IR 05000244/1988004
| ML17251B059 | |
| Person / Time | |
|---|---|
| Site: | Ginna  |
| Issue date: | 04/14/1988 |
| From: | Dudley N, Finkel A, Johnston W, Mcbrearty R, Oliveira W, Prell J, Strosnider J, Winters R NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION I) |
| To: | |
| Shared Package | |
| ML17251B056 | List: |
| References | |
| 50-244-88-04, 50-244-88-4, IEB-85-003, IEB-85-3, NUDOCS 8804270199 | |
| Download: ML17251B059 (66) | |
Text
U.S.
NUCLEAR REGULATORY COMMISSION
REGION I
Report No.
50-244/88-04 Docket No.
50-244 License No.
OPR-18 Licensee:
Rochester Gas and Electric Company 49 East Avenue Rochester, New York Facility Name:
R.
E. Ginna, Nuclear Power Station Inspection At: Ontario, New York Inspection Dates:
February, 22-26, 1988 Team Leader:
J. Strosnider, Chief, Materials
& Processes ection, Engineering Branch, DRS, RI Inspectors:
0 r 8%Qa N.
F. Dudley, Senior Operations Engineer, PWRS, OB, DRS, Region I 4/II/B8 date da e
A.
E. Finke, nior Reactor Engineer, OPS, OB, DRS, egion I d te R.
A. McBrearty, Reactor E
ineer, MPS, EB, DRS, Region I date
- W. Oliveira, Reactor Engineer, OPS, OB, RS, Reg'
date
. A.
rell, Reactor'ngineer, OPS, OB, ORS, Region I R.
W. Winters, Reactor Engineer, MPS, EB,
,
DRS, Region I Approved By:
William V. Johns
, Acting Director Division of Reactor Safety date date 3304270199 830413 PDR ADOCK 05000244
Ins ection Summar
Special announced inspection to determine readiness for restart on February 22-26, 1988 (Report No. 50-244/88-04)
Areas Ins ected:
A special team inspection was conducted to evaluate the licensee's program for controlling outage activities.
Particular emphasis was placed on the licensee's systems for completion and closeout of activities to assure plant readiness for restarts Five areas of outage activities were inspected.
These were 1) modifications; 2) operations; 3) maintenance and inservice testing; 4) inservice inspection; and 5) steam generator maintenance and surveillance.
SUMMARY:
The inspectors concluded that control of outage activities through various systems, such as Nonconformance Reports (NCRs), Maintenance Work Requests (MWRs), Procedure Change Notices (PCNs), etc.
appeared to be effective and that systems were in place to ensure activities that could affect plant safety would be completed prior to plant restart.
The major strength identified during the outage was the Liaison Engineers for various outage activities.
These individuals have the primary responsibility for monitoring and conducting modification and other outage activities.
They were found to be well qualified (e.g.
each Liaison Engineer held an operating license)
and displayed a thorough knowledge of the activities for which they were responsible.
The major weakness identified during the inspection was a large backlog of modification packages requiring closeout.
Approximately 88 packages dating back to 1984 were still open.
Most of the activities required for closeout of these packages are administrative in nature but include such activities as updating lower tiered as-built drawings.,
However, no incomplete activities dir'ectly affecting plant operability were identified.
The root cause of this problem is apparently inadequate staffing.
Early in the inspection the inspectors identified some areas where improvements in housekeeping could be made and improvements were noted during the inspection.
One violation was identified with regard to lack of documentation for certification of quality control visual inspection personnel.
In addition, two previous violations regarding temporary structures (87-09-01)
and inoperability of fire suppression systems (87-08-01)
and an unresolved item (87-04-05)
regarding the test tag control program were close DETAILS Persons Contacted kC
- J
- D C.
- A.
J.
G.
AD A.
"A.
N.
- T
- G L.
J.
Alexander, Liaison Engineer Anderson, Quality Assurance Manager Bodine, Manager, Nuclear Assurance Manager Bryant, Quality Control Engineer Catalano, Maintenance Department.
Curtiss III, Manager, Materials Engineering
& I Derleth, Pipefitter Eng, Liaison Engineer Fi lkens, Health Physics 8 Chemistry Supervisor Gellert, Nondestructive Examination Technician Goetz, Manager, Construction Engineering Goodenough, Technical Assistant Harding, Modifi.cation Liaison Coordination Henry, Nondestructive Examination, Level III Hubbard, Training Specialist Huff, Maintenance Training Manager Jonas, Staff Assistant nspection Services
- G. Link, Manager, Electrical Design
"T. Mar low, Maintenance Manager K. Masker',
Liaison Engineer M. Meleca, Instrument and Control Technician K. Nassauer, Quality Control Supervisor J. Neis, Liaison Engineer J.
Neopak, Construction Engineer T. Newbury,'esponsible Engineer
- R
- R
- G
- T C.
M.
M.
"M.
M.
- S
- J J.
J.
- J p.
- R Marchionda, Ginna Training Manager Mecredy, Director of Engineering Services Meier, Division Training Manager Meyer, Superintendent Support Services Pham, Engineer Rioch, Liaison Engineer Ruby, Shift Supervisor Saporito, Supervisor Materials Engineering Labo Schuller, Operations Manager Smith, Liaison Engineer Spector, Superintendent Ginna Station St. Martin, Station Engineer Stewart, Construction Supervisor Sweet, Temporary Modification Wahl, Manager Mechanical/Preventive Maintenance Widay, Technical Manager Wilken, Liaison Wood, Supervisor Nuclear Security ratories, ISI
United States Nuclear Re viator Commission
"C. Cowgill, Chief Reactor Projects Section lA, PB1, DRP
- C. Marshall, Senior Resident Inspector, Ginna Station
- N. Perry, Resident Inspector, Ginna Station
" Oenotes those attending the exit meeting.
The inspectors also contacted other administrative and technical personnel during the inspection.
2.0 Objective and'co e of Ins ection The objective of this inspection was to evaluate the licen'see's program for controlling outage activities.
Particular emphasis was placed on the licensee's systems for completion and closeout of activities to assure plant readiness for restart.
Five areas of outage activities were inspected.
These were:
1) modifications, 2) operations; 3) maintenance and inservice testing, 4) inservice inspection, and 5) steam generator maintenance and surveillance'nspectors assigned to each of these areas reviewed the progress of specific wor'k activities and nonconformances to ensure that they were properly tracked and resolved.
Specific inspection activities included:
Review of field change requests and their closeouts Review of nonconformance reports and their closeouts Review of gA/gC findings and their closeouts Review of post modification and post maintenance testi'ng procedures Observation and review of post modification and post maintenance testing Confirmation of configuration control Review of post modification and post maintenance training for licensed and nonlicensed personnel Rev'iew of procedure changes and implementation Review of On Site Review Committee activities Physical inspection of plant hardware Review of final closeout packages Since some outage activities were still in progress during the inspection, it was not possible to confirm final closeout of each item.
However; the status of items within the work control system was determined to assure
they had progressed properly through the system up to the time of the inspection and that proper procedures were in place to assure closeout prior to plant restart.
In addition, documentation for activities from the previous outage were reviewed to confirm proper closeout.
Section 3.0 of this report addresses the licensee's outage modification program including a description of the modification management program and inspection of specific modification activities.
Section 4.0 discusses the operations department interface with outage activities, Section 5.0 addresses maintenance and inservice testing activities, Section 6.0 addresses inservice inspection activities and Section 7.0 addresses the licensee's steam generator maintenance and surveillance program.
A summary of the inspection scope and findings is provided with the report cover sheet.
3.0 Ins ection of Outa e Modification Activities This section addresses
.outage modification activities.
