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101 MARIETTA STREET,N.W.
ATLANTA, GEORGI A 30323
'%,,Y,ktN September 21, 1992 oc e os. 50-424 and 50-425 Georgia Power Company ATTN: Mr. W. G. Hairston, III Senior Vice President
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l Nuclear Operations -
i P. O. Box 1295 j
Birmingham, AL 35201 1
Gentlemen:
SUBJECT:
LICENSED OPERATOR POSITIVE DRUG TEST Your staff at Vogtle reported on September 14, 1992,- that an NRC-licensed
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operator tested positive for marijuana following a random drug test taken on 1
September 7, 1992. Please provide the details of the circumstances pertaining j
to this occurrence, answers to the questions listed in the Enclosure to this letter,.and other records and information on.this employee's past fitness for duty which are relevant to the occurrence. Any other information that you think is pertinent or useful regarding this incident would be appreciated.
i 1
You are requested to submit a written reply to the U. S. Nuclear Regulatory i
Commission, Region II, within 20 days of the date of this letter. The infor-i mation in your reply to this letter will be evaluated to see if further action by the NRC pursuant to 10 CFR Parts 50 or 55 is warranted. The information 2
j supplied will be maintained in NRC Privacy Systems of Record-16, and will be subject to the Privacy Act.
If you have any questions, please feel free to contact me at (404) 331-5680. Your cooperation is appreciated, j
l Sincerely, (original signed by A. F. Gibson)
Albert F. Gibson, Director Division of Reactor Safety I'
Enclosure:
Licensed Operator Fitness-For-Duty Questionnaire 4
i cc w/ encl:
l W. B. Shipman, General Manager,
. Nuclear Operations, Vogtle
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Electric Generating Plant t
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R. E. Dorman, Plant Training and
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j Emergency Preparedness Manager,
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Vogtle Electric Generating Plant
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a ENCLOSURE Licensed Operator Fitness-For-Duty Ouestionnaire Georgia Power Company is requested to provide the followilig information concerning the Fitness-for-Duty concern of September 14, 1992, regarding the involved licensed operator:
1.
Name and responsibilities of the operator involved.
2.
Whether the operator was identified by the facility's random testing program.
3.
The date(s) the operator was tested, and the date(s) that the test (s) were confirmed positive for marijuana under your Fitness-for-Duty Program.
l 4.
Whether the operator used/ consumed, sold, or possessed marijuana within j
the protected area.-
i S.
Results of previous fitness-for-duty testing involving the operator.
6.
Whether the operator was at the controls or supervising licensed activities while under the influence of marijuana.
l-7.
Whether the operator was involved, while under the influence of marijuara, in procedural errors which resulted in, or exacerbated the consequences of, an emergency classified as an Alert or higher.
8.
Your intentions with regard to the operator's resumption of duties under j
the Part 50 and Part 55 license.
9.
If the operator is no longer enployed by Georgia Power Company, you are expected to provide notification to the NRC of licensee termination per g
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OCT 19199f
. a Docket No. 55-20829 l
i License No. OP-20583 l
EA 92-174 1
Mr. Phillip A. Humphrey' HOME ADDRESS DELETED UNDER 10 CFR 2.790 i
i
Dear Mr. Humphrey:
SUBJECT:
NOTICE OF VIOLATION AND EXPIRATION OF LICENSE i
The Nuclear Regulatory Commission (NRC) received a letter dated September 29, l
1992, from the Georgia Power Company (GPC) informing us that they no longer have a need to maintain your operating license for the Vogtle Electric Generating Plant. We also received a letter dated October 9,1992, from GPC containing information about your second confirmed positive test for 4
marijuana. We plan to place both of the referenced letters from GPC in your 10 CFR Part 55 docket file.
I In accordance with 10 CFR.55.55(a), the determination by your facility that you no longer need to maintain a license has caused your license, OP-20583, to i
expire as of September 29, 1992.
In addition, the following violation is i
being issued on your docket:
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10 CFR 55.53(j) prohibits the use of marijuana and prohibits the licensee from performing activities authorized by a license issued under
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10 CFR Part 55 while under the influence of marijuana.
"Under the 4
influence" is defined in 10 CFR 55.53(j) to mean that the " licensee exceeded, as evidenced by a confirmed positive test, the lower of the i
cutoff levels for drugs'or alcohol contained in 10 CFR Part 26, Appendix A, of this chapter, or as established by the facility l --
licensee."
i Contrary to the above, the licensee violated 10 CFR 55.53(j) as i
evidenced by the following examples:
a.
The licensee used marijuana as evidenced by a confirmed positive test for that drug resulting from a urine sample submitted on September 7, 1992; l
b.
The licensee performed licensed duties on September 7,1992, following the submission of a urine sample which indicated he was under the influence of marijuana.
j This is a Severity Level III violation (Supplement VII).
l 3
L fi Certified Mail No.
P 258 014 672 f
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4
t OCT 191992 Mr. Phillip A. Humphrey 2
d The purpose of the Commission's Fitness-for-Duty requirements is to provide reasonable assurance that nuclear power plant' personnel work in an environment -
4 that is free of drugs and alcohol and the effects of the use of these substances. The use of illegal drugs is a serious matter which undermines the special trust and confidence placed in you as a licensed operator. This j
violation is categorized as a Severity Level III violation in accordance with the " General Statement of Policy. and ProcedureL for NRC Enforcement Actions,"
l 10 CFR Part 2, Appendix C, because the "use of marijuana" by licensed 1
i operators is a significant regulatory concern.
Because your license has expired, you are not required to respond to the Notice of Violation at this time unless you contest the violation.
