IR 05000259/1990029
| ML18033B556 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 11/06/1990 |
| From: | Carpenter D, Kellogg P, Patterson C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18033B554 | List: |
| References | |
| 50-259-90-29, 50-260-90-29, 50-296-90-29, NUDOCS 9011190187 | |
| Download: ML18033B556 (50) | |
Text
UNITED STATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323 Report Nos.:
50-259/90-29, 50-260/90-29, and 50-296/90-29 Licensee:
Tennessee Valley Au hority 6N 38A Lookout Place 1101 Market Street Chattanooga, TN 37402-2801 Docket Nos.:
50-259, 50-260, and 50-296
'License Nos.:
DPR-33, DPR-52, and DPR-68 Facility Name:
Browns Ferry Units 1, 2, and
Inspection at Browns Ferry Site near Decatur, Alabama Inspection Conducted:
September 16 - October 15, 1990 ra Dm Inspector:
C. A. Patterson, NRC Restart Coordinator D.
R. Carpenter, Manager for Hodifications and Unit 3
~'<
~'/
~ P ul Kellogg, Section Chief, Inspection Programs, TVA Projects Division S~h.
'ate Signed SUMMARY Scope:
This routine resident inspection included surveillance observation, maintenance observation, operational safety verification, essential design calculations, power ascension testing, restart test program, electrical cable installation, SPOC, reportable occurrences, action on previous inspection findings, TYiI Action Items, modifications and Unit 3, and site management and organization.
PD~
Okey yo)20 ADOCI; n;OOO2~~
F5 Results:
A violation was identified for failure to comply with procedures for an adequate hold order boundary while conducting modifications and maintenance work activities, paragraph 4.
Several recent events have occur red in this area.
One example resulted in an unplanned ESF actuati on when a worker cut into energized electrical leads.
An unresolved item was identified concerning deletion of ECN/DCNs, paragraph 13.
During a
two week period in September 1990, the licensee deleted
ECN/DCNs from the Unit 2, Cycle 5 required list.
Eight of these. were termed NRC commitments.
An IFI was identified for two drywell penetrations that appeared to have been leaking, paragraph 9.
This was identified during a final system walkdown in RWCU heat exchanger
'room.
An IFI was identified concerning inadequately identified and trained conductors in safety related panels, paragraph 9.k.
This. was observed during a system walkdown in the diesel generator rooms.
General plant housekeeping was poor during this inspection period, paragraphs 9.d,e and 9.g.
During the system return to service final walkdowns some areas were observed not to have been cleaned up after completion of the modifications work.
Other general area walkdowns found that housekeeping Has poo'r.
The licensee management acknowledged this concern and took action to improve housekeeping.
A memorandum was issued to all modifications employees.
An incident investigation was initiated to formally review and correct the problem.
The backlog of work orders continues to be reduced, paragraph 3.
The maintenance department has made steady progress reducing the total from over 7000 in May 1989 to 2610 in September 1990.
The timeliness of performing preventive maintenance has also improve Cl
REPORT DETAILS 1.
Persons Contacted Licensee Employees:-
"0. Zeringue, Site Director
"L. Myers, Plant Manager M. Herrell, Plant Operations Manager.
J.
Rupert, Project Engineer R. Johnson, Modifications Manager
- R. Jones, Operations Superintendent A. Sor rell, Maintenance Manager G. Turner, Site guality Assurance Manager
"P. Carier, Site Licensing Manager
,*P. Salas, Compliance Supervisor J.
Corey, Site Radiological Control Manager R. Tuttle, Site Security Manager Other licensee employees or contractors contacted included licensed reactor operators, auxiliary operators, craftsmen, technicians, public safety officers, and quality assurance, design, and engineering personnel.
NRC Personnel:
"P. Kellogg, Section Chief
"C. Patterson, Restart Coordinator D. Carpenter, Manager for Modifications and Unit 3
"E. Christnot, Resident Inspector W. Bearden, Resident Inspector
"K. Ivey, Resident Inspector
"G. Humphrey, Resident Inspector
"Attended exit interview Acronyms used throughout this report are.listed in the last paragraph.
Surveillance Observation (61726)
The inspectors observed and/or reviewed the performance of required SIs.
The inspections included reviews of the SIs for technical adequacy and conformance to TS, verification of test instrument calibration, observa-tions of the conduct of testing, confirmation of proper removal from service and return to service of systems, and reviews of test data.
The inspectors also verified that LCOs were met, testing was accomplished by qualified personnel, and the SIs were completed within the'equired frequency.
Three of the SIs reviewed during this reporting period were:
( I) Standard Calibration Instruction, SCI-204, Rev.
7, Standard Calibration Instruction For General Electric Type 555 and 557, (2)
Instrument Maintenance
0
Surveillance Instruction, IMSI-3014, Rev.
0, Diesel Generator Storage Tank Level Channel Calibration, and Electrical Maintenance Instruction, EMI-46, Rev.
6, Freeze Protection.
The inspector found the documentation for the portion of the work effort to be correct and procedures were being followed'.
No violations or deviations were identified in this area.
Maintenance Observation (62703)
Plant maintenance activities were observed and reviewed for selected safety-related systems and components to ascertain that they were conducted in accordance with requirements.
The following items were considered during these reviews:
LCOs were met; activities were accom-plished using approved procedures; functional testing and calibrations were'erformed prior to returning components or systems to service; gC records were maintained; activities were accomplished by qualified personnel; parts and materials used were properly certified; proper tagout clearance procedures were followed; and radiological controls were imple-mented as required.
Work documentation (NR, WR, and WO)
was reviewed to determine the status of outstanding jobs and to assure tnat priority was assigned to safety-related equipment maintenance which might affect plant safety.
The plant maintenance department continues to make good progress toward reducing the backlog of maintenance work orders.
The progress is as follows:
Nay 30, 1989 Nay 29, 1990
'ept.
11, 1990 Startu W.O.
4469 1905 1606 Total 7335 3156 2610 The site goal for WO backlog is 600.
Continued progress has been made on timely performance of preventive maintenance items.
For late PMs the industry median is 4.3.percent.
TVA goal is to be less than or equal to 4.7 percent and only 1.9 percent were overdue in September 1990.
No violations or deviations were identified in this area.
Operational Safety Verification (71707)
The NRC inspectors followed the overall plant status and any significant safety matters related to plant operations.
Daily di scussions were held with plant management and various members of the plant operating staff.
The inspectors made routine visits to the control rooms.