Section 3. 1 describes the licensee's modification management program and section 3.2 discusses the NRC inspection of specific modification activities.
3. 1 Modification Mana ement Pro ram The Ginna outage organization is managed by an outage coordinator who reports to the plant superintendent.
Three assistant coordinators were assigned to the outage coordinator during this outage.
The
.
managers of construction, maintenance, technical engineering, and other site departments, including Nuclear guality Assurance report to the outage coordinator.
Prior to the outage, the coordinator working with the Ginna staff develops the plans and schedules for modifications, maintenance, and other outage activities.
The detailed planning and implementation of outage work is the responsibility of the assigned engineer.
For plant modifications the Liaison Coordinator, who reports to the Technical Engineering Manager,, is responsible for assigning the Liaison Engineers to the modification program.
The. Liaison Engineers provide the interface between the project organization and the station staff to ensure that the various organizations such. as engineering, maintenance, operations, procurement, construction, etc.
complete their portion of the modification as assigned.
The liaison engineer is responsible for coordinating, monitoring, and controlling activities related to his assigned modification and for assuring that the scheduled modification is completed and verified ready to be turned over to operations.
To obtain this control the Liaison Engineers have the following responsibilities:
Provides the operating staff organization with a description of the modification and provides the outage schedul Prepares a Preliminary Plant Operating Review Committee review of design inputs for operability, maintainability, testability, and ALARA criteria.
Also provides construction and materials engineering comments during the preliminary Plant Operating Review Committee review.
Establishes a Modification Follow Group (MFG) which includes representatives of each affected plant group.
The Modification Follow Group is comprised of individuals assigned from groups within the station to provide plant support for a particular modification.
The Modification Follow Group is directed by a Liaison Engineer representing a specific modification, however, the Liaison Engineer does not have administrative authority over the Modification Follow Group members, he does have the responsibility to assign tasks and to ensure that Modification Follow Group department management and Station management are aware of those tasks that are not being performed in a timely manner.
The Modification Follow Group is responsible for.the performance of their assigned activities as established by the Liaison Engineers.
The Liaison Engineers coordinate the review of drawings, specifications, and other design information for operability, maintainability, testabi lity, and ALARA criteria with plant personnel.
The Liaison Engineers are responsible for preparing and maintaining the Station Modification files, procedures, Field Change Requests, Station Work Authorizations and assuring that Corrective Action Requests (CAR) and Nonconformance Reports (NCR) that are written are closed in a timely manner.
In addition to the specific tasks described above the Liaison Engineers have the overall responsibi.lity for ensuring the the Modification Follow Group has reviewed the completed modification and have accomplished all items (procedure changes, training, interim as-builts, testing, tagging, etc.)
necessary for operation of the modification.
The completed data modification package is presented to the Plant Operating Review Committee by the Liaison Engineers for review and approval prior to turnover to operations.
Modification Control S stems Deviations or noncompliances to drawings, specifications or materials are identified using Nonconformance Reports, Field Change Notices, Corrective Action Requests, Procedure Change Notices or. Maintenance Work Request and Trouble Report.
These systems are described in the site guality Assurance Plan.
The Liaison Engineers are responsible for assuring that all changes to the original modification have been reviewed by the responsible Modification Follow Group personnel and that any NCRs, CARs, etc are dispositioned before presenting the
package to the Plant Operating Review Committee for their final review.
The Plant Operating Review Committee review of the modification package consists of reviewing remaining open items to assure that they do not impact system, operability and that the system can be turned over to operations.
As i ndicated above, the Liaison Engineer is assigned responsibility for tracking and verifying disposition of Nonconformance Reports, Corrective Action Requests and other work control forms.
The items are tracked manually by the Liaison Engineers.
A list of issued Nonconformance Reports is maintained by the Nuclear Assurance Group and a Nonconformance Report status is issued to the Plant Superintendent monthly.
Also, operations has a tracking system for Maintenance Work Requests and Procedure Change Notices.
Integrated tracking systems were not found to exist for lower tier documents such as Drawing Change Requests which are tracked by the Liaison Engineer as part of each modification package.
Communications The Liaison Engineers through the Liaison Engineer Coordinator, establishes the Modification Follow Group which generally consists of the following organizational personnel:
Liaison Engineer and backup Respon'sible Engineer Construction Enginee'r Operations Engineer Maintenance Training Instrument and Control Health Physics equality Control Fire and Safety Coordinator Planning Personnel The original meeting held by the Liaison Engineers defines the description of the modification and a general description of the proposed modification schedule.
The results of the meetings are documented by the Liaison.Engineers and form a part of the modification data packag Trackin of Commitments to NRC Open items,, violations and commitments made to the NRC, Region I are tracked by the Plant Superintendent's assistant.
Commitments made to the NRC, Office of Nuclear Reactor Regulation are tracked by the Corporate Engineering office.
These commitments are identified to the Plant Superintendent and assigned by the Liaison Coordinator to a
Liaison Engineer to be followed as part of their normal work load.
The 1988 NRC. startup commitments in addition to work related to DC fuses (EWR 3341) consisted of a change to certain control board pushbutton colors and a boric acid flow control modification (EWR 4375).
The inspector verified that the Station Modifications EWR 3341 and EWR 4375 were listed on the "Liaison/Construction Assignment for Active/Planned Modifications" memo and tracked by a Liaison Engineer.
The DC fuse program modification EWR 3341 was in progress during this inspection period, but appeared to be approximately 90%
completed.
The control board pushbutton color task was not listed in the EWR program, but was followed by a Liaison Engineer and identified in the master tracking system.
Daily outage meetings are held with key members of the licensee staff and the Liaison Engineers has daily meetings with the Modification Follow Group.
3.2 Ins ection of Modification Activities Sco e of Ins ection The licensee performed 35 major system modifications during this outage.
Five of these modifications were reviewed in detail:
1) the Steam Generator Blowdown System modification; 2) the Boric Acid
.Piping Replacement modification; 3) the A and B Battery Room Ventilation modification; 4) the Standby Auxiliary Feedwater Pump Cell Switch - Buses 14 and 16 modification; and, 5) The Emergency Diesel Generator Fuel Oil System Upgr'ade.
In addition, a modification completed prior to the o'utage was reviewed:
Salem Trip Breaker modification.
These modifications were reviewed to determine" the a'dequacy of the licensee's modification program.
For each modification the inspectors performed portions or all of the following:
Reviewed the licensee's administrative procedures used for controlling the modification design, review, installation, and turn over phases.
These were reviewed against appropriate regulatory requirements and standards.
Reviewed the modification packages for compliance with administrative and regulatory requirements.
Those items reviewed in the modification packages were the design criteria,
Cl
'safety analysis, installation and work procedures, testing procedures, installation drawings, Facility Change Requests, Engineering Change Notices, Procedure Change Notices, Nonconformance Reports, punchlists for tracking open items, gC surveillance reports, and the Plant Operating Review Committee review of the design input.
Reviewed the management controls used by the Liaison Engineer for monitoring the modification installation and close out.
Walked down portions or all of the modifications installed at the time of this inspection in order to verify that installation procedures and controls were being followed.
Held discussions with the Responsible Corporate Engineer, Liaison Engineering Coordinator, Liaison Engineer, Construction Engineer, Contractor Supervisor, contractor and gC personnel.
Those specific documents reviewed in making this assessment are identified in Attachment A.
Descri tion of Modifications Ins ected Steam Generator Blowdown S stem SGBS Modification The purpose of this modification was to improve the reliability of the SGBS and reduce the potential fo'r blowdown recirculation to the Steam Generators.
The modification was performed as part of the licensee's plant betterment programs.