Should you contest the Notice of Violation, a response is required within 30 days of the date of this letter addressing the specific basis for disputing 'the violation.
This response i
should be sent to the Regional Administrator, NRC Region II,101 Marietta Street, N.W., Suite 2900, Atlanta, Georgia 30323, with a copy to Branch Chief-1 j
Operations at the same address.
t The purpose of this letter is to make clear to you the consequences of your violation of NRC requirements governing fitness-for-duty as a licensed operator, in accordance with 10 CFR Part 55.
If you reapply for an operating I
license, you will need to satisfy not only the requirements of 10 CFR 55.31, j
but also those of 10 CFR 2.201, by addressing the reasons for the violation and the actions you have taken to prevent recurrence in order to ensure your ability and willingness to carry out the special trust and confidence placed j
4 in you as a. licensed operator and to abide by all fitness-for-duty and other i
license requirements and conditions.
In accordance with Section 2.790 of the NRC's " Rules of Practice," Part 2, Title 10, Code of Federal Regulations, enforcement actions are placed in the NRC Public Document Room (PDR).
A copy of this letter with your address removed will be placed in the PDR unless you provide a sufficient basis to withdraw this violation within the 30 days specified above for a response to this Notice of Violation.
Should you have any questions concerning this action, please contact Mr. Thomas A. Peebles of my staff. Mr. Peebles can be reached at either the address. listed above or telephone number (404) 331-5541.
Sincerely, Or:'ginal signed by Albert'F. Gibson Albert F. Gibson, Director Division of Reactor Safety 9
(cc w/o address:
seepage 3)
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b Mr. Phillip A. Humphrey 3
007 1 9 1992 s
cc w/o address:
j W. B. Shipman, General Manager Nuclear Operations, VEGP Georgia Power Company P. O. Box.1600 Waynesboro, GA 30830 bec w/o address:
J. Lieberman, OE R. M. Gallo, NRR N. Hunemuller, NRR L. L. Lawyer, DRS FFD OL file Public Document Room 55 Docket File j
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Mr. Phillip A. Humphrey 3
a! ! S 1992 cc:
W. B. Shipman General Manager Nuclear Operation, Vogtle Electrical Generating Plant bec
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' 101 MARIETTA STREET, NM.
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ATLANTA, GEORGI A 30323 e
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MAR 011993 Docket Nos. 50-424, 50-425 License Nos. NPF-68, NPF-81 4
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Georgia Power Company i
ATTN: Mr. W. G. Hairston, III Senior Vice President -
Nuclear Operations i
P. O. Box 1295 Birmingham, AL 35201 Gentlemen:
SUBJECT:
SYSTEMATIC ASSESSMENT OF LICENSEE PERFORMANCE (NRC INSPECTION REPORT NOS. 50-424/93-01 Alm 50-425/93-01)
The Nuclear Regulatory Commission (NRC) Systematic Assessment of Licensee Performance (SALP) has been completed for your Vogtle facility. The facility was evaluated for the period September 29, 1991 through January 2, 1993. The i
results of the evaluation are documented in the enclosed Initial SALP Report.-
This report will be discussed with you at a public meeting to be held at the Vogtle facility on March 18, 1993, at 10:00 a.m.
'The performance of your Vogtle facility was evaluated in the functional areas of Plant Operations, Radiological Controls, Maintenance / Surveillance, Emergency Preparedness, Security, Engineering / Technical Support, and Safety Assessment / Quality Verification. Overall performance has improved substantially during this period. Management involvement in plant evaluations and attention to detail has resulted in significant improvements in many of the functional assessment areas. Five of the functional areas were assessed as superior performers, with the other two areas assessed as good. Further, it is encouraging to note that in general, management's response to ist,ues --
both NRC and licensee identified - has been prompt and effective. The increase in performance indicates enhanced teamwork and communications between the plant departments and the corporate organizations. Your entire staff is to be commended for their efforts which have resulted in the improved-performance.
Any comments you have concerning our evaluation of the performance of your Vogtle facility should be submitted to this office within 30 days following the date of our meeting. These comments will be considered-in the deveTopment of the Final SALP Report.
Your couaients and a summiary of our meeting will be issued as an appendix to the Final SALP Report.
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Georgia Power Company 2
O Should you have any questions concerning this letter, we will be glad to discuss them with you.
l Sincerely, p
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i Stewart D. Ebyn!t
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Regional Admin trator
Enclosure:
Initial SALP Report 4
cc w/ encl:
R. P. Mcdonald Executive Vice President Nuclear Operations i
Georgia Power Company P. O. Box 1295 l
Birmingham, AL 35201 I
C. K. McCoy Vice President-Nuclear Georgia Power Company P. O. Box 1295 j
Birmingham, AL 35201 i
W. B. Shipman General Manager, Nuclear Operations i
Georgia Power Company P. O. Box 1600 i
Waynesboro, GA 30830 J. A. Bailey Manager-Licensing Georgia Power Company l
P. O. Box 1295 Birmingham, AL 35201 4
i ec w/ enc 1: Continued page 3 i
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2 II.
SUlgiARY OF RESULTS l
Vogtle continued to operate in a safe and conservative manner and exhibited good overall operational perfomance. This operational I
l perfomance was characterized by long periods of operation at power with few transients or problems requiring shutdowns to resolve. However, 4
l instances of insdmate procedure implementation, personnel error and j
inattention to detail during plant activities occurred. Management continued to perform an active and visible role in plant operations, and i
the professionalism of the licensed control room staff continued to be j
good.
i 5
Perfomance in the area of Radiological Controls remained excellent.
