IR 05000259/1988035
| ML20247E704 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/17/1989 |
| From: | Carpenter D, Little W NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20247E703 | List: |
| References | |
| 50-259-88-35, 50-260-88-35, 50-296-88-35, IEB-83-08, IEB-83-8, NUDOCS 8904030220 | |
| Download: ML20247E704 (19) | |
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NUCLEAR REGULATORY COMMISSION
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REGION H j
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101 MARIETTA STRE ET, N.W.
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Report Nos.:
50-259/88-35, 50-260/88-35, and.50-296/88-35 Licensee:
Tennessee Valley Authority 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 1, 2, and 3
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Inspection Conducted:
December 1-31, 1989 Inspector:
h $]& W N $ r 3/t7/R 9
[D.R. Carpenter,NRCSiteManager Date Signed Accompanying Personnel:
E. Christnot, Resident Inspector W. Bearden, Resident Inspector K. Ivey, Project Engineer A. Johnson, Project Engineer J. York, Senior Resident Inspector, Bellefonte Approved by:
d)$Y Sf/7fB9 W. S. Lfttif, Section Chief (fate' Signed Inspection Programs TVA Projects Division
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d SUMMARY
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Scope:
This routine resident inspection included the areas -of. operational safety verification, surveillance observation, system return to I
service, reportable occurrences, restart test program, followup of NRC Bulletins, licensee action on previous enforcement matters, and followup of open inspection items.
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Results:
Two unresolved items were identified:
260/88-35-01:
Surveillance Testing Concerns, paragraph 3
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(Restart Item)
260/86-35-02:
Missed Surveillance on SGTS, paragraph 3 (Restart Item)
The NRC staff is concerned about three separate occasions during the month of December 1988, when safety related components were unin-tentionally onerated.
These occurred due to either inadequate procedures or failure to properly follow Instructions.
Additionally,
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the licensee failed to identify that the "C" Train SBGT system should have been declared inoperable due to an expired surveillance.
Subsequent to this inspection, TVA initiated actions to correct problems identified in the surveillance program.
In other areas, such as the activities in preparation for beginning the reloading of fuel into the Unit 2 vessel, TVA management and staff performed well and showed sensitivity to NRC questions and concerns.
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REPORT DETAILS 1.
Persons Contacted Licensee Employees:
0. Kingsley, Jr., Senior Vice President, Nuclear Power C. Fox, Jr., Vice President and Nuclear Technical Director
- J. Bynum, Vice President, Nuclear Power Production
- C. Mason, Acting Site Director
- G. Campbell,. Plant Manager H. Bounds, PrQect Engineer
- J. Hutton, Opt. rations Superintendent
- D. Mims, Technical Services Supervisor G. Turner, Site Quali ty Assurance Manager
- P. Carter, Site Licensing Manager
- J. Savage, Compliance Supervisor A. Sorrell, Site Radiological Control Superintendent R. Tuttle, Site Security Manager L. Retzer, Fire Protection Supervisor H. Kuhnert, Office of Nuclear Power, Site Representative T. Valenzano, Restart Director Other licensee employees or contractors contacted included licensed reactor operators, auxiliary operators, craftsmen, technicians, and public safety officers; and quality assurance, design, and engineering personnel.
NRC Attendees
- D. Carpenter, Site Manager
- E. Christnot, Resident Inspector
- W. Bearden, Resident Inspector
- A. Johnson, Project Engineer
- Attended exit interview Acronyms used throughout this report are listed in the last paragraph.
2.
Operational Safety Verification (71707)
The NRC inspectors were kept informed of the overall plant status and any significant safety matters related to plant operations.
Daily discussions 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; onsite and offsite emergency power sources available for automatic operation; purpose of temporary tags on equipment controls and switches; annunciator alarm status; adherence to procedures; adherence to
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limiting conditions for operations;. nuclear instrument operability;-
temporary alterations in effect; daily jcurnals and logs; stack monitor
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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, l
each reactor building, and other plant areas were visited.
Observations included valve positions and system alignment; snubber and hanger condi-
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tions; containment isolation alignments; instrument readings; house-keeping; proper' power supply.and breaker alignments; radiation area controls; tag controls on equipment; work activities in progress; and radiation protection controls.
Informal discussions were held with
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selected plant personnel in their functional areas during these tours.
An NRC inspector reviewed the TACF file located in the main control room area, and noted that the number of open and outstanding Unit 2 and common TACFs was continuing to decrease in accordance with the schedule published by the licensee as part of an ongoing management program.