Inspection observations included instrument readings, setpoints and recordings, status of operating systems, status and alignments of emergency standby systems, verification of onsite and offsite power supplies, emergency power sources available for automatic operation, the purpose of temporary
tags on equipment controls and switches, annunciator alarm status, adherence to procedures, adherence to LCOs, nuclear instruments operability, temporary alterations in effect, daily journals and logs, stack monitor recorder traces, and control room manning.
This inspection activity also included numerous informal discussions with operators and supervisors.
General plant tours were conducted.
Portions of the turbine buildings, each reactor building, and general plant areas were visited.
Observations included valve position and system alignment, snubber and hanger condi-tions, containment isolation alignments, instrument readings, house-keeping, power supply and breaker alignments, radiation and contaminated area controls, tag controls on equipment, work activities in progress, and radiological protection controls.
Informal discussions were held with selected plant personnel in their functional areas during these tours.
a.
Control Room Layout The inspector reviewed the'ontrol room layout planned for Unit 2.
Individual work station areas are planned for the UOs, ASOS, STA, and SOS.
Tne SOS will be in tne center of the combined Unit I and
control rooms.
The implementation time for these changes is under consideration.
The inspectors reviewed the detailed control room design review project efforts for Unit 3.
Several restart CRDR items are being worked for Unit 2.
A major human factors reshuffling of switches and indicators on the control room panels wi 11 be implemented when Unit 3 restarts.
This effort is being conducted in the licensee training facility utilizing a photograph mockup or simulator of the control room panels.
b.
Failure to Control Work Activities.
SDSP-7.9, Integrated Schedule and Work Control, establishes the responsibilities for work control and the work control process at Browns Ferry.
Section 6.4. 1 requires that an impact evaluation be performed on all activities that have potential for affecting equip-
, ment operations which may affect tFie safe and reliable operation of the unit except as specifically exempted.
Attachment C to SDSP-7.9 is used to document that evaluation.
Section 7.0 specifies the Tagging and Support Coordinator's duties in reviewing, correcting, and issuing hold order boundaries.
SDSP-14.9, Equipment Clearances, provides for the protection of employees and plant equipment while work on or around electrical or mechanical equipment.
Section 6. 1 provides both general and special requirements for removing equipment from service to allow work to safely procee The inspectors noted several recent events which were associated with failure to adequately control work activities withi n the documented requirements of both of the above procedures.
During the SPOC walkdown on System 92, Neutron Monitoring, licensee personnel identified that a clearance'ag associated with Hold Order 2-89-871 was installed on panel 2-25-14.
Although the tag specified that leads were lifted in the panel, no lifted leads could be found.
This resulted in issuance of URI 260/90-27-02.
As the result of this event, the licensee reviewed all outstanding clearances that were associated with lifted electrical leads.
The licensee informed the inspector that 32 clearances of this nature were identified of which
had no discrepancies.
The clearance problem described below was identified during this licensee audit:
On September 6,
1990, during an audit of the equipment clearance tags performed as the result of the above review, the licensee identified that the C
EECW Strainer Drain Valve, O-FCV-67-8, had been replaced under Workplans 0057-90 and 0050-90 (occurred in July-August 1990)
with a
new valve and actuator with the electrical leads landed.
In addition to these events the inspector noted that two additional recent events have occurred which are related to this subject.
The events are as follows:
On August 25, 1990, during an effort to lift Hold Order 2-90-419 to allow system testing, wires were found unterminated for valve 2-FCV-74-100.
The licensee's investigation determined that Workplan 0252-90 was not complete and that modifications electrical craft personnel had been allowed to perform work without appropriate supervision having signed onto an existing hold order.
On September 16, 1990, an unplanned ESF occurred when the D D/G automatically. started while a
lead to level indicating switch, Z-LIS-3-58D, was undergoing Raychem splicing.
The splicing activities were performed while the associated circuits were thought to be deenergized.
The Impact Evaluation Sheet asso-ciated with the work order had incorrectly specified the work as RPS related with Hold Order 2-90-571 listed.
The actual circuits were ECCS related instruments rather than RPS.
This resulted in the work activity being attempted without adequate protection under an equipment clearance.
This event resulted in a
NRC reportable occurrence in accordance with 10 CFR 50.72.
The inspector discussed these events with members of plant management.
The inspector was informed that although the licensee considered their program sound, steps had been taken to conduct sensitivity training for modifications personnel concerning the clearance procedure.
This training was being conducted by an
experienced SOS.
These events constitute a failure to adequately control work activities which wi 11 be Violation (260/90-29-01)
Failure to Control Work Activities.
URI 260/90-27-02 is closed.
5.
Essential Design Calculation (37700)
The inspectors reviewed various completed essential calculations from the licensee's Calculation Cross Reference information System printout.
A brief summary review of the selected licensee's calculations for suitability of system and equipment performance was performed by the inspectors.
The selected calculations were reviewed for a partial check of mathematical equations for errors, to determine adequacy of the licensee's methodology and approach.
Additionally the calculations were reviewed to verify that the respective assumptions and inputs were current and valid and that those assumptions and inputs reflected the controlling condition with results reasonable considering the inputs, method, and objectives.
Any inputs or assumptions requiring later verification were verified to be identified and tracked.
Those reviewed were as follows:
Calculation, CD-Q2003-885437, Reactor Pressure Vessel Water Level instrument Line From Reactor Pressure Vessel Connection 12A (Seismic Analysis).
This calculation was performed by the licensee to document the piping analysis for 1 inch and 2 inch sensing lines from the reactor pressure vessel nozzle, N12A, to anchor point located outside the drywell in the reactor building.
No discrepant areas were noted.
b.
Calculation, CD-Q003-894604, Seismic Qualification of Valves 2-FCV-3-188A and B.
The calculation justifies seismic qualification of valves 2-FCV-3-188A and B which exceeded standard allowable acceleration limits and allowable nozzle loads.
No discrepant areas were noted.
Calculation, CD-Q0248-891321, Seismic Evaluation of Battery Charger Support.
The calculation documents structural integrity of the battery charger to the spare shutdown board (Units
and 2 diesel generator).
No discrepant areas were noted.
Calculation ED-Q0000-884-89, Appendix R - Tabulation of Equipment Power Supplies and Detailing Criteria for the Auxiliary Power System.
This calculation was performed for a
detailed criteria of the auxiliary power system that will satisfy the BFN applicable require-ments of
CFR
Appendix R.
No discrepant for Unit 2 restart areas were noted.
e.