The radiation protection and chemistry staffs were well qualified and effectively supported routine activities. The licensee's aggressive 4
control of contamination continued during the assessment period.
Programs for monitoring and controlling Liquid and gaseous radioactive effluent, as well as for generation, storage, and shipment of solid l
radioactive waste.(radwaste),. wore implemented effectively.
S l
Maintenance activities were very effective this assessment period and l
did not result in any reactor trips or unplanned outages.
Improvements this assessment period were noted in the predictive and preventive i
maintenance programs, in the reduction of maintenance backlog, and the general materia. condition of the plant. Overall performance in the l
surveillance area was good. A recurring concern was noted with the j
failure to perfom special condition surveillances.
1 In the area of Emergency Preparedness, the licensee demonstrated an effective response capability for dealing with emergency situations i
during the annual NRC evaluated exercise, with the exception of l
notifications to State and local agencies. During the exercise, the licensee failed to make accurate and timely notifications to State and-l local agencies concerning emergency classification status, release conditions, and follow up information. The NRC staff recommands 2
continued management attention to the accuracy ad timeliness of notification to offsite agencies.
I I
Site management continued to support the Sec'Jrity program resulting in a high level of performance by its staff and proprietary security force.
Management's involvement was evider.t by efforts in areas of hardware i
upgrades, improved cameras, timely maintenance of failed systems, and i
the elimination of long-tera compensatory measures. During i.he first part of this assessment period, the licensee continued to expsrience a i
high number of vital area doors being left unsecured from personnel errors or mechanical problems, m.
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ihe licensee's perfomance in providing engineering and technical support was very effective during this assessment period.
Improvements i
from the pretous assessment period were demonstrated in plant j
modifications, syItem engineering, and design control.
Effective j
performance continued in technical interf, ace support and communications, l
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system engineer training, qualification and certification and licensed operator training. The licensee's response to Generic Letter 89-10
" Safety-Related Motor-0perated Valve Testing and Surveillance" l
(GL 89-10), was not sufficiently comprehensive.
l Performance was excellent in the area of Safety Assessment / Quality l
Verification. Management was actively involved in assessing plant performance and safety issues, and fully supported organizations and l
programs that identify and assess problems. Overall, the Safety Audit j
- .nd Engineering Review (SAER) audits were thorough and effective in identifying deficiencies. The Safety Review Board and Plant Review j
Board (PRB) continued to be effective in reviewing plant safety matters.
i Overview Performance ratings for the last rating period and the current rating j
period are shown below.
Rating Last Period Rating liin Pert'cd' l
Functional Area 10/01/90 - 09/28/91 99/29/91 - 01/02/93 i
Plant Operations 2
2 J
j Radiological Controls 1
1 Maintenance / Surveillance 2
1 i
2 l
Security and Safeguards 2 (Improving) 1 l
Engineering / Technical Support 2 (Improving) 1 l
Safety Assessment / Quality 2
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Verification i
III.
CRITERIA 4
The evaluation criteria which were used to assess each functional area are described in detail in NRC Manual Chapter 0516, which can be found j
in the Public Document Room. Therefore, these criteria are not repeated here, but will be presented in detail at the public meeting held with a
l the licensee management on March 18, 19g3.
i l-IV.
PERFORMANCE ANALYS'.S f
A.
Plant Goerations
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1.
Analysis i
i This functional area addresses the control and performance of activities directly related to operating the facility, j
including fire protection.
Vogtle continued to operate in a safe and conservative i
manner and exhibited good overall operational performance.
This operational performance was characterized by long j
periods of operation at power with few transients or v
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y.
4 IO' problems requiring shutdowns to resolve. During this i
assessment period Unit 1 experienced no automatic reactor j
trips, one unplanned manual reactor trip, one forced i
shutdown due to high unidentified reactor coolant system l
(RCS) leakage, and one voluntary shutdown due to identified leakage approaching TS limits. Unit 2 experienced two automatic reactor trips. The total number of trips was l
comparable to the number of trips during the previous i
assessment period. The reactor trip on. Unit I was due to a j
failed controller on a main feedwater pump (MFWP). The two j
Unit 2 automatic reactor trips were both caused by personnel i
error. One trip was caused by a plant equipment operator who inadvertently tripped a DC breaker while checking switchgear alams. The second trip was caused by a l
technician momentarily grounding the main generator excitation circuitry which caused a main generator trip and resultant reactor trip.
In each of these cases the control i
room operators responded effectively.
During this assessment period instances of inadequate l
procedure implementation, personnel error and inattention to detail during plant activities occurred.
Examples included 4
a failure to exercise adequate control over a Unit I reactor-cavity draindown evolution and inadequate procedures which -
resulted in the interruption of decay heat removal, an l
inadvertant entry into a condition prohibited by TSs i
associated with the hydrogen monitor containment isolation i
valves, and a failure to follow procedure during a Unit 1 l
startup which resulted in a failure to correctly withdraw control rods.
Plant management recognized'the weaknesses in j
these areas and has attempted to foster a greater awareness f
of procedural compliance and management expectations.
Nrsonnel, including managers and supervisors, are held accountable for following procedures.
Performance is monitored through trend reports and by management observation and involvement. Although improvement has been l
observed, these ongoing activities have not been fully i
successful. These types of problems were also observed in j
the previous assessment period.
The professionalism of the licensed control room staff was generally good. This was evident in observation of control room demeanor, watchstanding practices, shift turnovers, and j
attentiveness to duties. A weakness, however, was identified in Control Room fomality. There was a tolerance 4
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of informality in one aspect of Centrol Room activities. As i
a result, an incident occurred where no personnel were in i
the 'at-the-controls' area of the control Room for a brief j
period.