The modification, as described in Engineering Work Request (EWR) 4324, involved replacing the SGBS heat exchanger with an existing flash tank based process, Because of the high temperature changes involved, the reliability of the heat exchanger'
tubes could not be assured.
By converting to a flash tank process, these high temperature
.changes would be transferred to the flash tank and associated piping.
In addition, the new piping installed to accomplish this conversion was sized to reduce erosion and corrosion concerns which had been identified.
Boric Acid Pi in Modification The purpose of this modification was to correct a number of leaks which had occurred in the piping used to carry 12% boric acid and to improve the system configuration.
The modification was performed as part of the licensee's plant betterment program.
As discussed=-in EWR 3092, the modification involved replacing all schedule 10 piping carrying 12% boric acid with schedule 40 piping and rerouting the eight, inch Safety Injection System pumps suction piping to avoid passage through-high radiation areas.
This modification was being performed over several outages
~
During this outage the eight inch piping between the boric acid header pipe and the suction side of the Safety Injection System pumps was replace Batter Room Ventilation Modification This completed modification installed a
new electric heating coil in the battery room air conditioning supply air ductwork to control battery room temperature within 74 + 3 degrees.
Standb Auxiliar Feedwater Pum Cell Switch - Buses 14 and
The purpose of this modification is to install a test cell switch to the D and C Standby Auxiliary Feedwater Pump logic. This modification was designed to modify the existing Standby Auxiliary Feedwater Pump pump start logic so that the pump can be started when the corresponding Auxiliary Feedwater Pump breaker is in the test position or when the breaker is removed from the circuit.
Emer ency Diesel Generator Fuel Oil S stem U
rade The Emergency Diesel Generator fuel oil system upgrade modification is, at the time of this inspection, 80% complete.
The modification included the following tasks:
Installation of flanges in the Day Tank fill and recirculation solenoid valves (SV 5907, 5907A, 5908, and 5908A).
Flushing and hydrostatic testing of drain connections were installed on the, new piping.
New load control and alarm switches were installed on the Emergency Diesel Generator Day Tank.
Local mechanical type pressure instruments were installed on the.
Emergency Diesel Generator Fuel Oil Pump discharge and a delta pressure instrument was installed on the duplex suction strainer.
Salem Tri Breaker Modification The purpose of this modification was to modify the reactor trip breakers in order to prevent an incident similar to Salem's Anticipated Transient Without Scram (ATWS).
This mo'dification allows redundant trip methods of the breakers on an automatic trip signal and was performed in three phases
- modification of the reactor trip breaker; modification of the reactor trip bypass breaker; and the addition of a test light to the reactor trip breaker.
The modification was completed on February 9,
1987.
At the time of this inspection the system had been turned over to operations but the modification package had not been closed ou ~Findin s
Based on the review of the above modifications, the following strengths and weaknesses of the Ginna Modification Program were identified:
~Stree ths Although final close out of modification packages takes a relatively long time period to accomplish, those modification packages which have been closed out appeared to be complete in all areas.
This included revisions to yll as-built drawings, copies of procurement records, complete sets of vendor drawings, etc.
Those'iaison Engineers interviewed displayed thorough knowledge of the modifications for.which they were responsible, had current lists of open items which needed to be closed or justified prior to turnover, and were deeply involved with the day to day progress of the modification.
To assist the Liaison Engineer in keeping track of all his responsibilities, a comprehensive Modification Management Form is used.
This form identifies all new and existing procedures and drawings which need to be developed or revised, all testing and maintenance activities which must be addressed, all training requirements which need to be satisfied, all vendor manuals which need to be obtained, and all other activities which must be accompli, shed, such as component labeling, FSAR update, close out of punchlist items, etc.
Based on discussions with contractors, licensee engineers and management it appeared good communications existed at all levels'icensee construction and liaison engineers were found to be highly knowledgeable of the status of the modifications for which they were responsible.
Through reviews of the job site modification packages it was evident that the responsible headquarters engineers were directly involved in the resolution of all engineering problems which arose during construction of the modification.
Weaknesses One case was identified where the Liaison Engineer was not fully using the Modification Follow Group to identify and monitor the status of procedures and drawings which needed revision.
Administra-tive Procedure A-203, Ginna Modification Project Organization, Section 3.4.S requires the Liaison Engineer to establish and assign tasks to the Modification Follow Group in support of the modifica-tion.
Included in those tasks required of the Modification Follow Group was the identification of procedures and drawings requiring revision.
The inspector found examples where no documentary evidence existed to show that the Modification Fol,low Group had identified those type A and 8 drawings which needed revision.
However, the
inspector determined that the subject drawings had been identified for revision by the responsible Corporate Engineer.
The time to close out the modification packages was excessively long.
As of the date of this inspection, there were 88 modifications where the system had been turned over to the plant but where the modifica-tion packages were still awaiting close out.
Some of these packages had been awaiting close out since 1984.
The average time to close out these packages was 18 months.
One of the reasons for this backlog is the delay in revising the appropriate type B Control Configuration drawings.
This is considered a weakness in the modification program in that Type B Control Configuration drawings may not be accurate if needed during an emergency.
4.0 0 erations Interface with Outa e Activities
~Sco e
During outages, the licensee maintains configuration and procedural control, in part, by using Nonconformance Reports (NCRs), Maintenance Work Request/Trouble Reports (MWRs), and Procedure Change Notices (PCNs).
The licensee identifies and provides training on modifications made during the outage.
In some cases, outage maintenance is controlled by hand written procedures developed by maintenance personnel which incorporate Engineering Change Notices (ECN) or other approved procedures.
An audit was conducted of the tracking systems for NCRs, MWRs and PCNs to evaluate the effectiveness of the management control systems.
A review was conducted of the process for determining the content of post outage training programs for licensed and non licensed personnel and'a classroom training session was observed.
Two maintenance activities were observed to evaluate the effectiveness of the interactions between different departments and to evaluate proper procedural compliance.
The training and certification of the individuals involved with maintenance were reviewed.
~Findin s
Nonconformance Re orts NCR Nonconformance Reports are issued and handled in accordance with administrative procedure A-1502.
Outstanding NCRs are tracked by the plant guality Assurance department and a status report is issued monthly which identifies the outstanding NCRs.
The January status report identified fifty five NCRs that had been open for over three months.
Twenty eight of these NCRs required a plant outage in order for.them to be closed out.
None of the remaining twenty seven NCRs appeared to affect system operability.
Fifteen of these outstanding NCRs were reviewed to determine their present status.
With the exception
of one NCR (G-87-254), all NCRs reviewed had a clear delineation of the responsibility for closure.
The licensee issued a Corrective Action Request to effect closure of NCR G-87-254.
The NCR tracking system appears to be'dequate.
However, the number of NCRs over three months old and those NCRs identified as being planning or document related, represent a significant backlog of unresolved deficiencies.
It should be noted that the open NCRs are related to the backlog of unclosed modification packages.
Maintenance Work Re vest/Trouble Re orts MWR The MWR form, which is also referred to as a trouble card or work order, is used by the operations and maintenance departments to correct minor facility problems or to identify potential problems which require review.
The MWR program enables timely actions to, be taken and allows recommended solutions to be reviewed by the shift supervisor and appropriate work group supervisor.
The MWRs are tracked on a computer based system main-tained by the Maintenance department.
A review of the MWR tracking log for MWRs which were being worked and of MWRs which had been completed identified no discrepancies.
the tracking program appears to be adequate for controlling minor plant maintenance'rocedure Chan e Notices PCN A PCN'is initiated to make either a temporary or permanent change to a
procedure or to a station modification procedure.
The PCN is reviewed by a responsible manager, plant staff, duty engineer or shift supervisor depending on the type of procedure and the type of change.
