~ .- ..
. .
.
18
The licensee followed management approved refueling procedures. The
procedures were- enhanced by monitoring up-to-date fuel status boards
! inside and outside containment. The licensee's safety audit engineer
a review group performed audits during the refueling period. The
licensee scheduled and followed the refueling outage with the aid of
^
'
flow and critical path charts. At the conclusion of each refueling
outage . the licensee conducted a complete review of completed work.
. Problem areas were identified and analyzed. Special attention was
given to these areas for future refueling outage scheduling.
The licensee's overall control and planning for refueling outages
results in a well planned and controlled evolution. All work is
planned with regard to scope, repair parts and work procedures.
! Planning for the next refueling cutage starts at the conclusion of the
'
present outage. There are in the order of 70 to 100 modifications
performed on each unit during a refueling outage with each refueling
,
outage typically scheduled for six weeks. Extensive operator training
'
is conducted to familiarize personnel with plant modifications.
The licensee's interface and control of contractors during refueling -
l
outages has become stronger. This is primarily due to business
meetings which the licensee has set up with the contractors prior.to '
the start of the work. This accounts for the licensee having better
'
control and understandi.ngs with the contractor.
j Licensee management involvement in inservice inspection activities
1 appeared to be adequate and decision making was at a level that assured
I adequate management review. Records were complete, well maintained,
,
and available. Procedure and policies were occasionally violated as
j evidenced by the violations listed below.
1
Six violations were identified. Violation (a) involved inadequate
> activities performed by a contractor quality control inspector.
Violations (b), (c), and (d) involved failure to comply with Technical
'
'
Specifications requirements. Violations (e) and (f) involved procedure
violations. These violations, while not indicative of a programmatic
l problem, indicate procedural adherence problems.
'
a. Severity Level IV violation for failure to follow the inspection
plan for inspection of steam generator welds (348/85-22).
b. Severity Level IV violation for failure to adhere to the
requirement of a procedure which violated Unit 1 containment
- integrity by having both inner and outer doors of the containment
auxiliary hatch open during refueling operation (348/85-20).
c. Severity Level IV violation for failure to adhere to procedure
requirements which caused the loss of both Unit 1 RHR trains for
- 52 minutes during a refueling outage (348/85-20).
1
I
,~ . - - . - - . .- _._ - -,-- , . . _ . . . _ - - . _ . ~ _ _ , .
. ..
.
19
d. Severity Level V violation for failure to have one charging pump
in the boron injection flow path operable as required by Technical
Specification during Unit 1 refueling operations (348/85-20).
e. Severity Level V violation for performing reactor - core video
inspection without a procedure to govern the activity (364/85-04).
f. Severity Level V violation for failure to fully implement fuel
handling procedure sequence in releasing the top fastener during
new fuel receipt and inspection (364/85-43).
2. Conclusion
Category 1
3. Board Recommendations
No changes in the NRC's reduced inspection resources are recommended.
I. Quality Programs and Administration Controls Affecting Quality
1. Analysis
During the assessment period, inspections -were conducted by the
resident and regional inspection staffs. The following areas were
reviewed by the regional staff: licensee actions on previous
enforcement matters, quality assurance / quality control (QA/QC)
administration, audits, document control, and licensee actions on
previously identified inspection findings.
Interviews with licensee personnel indicated that the QA program was e
adequately stated and understood. Frequent site communication was
evident and indicated that corporate QA management was actively
involved in onsite activities.
Key staff positions had been identified and authorities and respon-
sibilities for these positions were procedurally delineated. Staffing
was adequate. During this assessment period, two senior reactor
- operators were assigned to -the audit staff. Their addition provided
depth and additional expertise to operational auditing activities.
l
Audits performed by onsite QA personnel are basically compliance
audits. Audits were written by the licensee in a professional and
adept manner. Although violation (a) was identified in this area, the
- violation was administrative in nature. Audits and their responses
t were completed in a timely manner, comprehensive checklists were
'
utilized, and all audit findings were reviewed by the Senior Vice
!