Calculation, ED-00219-87313, Thermal Overload Heater Calculations-480V Diesel Auxiliary Board A.
The TOLs at BFN were not documented.
This calculation was performed for ECN E-1-P7009 to determine the design and evaluate the installed TOL heaters.
However this calcula-tion was voided and superseded by calculation ED-Q29999-890158, Thermal Overload Heater Calculations - Continuous Duty Motors.
This
I calculation was issued to supersede all or portions of 13 TOL heater calculations.
No discrepant areas were noted.
f.
Calculation BFEP-E2-86135, Review of Reactor Building Harsh Environment Cables Required for System
MCEL Equipment Function.
This calculation was performed to analyze and evaluate the Reactor Building Harsh Environment Cables to determine if the cables are required for the associated components to perform their Master Component Electrical List functions.
No discrepant, areas were noted.
g.
Calculation ED-N2068-890304, Cable Size Verification of Replacement/
Rerouting Cables.
This calculation was performed to verify that replaced and/or rerouted cables and new cables scoped in DCN H5860 were adequate for ampacity, voltage drop, short circuit and Appendix R high and low impedance faults considerations.
No discrepant areas were noted.
h.
MD-f0074-87017, RHR Pump Operation without EECW Cooling of Seal Heat Exchangers.
This calculation was intended to provide justification for RHR pump operation without EECW flow to the seal heat exchangers for all operazing modes of the RHR system.
Additionally tne calcula-tion would serve as a technical basis for maintaining RHR operability whenever EECW flow to the seal exchangers is unavailable.
The effects on service life of pump mechanical seals during pump operation without seal exchanger EECW flow (provided continuous seal flow is maintained) is evaluated for various modes of operation for process temperatures up to and including 300 degrees F.
The inspector noted that the results of this calculation are consistent with a letter from GE to TVA dated August 14, 1986, which states that even in worse case failure of the pump seals, leakage would be limited to 25 gpm, which is small compared to the pump capacity, and would have negligible effects on the ability of the RHR pumps to accomplish their safety function.
No violations or deviations were identified in this area.
6.
Power Ascension Testing The inspectors reviewed the licensee's reported status in preparation for the Power Ascension Test Program.
Procedures for this program have been completed and approved except for four which are:
(1) 2-TI-190, Thermal Expansion, (2) 2-TI-180, Cooldown from Outside the Control Room, (3)
2-TI-130, Pressure Regulator, and (4) 2-PMT-201, Drywell Vibration.
Walkdowns for the expansion points used for thermal expansion were in progress during the reporting period.
The licensee reported that 72 of 99 points have been walked down.
In addition, some efforts have been performed to evaluate and tune the hydraulic control systems on the turbines for the main turbine generator and the reactor feedwater pump A third major effort has been in progress to get the Transient Analysis Recording System (TARS) installed to record the vital data during restart.
One-hundred and nine signals ha've been powered up and 100 of the 117 have undergone preliminary amplifier calibration and are ready for final calibrations.
A General Electric Transient Analysis Recording System (GETARS) is also to be installed and that effort remains at the end of this reporting period.
No violations or deviations were identified in this area.
Restart Test Program (37701)
RTP-065 identified a
lack of stack effect.
Consequently, a
major modification was issued for the stack, DCN W11053 which was implemented by WP 0467-90.
The specific work activity observed involved the terminations of cable conductors at the various damper position limit switches, the damper solenoid valves, and terminal boards.
The inspector also observed the installation of electrical splices in the manholes C
and D.
The inspector noted during the ongoing work activity observation that manhole C was subjected to flooding during rain storms.
The inspector inquired about the qualification for submergence of the splices installed in manhole C.
The licensee initiated 2 DCN 14294 which adequately addressed this item.
All work activities were supervised, the WPs were at the work area, and all activities were observed by gC Inspectors.
No violations or deviations were identified in this area.
Electrical Cable Installation (37701)
During the preparations for fuel reconstitution in the summer of 1988 the licensee discovered that the electrical duct bank between the Units 1/2 DG building and the SBGT building had collapsed.
The licensee issued a major modification W10725 to install a
new duct bank routed from the top of the Units
and
DG building to the top of the SBGT building.
This will facilitate running new cables to SBGT systems Trains A and B.
During this and the previous reporting period the inspector observed the installation of rebar, concrete forms, and conduit.
The inspector also observed the concrete pours.
Several WPs were involved in implementing this modifica-tion.
The inspector reviewed WP 2393-90, Cable Pull to Train A;
WP 2399-90, Cable Pull to Train B;
and WP 2400-90, B Train Cable termina-tions.
All work activities were supervised, the WPs were at the work area, and all activities were observed by gC inspectors.
No violations or deviations were identified in this area.
System Pre-operability Checklist The inspectors continued to monitor the licensee's activities to evaluate and upgrade both plant equipment and documentation as necessary to insure that plant systems are in compliance with applicable standards and commitments to meet their required functions.
Of the
systems to be
evaluated and upgr'aded as necessary, 31 were indicated as completed by the licensee as of October ll, 1990.
Those systems reviewed by the inspectors during this reporting period are listed as follows:
a.
Auxiliary Boilers (System 12)
The system checklist was compl.eted on August 5, 1990.
An inspector reviewed the completed SCL on August 20, 1990, and identified no deficiencies.
The inspector noted that no'xceptions or deferrals were taken for this system.
Raw Cooling Water (System 24)
An inspector accompanied the systems engineer on a final walkdown of this systems'nterfaces with the EECW system on September 27, 1990.
No deficiencies were identified during the walkdown.
The system engineer noted that hanger work being performed on the interfaces were being tracked under the EECW system.
The SPOC for this system was completed on October 5, 1990 C.
120V DC Battery Distribution (System 57-1)
An inspector accompanied the systems engineer on a final walkdown of this system on September 28, 1990.
Members of the operations and maintenance groups were also present.
Only minor work and house-keeping items were identified.
The inspector noted that the undocumented splices found during the preliminary walkdown of the system (see IR 90-27)
were still in place in the battery chargers.
The system engineer noted that the splices were vendor installed and the licensee had no plans to replace them.
This issue is being tracked by an IFI issued in IR 90-27.
The inspector noted that modifications involving fuses and fuse holders were still ongoing in one DG panel.
The SPOC for this system was completed on October 8, 1990.
d.
Emergency Equipment Cooling Water (System 67)
Resident inspectors accompanied the cognizant systems engineers on a
preliminary walkdown of this system on September 25-27, 1990.
Members of the operations and maintenance groups were also present.