All PCNs are reviewed and approved by the Plant Operations Review Committee.
The Technical Engineering department tracks their assigned PCNs under two numbering systems, one for changes initiated by engineers and a second for changes initiated through the shift supervisor.
The Maintenance depart-ment tracks their assigned PCNs under two additional numbering systems, one for permanent changes and one for chan'ges initiated through the shift supervisor.
All PCNs are reviewed by the PORC and the dispositions are documented in the PORC meeting minutes.
Over 150 PCNs, which were initiated through the shift supervisor, were audited to ensure proper resolution and documentation in the PORC meeting minutes.'ll PCNs were
.
properly resolved and documented with only 'one minor administrative error.
The management control system appears to be adequate for tracking PCNs from initiation to final disposition.
Post Outa e Trainin The Training department maintains a standing member on the modification follow group to identify modifications being conducted during the outage.
A brief summary of each modification is prepared and presented to the various curriculum committee The Operations Manager chaired the curriculum committee meeting for licensed and nonlicensed training on January 29, 1988.
The committee decided which modifications would be included in the post outage training session.
Classroom and simulator training was pr'ovided to all operators prior to the end of the outage.
One hour of classroom training was observed and was found to provide adequate information to the operators on the plant, modifications.
The Mechanical Maintenance, Instrument and Control, and Electrical supervisors chair separate curriculum committees after the plant reaches 100
% power.
Each committee reviews the brief summaries of the outage modifications and decides which modifications will be covered in future training sessions.
Although training is provided after the completion of the outage, members of the plant staff would be familiar with each modifi-cation due to their involvement in the installation of the modification and due to the distribution of modification summaries to each shop during the outage.
It appea)
s that adequate training will be provided on outage modifications.
Conclusions The T'raining department has developed a computerized tracking system for identifying and following all modifications which affect the simulator.
The Training department is expanding the computerized system to include identification of training material that is affected by plant modifications.
The computerized tracking system was found to be effective.
Maintenance The.replacement of a resistance temperature detector (RTD) by I8C technicians was observed.
The job was well planned and was performed efficiently.
The job required the support of operations, health physics and quality control personnel.
Procedural compliance and departmental interfaces appeared to be adequate.
The removal and replacement of service water'ontainment isolation valves by the mechanics was also observed.
A hand written proce'dure for the overall job was prepared by the maintenance foreman which detailed the gA check points, the maintenance procedures and the engineering change notice (ECN) steps required to perform the work.
The cognizant engineer was present during the maintenance and concurred in some deviations from his ECN procedure which was not modified to reflect the deviations'he maintenance foreman planned to annotate the procedure after the completion of the work to identify the procedural deviations.
Procedural steps for removal of piping snubbers were not signed as the steps were completed.
This did not adversely affect the activity.
However, unapproved deviations from procedures and failure to promptly annotate completion of procedural steps could lead to improper maintenance and the reduction in availability of safety related equipmen Maintenance work appeared to be conducted in a controlled manner with effective divisional interfaces.
However, strict procedural compliance was not always followed.
The training and certification of site individuals performing maintenance tasks are documented by an incumbent review which was performed.to meet the requirements of the facility proposed INPO training program.
The incumbent review was an evaluation by supervisor s of the experience, train-ing and competence of all their personnel.
In the future, classroom and on the job training will be formally documented.
The IEC technicians were qualified, by the incumbent review, to perform the RTD replacement, how-ever, the gA inspector had to explain to the technicians how to use an 0-r ing insertion tool during the replacement of the RTD.
Nonfaci lity personnel who support outage activities are drawn from Rochester Gas and Electric fossil stations and from outside contractors.
There are no training or certification records maintained for nonplant personnel even though some workers have been supporting Ginna outages for several years.
It should be noted that this was not the case for steam generator main-tenance personnel who do receive significant documented training.
Non-facility personnel are directly supervised by Ginna facility personnel who do have documented certifications.
Although no examples were found, the lack of documented training and experience of nonfacility personnel could result in"a situation where individuals are assigned tasks for which they may have insufficient training or experience.
The approval and implemen-tation of an INPO accredited training program should eliminate this potential problem.
Conclusions Even though some concerns were identified, the training and certfication of maintenance workers appears to be adequate'he licensee's programs for tracking NCRs, MWRs and PCNs are adequate.
However, the number of NCRs open for over 3 months represent a significant backlog of unresolved deficiencies.'he licensee's training programs on plant modifications were adequate.
However, the implementation of the proposed INPO accreditation program for maintenance personnel should result in improvements in the training and ti aining documentation of maintenance workers.
The maintenance work which was observed was conducted in a controlled manner with effective divisional interfaces.
It appears an increase in staffing would reduce the backlog of outstanding NCRs and expedite the implementation of INPO accreditation programs.
5.0 Maintenance and Inservice Testin Sco e of Ins ection During this outage, the licensee performed Inservice Testing (IST) on:
I) 12 mechanical and 3 hydraulic snubbers; and, 2) 30 motor operated valves (MOVs) in the Motor Operator Valve Analysis and Test System (MOVATS)
Program.
The Safety Evaluation Summary Reports were prepared under the
Engineering Work Request (EWR) 2512 for snubbers and EWR 4539 for MOVATS testing.
Activities related to the testing of the three mechanical and two hydraulic snubbers and six MOVATS tested MOVs were selected for inspection.
~Findin n
Snubbers Test Pro ram In accordance with Technical Specification Section 4. 147, 10% of the snubbers are to be.selected and functionally tested during each refueling outage.
Technical Specification 3. 13. 1 requi res that all snubbers be operable prior to the reactor being made critical.
In compliance with Technical Specifications 3. 13. 1 and 4. 14.7, 10% representative samples of mechanical and hydraulic snubbers were select'ed for this outage, twelve mechanical snubbers and the three hydraulic snubbers.
Additionally, a number of hydraulic snubbers that were previously inspected and found deficient also were functionally tested during this outage.
The licensee initiated a Seismic Upgrade Program (EWR 2512) in January 1986.
As part of the program a number of snubbers were changed to mechanical snubbers.
A proposed amendment to the Technical Specifications to include mechanical snubbers in Technical Specifications Table 3. 13. 1 is currently under NRC review.
The licensee established a formal Snubber Inspection and Test Program described in procedure ME256.
The documented program was reviewed and the inspector had the following findings:
Acceptance criteria for operability as defined in ASME Section NI, in the Technical Specifications, Vendor Technical Manuals and Industry Standards were satisfied.
Responsibil.ities of plant personnel for performing the functional tests are well defined.
Closeout actions'to assure all requirements and commitments are tracked on the licensee's Plan-0-Log P-2 system.
Methods and instructions for functional testing are prescribed and maintained current.
Provisions are available to verify that the functional tests are complete, reviewed, approved and documented and that any lessons learned are applied to future testing.
A shock and sway suppression tester, owned by the licensee and operated by qualified licensee personnel, is used for functionally testing the hydraulic snubbers in accordance with procedure M-40.8.
A validator (device used for validating test results),
bought by the licensee, was
used by a contractor to perform the functional tests for the mechanical snubbers.
Licensee personnel are being trained in the use of the validator in accordance with licensee procedures ME256, M-40. 11 and M-40. 12.
These procedures and other documentation reviewed are listed in Attachment A.
A matrix for the Mechanical Snubber Test Schedule for 1988-1999 was discussed with the inspector.
The Mechanical Snubber Inspection Criteria categories list was used to determine the schedule.
The criteria includes environment, sizing, thermal movement, and other criteria.
In addition to the testing pf snubbers described above,'team generator snubbers CB 6, 7 and 8 were rebuilt and passed their Ten Year Program tests.