1 i
As a result of successfully licensing a relatively large j
number of personnel and a reduction in the attrition rate of i
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licensed operators, the licensee reorganized the Operations Department to more effectively use their resources. This resulted in several licensed reactor operators being i
assigned to non-licensed duties and changes in responsibilities for some Senior Reactor Operator (SRO) j licensed personnel. These additional personnel resulted. in l
better definition of responsibilities for shift personnel, more involvement in work pinnning activities by shift l
personnel, and more opportunity for field observation and j
involvement by supervisors. A Control Room staff exceeding l
TS requirements was maintained.
In addition, most j
department managers now hold an SR0 license. This has resulted in enhanced coemunication and understanding of l
plant issues.
i l
Management continued to perfom an active and visible role in plant operations.
In routine operational activities plant management actively participated in morning meetings, j
shift turnover activities, and management of the equipment-j out-of-service list that prioritized repairs to equipment.
l In an effort to more clearly communicate management l
expectations, management enhanced the process for control of i
infrequently perfomed evolutions. These evolutions
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included integrated engineered safety feature actuation systems tests, reactor vessel head lifts, refueling, reactor l
startup, etc.
Briefings of the personnel involved in these evolutions were held by the manager responsible for the 4
activity, prior to the activity being perfomed. These briefings covered various aspects of the evolution and i
management's expectations.
Management also initiated several innovations to improve the perfomance of the Operations Department. This included use of the computer to prioritize daily work activities by the shift supervisors, an integrated computer system and local area network that provided better access to data and I
infomation, and the implementation of some computerized l
rounds using a hand-held digital data entry device. The infomation obtained by the data entry device can be used by l
system engineers and others to track and trend data.
Improvements in the area of configuration control were noted j
during this assessment period. During the refueling outags i
in the early part of this assessment period, severaL problems occurred in the area of configuration control, such t
as the reactor cavity drain doun-incident. Corrective actions for these problems resulted in significant 1
improvements in configuration contro1~ishich were reflected j
in the most recent refueling outage.
j During this assessment period, plant housekeeping and j
materini condition continued to improve and were considered b
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excellent. This was achieved through the continuation of a broad painting and preservation program, management walkdowns, a leak walkdown program, and greater i
accountability by plant personnel for plant appearance and material condition.
l l
Routine observation of activities associated with the fire protection program indicated the program requirements were l
well implemented. The Operations Department now has responsibility for the fire protection program and the fire l
brigade is composed of trained Operations personnel.
Fire l
protection surveillances that require operation of fire protection equipment are perfomed by Operations personnel.
l A fire protection technician is on shift at all times in the l
Control Room to assist in implementing program requirements.
Fire brigade responses to drills were satisfactory.
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j During the assessment period, nine violations were cited.
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2.
Perfomance Ratina j
Category:
2 3.
Recommendations Although the board has noted improvement in the area of l
procedural compliance, management attention in this area should continue to assure procedures are correctly i
implemented.
I B.
Radioloaical Controls l
l 1.
Analysis This functional area addresses those activities directly i
related to radiation safety, radiological effluent. control and monitoring, and primary / secondary chemistry control.
l The radiation protection and chemistry staffs, consisting of licensee and contractor personnel, were well qualified and effectively supported routine activities.
Personnel training was comprehensive, in-depth, and was a program j
strength.
j The licensee's aggressive control of contamination continued 4
during the assessment period.
Contaminated square footage continued to be well-maintained, averaging less than 2 % of the total radiologically controlled area. The number of personnel external contamination events decreased from 53 i
during the previous assessment period to 42 during the current assessment period. This represents a significant improvement given that this assessment period included two i
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i total activity from fission and activation products in i
liquid effluents released during the assessment period 1
continued to decline from 0.9 Curies (C1) for 1990, to 0.3 i
Ci for 1991, and remained low (0.1 Ci) during the first half j
of 1992. One unplanned liquid release occurred involving i
approximately 15 microcuries of radioactive material. No dose limits were exceeded as a result of this release.
The Environmental Monitoring Program continued to effectively monitor releases to the environment. As a result, 1991 data indicated that there was no adverse a
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radiological impact to the environment resulting from plant releases. Also, there was excellent agreement between the Georgia Department of Natural Resources' environmental monitoring program data and the licensee's program data.
In addition, the results of the licensee's Environmental Protection Agency's interlaboratory crosscheck program j
verified that the licensee was able to perform accurate j
analyses of environmental samples.
The licensee's programs for generation, storage, and shipment of solid radwaste were implemented effectively.
j For 1992, the total quantity of radwaste generated and j
processed due to routine and-outage activities remained low.
Contaminated sludge, which had been previously stored
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j onsite, was also disposed of during this assessment period.
j No incidents involving the transportation of licensed i
material were reported or identified.
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During the assessment period, four violations were i
identified.
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2.
Perfomance Ratina Category:
1 i
3.
Recomanndations None.
C.
Maintenance / Surveillance 833.21.11 1
1.
This functional area addresses those activities related to equipment condition, maintenance, surveillance perfomance, and equipment testing.
Overall maintenance and surveillance activities were very effective. Maintenance activities did not result in any Q
reactor. trips or unplanned outages.
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Maintenance management increased focus on the reduction of corrective maintenance through a strong preventive and predictive maintenance program, efforts to effectively diag-nose the root cause of recurring maintenance problems, and by effectively using opportunities to perfom maintenance.