Numerous minor work items and general housekeeping items were identified.
Various portions of the system were obstructed by scaffolding and ongoing work throughout the plant.
These areas were not covered in detai
during thi s walkdown and will" require particular attention during the final walkdown
~
Several modifica-tions remain to be completed before the SPOC can be performed.
The SPOC for this system is forecast for completion on October 17, 199 Reactor Water Cleanup System (System 69)
On October 2,
1990, the inspector observed portions of the final walkdown for this system.
Observations were conducted in the RWCU heat exchanger room, RWCU valve operating room, main steam tunnel, and RWCU pipe chase.
In general, the housekeeping in the areas was poor.
The areas had not been cleaned up and various debris was in the areas.
Scaffolding was still in place in most of the areas.
In the RWCU heat exchanger room, there was a wall mounted phone that was pulled from'he wall and lying on the floor.
Insulation was stacked in one corner of the room.
The system engineer stopped the planned starting of a RWCU pump due to a large amount of concrete dust in the room which might be drawn into the motor.
The inspector discussed these items with the Plant Manager, Site Director, and a'odifications Supervisor.
A modifications supervisor was assigned to monitor housekeeping and direct the removal of unnecessary scaffolding.
However, most scaffolding was to be left in place until system testing is complete to prevent having to erect scaffolding again for inspections or possible corrective maintenance.
During the tour of the RWCU heat exchanger room, the inspector observed two drywell penetrations that appeared to have been leaking.
One was for reactor water level instrument lines, and the other was a
penetration which was not being used.
The penetrations were 2X15 and 2X28.
This will be an inspector followup item IFI 260/90-29-03, Leaking Drywell Penetrations.
The penetrations were shown to the system engineer.
Reactor Building Closed Cooling Water (System 70)
The inspector reviewed the results of the RBCCW tube inspections discussed in IR 90-25 for heat exchanger 2A.
The licensee completed the inspections on 2B and the spare heat exchanger on Unit 1, 1C.
The results of the inspections and plugging are as follows:
HTX 1C 2A 2B PLUGGED
163*
59%
PREVIOUS PLUGGED
8
TOTAL PLUGGED
171
TOTAL BUNDLE (740)
<I 23.11 9.73
" Thru wall cracks greater than 60%
The licensee is developing a corporate plan to retube 'the heat exchanger g.
RHR (System 74)
An inspector accompanied the system engineer on a
preliminary walkdown of the system on October 1,2,and 3,
1990.
Members of the operations and maintenance groups and the site quality organization were also present.
The following discrepancies were noted by the inspector during the walkdown:
Housekeeping and cleanliness in the pump and heat exchanger rooms was poor.
Work was ongoing and much scaffolding was still located in the rooms.
Large portions of piping insulation was removed and stacked in the rooms.
Major portions of the permanent lighting was not functional.
Examples of damaged flexible conduit, damaged or missing insulation and valve handwheels, inoperable MOV manual controls, were noted in several of the rooms.
These items were noted by licensee personnel present on the walkdown and were documented in the walkdown discrepancy list.
The most significant observation was the poor overall condition in the pump and heat exchanger rooms.
This was especially true in those Unit
portions of the RHR system, needed to support Unit 2 restart.
Since there is continuing work in these areas, the final removal of the large amount of construction materials when work is completed may result in some damage to permanent equipment.
The inspector believes that another walkdown will be necessary prior to completion of the SPOC process.
h.
Core Spray System (System 75)
The inspector reviewed final walkdown of the system and the completed SPOC package.
A total of 10 deferrals were taken to the system.
In the deferrals, all, physical work except for two items was completed.
The work packages were generic to other systems and remained open because of the outstanding work remaining on these other affected systems.
The two items noted above that have physical work remaining were:
(1)
ASME Section XI requirements to be performed within 30 days of declaring the system operable for plant restart.
(2)
Revising set-points, scaling documents, calibration instructions and surveillance tests and performing the associated work.
This effort and tracking were in progress.
Those portions of work and documentation reviewed by the inspector were found to be acceptabl Fuel Pool Cooling (System 78)
This system
'was returned 'to service on September 17, 1990.
The inspector reviewed the completed SPAE and SPOC packages.
There were five deferrals and no exceptions for this system.
Two of the deferrals were because two ECNs were not closed.
The work for system
was completed but per the applicable procedure a deferral was necessary until the ECNs are closed.
Two of the deferrals were maintenance items which would.require
%he reactor cavity to be drained in order to perform them.
The deferral was tied to the cavity drain down which will occur after
'fuel loading.
The last deferral, was performance of a SI prior to declaring supplemental fuel pool valve operable.
This is tracked in the LCO tracking system and is not required prior to startup.
The inspector concluded the deferrals were in accordance with the SPOC procedure and were items which were logically deferred The inspector reviewed the system status file on September 17, 1990.
A completed valve lineup checklist was in the status file for this system.
The valve lineup checklist was reviewed.
It covered a two week span but the performers did not date the checklist.
This was discussed with the Operations Nanager.
Fuel Handling and Storage (System 79)
The inspector accompanied plant personnel on a final walkdown of the system on September 16, 1990.
No deficiencies were identified.
The SPOC for this system was completed on September 16 and the inspector reviewed the completed package on September 17.
The inspector noted there was one exception and one deferral taken for this system.
The exception involved the completion of PMTs on the refuel bridge which cannot be conducted until secondary containment is returned to service.
The deferral was for replacement of the frame mounted hoist motor heater which is awaiting.receipt of a
new heater.
No defi-ciencies were identified during the review of the SPOC package.
Diesel Generators (System 82)
The inspector accompanied and observed a preliminary walkdown of the eight DGs.
The walkdown included the DGs rooms, the DGs, the control cabinets, and the air inlet plenums.
During the inspection of the control cabinets several unterminated cable conductors were noted that were not labeled spare or abandoned and Time Delay Relay TD 0-02-211822A in Shutdown Board B
had broken conductors.
This is identified as IFI 259,.260, 296/90-29-03, Inadequately Identified and Trained Conductors in Safety Related Panels.
Also noted was a non-conformance tag on a time delay relay in the 3EB shutdown board output breaker cubicle for the 3EB DG.
The tag referenced CAQR BFP 880476.
The system engineer stated this CAQR
would be researched and additional information would be given to the inspector.
1.
Reactor Protection System (System 99)
An inspector accompanied the systems engineer on a preliminary walkdown of the system on September 20, 1990.
Members of the operations and maintenance groups were also present.