Also tested were snubbers reported deficient by walkdown ins'ervice inspections, trouble cards and nonconformance reports.
The inspector confirmed with plant personnel that snubbers declared operable, with any outstanding items such as Nonconformance Reports, trouble cards, etc.=
would not affect snubber operability.
Test packages for snubbers AFU-3, FWV-47, AHV-30 were reviewed and found complete in accordance with procedure ME-256 and the individual M-40.8 (Hydraulic) and M-40. 11-(Mechanical)
snubber test procedures.
The test packages reviewed were being routed for final review by the Mechanical Foreman, Maintenance Supervision, gC Supervision and the Plant Operating Review Committee.
The inspector observed snubber velocity testing conducted in accordance with paragraph 3.6 of procedure M-40.8.
The tester failed
'ts calibration tests and a Maintenance Work Request and Trouble Report was prepared to have the tester calibrated.
The test technician also prepared a procedure change notice to revise paragraph 3.6 to agree with the vendor technical manual.
The inspector noted that the vendor technical manual used for the test was not a controlled document.
A controlled vendor technical manual was provided.
The inspector confirmed that both the controlled, and uncontrolled vendor technical manuals were the same.
A quality control inspector was present and verified some of the test results.
The inspector observed a technician preparing the results of the tests for hydraulic snubber PS-9 in accordance with MWRTR 88-0475 and Nonconformance Report G-88-073.
The technician explained that contrary to the Nonconformance Report there was nothing wrong with the snubber.
The preparer of the NCR erroneously read a red indicator on the snubber as a deficiency.
The Mechanic'al/Preventive Maintenance Analyst confirmed the technician's explanation.
The inspector toured the auxiliary building to observe some of the installed snubbers.
One snubber (SWV370) was temporarily disconnected at one end of the support.
Maintenance personnel showed the inspector that procedure M-93 provided assurances that the
snubber SWV370 would be reinstalled correctly prior to the system being declared operable.
Motor 0 erator Valve Anal sis and Test S stem MOVATS Pro ram In response to IEB 85-03, the licensee has established a documented MOVATS Program.
The program uses a signature analysis technique, which has been proven effective in diagnosing the overall performance of Motor Operated Valves (MOYs).
Engineering, Work Request (EWR) 4539 provided the Design Criteria and Safety Evaluation Report for the safety related MOV program.
For this outage, all 30 MOVs in the MOVATS program were to be tested.
At the end of this inspection,
MOVs had been tested.,
The inspector determined the following based on review of the program:
Acceptance criteria for operability of MOV switch settings were established as discussed in IEB 85-03.
'
Responsibilities are defined in procedure M-64-12 for performing MOVATS.
These responsibilities include planning, training and qualification, conducting, controlling, post maintenance testing and closeout activities.
Step by step procedures existed for testing, recording data and preliminary evaluations of the test results.
Procedures included verifications to ensure that the tests were properly performed, documented, reviewed and approved.
Trained and qualified l-icensee and contractor personnel performed the MOVATS testing observed and reviewed by the inspector.
The licensee has its own equipment to perform MOVATS, and also uses mockups to train and qualify licensee personnel.
Appropriate team members have received advance MOVATS training.
Planning and scheduling of MOVATS is the responsibility of the Liaison Engineer in the Technical Department.
The records of the MOVATS program were reviewed and found to be satisfactorily'aintained, filed and stored.
MOVATS files for MOV 871 AEB, 3505A, 4013 and 9701A were reviewed and found complete in accordance with procedure 64. 1.2 and their individual test procedures, e.g.
T-64.871 A&B and T-64.3505A.
Section VIII in each of MOV test procedures contains status of NCRs
,
PCNs, Material Requisitions and trouble cards written against the MOV during the.tests.
The disposition of all these items was not complete at the time of this inspection.
However, the Liaison Engineer was well aware of the items not completed and cognizant of the need 'for their disposition prior to establishing system operabilit During the inspection the inspector observed post maintenance testing of MOV 871A.
The post maintenance testing was successfully conducted in accordance with test procedure T-64-871A.
At the end the post maintenance testing, a
QC inspector prepared a
QC Report+QCR 88-0506)
regarding the motor leads not being labeled.
Maintenance is reviewing the QCR.
When the QC inspector was questioned by the NRC inspector regarding his training and qual.ification to survey MOVATS work, he responded that he had not received specific training on MOVATS.
However, he indicated'hat his job was to survey (watch)
and ensure procedural adherence and that he would do this based on his electrical experience.
The QC supervisor confirmed that this is the licensee policy.
During a plant tour, the inspector noted that the motor was removed from MOV-3996.
Further investigation indicated that MOV-3996 experienced a locked rotor during a
MOVATS test on January 6,
1988 (NCR G 88-002).
The motor was removed because of "solder melted on the motor armature and it was impregnated with oi 1/grease".
MOV-3996 had failed MOVATS testing on December 21, 1987 for back seating problems'(NCR G 87-334)
and on December 23, 1987 for over thrust problems.
A maintenance engineer investigated the locked rotor problem and found that the torque switch caused the failure.
The documented evidence did not show that the back seat and over thrust problems were also considered in the investigations The licensee, however, assured the inspector that these problems were considered and they did not contribute to the locked rotor problem.
The motor was repaired and was being reinstalled.
Conclusions The licensee has documented its Snubber Inspection and Test Program and its MOVATS Program.
These programs are being implemented during this outage.
The licensee's staff and qualified contractor personnel performed these programs.
Qualified contractors are also used to train and give licensee personnel experience in these activities.
Evidence of management involvement includes snubber validators bought for future training and a
MOVATS mockup built to conduct training of licensee personnel.
No violations or deviations were identified in the review of these program areas.
6.0 Inservice Ins ection Sco e of Ins ection The licensee performed inservice inspection during this outage to comply with requirements of the ASME Boiler and Pressure Vessel Code,Section XI, and with its inservice inspection schedule for the 1988 outage.
The licensee additionally performed examinations in accordance with its maintenance inservice inspection examination schedule (plant erosion-corrosion inspection program) for the second interval 1988 outag The following areas were selected for inspection:
Examination data related to the steam generator lower head-to-tubesheet weld No.
LHTSW-B, control rod drive housing No.
and 32, pressurizer lower shell vertical weld No. V-2, and the reactor pressure vessel head-to-flange circumferential weld No.
RVP-HFW.
Inservice inspection related nonconformance reports Visual examination personnel qualification/certification records Maintenance inservice inspection plan (plant erosion - corrosion inspection program)
~Findin s
Inservice inspection is mandated by the ASME BEPV Code,Section XI, and the Code edition applicable to a specific facility is identified in 10 CFR 50 '5a(g)
based upon the issuance date of its construction permit.
The Ginna facility is committed to the 1974 edition through the Summer 1978 Addenda.
The inspector determined that the examinations represented by the reviewed data met the applicable code and regulatory requirements regarding test method, recording and evaluation of results.
The nonconformance reports regarding ISI items indicated that the disposition of each item was in progress although not completed at the time of the inspection.
The gA/gC department acts as a coordinator for NCRs in that it (1) provides a unique number for each NCR, (2) submits the NCR to the department responsible for providing a disposition, (3) assures that the dispositioned NCR is assigned to the responsible department for completion, and (4) maintains a
file of closed NCRs.
The inspector. traced the open NCRs through the system and verified that guality Control and Maintenance records (NCR logs arid status sheets)
were complete and identified the current status of each item.
The ASME Section XI, 1977 Edition through Summer 1978 Addenda references ANSI N45.2.6, 1973 for the qualification of personnel performing visual examination VT-2, VT-3, and VT-4 of IWA 2212, IWA 2213, and IWA 2214, respectively.