Examples of this were the licensee's persistent and i
successful efforts to resolve longstanding problems with the j
emergency diesel generators (EDG) and digital radiation monitoring system, and modifications and replacement of non-l 1E 4160/480 volt transformers, resulting in greater reliability of these components. Management oversight resulted in an effective inservice inspection (ISI) program l
and an improvement in the ratio of preventive maintenance to corrective maintenance from the last assessment period. The l
number of non-outage corrective maintenance work orders also i
decreased during the period.
Improvements this assessment period were noted in the i
predictive and preventive maintenance programs, in the reduction of maintenance backlog, and the general material condition of the plant. The snubber reduction program was l
completed which will reduce manpower and exposure for subsequent examinations. Recently procured infrared equimt was useful in performing functional tests and l
trousleshooting problems. Predictive maintenance i
enhancements were implemented which included a major i
revision to vibration analysis computer software, 4
utilization of the main frame computer to coordinate repetitive predictive maintenance tasks with the Operations surveillance testing schedule, and implementation of lubricating oil ana ysis optical examination for Engineered Safety Feature Actuation System equipment.
Staffing and training were sufficient to accomplish the maintenance program. The onsite Maintenance Engineering group worked with the Work Planning group to reduce the preventive maintenance backlog. Corporate engineering and maintenance personnel were used to resolve complex problems and to assist in various maintenance evolutions.
ISI personnel were knowledgeable and well qualified. Mainte-nance management made a significant commitment to training over this assessment period. Maintenance foremen and supervisors aceived a three week supervisory course.
Instrumentation and Control technicians received a substantial increase in training over the previous year in specialty areas of the Nuclear Steam Supply System, the Balance of Plant, and niemprocessor based systems. This resulted in a greater number of technicians qualified to work on specialty systems such as the Digital Rod Positioning Indication System, the turbine Electrohydraulic controls, and the Reactor Vessel Level Indication System.
Additional training was provided to some maintenance
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engineers in vibration analysis, nondestructive examination j
of welds, and motor-operated valve (MOV) test data evalua-i tions.
l Various maintenance areas throughout the plant were i
improved. Cleanliness and access control to the hot shop i
and hot tool room were improved. These areas were previously identified as poorly-controlled. The weld test shop was renovated to improve the ability to qualify welding procedures, test / qualify welders and conduct specialized training. Auxiliary building leak inspection walkdowns were j
effective in reducing the number of leaking contaminated i
~ systems.. A weakness noted in the previous assessment period was a high threshold for the identification of poor material condition and housekeeping problems, and the material condition of sampling system valves in the penetration rooms. This weakness was improved during this assessment period.
Overall performance in the surveillance area was good. The surveillance tracking program continued to be well managed i
for routine surveillances. However, a concern was noted i
with the failure to perfom special condition surveillances.:
Management recognized this area of weakness and took i
actions, including assigning specific responsibilities to operations shift supervision for ensuring completion of i
these types of srJrveillances. Although improvements were observed with the adequacy of surveillance procedures, j
problems continued such as, inadequate procedures which resulted in an unplanned safety injection and an improperly i
calibrated residual heat removal suction valve interlock.
i Although safety systems were challenged, there were no significant examples identified which had a negative effect d
on the operability of safety-related equipment. Management recognized these problems and continued their commitment to improvements in this area by emphasizing strict c.ompliance j
and correcting procedural deficiencies.
During this assessment period, three violations were identified.
2.
Performance Ratino Category:
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3.
Recommendations None.
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t e(.dM gay U.1993 Docket Nos. 50-424; M-425 License Nos. NPF-68, NPF-81 Georgia Power Company ATTN: Mr. W. G. Hairston, III i-Senior Vice President -
. Nuclear Operations
.P. O. Box 1295.
Birmingham, AL 35201 i
Gentlemen:
SUBJECT:
(NRC INSPECTION REPORT NOS. 50-424/93-07 AND 50-425/93-07)
This refers to the inspection conducted by B. Bonser of this office on March 28 - May 1, 1993. The inspection' included a review of activities authorized for your Vogtle facility.
At the conclusion of the inspection, the i
findings were discussed with those members of your staff identified in the enclosed report.
l Areas examined during the inspection are identified.in the report. Within these areas, the inspection consisted of selective examinations.of procedures
-and representative records, interviews with personnel, and observation of activities-in progress.
i Based on the results of this inspection, certain of your activities appeared to.be in violation of NRC requirements, as specified in the enclosed Notice of Violation (Notice).. In addition, the enclosed Inspection Report identifies activities that violated NRC requirements that will not be subject to I.
enforcement action because the licensee's efforts in identifying and/or correcting the violation meet the criteria specified in Section VII.B. of the l
You are required to respond to this letter and should follow the instructions specified'in the enclosed Notice when preparing your response.
In your response, you should document the specific actions taken and any additional actions you plan to prevent recurrence. After reviewing your response to this Notice, including your proposed corrective actions and the results of future inspections, the NRC will determine whether further NRC enforcement actions is necessary to ensure compliance with NRC regulatory requirements.
The responses-directed by this letter and the enclosed Notice are not subject to the clearance procedures of the Office of Management and Budget as required by the Paperwork Reduction Act of 1980, Pub. L. No.96-511.
In accordance with 10 CFR 2.790(a), a copy of this letter and its enclosure will be placed in the NRC Public Document Room.
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5 ENCLOSURE I NOTICE OF VIOLATION i
Georgia Power Company Docket No. 50-424 Vogtle Unit 1 License No. NPF-68 During the NRC inspection conducted on March 28 - May 1, 1993, a violation of i
NRC requirements was identified.