President. However, the site internal audit organization lacked
sufficient expertise in the area of health physics to perform
} meaningful evaluations.
. _ __ . . - -
. ..
.
20
,
The audits examined contained two types of findings, noncompliance and
'
comments. The noncompliances were licensee identified violations of
- regulatory and site procedural requirements. Comments appeared to be
used by the audit group to identify weaknesses in the site's methods or
procedures to management. Of the comments examined by the NRC, some of
those appeared to be in grey areas which were not strictly defined by
regulation; a subset of those comments appeared to be borderline
- noncompliance. Overall, the comment concept is acceptable for its
feedback potential.
Based on the samples selected by the inspectors,. the process of-
releasing'and controlling documents for the purpose of maintenance and
operation of the plant was effective. Aside from some minor filing
problems at the user level, the document control program met regulatory
requirements and also met requirements that site personnel had placed
,
on the system.
.
'
The procurement of safety-related equipment and services and the
receipt, storage, and handling of materials met regulatory require-
. ments. Procurement documents were complete, accurate, and equipment
storage areas were well organized and clean. The licensee constructed
a new warehouse which is capable of containing all safety-related parts
!
'
and equipment under one roof in an environmentally controlled
atmosphere.
The licensee has been responsive to NRC concerns as evidenced by
successfully taking corrective actions for previously identified
enforcement matters. Two previously identified -inspection findings
were also reviewed and although they could not be closed, corrective
- action was ongoing.
The special tests and experiments program was adequate; however, a
violation was identified because one special test package was not
,
reviewed by the plant manager prior to implementation. The apparent
! root cause of this problem was that site procedures had not - been
updated to raflect this technical specification requirement. Records
- for this program were readily retrievable. Safety evaluations were
'
thorough and technically adequate.
.
'
The Quality Program and Administrative Controls Affecting Quality
section of this SALP report includes an assessment of the licensee's
ability to identify and correct his problems. As such, each specific
SALP functional area provides input for judging QA program effective-
ness. As previously mentioned in this report, the licensee takes pride
in the'c plant as evidenced by standards set for cleanliness, radwaste
> co itrol, general plant operations, and maintenance. These attributes
reflect positively on QA program effectiveness. However, the problems
addressed in Section K, Training and Qualification Effectiveness
pertaining to wrong unit / wrong train reflects negatively on the QA
program effectiveness.
2.
-_ _ _ , ._ _ _. ___ _ _ - . . _ _ _ . _ __
. ..
-
l
21
Five violations were identified. Violation -(d) resulted when site
procedures were not updated to the technical _ specification require-
ments.
a. Severity Level IV violation for failure to establish measures to
verify correct item replacement which allowed incorrect fusible
links to be installed in 2B and 2C containment air. coolers
(364/85-05).
b. Severity Lavel IV violation for failure to have a procedure
reviewed independently of the group that wrote the procedure (348,
364/85-34).
c. Severity Level V violation for failure to perform evaluations on
test equipment and prompt assessment of safety significance for
measuring and test devices found out of tolerance (348,
364/85-25).
d. Severity Level V violation for failure to have measures
established to assure that the plant manager approves tests and
experiments prior to implementation (343, 364/85-32).
e. Severity Level V violation for failure to list persons contacted
during the audit (348,'364/85-21).
2. Conclusion
Category 2
3. Board Recommendations
No cha'nges to the.NRC's inspection resources are recommended.
J. Licensing Activities
1. Analysis
Performance in the area of licensing continues to demonstrate a high
level of management involvement in assuring quality in licensing
activities. Corporate management is frequently involved in site
activities. This attribute was most certainly . evidenced by the-
containment tendon problem on Unit 2 described below.
During the third refueling outage on Unit 2, a degraded vertical tendon
was found during preparation for the containment building integrated
leak rate test. Licensee senior management organized and directed an
aggressive program of inspection and repair of -the tendons because of
an uncertainty in the licensing basis for the containment structural
_. . . - - . __ . .
. ..
.
.