In addition to invoking the qualification provisions of ANSI N 45.2.6, 1973, ASME Section XI, IWA 2300 requires that personnel performing nondestructive examinations using methods not covered by SNT-TC-1A shal'l be trained and qualified to comparable levels of competency by subjection to comparable examinations on the particular method involVed. It further requires that the practical examinations shall be performed using procedures and parts representative of the Owners plant.
ANSI N45.2.6, 1973 specifies the minimum work experience an
'ndividual must have to be considered for certification, and requires that records shall be maintained which include the individual's work experienc Records of the ISI group visual examination personnel were found to be complete.
These personnel were qualified and certified in accordance with procedure QRLS 911, Revision 1, which was found to meet the intent of ANSI N45.2.6, 1973 Edition.
Work experience records of certified personnel were maintained and identified daily activities in which an inspector participated and time spent in each activity on an hourly basis.
Practical examinations were performed using an ISI Department mockup comprised of parts representative of the plant and which contained various types of defects.
In addition to the ISI personnel, certain Quality Control inspectors were certified by the licensee as Level II visual inspectors qualified to perform VT-2, VT-3, and VT-4.
These individuals were identified as having been qualified to the provisions of the applicable sections of procedure A1002, Revision 8, Qualification of Inspection Personnel.
The inspector found that evidence of each individual's work experience to support the certification was not part of the record, and he was advised that the information was not available.
The inspector further found that the practical test required by the certification process was not performed using parts representative of the Owner's plant as required by ASME section XI, but were performed by answering test questions which referenced sketches of parts.
Failure to establish certification records as required by ANSI N45.2.6, 1973 and failure to conduct practical certification examinations consistent with the ASME Code,Section XI requirements for visual inspectors is a violation (50-244/88-04-01).
Concern regarding erosion - corrosion in balance of plant piping systems has been heightened as a result of the December 9,'1986 feedwater line rupture that occurred at Surry Unit'.
This event was the subject of NRC Information Notice 86-106 issued on December 16, 1986 and its supplement i ssued on February 13, 1987.
The licensee's erosion - corrosion program for this outage included approximately 350 scheduled examinations of which 45 were not completed at the time of this inspection.
Acceptance criteria are included in procedure MDG-9, Revision 1,
Secondary System Erosion Corrosion Guidelines and the minimum acceptable wall thickness is'calculated using an equation found in, ANSI B31. 1, 1980.
Examination results identified approximately 90 components which will be scheduled for reexamination in 1989.
Examina-tion of steam extraction components identified 10 items which were replaced this year, and examination of items associated with the MSR 2nd pass to the 5B high pressure heater resulted in the replacement of 11 items including short sections of associated piping.
7.0 Steam Generator Maintenance and Surveillance Sco e'f Ins ection The inspector reviewed the eddy current test data collected during the current outage, the water chemistry results for the preceding year, and
current outage, the water chemistry results for the preceding year, and the radiation records for the work done in the inspection and repair of the steam generator tubes during this outage.
Secondar Water Chemistr Control
~Sco e
This inspection was performed to determine the licensee's compliance with the requirements of the Technical Specifications for inservice inspection and the Steam Generator Owners Group and EPRI recommendations for water chemi stry.
The methods of analysi s or operation of the chemi stry laboratory were outside the scope of this inspection.
Ginna Station's steam generators are Westinghouse Series 44 vertical shell, U-tube type heat exchangers that operate with a recirculating design.
They are rated at 3,130,000 lbs/hr.
steam flow at 725 psig.
The steam generator tubing is mill annealed Inconel 600 conforming to ASTM SB-163-61T.
The 3260 tubes are partially rolled into the tube sheet and seal welded, leaving an approximately 19 inch deep crevice in the 22 inch thick tube sheet.
The support plates are carbon steel of the drilled hole design.
Initial secondary water chemistry control recommended by the steam generator manufacturer was phosphate buffering control.
In November 1974 the secondary water chemistry control was converted to all volatile treatment and efforts made to assure that the cation conductivity was kept as low as achievable.
Water Chemistr Table 1 summarizes the water chemistry history at Ginna station for the past 13 years.
TABLE 1 Water Chemistr Histor at Ginna'Station Parameter 1974-1977 1978 1981 1985 1987 (1)
EPRI Guide Cat Cond.,
umhos Chloride, ppb Sodium, ppb Silica, ppb pH
.7-2.5
5-15 20-50 8.6-9.0
.2-.4
5-15 15-30 8.7-8.9
.12-.2 3-5 3-8 5-10 8.7-8.9
.09-.13 4-6 1-2 5-10 8.7-8.9 01.2 1-12 1-4 0.8
20 300 8.5-9.2 Single values are maximum (1) At steady power operation
Conclusions The results of the licensee's efforts to control water chemistry have been effective in meeting the requirements of the Owners Group and EPRI. Since 1977 the control has been within the guidelines during periods of operation above -30% power.
The licensee maintains close contact with the Steam Generator Owners Group and EPRI to keep abreast of developments to improve secondary water chemistry quality.
Edd Current Testin of Steam Generator Tubes I
Eddy current testing was performed using multifrequency eddy current examination techniques on both the the "A" and "B" recirculating steam generators.
These are Westinghouse Series-44 design, containing 3260 Inconel Grade 600 tubes having a nominal outside diameter of 0.875" and a
nominal wall thickness of 0.050".
The eddy current testing was performed by licensee personnel that have been trained and qualified in the eddy current examination method.
Support services such as opening and closing the 'generators, installing the manipulator, and probe changing were also performed by licensee personnel.
These support personnel are volunteers from other licensee divisions such as the Gas Construction and the Transmission and Distribution Divisions.
Most of these individuals have been performing these functions for a number of outages.
The licensee maintains two steam generator mockups for training purposes.
These are used for refresher training when the support crews come on site and for training in any special techniques that might be required.
One such technique was the method used by the licensee to inspect for loose parts by using a camera mounted on a sled and pulled through the generators.
The licensee also maintains accountability of tools and equipment,to control loose parts.
In interviews with crew leaders the inspector was told that each of these crew leaders had a minimum of five years experience in this work and one had worked at eight outages.
The scope-of inspections performed by.the licensee is as shown in Table 2.
TABLE 2 Sco e of Edd Current Ins ection Location Steam Generator
'8 Hot Leg - inlet Sleeves inlet Full Length Hot Leg - inlet to 1st support 100%
to 1st support 10%
40%
to 6th support 100%
10%
40%
20%
Data collection for the eddy current testing was controlled from a remote trailer using video equipment to watch the operation thereby reducing radiation exposure.
The licensee also videotaped virtually all of the operations performed during the inspection including installation of the dams, tube marking, plug installation, and plug expansion, as well as the eddy current test operations.
The data were first analyzed by computer to determine if any indications existed, this analysis was then printed and verified by a review by the licensee's trained analysts.
All indications reported by either the computer or the analyst as being greater than 20% through wall were then reviewed by the licensee's Level III analyst.
Any differences between the Level III and the original analyst were resolved through discussion between these individuals.
Eddy current testing revealed that steam generator
"A" had a total of eight tubes that required plugging due to defects and steam generator
"B" had thirty tubes plugged due to defects.
In addition, seven tubes in "A" and eighteen in "B" steam generator were plugged due to the inspection for anti-vibration bass (AVB) location determination or as a conservative
'action to box tubes without AYB support.
A summary of the results of the last three inspections including the cause of apparent tube degradation is given in Table 3.
"A" now has 150 tubes plugged and
sleeved and steam generator
"B" now has 332 tubes plugged and 281 sleeved.