Only minor work and housekeeping items were identified.
Some modifications remain,to be completed on this system before the SPOC can be performed.
The SPOC for this system is scheduled for completion on November 23, 1990.
No violations or deviations were identified in this area.
10.
Reportable Occurrences (92700)
The LERs listed below were reviewed to determine if the information provided met NRC requirements.
The determinations included the verifi-cation of compliance with TS and regulatory requirements, and addressed the adequacy of tne event description, ine corrective actions taken, the existence of potential generic problems, compliance with reporting requirements, and the relative safety significance of each event.
Additional in-plant reviews and discussions with plant personnel, as appropriate, were conducted.
(CLOSED)
LER 259/86-21, Reactor Building Flood Level Switches Not Seismically (}ualified Due to Design Deficiencies.
FSAR Section 10. 11.5, Fire Protection System Analysis, takes credit for two seismically qualified means of flood detection inside the reactor building.
One of these systems was the reactor building flood level switches located on the elevation 519 foot level.
The design review identified that in fact this system was not seismically qualified.
The licensee has revi sed the FSAR to eliminate this statement and takes credit for existing operator action.
The inspector reviewed the 'revised FSAR section, the Safety Evaluation and several Alarm Response Procedures and determined that they are consistent.
While the LER and subse'quent FSAR revision were written to address the seismic
- issue, during his review and field verifica-tion the inspector identified that both the LER and FSAR revision contained errors.
The FSAR states that the flood detectors are six inches off the floor and indicates to the detector locations as two in the HPCI room, two in the RCIC room, and two in the torus area.
The LER in the analysis section clearly states the detector locations as above.
The inspector walked the system down and determined that the detectors are in fact two inches off the floor.
The detector locations are one each in the HPCI room, the four corner rooms (one of which is the RCIC room),
and one is in the torus area.
This does not affect the results of the Safety Evaluation but does affect the technical accuracy of the LER and FSAR.
The licensee has corrected
these problems by updating the FSAR for inclusion in the next annual update and will revise the LER to be accurate.
The inspector reviewed all of the above documentation, walked the system down, and reviewed all of the affected Alarm Response Procedures to insure the commitments have been implemented.
(CLOSED)
LER 259/89-16, Discovery of Incore Detectors unaccounted for in BFN's Special Nuclear Material Inventory.
During the performance of housekeeping activities in the Radwaste Evaporator Building, on November 4, 1989, RADCON personnel identified three buckets containing seven lead bricks presumed to contain neutron instrumentation detectors and one polybag presumed to contain SNM material.
These items were placed in a
55 gallon drum.
On November 8,
1989 the SNM custodian was notified and subsequently determined that the items stored in the 55 gallon drum were not listed on current SNM inventory list.
The licensee identified
additional drums containing SNM in the Radwaste Evaporator Building in which discrepancies existed.
Based on the discrepancies iden-tified LER 50-259/89016 was issued.
The following corrective actions were committed to by the licensee:
(1)
Develop and implement a
SNM training program for personnel involved in SNM control.
(2)
Review previous NRC SNM commitments (3)
Overview by equality Assurance (4)
SNM Physical commitment (5)
SNM only storage area for unusable SNM items (6)
Provide a final SNM accountability report The inspector reviewed the licensee closure package and associated documentation.
Four of the six corrective actions have been completed.
Those items not completed at the present time are the complete physical inventory and the accountability report of SNM.
These items are scheduled to be completed following and ILRT mile-stone (Drywell Head move into place).
The two uncompleted items are also addressed in a notice of violation and proposed imposition of civi 1 penalty issued per NRC Inspection Report No.
50-259/89-55, 50-260/89-55, and 50-296/89-55.
This NOV provides adequate tracking for the uncompleted items, therefore the LER may be closed.
(CLOSED)
LER 259/90-13, Unplanned ESF Actuation Due to a Blown Fuse Caused by a Failed Relay.
On August 7,
1990, a
blown fuse in the Unit
PCIS logic panel resulted in the actuation of engineered safety features.
Investigation
revealed that the blown fuse was the result of a failed relay in the Unit
secondary containment isolation circuitry.
This event is similar to one that occurred in Unit 3 on March 14, 1990, (LER 296/90-03).
The inspector reviewed the LER dated September 6,
1990, and verified that it met the requirements of
CFR 50.73.
TVA reviewed records for relays of this type (Westinghouse type MG-6) and noted that only one prior failure had occurred.( LER 296/90-03).
At that time, the INPO NPRDS database indicated good experience with this type of relay.
The licensee considered this to be'n isolated occurrence.
The inspector identified no deficiencies or concerns.
(CLOSED)
LER 259/90-14, Unplanned ESF Actuation Due to a Blown Fuse.
On August 10, 1990, the Unit
RPS bus-1A was deenergized resulting in the actuation of engineered safety features.
A control power fuse for the lA RPS MG set failed causing the 1A RPS bus to trip.
This resulted in a
PCIS actuation.
All other related actuations and isolations w'ere disabled at the time of the event.
The inspector reviewed the LER dated September 10, 1990, and verified that it met the requirements of
CFR 50.73.
The licensee's investigation of this event identified no root cause for the blown fuse or abnormalities in the system.
A review of the records found no previous occurrences of failure of a fuse of this manufacturer and type.
This event was attributed to a
random fuse fai lure.
The inspector identified no deficiencies or concerns.
(CLOSED) Part 21 259, 260, 296/P21 89-06, Concerns With Core Neutron Flux Monitoring and RPS During Refueling.
P The inspector reviewed the licensee's closure package for this item and the Part 21.
These issues were addressed by closure of VIO 259, 260, 296/89-18-01, VIO 259, 260, 296/89-18-02, and IFI 259, 260, 296/89-18-06 closed in IR 90-27.
The inspector reviewed the GOI, TI, TS, and SI revisions which addressed neutron monitoring and RPS configuration during refueling.
No violations or deviations were identified in this area.
ll.
Action on Previous Inspection Findings (92701, 92702)
a.
(CLOSED) IFI 260/90-18-03, RHR Valve Body Erosion.
This item was opened to follow the licensee's corrective actions associated with CARR BFP900163 which documented valve body erosion on the Unit 2 RHR Outboard Loop Injection Isolation Valve, 2-FCV-74-66.
On May 21, 1990, the licensee had determined by visual inspection that the wall thickness was less than the minimum required thickness of 1.57 inches.
Additionally, a finger size area in the valve body
was found to have eroded to approximately 0.75 inches thickness.