22
integrity. Licensee management briefed the Commission staff on the
proposed action plan during a meeting on February 7, 1985, in Bethesda,
i Maryland. Later, on March 1,1985, licensee management again briefed
,
the Commission staff on the inspection and repair program for Unit 1.
The initial repair phase was completed on Unit 2 in April 1985,-and on
'
Unit 1 in June 1985. Followup licensee actions have been identified by
- the licensee to assure continued reactor containment- building
i
structural integrity. Structural integrity was maintained-at all times
during the entire test and. repair program.
- Licensee planning and prioritizing methods for license amendment
- requests has continued in a satisfactory manner as during the previous
-
assessment period. Meetings were held at the Farley site in June,1985,
and at the licensee's headquarters in January 1986, between the
Operating Reactors Project Manager and the licensee's staff, for
discussions of licensing priorities and schedules. These meetings were
, fruitful, resulting in a clearer understanding of the licensee's
requests as well as Commission initiated licensing actions. The
licensee provides quarterly updates in the form of a " Status of
Licensing Items" which is a helpful tool. These updates show
consistent evidence of the licensee's planning and assignment of
, licensing priorities.
,
! The licensee usually demonstrates a clear understanding and approach to
,
resolution of technical issues. The example, noted above, relating to
the resolution of the containment tendon anchor failures shows how the
..
licensee solved a very complex technical problem in a timely manner.
l However, for another technical issue relating to the analyses provided
4 by the licensee to support changes to the heatup/cooldown curves for
each unit, a weakness was evident. In the Unit 2 application for
changes to the curves, the NRC staff noted that the licensee's
submittals were not technically sound and did not exhibit conservatism
when considering safety significance. A reanalysis was required for
the Unit 2 submittal.
- The responses to NRC initiatives are generally timely. During our
review of licensee requests for 76 specific fire protection exemptions,
i the licensee revised their submittals as requested for 21 exemption
i
'
requests to document additional fire protection commitments. Their
proposals to resolve our concerns were viable, technically sound, and
are being accepted.
In the licensing support activity, the licensee has increased the
number of qualified senior reactor operators on the corporate nuclear
support staff. These trained and qualified managers, associated with
licensing support, provide a positive contribution in understanding
.
operations and in coordinating license amendment evaluations with the
l NRC staff.
,
,_.~r-r .-m., ,_.--m- --r,,,---,,,-,,,.-.m,o.,, ,-_e,..-----.--.c2-e-
. _ - _ . .- . . -- . -. . . -.
. ..
, .
23
.
I
- The licensee's staff for licensing actions is quite adequate.
j Operations qualified personnel are integrated into the corporate levels
4
of licensee management. This is a positive factor which enhances the
- . licensee's ability at the corporate level to evaluate licensing matters
which frequently involve operations.
One violation was identified,
j
i * A violation of anti-trust licensing condition No. F.2 initiated an
'
enforcement action to require compliance. The violation was
i issued pursuant to 10 CFR 2.206 against both Units 1 and 2.
- 2. Conclusion
!
Category 1
3. Board Recommendations
! None.
K. Training and Qualification Effectiveness
,
1. Analysis
During the assessment period, inspections were conducted by the
resident and regional staffs. Inspections included three licensing
examination site visits and a two week training assessment.
.
- Farley's training center consists of lecture rooms, maintenance labs,
! and a site specific simulator. This versatile and professional
j facility provides an atmosphere conducive to the proper training of
'
licensed operators and plant staff. The site specific simulator has
become a valuable tool in replacement and requalification licensed
l
operator training.. The instructors appear to be quite proficient and
! abreast of the latest plant modifications. Commensurate with the plant
i
modifications is a procedure which keeps the simulator updated to
i reflect current plant layout hardware and operating parameters.
Licensed operators felt that the simulator was an important aspect of
'
l
( their training.
l
l The licensing examinations for replacement and upgrade operators were
l administered in January and August of 1985 and February and July of
1986. In spite of the excellent training facilities noted above,
examination results have yielded a failure rate which is above the
industry average. January 1985 examination results yielded a 3 694
(5 of 14) failure for SR0s and 40?; (2 of 5) failure for R0s.