TABLE 3 TUBES REQUIRING CORRECTIVE ACTION Cause 1986 Steam Generator 1987 Steam Generator 1988 Steam Generator A
B
A B
SCC IGA WASTAGE OTHER No AVB Support
23
8
2 (2)
19
72 4O (3)
18 (4)
(1)
Plugs replaced with sleeves in 3 tubes (2)
Cause not reported (3)
Crevice indications either SCC or IGA also one leaking sleeve was plugged (4)
Plugged with'sentinel plugs.
"B" had 8 tubes plugged with sentinel plugs to box tubes without AVB suppor The improved water chemistry appears to have effectively reduced the frequency of stress corrosion cracking and intergranular attack except in the area of the tubesheet crevices The licensee believes the cause of the continuing intergranular attack in steam generator B is related to the early operating procedures used in startup and hot shut down.
During the early life of the plant, the plant was heated up.and brought to hot shutdown conditions by using the A Reactor Coolant Pump and temperature control by steaming off the A Steam Generator while keeping the B
Steam Generator bottled up until starting the second reactor coolant pump and criticality of the reactor.
During this method of operation, a deeper sludge pile was formed in the A Steam Generator and significantly different chemical conditions were found in the sludge piles of the A
and B Steam Generators.
During this period and f'r several years after conversion to all volatile treatment in November 1974, wastage and SCC were experienced in the tubesheet crevice and sludge pile of the A Steam Generator.
After conversion to all volatile treatment, the occurrence of these defects was arrested.
The B Steam Generator did not experience these defects.
However, the B Steam Generator began to experience intergranula>
attack in these regions several years after conversion to all volatile treatment was started.
This condition is still apparent.
Conclusions The licensee performed the inspection in accordance with the Technical Specification requirements.
The training provided to employees from other divisions of the company was adequate.
Having two mockups of the steam generator assists in this training.
Use of licensee employees lends continuity to the inspections"and provides a high level of experience at Ginna.
Radiation Ex osures Radiation exposures were reviewed as part of the steam generator inspection and maintenance inspection.
The methods of collecting and verifying the accuracy of these exposures was not included in the scope of this inspection.
Initial surveys of the steam generators showed the average fields in the generator plenum area prior to the start of work were as follows:
'A'B Units are R/Hr Hot Side Cold Side 10.2 4.8 11.0 The licensee reported that at the time of the inspection radiation exposures to the individuals involved with the steam generator inspection and maintenance was as shown.,in Table
~
This total was based on the pocket dosimetry and is not the official record.
The largest item of exposure not included in this total was the removal of the dams in steam generator
"B".
TABLE 4 Units are Man-Rem
~0eration Steam Generator Opening Generator Initial Survey Nozzle Oam Installation Eddy Current Testing Tube Plugging Sludge Lancing All Other Operations (both units)
0.66 1.09 0.06 0.06 3.78 2.68 9.2 4.66 4.48 22.15 6.93 4.64 33.71 Total 92.0 Conclusions The higher exposure in steam generator
"B" for the tube plugging operation was the result of the significantly higher number of tubes that required plugging (7 in "A" - 58 in"B").
8.0 Licensee's Actions on Previous NRC Concerns Closed Unresolved Item 87-04-05
- Licensee to evaluate the need to document'removal of test equipment when tests are delayed.
The licensee issued corrective action report (CAR) No.
1798 to investigate and correct the concern identified in IE 87-04-05.
The CAR identified the
.
following areas that had to be addressed'to resolve the concern of IE 87-05-04 and the findings of the licensee.
I Review the test tag control program requirements with Lead Test Supervisors',
and Review the Test Tag Control Program requirements with maintenance personnel On July 6, 1987, meetings were held with the.Lead Test Supervisors to inform them of the problem and discuss with them actions to take to correct the problem of tagging equipment.
A training course was held on September 23, 1987 at which time the requirements of procedure A-1103,
Test Tag Control Program, were discussed with the class.
The training attendance record reviewed by the inspector included approximately sixty licensee personnel from all areas of their organization.
The course given to the personnel discussed the concern referenced in IE 87-04-05 and the requirements of procedure A1103.
During this inspection period, the inspector discussed the requirements of procedure A-1103 with various craft personnel and determined they had received the September 23, 1987 training and understood the concerns of IE 87-04-05.
The inspector verified that the present tag control program had corrected the problems identified in the NRC unresolved item of IE 87-04-05.
This item is closed.
Closed Violation 87-08-01 On March 16, 1987; fire suppression,,
systems in areas necessary for safe shutdown were rendered inoperable for approximately one hour and twenty minutes without continuous fire watch with backup fire suppression equipment being stationed in these areas.
On March 16, 1987, a major portion of the Fire Detection and Automatic Suppression System was found inoperable by the licensee.
The areas affected were the auxiliary building, intermediate building, diesel building, turbine building, containment and control room.
It was determined that this condition had existed for one hour and twenty minutes.
The event was caused by a nonlicensed operator's failure to follow procedure SC-3. 16.2.4, Fire Signaling System/Component Disconnections-Reconnections.
The procedures precautions identify the significance of dep'ressing the alarm off button.
Prior to the violation procedure SC-3. 16.2.4 had been modified to include extra signoffs as a result of LER (84-010);
however, the licensee's corrective action only addressed the reconnection portion of the procedure and not the disconnection portion.
The inoperable condition discovered March 16, 1987 was caused by the nonlicensed operator when performing the disconnection portion of pro-cedure SC-3. 16.2.4.
On July 9, 1987 the licensee revised procedure SC-3. 16.2.4 to require system verification by a licensed operator or knowledgeable Fire Control and Safety personnel.
The inspector discussed procedure SC-3. 16.2.4 with nonlicensed and licensed operators and fire safety personnel and determined them to be knowledgeable of the latest revisions of the procedure and the specific problem that occurred on March 16, 1987.
Based on the procedure revision and the training given to site personnel the inspector concluded that this item is closed.
Closed Violation 87-09-01
- Administrative Procedure A-1406. 1, Installation and Removal of Temporary Structural Features, steps 3.4.2 of A-1406. 1 states that the Technical Manager or alternate shall document his review providing a determination whether the temporary installation will:
1) result in a change to the facility or its operation as described in the Safety Analysis Report, 2) provide a change to the plant Technical Specifications, or 3) involve an unreviewed safety question.
Contrary to
the above, Administrative A-1406. 1 was not followed, in that, the required documented review was not attached to the following temporary structural features authorization forms:
1.
From April 12 - October 9, 1986 scaffolding was installed in the auxiliary building over safeguards bus 14.
2.
From November 17 - December 19, 1986, scaffolding was installed in the control room above and around the main control board while at power.
3.
From April 7 Hay 8, 1987, scaffolding was installed in the auxi liar y building over bus 14.
To correct the condition identified in violation 87-09-01, the licensee issued Corrective Action Report (CAR) No.
1790 which required the" following actions:
a review of the A-1406 series procedures and changes to provide clearer direction on proper documentation of reviews an upgrade of the review documentation for activity authorizations perform a guality Assurance audit of the program for control of temporary modifications to assess the effectiveness of the changes'he licensee issued Procedure Change Notice (PCN)
No. 87-5220 to revise the A-1406 seri'es of procedures which did the following:
added a requirement for independent review enhanced the detail 'portion of the documentation added clarification in justifying documentation involving credit being taken for operability of alternate equipment Incorporated criter'ia in response to AFCAR 86-12 RD The following procedures that were revised to include the above requirement are listed below:
A1406, Control of Temporary Hodifications A-1406-1. 1, Control of Temporary Lead Shielding A-1406.2, Installation and Removal of Temporary Fluid System Provisions The inspector reviewed the above procedures and verified that the criteria of PCN No. 87-5220 had been incorporated into the A1406 procedures.