UT data on the same valve in the opposite loop, 2-FCV-74-52, indicated areas in the seat area which were also below minimum required wall thickness.
The inspector held discussions with the RHR system engineer and other members of licensee management and reviewed the completed CARR and other documentation.
The inspector determined from this review that the licensee had completed the.disassembly and visual inspection of 2-FCV-74-52.
Although similar wall thinning had occurred in that valve it was less severe than the problem observed in 2-FCV-74-66.
The licensee has made weld repairs to both valves to restore the valve bodies to design specifications.
Additionally mechanical maintenance has issued a preventive maintenance item to periodically inspect the affected areas of both valves during future refueling outages to identify if further degradation is occurring.
This item is closed.
(CLOSED) IFI 259, 260, 296/90-27-04, Potential Degradation of CST Bottoms.
During the final SPOC walkdown performed for System 2,
Condensate, Condensate Storage and Demin Water Transfer System, a
concern was identified associated with potential undetected degradation of the tank bottoms for the five CSTs.
Each of the tanks had a visible gap of approximately
.375 inch which completely encircled the outer rim of the tank at the base next to the tank foundation.
The gaps were filled with dirt, moss, or other foreign material which could support accelerated and undetected corrosion of the tank bottom.
The licensee agreed to further evaluate the condition and perform an inspection of the gap area to determine exact material condition.
The inspector held additional discussions with licensee engineering personnel concerning this subject.
Additionally the inspector reviewed Civil Structure Drawing, 10N315, Rev.
B, Foundations
&
Trenches for Water.
& Oil Tanks.
This drawing illustrated that the CST foundation/basemat original design consisted of an outer circular concrete base filled with a compacted mixture of sulfur free sand and oil over gravel.
The licensee stated their position that this type design should reduce the chance of undesirable buildup of moisture and resultant growth in the area immediately under the tank bottom.
Additionally, the inspector reviewed completed work requests C032979, C032980, C032981, C032982, and C032983.
These WRs documented the licensee's cleaning and inspections of the gaps and CST bottom rims.
The licensee determined during those inspections that no significant corrosion was present on the bottoms of the tanks.
Based on the above additional information provided by the licensee the inspector determined that the licensee 'has adequately addressed the inspector's concern.
This item is close Ci
(CLOSED)
URI 50-259,260,269/89-34-01,
"Introspect-98 software has apparent circular scan depth calculation error".
This item was identified during BFN Unit
IGSCC weld overlay examinations during NRC Inspection 50-259, 260, 269/89-34.
The NRC inspector and a
TVA Level III Examiner determined that depth of the calibration reflectors depicted on the computer's hardcopy graphics for the initial calibration of weld GR-2-59 were in errors Subse-quent investigation revealed that an error was made by AMADATA in a
revision of its computer software program.
This error was determined to be in the conservative direction.
The inspector reviewed the licensee closure package for this URI.
The licensee sent out an information notice on Nuclear Network to advise other utilities of the software problem.
TVA performed a
review of the 1989 examination data previously performed at BFN for Unit 2,, and found no adversely affected data.
TVA personnel involved with the I/98 computer system have been instructed not to use the feature during future examinations.
(CLOSED) URI 260/90-27-02, e.
The closure of this URI is discussed in paragraph 4.b of this IR.
(CLOSED)
VIO 259,290,296/89-56-01, Control of Measuring and Test Equipment (Failure to meet the 4: 1 calibration requirement)
During a previous inspection, the inspector identified cases where the licensee used measurement and test equipment without proper documentation to allow deviation from the 4: 1 accuracy requirement per SDSP 29. 1, Control of Measuring and Test Equipment.
The inspector reviewed the corrective actions and associated documentation.
Interviews were also conducted with licensee personnel, both site and corporate.
The licensee initiated CARR 900035 to address
.the items identified in the violation.
A review of records/documents indicates that the corrective actions were complete.
The revision to the licensee's procedure (SDSP 29. 1)
contains sufficient information to address standard requirement, and instructions on what is required should an instrument not meet the
,4:1 ratio requirement.
No violations or deviati ons wer e identified in this area.
12.
TMI Action Items (TI2525/065)
A review of the following Three Mile Island (TMI) Action Items was performed by the inspectors during the reporting period.
The status of this review is listed belo (OPEN)
TMI III.D.3.4.3, Control Room Habitability.
This item resulted from the Three Nile Incident and dealt with the evaluation of the control room habitability analysis following a
postulated hazardous chemical release.
The licensee had previously concluded that the chemicals stored onsite or offsite within a 5-mile radius or transported near the site by barge, rail, or road within a 5-mile radius, that only chlorine traveling by barge could present a
hazard to personnel ia the control room."
The licensee further concluded that the impact from the barge source was negligible since the information received from the Army Corps of Engineers ( 1979 data)
did not indicate that chlorine barges passed the BFN site.
During 1986, the inspector contacted the Army Corps of Engineers and learned than more recent data indicated that approximately four barge loads of chlorine per month passed BFN due to the completion of the Tennessee-Tombigbee Waterway.
As a result, Unresolved Item (URI 259,260,296/86-40-10)
was initiated to re-evaluate the issue.
The licensee re-evaluated the issue and determined that the barge shipments were below the 50 per year allowed per Regulatory Guide, RG 1.)8, Assumptions for Evaluating tne Habitability of a Nuclear power Plant Control Room During a Postulated Hazardous Chemical Release, and therefore the plant was in compliance with the requirement.
However, the NRC staff in a letter dated December 19, 1989, requested that the licensee evaluate th'e release of all hazardous chemicals, including chlorine, from a potential barge accident for impact on control room habitability, as opposed to shipments of specific chemicals exceeding the limit of 50 per year.
A response from the licensee, dated Nay 31, 1990, listed the hazardous chemicals that were barged past BFN and an evaluation of an accident release and, affect on the control room habitability for each.
Five chemicals were considered to impact control room habit-ability during a release.
Three of these were referenced in RG 1.78 and two were referenced in the OSHA standard.
Chemical i
Evaluation (worst case accident) toxicity limit exceeded in the control room Acrylonitrile (RG 1.78)
Limit would be reached in slightly less than three minutes after the odor threshold is reached.
Benzene (RG 1.78)
Limit would be exceeded within 15 seconds after the arrival of the chlorine puff.
Limit would not be reached less than 12 minutes after the odor threshold is reache Vinyl Acetate (OSHA)
Toluene (OSHA)
Odor threshold value for vinyl acetate is 0.12 ppm and can be detected immediately.