'
August 1985 results yielded a 67?s (4 of 6) failure for SR0s and a 40?s
, - - . - - - ___ - -.-_- , - - , - - . - . . - - _ . _ , . _ - _ .
. .. 1
.
24
(4 of 10) failure for R0s. February 1986 results yielded no failures
for two SR0s and two R0s. July 1986 results yielded an overall failure
rate of 40% (4 of.10) for SR0s and no failure for one R0. Areas of
generic weakness noted during the candidate's operating examinations
were as follows:
Difficulties in-classifying emergency plan levels
Inadequate use of procedures during simulator exams
Inability to diagnose minor malfunctions and abnormal situations
on simulator exams
Inconsistent use of abnormal operating procedures
During inspection (85-15) conducted in March 1985, nine apparent
violations were identified; however, as a result of the current NRC
policy statement and agreement with INPO on training and qualification
of nuclear power plant personnel, these apparent violations are being
carried as unresolved items. The following summary describes the
corrective actions taken by the licensee with regard to these
unresolved items. (It should be noted that the NRC has not reinspected
these items but is taking steps to determine whether appropriate
correctiv.e actions have been taken.)
(a) In December 1984, the Accreditation Board of the Institute of
Nuclear Power Operations (INPO) awarded Farley accreditation for
several training programs including Operator License, License
Upgrade, an'd Shif t Supervisor Training. One of the unresolved
items. pertains to Farley's failure to implement the INPO
accredited SRO Upgrade Training program. The licensee has stated
this training is now specifically addressed in procedures and is
implemented in their program.
(b) The licensee conducts the annual procedure review simultaneously
with control manipulations. This practice has not ensured that
all procedures are reviewed, or that a procedure is utilized in
its entirety as required by 10 CFR 55, Appendix A, 3.d. The
licensee stated current training specifically addresses this
matter.
(c) Since completion of the initial training in mitigating core damage
in May of 1981, replacement licensed operators have not received
the equivalent training pursuant to NUREG 0737, II.B.4, nor had
the training been specifically conducted as part of licensee
requalification training. Additionally, the licensee had failed
to provide nitigating core damage training to all I&C technicians
as committed to in their letter dated February 9, 1981. The
licensee has stated that current training is now provided to these
individuals.
(d) In the area of operational feedback experience, it was noted that
the distribution of pertinent information to the individual
mechanics and I&C technicians was informal, uncontrolled, and not
. .
.
25
documented. Operational experience was not incorporated into the
training and retraining programs for mechanics or I&C technicians.
The licensee has stated that more operational feedback experience
is being incorporated in their training and is being documented.
(e) Operational feedback experience provided in operator requali-
fication training was in the form of reading the event reports
verbatim to the students which leaves the interpretation of the
event and its applicability to Farley Nuclear Station to the
student. The importance of this lack of operational feedback
experience training is best exemplified when reviewed in
conjunction with the number of wrong unit / wrong train and tagging
event errors occurring at Farley as documented in their internal
incident reports. It should be noted that while a majority of
these incidents are minor in nature, several have caused entry
into limiting conditions of operation.
(f) In addition certain other unresolved items not described in detail
are listed herein for completeness: 1) require vendor licensed
operator instructors teaching SRO/R0 requalification to have a NRC
SRO certification or license, attend requalification lectures, and
take the annual requalification examination; 2) establish training
program for quality control inspectors; 3) provide management
training to STA candidates; 4) provide General Employee Training
to members of plant management. The licensee has stated these
items have been adaressed and corrected.
No violations or deviations were issued.
2. Conclusion
Category 2
3. Board Recommendation
No changes to the NRC's inspe, tion resources are recommended.