Audit No. 87-68 was performed on the temporary modifications program and verified that the recommendations of CAR No.
1790 had been implemented.
The audit report described the actions. taken by the Temporary
'
Modifications Supervisor (TMS) to bring his program in line with the latest changes of the A-1406 series of procedures.
Inspection of selected temporary modifications performed during this outage by the inspector indicated that the revised requirements of A-1406 procedures were complied with during this outage.
This item is closed.
9.0 Unresolved Items Unresolved items are matters about which more information is required in order to ascertain whether they are acceptable items or violations.
Unresolved items are discussed in paragraph 8.0.
10.0 Mana ement Meetin s
Licensee management was informed of the scope and purpose of the inspection at the entrance interview on February 22, 1988.
The findings of the inspection were discussed with licensee representatives during the course of the inspection and presented to licensee management at the February 26, 1988 exit interview (see paragraph 1'for attendees).
At no time during the inspection was written material provided to the licensee by the inspector.
The licensee did not indicate that proprietary i.nformation was involved within the scope of this inspection.'
ATTACHMENT A Documents Reviewed Procedures Maintenance Inservice Inspection Plan (Plant er'osion - Corrosion Program)
Maintenance Inseryice Inspection Plan Examination Schedule for Second Interval, 1988 Outage A-25.1 Ginna Station Event Report A-102.9 Maintenance Training Program A-203, Rev 7, 3/4/87 Ginna Modification Project Organization A-301, Rev 20, 9/27/84 Control of 'Station Modifications A-301. 1, Rev 4, 11/18/87 Station Modification Classification and Review A-301.3, Rev 2, 2/19/87 Station Modification Installation and Acceptance A-503 Plant Procedure Adherence Requirements A-601. 1 Procedure Control - New Procedures A-601.2 Procedure Control - Permanent Change A-601.3 Procedure Control - Temporary Changes
'A-601.4 Procedure Control - Periodic Review A-1002, Rev 8, 12/8/86 gualification of Inspection Personnel A-1101, Rev 11, 1/27/88 Performance of Tests A-1406 Temporary Modifications A-1501 Control of Nonconformance Items A-1502 Nonconformance Report A-1601 Corrective Action Report A-1603 Maintenance Work Request and Trouble Report (MWR)
c
ATTACHMENT A M-40.8, Rev 13 Functional Testing of Hydraulic Snubber,
"B" Steam Generator Inner Sleeve Removal and/or Split Ring Assembly Removal South Handhole
"A" Steam Generator Secondary Side Handhole Cover Removal North Handhole M-43.2.9, Rev
M-43.27, Rev 8 M-43.29, Rev
"A" Steam Generator Final Cleanup and Inspection M-37.21, Rev ll, 5/23/86 Inspection and Repair of Main Steam Check Valves M-43.30, Installation of Tube Lane Blocking Device (TLBD) for
"A" Steam Generator North Handhole M-43.33, Rev
Examination and/or Removal of Objects from Steam Generator Secondary Side Through Inspection Holes "A" Steam Generator M-43.33, Rev
Examination and/or Removal of Objects from Steam Generator Secondary Side Through Inspection Holes "B" Steam Generator M-43.54, Rev 8 Steam Generator Sludge Lancing (RG5E System)
M-72. 1 Replacement Primary Loop RTD M-93, Rev
Removal and Reinstallation of Seismic Pipe Supports MDG-9, Rev 1, 8/3/87 Secondary System Erosion - Corrosion Guidelines ME-256, Rev 0, 2/9/88 Snubber Inspection and Test Program ME-64. 1.2, Rev 3, 2/13/88 MOVATS Testing of Limitorque Motor Operated Valves QR&S-911, Rev 1, 2/10/88 Qualification of Visual Examination Personnel RGE-66, Rev
S/G Blowdown and Blowdown Recovery System T-64.826A, Rev
MOV-826A Isolation and Restoration for MOVATS Testing T-64.871A, Rev
MOV-871A Isolation and Restoration for MOVATS Testing T-64.3505A, Rev
MOV-3505A Isolation and Restoration for MOVATS Testing
e J
ATTACHMENT A Modification Packa es EE 80, Rev 0, 12/19/85 Installation, Inspection and Testing Specification for Non Class 1E Electrical Equipment, Cable and Raceway
.EET 150, Rev 1, 1/15/88 Preoperational Test Specification S/G Blowdown System EWR -
4324 EWR 2512 Seismic Upgrade Program EWR 3092 Modification Management Form EWR 3092, Rev 0, 5/6/86 Design Criteria Boric Acid Piping EWR 3092, Rev 0, 5/20/86 Safety Analysis Boric Acid Piping EWR 4323 A and B Battery Ventilation EWR 4324, Rev 1, 1/6/88 Design Criteria Ginna Station S?G Blowdown System EWR 4324, Rev 1, 1/6/88 Safety Analysis Ginna Station Steam Generator Blowdown System
/
EWR 4324, 1/19/87 Modification Management Form EWR 4324A Test Plan EWR 4269 SAFW Pump Interlocks EWR 4526 D/G Fuel Oil System EWR 4539 Design Criteria Ginna Station Safety Related Motor Operated Valve
'rogram SM 3092. 12, Rev Rev 0, 2/5/88 Boric Acid Piping Upgrade, Phase,II, Electrical Removals, Reconstruction, and Relocation SM 3698. 1, Rev 0, 3/6/85 Reactor Trip Breaker Upgrade SM 3698.2; Rev 0, 3/27/85 Reactor Trip Breaker Preoperational Testing SM 3698.3, Rev 0, 3/1/86 Reactor Trip Bypass Breaker Upgrade SM 3698.4, Rev 0, 3/8/86 Reactor Trip Modification Testing SM 3698.5, Rev 0, 2/6/87 Test Light Addition to Reactor Trip Breaker SM 3698.6, Rev 0., 2/9/87 Testing of Reactor Trip Breaker Test Light Addition I
k t
ATTACHMENT A SM 3797. 1, Rev 0, 2/2/87 MRPI Installation Outside Containment SM 4324. 1, Rev 0, 1/7/88 Pipe Support Installation for Steam Generator Blowdown System Modification SM 4324.2 Piping Installation for Steam Generator Blowdown System Modification SM 4324.5, Rev 0, 2/17/88 Steam Generator Blowdown System Modification Functional Testing SWAR 4324.2 'emporary Change Work Authorization Request WP 4324. 1, Rev
Work Procedure of Steam Generator Blowdown System-Pipe Supports WP 4324.2 Work Procedure for the Steam Generator Blowdown System Piping WP 4324.4 Work Procedure for Installation of Instruments and Airline I
Tem orar Modification Status 88-19 A Emer.gency Generator Start Air Scaffold 88-25 Containment Purge Supply Blank Flange Scaffold 88-33 A Lance Platform (A Loop)
88-34 B Loop Platform (B Loop)
~Drawia a
DWG 33013-1235, Rev 1, 12/85 Condensate System DWG 33013-1277, Steam Generator Blowdown System, - P&ID DWG 33013-1982, Rev
Steam Generator Proposed System Modification -
P8(ID DWG 33013-1986, Rev 1, Steam Generator Existing System Modification PB ID Other Documents Curriculum Committee'eeting Minutes, January 28, 1988 NCR Status Report for the Month of January Plant Operations Review Committee Meeting Minutes, February 10, 1988 gA'anual, Appendix C, Rev 8 Ginna Station Inservice Pump and Valve Testing Program - January 1,
1981 through December 31, 1989 Period