In at.cordance with the " General Statement of l
Policy and Procedures for NRC Enforcement Actions," 10 CFR Part 2, appendix C, the violation is listed below.
i 10 CFR 50, Appendix B, Criterion XVI, Corrective Action, requires I
measures to be established to assure that conditions adverse to quality, such as failures, malfunctions, deficiencies, deviations, defective material and equipment, and nonconformances are promptly identified and 4
corrected.
l Contrary to the above, on March 16, 1993, a Design Change Package and other associated reference material were not adaquately reviewed as part of the corrective action to troubleshoot.an illuminated bistable light i
on the Main Control Board. This resulted in a momentary loss of the i
only available train of decay heat removal.
This is a Severity Level IV violation (Supplement 1).
Pursuant to the provisions of 10 CFR 2.201, Georgia Power Company is hereby required to submit a written statement of explanation to the U. S. Nuclear l
Regulatory Commission, ATTN: Document Control Desk, Washington, D. C. 20555, with a copy to the Regional Administrator, Region II, and a copy to the NRC i
Resident Inspector Vogtle Nuclear Plant, within 30 days of the date of the j
letter transmitting this Notice of Violation (Notice).
This reply should be i
clearly marked as 'a " Reply to a Notice of Violation" and should include for each violation:
(1) the reason for the violation, or if contested, the basis for disputing the violation, (2) the corrective steps that have been taken and i
the results achieved, (3) the corrective steps that will be taken to avoid further violations, and (4) the date when full compliance will be achieved.
If an adequate reply is not received within the time specified in this Notice, l
an order or demand for information may be issued as to why the license should i
not be modified, suspended, or revoked, or why such other action as may be proper should not be taken. Where good cause is shown, consideration will be given to extending the response time.
i Dated at Atlanta, Georgia this 25thday of May i
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UNITED STATES j
g NUCLEAR REGULATORY COMMisslON
,f(.n REGION 11 y.
j g/' j 101 MARIETTA STREET, N.W.
1 3
g ATLANTA, GEORGIA 30323 i
%[ E.VM,E:
l Report Nos.:
50-424/93-07 and 50-425/93-07
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Licensee: Georgia Power Company i
P. O. Box 1295-Birminghain, AL 35201 4
4 Docket Nos.:
50-424 and 50-425 License Nos.:
NPF-68 and NPF Facility Name:
Vogtle 1 and 2 Inspection Conducted: March 28 May 1, 1993 Inspector: O8 f 2o 93 p B.
Bons nior Resident Inspector Date Signed
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6~ 2.o 9.3 K. D. Stark ident Inspector Date Signed bl kW G. 73 93 P. A. BTima dent Inspector
'Date Signed I
Accompanied by: J.L Starefos.
A Approved by:
f 2N P.' Skinnet, Chief Date Signed 1
Reactor Projects Section 38-Division of Reactor = Projects
SUMMARY
Scope:
This routine, inspection entailed inspection in the following areas: plant operations, surveillance, maintenance, plant modifications, refueling activities, review of overtime records, and follow-up of open items.
Results:
One violation, three non-cited violations (NCV), and one unresolved item (URI) were identified.
The violation occurred from a failure to take adequate corrective action when troubleshooting an illuminated bistable light on the main control board.
Inadequate preparation of the troubleshooting activities subsequently resulted in a Unit I loss of decay heat removal (paragraph 8b).
One NCV involved the performance of a procedural step out of
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sequence during performance of a Unit 1 Engineered Safety Features C%0bjGLl?ll -
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l-completed and the unit entered Mode 5 on April 15.
The unit l
entered Mode 4 on April 21, Mode 3 on April 22, and Mode 2 on April 24.
The reactor was taken critical during physics testing on April 24 but reentered Mode 3 following a manual reactor trip
- on April 25 due to an' unexpected large negative reactivity.
i insertion.
The unit was taken critical and returned to mode 2 on April 26.
Physics testing was completed and the unit entered Mode 1 on April-26.
At the close of the inspection period the unit had l
reached 94% power.
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. Unit 2 Summary L
The unit began the period operating at 100% power and operated at j'
full power throughout the inspection period.
4 d.
Loss of Unit 1 Train B IE Switchgear During ESFAS Test On April 15, the licensee was performing procedure 14667-1, Train
-B DG and ESFAS Test, section 5.2, Loss-of-offsite Power / Safety Injection. The unit was in Mode 6. Refueling, with the reactor head set, and RHR A was providing shutdown cooling.
Prior to the l
manual initiation of the UV/SI signal.as directed in the procedure, the IB DG had been paralleled to the safety-related 4160 V switchgear, IBA03, and "B" Train RHR, CS, AFW, NSCW, and CCW pumps-had been started.
When the UV/SI signal was manually initiated the normal incoming supply breaker from the. "B" RAT to IBA03 tripped as expected, the loads on'IBA03 were shed and then sequenced back on to IBA03 which was still being powered by the IB DG'.
All loads sequenced automatically on to IBA03 as expected.
At 10:28 a.m., approximately five minutes after all loads had been automatically sequenced onto 1BA03, the IB DG output breaker tripped open resulting in all power being lost to IBA03.
Consequently, all the "B". train safety related pumps tripped.
Of particular concern to the operators was the loss of NSCW which supplies cooling water to the DG and they manually initiated an emergency trip of the IB DG.
During this entire evolution A Train safety-related equipment was not affected and RHR A continued to provide shutdown cooling. Operators also entered TS Action Statement 3.9.8.2, since less than two trains of RHR were operable.
TS 3.9.8.2a requires that the inoperable train be restored as soon as possible. At 10:40 a.m., power was restored to IBA03 from the normal supply via the "B" RAT.