V. Support Data and Summaries
A. Licensee Activities
During the assessment period, the licensee conducted major activities
during the one refueling outage for Unit 1 and two refueling outages
for Unit 2. The anti vibration bars for steam generators 1B, 2A and 2B
were replaced. The licensee detected, evaluated and repaired as
required the containment building tendons; replaced the high pressure
turbine rotor, blade and nozzle blocks; and replaced all feedwater
. - - - - - _ - - _-_- - - _. - , _ _ . _.
l
. ..
.
26
1
l
heaters for both units. In Unit 2, the licensee conducted eddy current
testing of steam generators; installed the reactor vessel level
monitoring system; and conducted containment local and integrated leak
rate testing.
INP0 conducted an operations evaluation during June 1986 and an
accreditation visit during July 1986.
B. Inspection Activities
During the assessment period, routine inspections were performed at the
J. M. Farley facility by the resident and regional inspection staffs.
A special team inspection was conducted by the fire protection /
prevention program as described in Section IV.E, above. A small scale
emergency preparedness exercise was conducted (Section IV.B.). Three
licensing examination site visits, a two week training assessment, and
a supplementary reactive inspection following the RHR inoperability
were conducted as described in Section IV.A.
C. Licensing Activities
UNIT 1/ UNIT 2 TITLE OF AMENDMENT DATE
--/48 Heatup Cooldown Curves 01/22/85
Capsule U Schedule Only
57/49 Reporting Requirements (GL 83-43) 02/19/85
58/-- Heatup/Cooldown Curves to 7EFPY 05/02/85
59/E0 Update Surveillance for DC Batteries 05/25/85
60/51 Organizational Changes 01/27/86
61/52 QPTR Changes to Allow Full Core Map 03/14/86
62/53 Turbine Trip Before Latching 04/15/86
63/54 Deletion of Shutdown for Cumulative 04/16/86
__/55 Heatup/Cooldown Curves for 8EFPY 04/21/86
__/56 Deletion of Fuel Rod Height 04/22/86
64/57 Deletion of Rod Bow Penalty 06/16/86
-. __.
..
27
EXEMPTIONS GRANTED
Appendix R, 33 Technical Exemptions from Section III.G. 11/19/85
(Unit 2 and shared Unit 1 areas)
ORDERS ISSUED
Confirmatory Order - Additional Commitment on Scheduling 07/25/85
Final Emergency Response Capability
RELIEFS GRANTED
ISI Relief from ASME Code for Reactor Vessel Ligaments, 12/27/85
and for Reactor Coolant Pump Interior and Flange Areas
ISI Relief from ASME Code for Certain Valve Body Welds 06/19/86
and Internal Pressure Boundary Surfaces
D. Investigation and Allegations Review
There is currently one significant investigation in progress. This
investigation is being conducted by the Office of Investigations.
E. Enforcement History
During this SALP period, 57 inspec tions resulted in one antitrust, 12
Unit 1 and 11 Unit 2 Severity Level IV violations, and 14 Unit 1 and 12
Unit 2 Severity Level V violations. One Severity Level III violation
was identified. The Sevgrity Level III violation is related to one
train of the RHR system on Unit 1 being unable of transferring pump
suction to the containment sump during recirculation phase. Control
board operators failed to assure operability during 12 shifts of
turnover operations. During an enforcement conference at Region II on
June 3, 1986, the licensee reviewed the details of the event. Because
of self identification and prompt and extensive corrective action the
Severity Level III violation was issued without a civil penalty.
F. Management Conferences Held During Appraisal Period
1. Farley 2 Containment Tendon Field Anchor Failures Resulting in
IN 85-10 - 2/7/85 and 3/1/85
2. Project Manager Meeting at Licensee Offices to Review 1986
Licensing Schedules - 1/15-16/86
3. Project Manager Site Visits:
Commissioner Asselstine's site visit 1/23-25/85
SALP Review Meeting with Licensee 4/10-12/85
Quarterly PM Visit and Tendon Review 6/22-27/85
Appendix R Fire Protection Audit 8/21-26/85
Regulatory Effectiveness Review Audit 2/5-7/86
_ . .
, . -
. ..
.