The concentration in the control room two =minutes after detection is
ppm.
OSHA provides a
threshold limit value (TLV) of
ppm for an eight hour weighted average.
Odor threshold value for this chemical is 0. 17 ppm.
The concentration in the control room after two minutes is only 4.4 ppm.
OSHA provides a TLV of 200 ppm for an eight'our weighted average.
The NRC issued an SER which identified six chemicals that could affect control room habitability and added Ethyl Benzene to those listed above.
The NRC evaluation agreed with the licensee's analysis that the probability of the control room becoming uninhabitable from chlorine gas as a result of a barge accident was below the risk value of 1E-6 criterion for the
CFR 100 guidelines.
However, the remaining five chemicals allow, at a
minimum, a two-minute warning after detection and this is sufficient for the operators to don their
.protective gear in tne event of an accident with a chemical shipment.
In further evaluation, the NRC staff requested that the licensee provide training for the control room personnel in recognizing and responding to the five hazardous chemicals (other than chlorine)
whose presence could reach toxic levels in the control room.
However, the option.remains open for the licensee to provide additional analysis concluding that these measures are unwarranted because the level of probabilistic risk is sufficiently low.
The licensee responded by letter to the Commission dated October 15, 1990, from E.G.
Wallace, that:
( 1)
less than
shipments of hazardous chemicals per year travel past the plant; this poses an extremely low probability of an offsite dose especially when coupled with the BFN procedures for control room, personnel actions upon detection of a
chemical hazard, and (2)
the commission should consider their differing interpretations under the standards of 10 CFR 50. 109.
(CLOSED) TMI Action Item II. F.1.2. D; Containment Pressure The inspector reviewed TMI Action Item (old number II.F. 1.4, new number II.F.1.2.D),
Containment Pressure Monitor, to determine the status of the licensee's efforts in this area.
The licensee had submitted a letter to the Nuclear Regulatory Commission, Notification Of Implementation Of NUREG-0737, Item II.F. 1.4, dated October 1, 1990, Containment Pressure Monitor stating'hat the Browns Ferry Nuclear Plant is now in compliance with the requirement.
In-plant inspections performed by the inspector determined that the installation of the applicable instrumentation had been completed and the Technical Specifications updated to include the instruments.
The
pressures are monitored on two sets of instruments with overlapping ranges and redundant channels.
Pressure indication of 0 to 300 psig is indicated in the control room on pressure indicator, 2-PI-64-160A, and pressure recorder, 2-PT-64-159B, and a pressure of 0 to
PSIA (approx.
-15 to
PSIG)
is indicated on pressure indicator, 2-PI-64-67B, and on pressure recorder, 2-PT-64-50.
Drawings and documentation have been updated to reflect the current configuration.
Surveillances were identified for each of the instruments and each was noted as required for entry. into a Limited Condition of Operation (LCO) when the instrument was out of tolerance.
In addition, the requirement for monitoring the containment pressure as specified in Regulatory Guide 1.97, Rev.
2, Instrumentation For Light-Water-Cooled Nuclear Power Plants To Assess Plant And Environs Conditions During and Following An Accident.
No deficiencies were noted.
(OPEN)
TMI Action Item, II.F. 1.2.B, Accident Monitoring-Iodine/
Particulate Sampling.
The licensee approved ECN P0354 to meet the requirements of NUREG 0737 Item II.F. 1.2.B, Accident Monitoring-Iodine/Particulate Sampling.
This engineering change resulted in the issuance of numerous WPs.
As of this reporting period, the ECN was still open.
The inspector reviewed the status of the WPs and noted that a total of
WPs were completed, such as WP 2086-85, Install Permanent High-Range Radiation Monitors for the 600 ft. Off-Gas Stack and WP 2142-85, Fabricate and Install supply and return sample lines from the Gull Radiation Monitor Skid to the Off-Gas Stack.
The majority of the WPs were completed from February 1988 to June 1989.
Addi-tional WPs were still outstanding such as WP 2662-90, Delete the Jumper Between Terminal Points 21 and 22 of TB1 in panel 25-418 and verify that on panel 25-418, the CKT Spare annunciator light was deenergized after the jumper was removed.
I (OPEN)
TMI Action Item, II.K.3.2.8, Qual ification for ADS Accumul ators/Modi fic'ati on.
The licensee approved DCN W 10399A.
The purpose of the DCN was to provide redundant source of backup containment atmosphere dilution nitrogen for the drywell control air system to fulfillthe commitment of having compressed air available to the automatic depressurization system relief valve accumulators for a
100 day period following an accident as required by NUREG-0737, Item II.K.3.2.8, Qualification for ADS Accumulators/Modification.
This DCN resulted in the issuance of seven WPs such as WP 2386-90, Divide Drywell Control Air Header into two Valves, and Install Containment Atmosphere Dilution Line in Unit 2 Reactor Building and WP 2406-90, Fabricate and Install Supports on Control Air Line in Unit 2 Reactor Building.
As of this
~ ~
t
20 S
reporting period all field work was completed for this DCN and the OCN was in the status of closeout review by NE.
e.
(OPEN)
TMI Action Item, Item II.3. 18.c, ADS Actuation/Modification.
The licensee approved ECN P 7116.
The purpose of this ECN was to replace the existing ADS timers, 105 sec setting..This ECN also relocated the Annunciation
"RCIC Relay logic power failure" from annunciation window ¹3 to window ¹1, annunciator XA-55-3C, Control Room Panel 9-3.
Thi s i s to fulfi1 l the comment to modi fy the Automatic Depressurization System to eliminate the need for manual actuation to assure adequate core cooling as required by NUREG-0737, Item II.3.18.c, ADS Actuation/Modification.
This ECN resulted in the issuance of two WPs, WP 2332-88, Modify ADS Logic in the Control Room and Auxiliary Instrument Room and WP 2614-90, Install Hand Switch/
Indication Lights on Various Panels
'and Install/Terminate four cables.
As of the end of this reporting. period WP 2614-90 was still outstanding.
No,violations or deviations were identified in this area.
13.
Modifications and Unit 3 (37700, 37828)
a.
Unit 2 DCN/ECNS During a
two week period in September 1990, the licensee deleted
ECN/DCNs from the Unit 2 Cycle 5 required list per, the computer list BFNP Work Item Information System Weekly U2C5 Deletions.
A review of this listing indicated that eight of these were identified on the.
printout as
"NRC Commitments."
Another five appeared to have safety significant impact on the restart of Unit 2.