The Hatch: Nuclear Power Plant had identiffed bent and broken bolts on the torus supports.
The 16 supports located under the Browns Ferry Torus were inspected and no evidence of bent or broken bolts or any other type of damage were identified.
No,-violations or deviations were identified in the~ 0perational Safety Verification area.
3.
Surveillance Testing (61726)
The inspectors observed and/or reviewed the SI procedures discussed below.
The inspections consisted of a review of the sis for technical adequacy and conformance to Technical Specifications (TS), verification of test instrument calibration, observation of the conduct of the test, confir-mation of proper removal from service and return to service of the system, and a review of the test data.
The inspector also verified that limiting conditions for operation were met, testing was accomplished by qualified personnel, and the sis were completed at the required frequency.
The NRC inspector observed and reviewed portions of the following sis performed on the Unit 1/2 DGs A thru D:
SI 4.9.A.1.b-1 Diesel Generator A Emergency Load Acceptance Test SI 4.9.A.I.b-2 Diesel Generator B Emergency Load Acceptance Test SI 4.9.A.1.b-3 Diesel Generator C Emergency Load Acceptance Test SI 4.9.A.I.b-4 Diesel Generator D Emergency Load Acceptance Test
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These four sis were performed prior to fuel load and were the last L
scheduled testing of the DGs.as part of the overall DG issues at the BFNP facility.
The initial NRC review of the SI results indicated that these activities, were adequate to support fuel load.
However, a more indepth review of the BFN DG package will be conducted prior to restart.
Prior.to observing the test performances, the NRC inspector reviewed the SI procedures.
The inspector-noted that these four sis were revised one week before they were scheduled to be performed, and additional revisionsE were being made one day and-the procedure performed the next day..The increased activity during the' final preparations for fuel load did not appear to leave enough-time for the licensee to adequately review the
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revised sis prior to their being performed.
The NRC inspector observed the performance of these sis and noted that in all observed cases, the licensee used second party and not independent verification to verify lifted leads, booted relay contacts, and test switches being placed. in test positions.
The NRC inspector was later informed by licensee management' that independent verification also applied to sis.
The proper application of independent verification in the performance of sis _ will be reviewed by the.NRC inspector, and _this is identified as Unresolved Item 260/88-35-01, Surveillance Testing Concerns.
This item must be resolved prior to Unit 2 rartart.
The following concerns were identified by the NRC inspector during the review of the completed sis:
On December 9, 1988, while operators were performing 2-SI-4.2.B-45.A,
"RHR Logic System Functional Test", a step in the procedure required that the stop pushbutton be depressed.
However, the operator did not follow the procedure and depressed the start pushbutton resulting in the RHR pump starting and running for five seconds until they stopped the pump.
On December 14, 1988, during the performance of 0-SI-4.7.B.7, " Flow Rate Testing on Standby Gas Treatment System (SGTS) Train C", the system failed to meet the minimum flow requirement and was declared inoperable.
Further investigation by the licensee revealed that an undocumented damper, i.e., not tagged with an identification number and not indicated on any drawing, was attached to the suction side of the train C exhaust blower.
The damper appeared to be skid mounted.
This train was supplied by a different vendor than trains A and B.
The licensee suspended further activity over the spent fuel pool rehen the train was declared inoperable.
However, further review by licensee personnel revealed that on November 30, 1988, an SI for train C had expired and the train should have been declared in-operable on that date.
Since November 30, 1988, during any fuel movements or the conduct of operations over the spent fuel pool, the i
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TS could have been violated if less than two SGTS trains.were operable.
This concern is identified as. Unresolved ' Item 260/86-35-02, Missed Surveillance on SGTS.
On December 17, 1988, during the performance of 0-SI-4.9. A.1.b-1,
" Unit 1/2 DG A Load. Acceptance Test", a start of the 2D RHR pump
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occurred.
This occurred because the procedure did not specify the-correct sequence.
~The procedure required the steps which initiate the' logic to start the pump to be performed prior to the steps to preclude a start of the pump. The pump was immediately stopped.
On December 18,.1988, during a performance of 0-SI-4.9.A.lb-2, " Unit 1/2 DG B Emergency Load Acceptance Test", the 2C core spray. pump started due to the wrong key. lock switch being placed in the test position contrary to the surveillance procedure.
The start occurred when a jumper was installed to simulate a core spray system initia-tion.