At 10:43 a.m.,
RHR "B" was available for operation and TS 3.9.8.2a was. exited.
The cause of this event was the failure of personnel directing procedure 14667-1 to perform procedure steps in the proper sequence.
Specifically, step 5.2.20, which requires that the SI signal be reset, was completed prior to step 5.2.19, which requires that the 151V phase A time-overcurrent relay at the IB DG be manually tripped.
The time-overcurrent relay trips the 186B lockout relay.
During a non-emargency start of the DG a trip of
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the 186B lockout relay will cause the DG to_ trip.
However, this_
trip is bypassed during an emergency start of the DG.
In this event, the SI signal-had been prematurely reset and therefore the i
UV/SI signal, which initiated the DG emergency start and would j
have prevented a DG trip on a trip of the 186 B lockout relay, had i
been removed.
Subsequently, when the Phase A time-overcurrent relay was tripped in step 5.2.19, which caused the 186B relay to trip, resulting in the DG tripping.
i The inspector was present in the control room during this ESFAS' test evaluation. The evolution was thoroughly briefed to all F
participants by the test director and the Manager Operations.
i During the test the test director became distracted when questioned by other personnel in the control room as to when the 4
SI could be reset.
He mistakenly directed that the SI be reset and by doing so performed a procedural step out of' sequence. The test director was counseled. The inspector witnessed the i
procedure again when it was successfully reperformed and was i
satisfied that it was adequate and that the mistake which was made l
previously was due to' personnel error in not performing the procedural steps in sequential order.
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Performing a procedural step out of sequence was a failure to implement procedure 14667-1, Train B DG and ESFAS Test, section i
5.2, Loss-of-Offsite Power / Safety Injection; and represents a 4
violation of TS 6.7.la.
This violation will not be subject to i
enforcement action because the licensees' efforts in identifying I
and correcting the violation meet the criteria specified in j
Section VII.B of the NRC Enforcement policy.
This violation is j.
identified as NCV 50-424/93-07-01, Failure To Follow Procedure.
e.
Unit 1 Containment Ventilation Isolation On April 11, with Unit 1 in Mode 6,.a CVI occurred due to high radiation on containment low range area radiation monitors 1 RE-002 and 1 RE-003, during placement of the reactor upper internals in the reactor vessel. The highest reading on 1 RE-002/003 was approximately 50 mR/Hr which was expected for the' evolution in progress.
The CVI actuation setpoint was 15 mR/hr.
Procedure 12000-C, Post Refueling Operations, requires that 1RE-002/003 be reset to a higher alarm setpoint during this evolution.
Several factors contributed to the setpoints not being increased as required. The USS was involved with several activities and his attention was diverted away from the task at hand which he felt was to complete the setting of the upper reactor vessel internals on his shift. The internals were set late in the shift while control room turnover activities were being conducted. The USS was using procedure 12007-C, Refueling Operations, and had delayed transition to procedure 12000-C because step 4.6.1 of 12007-C could not be signed off until completion of " core alterations" which includes setting of the upper internals. The USS felt that
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potential defect mentioned in the Part 21 notification.
- However, l
the licensea, u. a precautionary measure, replaced the gear with i
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an improved design gear. During the next Unit 2 refueling outage the Unit 2 DG jacket water pumps will'be inspected and the gears replaced if necessary.
Unit 2 jacket water pumps have' operated in L
excess of twenty hours each with no indication of pump gear l
problems. The inspector was satisfied that the licensee had i
appropriately addressed the concern identified in the Part 21 j
notification.
b.
(Closed) URI 50-424/93-04-02,. Review Causes of Loss of Decay Heat Removal Event The URI addressed inadequacies in design change package reviews and the performance of a reactor protection system modification with only one train of. decay heat removal in service. 'This event, which resulted in a momentary loss of decay heat removal, was caused by the RHR inlet isolation valve,1HV-8701B, closing. As discussed in NRC IR 50-424,425/93-04, the event occurred when an I&C technician, troubleshooting the cause of an illuminated bistable' light prior to performing a functional test on the design change to the reactor protection circuitry, removed a card in the RHR valve autoclosure circuitry. The URI was opened pending completion of the review of this event.
4 The inspectors concluded, after further. review, that the cause of this loss of decay heat removal was anvinadequate' review by'I&C-
- personnel of the modification ^and other available material which described the circuitry being modified.
Prior to performing the troubleshooting'to clear the bistable light, I&C personnel did not identify from their review, that-the RCS wide range pressure loop SP-408, shared the circuit card that was removed. A more thorough review by the I&C foreman before directing the technician to continue troubleshooting'should have identified.the potential problem. As a result of the inadequate review, the technician was poorly prepared to troubleshoot, creating additional problems which challenged core safety.
4 The inspectors also concluded that there were other contributing causes to this event. The initial review of the DCP by I&C during the implementation planning stages did not identify all the circuit cards to required to be removed to deenergize each process loop prior to performing the DCP.
In this case, all power was not removed from the lo_op before work was performed. The RCS wide range pressure loop was not identified as sharing part of the process loop on which the design change was being performed and as a result work was performed on~a partially energized loop. The cause of the bistable light which initiated the troubleshooting process and resulted in this event, was suspected to be a contact between two pins on a multi-pin plug, one of which was still energized. Also this DCP was scheduled during a period of higher risk since, at the time of the event, RHR Train A was in the
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i 15 service for maintenance. Two SGs were available as a heat ~ sink in accordance with TS 3.4.1.4.1. 'The tie between the DCP work and the single train of decay heat removal was not considered.