28
4. Enforcement Conference at Region II relating to the RHR valve
being inoperable - 6/3/86
G. Operational Events
During this 19 month period, the licensee reported 42 non-security
events to the NRC Operations Center to comply with 10 CFR 50.72
requirements. Ten events involved losses of Emergency Response
Capabilities such as failures of communications or meteorological
equipment. One involved damage to a fuel assembly during installation
and one involved failures of the containment vertical tendon field
anchors.
Review of the 10 CFR 50.72 reports for these events indicates
appropriate hardware and operator response subsequent to scrams, prompt
and clear reporting by the licensee, and appropriate repairs prior to
returning to power. None of the events involved a radiation release.
Eleven events involved human error, three by operators and eight by
technicians. With the exception of the vertical tendon failures (See
Section IV.J), none of the events were considered to be significant,
especially in terms of being generic, recurring, or precursory in
nature.
During the SALP period, Farley 1 exhibited nigh availability. In 1985
the reported reactor availability was 84.3?; and for the first 6 months
of 1986 the availability was 98.8*4. The 1985 figure includes a 51-day
refueling outage. During the same period, Farley 2 exhibited slightly
above-average availability. In 1985, the reported reactor availability
for Unit 2 was 77.8%. For the first 6 months of 1986 the availability
was 72.2?s, yielding an overall average for both units during the
reporting period of about 82.5?;, or 14?; above the 1985 national average
for availability, 68.5?s.
H. Review of Licensee Event Reports and 10 CFR 21 Reports submitted by the
Licensee
During the assessment period, there were 27 LERs reported for Unit 1
and 22 LERs reported for Unit 2. The distribution of the first 44
events analyzed by cause, as determined by the NRC staff, was as
follows:
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29
Cause Unit 1 Unit 2
Component Failure 4 8
Design 4 1
Construction, Fabrication, or 4 1
Installation
Personnel
- Operating Activity 6 4
- Maintenance Activity 2 2
- Test / Calibration Activity 2 2
- Other 2 -
Out of Calibration - -
Other 1 1
TOTAL 25 19
I. Inspection Activity and Enforcement-
.
FUNCTIONAL NO. OF VIOLATIONS IN EACH SEVERITY LEVEL
AREA V IV III II I
Unit No. 1/2 1/2 1/2 1/2 1/2
Plant Operations 2/2 1/3 1/0
Radiological Controls 0/0 2/2
Maintenance 4/2 2/0
Surveillance 3/2 0/1
Fire Protection 0/0 0/0
Emergency Preparedness 0/0 1/1
Security 1/1 2/2
Outages 1/2 3/0
Quality Program and 3/3 1/2
Administrative Controls
Affecting Quality
Licensing *
Training
TOTAL 14/12 12/11 1/0
- 0ne violation with no s eve r i t.- level issued against licensee
activities of anti-trust licensing condition No. 2.
Unit 1
3/13/85 Reactor trip due to low-low water levels in steam generators
(SGs) 18 and IC. This was caused by closure of the turbine
governor and intercept valves due to spurious actuation of a
limit switch on a main steam isolation valve (MSIV). The
. ..
30
limit switch on the MSIV was replaced. Additionally,
operations personnel now verify the proper MSIV limit switch
positions prior to each unit start-up.
6/08/85 Reactor trip occurv ed due to underfrequency on the reactor
coolant pump (RCD) electrical buses. The underfrequency
condition occurred because the RCP bus power sources had not
been realigned prior to tripping the main turbine. The event
was caused by personnel error and procedural inadequacy.
Subsequently, corrective actions were taken which included
personnel counseling and revisions to appropriate plant
procedures.
6/23/85 Reactor trip occurred due to an electrical short between two
control rod drive system cables which were routed through the
same containment electrical penetration. The cables have
been repaired and rerouted. Additionally, all CRDM
- penetration modules are undergoing replacement with an
improved module manufactured by Conex (Unit 1, 7th refueling
outage (R.O.) and Unit 2, done in 4th R.0).