A review of the licensee's governing procedures, SDSP 7. 1, Control of Work Scope and the Planning and Estimating Process and SDSP 8. 1, Plant Modification/
Design Change Approval, left questions as to how the process was being implemented.
In SDSP 7. 1 there is a form, SDSP-63, Schedule Justification, BFNP that was used as the agent to delete the ECN/DCNs from U2C5.
There are no directions for the completion of this form and its use in the body of SDSP 7. 1.
The completed forms reviewed by the inspector appeared, incomplete with only Sections 1 and 3 of the form completed.
None of the forms had adequate justification for the DCN/ECN deletions.
None of the deletions appeared to have been revi ewed by the CCB or middle management, they went from the system engineer to the Site Director, who is the CCB chairman.
None of eight deletions that were identified by the licensee as NRC commit-ments were discussed with the NRC resident staff.
The licensee is reviewing the use of SDSPs 7. 1 and 8. 1, the control of ECN/DCN deletions, and the deletion of the NRC commitments.
The results of this review will be considered an Unresolved Item 259, 260, 296/90-29-04, Deletion of ECN/DCNs, and must be resolved for restart of Unit b.
Unit 2 Modification The modifications schedule's have been revised to coincide with recent TVA Unit 2 restart schedule.
Despite the new schedule the unit work off rates do not appear to support the early fuel load/restart dates.
As of October 11, 1990, all but eight of the 1,610 ECN/DCNs required for restart have been star ted through the design process.
A total of 1,324 of 1,610 ECN/DCNs have been closed.
Little new work is expected to be identified.
What remains-is to work off remaining 287 packages.
For core refueling only 127 DCN/ECNs remain open.
c.
Unit 3 The Unit 3 effort continues on the Scope Development Phase but is constrained by the unavailability of key personnel that are still obligated to the Unit 2 effort.
The licensee is planning a meeting with NRC-H(} in mid to late November 1990, to present their Unit 3 plan and discuss any regulatory relief that would be needed to support that plan.
The Unit 3 effort is directly coupled to the Unit 2 effort.
As Unit 2 approaches restart the key personnel will be reassigned full time to Unit 3 effort and tnat schedule can then start.
No violations or deviation were identified in this area.
14.
Site Management and Organization (36301, 36800, 40700)
The licensee has made progress toward integration of all engineering, modifications, maintenance, and system turnover activities into a
more credible restart schedule.
The utilization of computer codes, industry consultants, and commodity work off curves has helped to quantify the remaining activities required for startup.
A comprehensive project plan to track the restart commitments and programs has been implemented.
These actions are a positive step toward correcting the unrealistic scheduling of the past.
The daily work activity focus has shifted from a punchlist of the next group of system turnovers to focus on an overall P-2 schedule.
This will give critical items to be worked by measure of float instead of working only items next in line for system turnover.
15.
Exit Interview (30703)
The inspection scope and findings were summarized on October 16, 1989 with those persons indicated in paragraph 1 above.
The inspectors described the areas inspected and discussed in detail the inspection findings listed below.
The licensee did not identify as proprietary any of the material provided to or reviewed by the inspectors during this inspection.
Dissenting comments were not received from the license,
Item Number Descri tion and Reference 259, 259, 259, 260, 296/90"29-01 260, 296/90-29-02 260, 296/90-29-03 Violation, Failure to Comply With Procedures for Control of Modifications and Maintenance Activities, paragraph 4.
Inspector Followup Item, Leaky Drywell Penetrations, paragraph 9.
Inspector Followup Item, Inadequately Identified and Tra'ined Conductors in Safety Related Panels, paragraph 9.
259, 260, 296/90-29-04 URI, Deletion of ECNs/DCNs, paragraph 13.
Licensee management was informed that four LERs, one Part 21 Report, seven IFIs, three URIs, and one Violation were closed during this inspection.
Acronyms ASME ASOS BFNP BWR CAQR CCB CCRIS CFR CRD CRDR CST DCN DG ECCS ECCW ECN EEB EECW EP ESF FCV FSAR GE GETARS GOI HP HPCI HQ HTX HVAC American Society oi Mechanical Engineers Assistant Shift Operations Supervisor Browns Ferry Nuclear Plant Boiling Water Reactor Condition Adverse to Quality Report Change Control Board Calculation Cross Reference Information System Code of Federal Regulations Control Rod Drive Control Room Design Review Condensate Storage Tank Design Change Notice Diesel Generator Emergency Core Cooling System Essential Component Cooling Water Eng,ineering Change Notice Electrical Engineering Branch Emergency Equipment Cooling Water Emergency Preparedness Engineered Safety Features Actuation Flow Control Valve Final Safety Analysis Report General Electric General Electric Transient Analysis Recording System General Operating Instructions Health Physics High Pressure Coolant Injection Headquarters Heat Exchangers Heating, Ventilation, 5 Air Conditioning
E'
IFI IGSCC ILRT INPO IR ISI LCO LER MCEL MG MOV NCV NPRDS NRC NRR PCIS PM PMT QA QC RBCCW RCIC RHR RPS RTP RWCU SBGT SCI SCL SDSP SER SI SNM SOS SPAE SPOC STA TACF TARS TD TI TMI TOL TROI TS TSC TVA UO URI VIO WO WP WR Inspector Fol 1owup Item Intergranular Stress Corrosion Cracking Integrated Local Leak Rate Test Institute of Nuclear Power Operations Inspection Report In Service Inspection Limiting Condition for Operation Licensee Event Report Master Component. Electrical List Motor Generator Motor Operated Valve Non-Cited Violation Nuclear Plant Reliability Data System Nuclear Regulatory Commission Nuclear Reactor Regulation Primary Containment Isolation System Preventive Maintenance Post Maintenance/Modification Test Quality Assurance Quality Control Reactor Building Closed Cooling Water Reactor Core Isolation Cooling Residual Heat Removal Reactor Protection System Restart Test Pro'gram Reactor Water Cleanup Standby Gas Treatment System Standard Calibration Instruction System Checklist Site Director Standard Practice Safety Evaluation Report Surveillance Instruction Special Nuclear Material Shift Operations Supervisor System Plant Acceptance Evaluation System Pre-Operability Checklist Shift Technical Advisor Temporary Alteration Change Form Transient Analysis Recording System Test Deficiency Temporary Instruction Three Mile Island Thermal Overload Tracking and Reporting of Open Items Technical Specifications Tech Support Center Tennessee Valley Authority Unit Operator Unresolved Item Violation Work Order Work Plan Work Request