Based on this review, URI 50-424/93-04-02, Review Causes of Loss of Decay Heat Removal Event, is closed. The loss of decay heat removal which resulted from a failure to take adequate corrective action when troubleshooting an illuminated bistable light on the main control board is identified as violation 50-424/93-07-05, Failure To Take Adequate Corrective Action Resulting In Loss Of Decay Heat Removal, q
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c.
(0 pen) LER 50-424/91-015, Rev. 1, Valve Manufacturing Defect Leads to Containment Isolation Valve Failing Open.
The casting mark on valve 1-2401-04-034 was removed, LLRT testing-was completed satisfactorily, and the valve was returned _to-service. The licensee was unable to find any similar occurrences of this type of event. Additionally, the licensee committed to inspect, during refueling outages 1R4 and 2R3, a representative sample of these valves which are not inspected as part of the IST program.or the valve' disassembly program.
During outage IR4, a sample of seven valves were inspected out of a total of twenty-one possible valves. None of the seven valves tested exhibited.the casting mark discrepancy described in the LER which would cause the disk to bind. During refueling. outage 2R3 later this year, the licensee will again perform a random sample of these
. identified valves which are not inspected in other inspection programs. The inspectors will review the results of those inspections following.2R3.
One violation was identified.
9.
Exit Meeting The inspection scope and findings were summarized on April 30, 1993, with those persons indicated in paragraph 1.
The inspector described the areas inspected and discussed in detail the inspection findings listed below. No dissenting comments were received from the licensee. The licensee did not identify as proprietary any of the material provided to or reviewed by the inspectors during the inspection.
Item No.
Descriotion and Reference NCV 50-424/93-07-01 Failure to Implement ESFAS Test Procedure Results in the Loss of Unit 1 Train B IE Switchgear (paragraph 2d)
NCV 50-424/93-07-02 Failure to Implement Low Power Physics Test Procedure For Controlling Reactivity (paragraph 2f)
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Georgia Power Company
-g 40 inv:< ness C.:nt:r Ptrxw y Post Ottco Box 1295 Birminghtm. Altb:ma 35201 Telechene 205 877 7122 L
Georeia Power l
- c. x. uccoy U
Vee Prescent. Nucles '
the southem e@CffC System j
Vogue Prp$ct' June 16, 1993 LCV-0033 Docket No.
50-424 U. S. Nuclear Regulatory Commission ATTN: Document ControlDesk Washington, D. C. 20555 4
Gentlemen-VOGTLE ELECTRIC GENERATING PLANT REPLY TO A NOTICE OF VIOLATION l
Pursuant to 10 CFR 2.201, Georgia Power Company (GPC) submits the enclosed response to the conducted by Mr. B. Bonser during the period of March 28 through May 1,1993.
violations identified in Inspection Reports 50-424/93-07 and 50-425/93-07 conceming the inspection 1
Sincerely, C. K. McCoy
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CKM/NJS Enclosure xc:
Geornia Power Comnany Mr. J. B. Beasley Mr. M. Sheibani I
NORMS U. S. Nadaar Ramdatory Commie = ion Mr. S. D. Ebneter, Regional Administrator 4
i Mr. D. S. Hood, Licensing Project Manager, NRR Mr. B. R. Bonser, Senior Resident Inspector, Vogtle f{ /
s J
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ENCLOSURE l
VOGTLE ELECTRIC GENERATING PLANT - UNIT 1 l
REPLY TO A NOTICE OF VIOLATION NRC INSPECTION REPORTS 50-424:425MbO7 The following is a transcription of the violation as cited in the Notice of Violation (NOV):
"10 CFR 50, Appendix B, Criterion XVI, Corrective Action, requires measures to be estabhshed to assure that conditions adverse to quality, such as failures, mahetions, deficiencies, deviations, defective material and equipment, and nonconformances are promptlyidentified and corrected.
l Contrary to the above, on March 16,1993, a Design Change Package and other associated reference material were not =d-ntah reviewed as part of the corrective action to j
troubleshoot an illuminated bistable light on the Main Control Board. This resulted in a j
momentary loss of the only available train of decay heat removal.
i This is a Severity level IV violation (Supplement 1)."
RESPONSE TO VIOLATION (50-424Mb07-05) l Admission or Denial of the Violation l
This example of the violation occurred as stated in the NOV.
Reason for the Violation This violation was attributed to an inadequate review by Instmmentation and Controls (I&C) i personnel of the affected circuitry prior to removing a circuit card for troubleshooting an indication on the trip status light board. An I&C technician and foreman performed these reviews, but they
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failed to identify all the components affected by the circuit card. The removal of the circuit card
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caused residual heat removal system suction isolation valve 1-HV-8701B to close, causing a loss of j
decay heat removal for approximately 2 minutes.
Corvective Stens Which Have Been Taken and the Results Achieved 1
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- 1. The I&C technician and foreman have been counseled regarding the importance of adequate l
reviews.
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- 2. Instrumentation and Controls personnel have been briefed i@.g this event and the enhawal
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specific work controls to use when performing this type of work.
ij Corrective Stens Which Will Be Taken to Avoid Further Violations All co Tective actions have been completed, and no further action is warranted at this time.
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l ENCLOSURE (CONTINUED)
- i VOGTLE ELECTRIC GENERATING PLANT - UNIT 1 REPLY TO A NOTICE OF VIOLATION NRC INSPECTION REPORTS 50-424
- 425/93-07 2
1 Date When Full Comoliance Will Be Achieved j
Full compliance was achieved on March 16,1993, at 1652 EDT, when valve 1-HV-8701B was reopened, thereby restoring decay heat removal capability.
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