07/17/85 Reactor trip occurred due to low-low SG level following a
trip of the 18 steam generator feed pump (SGFP). The SGFP
tripped when a technician accidentally bumped a cable and
broke a connection on a wire leading to the SGFP thrust
bearing wire protective unit. The broken wire caused the
protection unit to indicate excessive thrust bearing wear.
Discussions are underway to either reroute or provide conduit
for the above cable.
02/28/86 A reactor trip occurred as a result of dropped control rod.
A short existed in the containment electrical penetration for
the control rod stationary and movable grippers and caused
the control rod to drop. The cable was rerouted to spare
terminals in the penetration. All CRDM penetration modules
are undergoing replacement with an improved module manu-
factured by Conax.
05/18/86 A main turbine trip occurred due to ruptured diaphragm of an
automatic stop oil / hydraulic system interface valve. The
ruptured diaphragm caused the turbine auto stop pressure to
decrease to the trip setpoint resulting in a turbine trip.
Replacement of the diaphragm is to be included as a
preventative maintenance item for subsequent refueling
outages.
07/02/86 Unit 1 tripped from 99*; reactor power on a negative rate
trip. The trip was caused by a short in control rod F14
stationary gripper coil circuit in the containment pene-
tration, resulting in blown fuses and control rod F14
_ ._, _._ _ _ _ _ _ _
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31
dropping into the core. All systems functioned as designed.
The cabling for control rod F14 stationary gripper coil
circuit was - rerouted through spare conductors in the same
Unit 2
03/28/85 Reactor trip due to 2A SG low-low level caused by the loss of
B main feed pump control. This was due to a printed circuit
card being removed incorrectly from the 28 SGFP control
cabinet. The event was caused by personnel error.
Appropriate personnel were counseled in the importance of
exercising caution while performing maintenance on operating
equipment.
03/30/85 Reactor trip due to low-low level in 2A SG following the loss
of SGFP 28. During instrument calibration, an isolation
valve leaked causing the SGFP to trip due to an incorrectly
indicated low vacuum condition. The isolation valves was
replaced.
07/15/85 Reactor trip occurred due to the loss of power in two rod
control system power cabinets. This was caused by lightning.
To eliminate the potential for future failures of this
nature, surge arrestors were installed on the auxiliary power
supplies to each CRDM power cabinet during Unit 2 4th R.0.
Arrestors will be installed durir.g Unit 1 7th R.0.
07/17/85 Reactor trip occurred due to low-low level in the 2C SG
following a SGFP and main turbine trip caused by high level
in the 2A SG. The high SG level occurred due to a main
feedwater regulating valve which apparently failed to respond
in manual control. Trouble-shooting of the control circuit
did not identify a problem. Operations personnel were
counseled in the importance of monitoring / maintaining SG
levels in manual.
08/02/85 Reactor trip occurred due to over-temperature-delta
temperature (OTDT). This event was caused by the failure of
the IB inverter while one channel of OTDT had been placed in
test for maintenance with the bistable in the tripped
condition. A faulty ferroresonant transformer in inverter 2B
was replaced. The inverter was returned to service.
01/17/86 A turbine trip was initiated manually following the loss of
both steam generator feedwater pumps (SGFP). This resulted
in a reactor trip. A short in the SGFP circuit control panel
resulted in the loss of the redundant power supplies which
serve both SGFPs.
_ _ _ - _ _ _ _ _ _ _ _ .- _-- _ _ --
_ _ . __ , - _ _ _ _ .
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32
05/13/86 During start-up from the Unit 2 Cycle IV refueling outage,
while performing test with the turbine generator at 1800 RPM,
feed flow from the steam generator feed pump went low due to
electro hydraulic (EH) fluid low pressure. This caused a
low-low level in the steam generators, resulting in a reactor
trip. Apparently the low EH pressure was due to a faulty
turbine valve actuator. These actuators are to be replaced
at each refueling outage as preventive maintenance.
06/08/86 A reactor trip occurred when both motor generator (MG) sets
malfunctioned. This allowed all control rods to fall in the
core resulting in a high negative flux rate causing the
reactor trip breakers to open.
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