The pump was immediately stopped.
On. December 18, 1988, during the performance of 2-SI-4.2.C-3(G), "IRM Channel C Calibration", an SI installed jumper came loose, shorted out a fuse, and tripped RPS scram channel A~ (half scram).
Approximately five seconds after this event, another IRM received a spike from an unknown cause which tripped RPS scram channel B.
With both RPS scram channels A and B tripped, a full scram was present.
.The ' instances of failure to follow surveillance procedures on December 9, and 18, 1988; the inadequate procedure found on December 17, 1988; and the concern about the proper application of independent verification to the SI program, are collectively identified as Unresolved Item 260/88-35-01, Surveillance Testing Concerns.
This item must be resolved prior to Unit 2 restart.
4.
System Return to Service (71711)
In the final phases of preparation for fuel reloading, the licensee continued with the system return to service program.
The NRC inspectors continued reviews of a sample of the licensee's return to service activities involving the final signoff of the system SP0C packages.
The inspector noted that increased emphasis was being placed on determining whether specific items were required for fuel load or not.
The NRC inspectors expressed their observations on a regular basis with system engineers, operations personnel and the return to service
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supervisors and manager.
The inspectors attended SP0C status meetings and remained cognizant of the day to day activity.
The licensee reported that on December 21, 1988, all systems identified as being required for fuel load were through the SP0C process and were under both status control and configuration control.
During this reporting period, the resident inspectors participated in a fuel loaa readiness assessment inspection (NRC Inspection Report 88-36) in which several concerns were identified with the licensee's method of establishing and maintaining configuration L
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These problems were discussed with BFN management and.
were documented in the fuel load readiness assessment inspection report.
(50-260/88-36)
No violations or deviations were identified in the System Return to Service area.
5.
Reportable Occurrences (90712, 92700)
The Licensee Event Report (LER) listed below was reviewed to determine if the information provided met NRC requirements.
The determination included the verification 'of compliance with TS and regulatory requirements, and addressed the adequacy of the event description, the corrective action 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 No. 260/86-06:
Linear Indications In RHR Pump Impeller Wear Rings After the licensee received Peach Bottom experience data concerning the failure of RHR pumps due to impeller wear ring failure (reference NRC Information Notice 86-39, Failure of RHR Pump Motors and Pump Internals),
dye penetrant inspections were performed on the upper and lower wear rings of two of the' four Unit 2 RHR pumps.
Pump 20 had linear indications in
.the lower wear ring and metallurgical analysis revealed the cracks were caused by intergranular stress corrosion cracking.
The RHR pump manufacturer recommended replacing the pump impeller with a case martensitic stainless steel impeller that has an integral wear ring; however, the licensee selected an alternative method where the wear rings would be formed from the same martensitic stainless steel (type 410)
previously used, but have a higher tempering temperature than those that were cracking due to intergranular stress corrosion cracking.
This would result in a softer wear ring with a recommended hardness of less than or equal to R 28 that would be more corrosion resistant than a harder (R of 31 to 47) Sing.
A metallurgical reference, " Heat Treatment of Ferro%s Alloys," McGraw-Hill,1979, states that tempering martensitic stainless steel 410 in the range of 700 degrees F to 1050 degrees F (producing hardnesses R 31 to 47) lowers stress corrosion cracking resistance.
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of the lower wear rings for the Unit 2 RHR pumps and the cross tie RHR pumps (Unit 1, pumps 18 and 1D; Unit 3, pumps 3A and 3C) were replaced.
The Unit 2 pump wear rings were replaced with in-stock rings having hardness values of Rockwell C 33-37.5 until the next refueling outage when the wear rings will be replaced with the lower hardness rings.
The wear rings on the cross tie RHR pumps were replaced with lower hardness (less than or equal to R 28), higher corrosion resistant wear rings.
In addition to the inspector's evaluation, NRC OSP Headquarters staff reviewed the licensee's action and plans and found them to be acceptable.
The lower hardness, higher corrosion resistant material is considered to be acceptable.
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-This item.is closed.
No' violations or deviations were identified in the area of Reportable
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6.
. Restart Test Program (99030B)
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The inspector maintained cognizance of ongoing restart test activities, q
and monitored particular items.in detail as appropriate.
The testing for fuel' load syste J had essentially been completed, and due to the increased pre-fuel loading activities, very little additional restart testing was performed during this reporting period.-
No violations or deviations were _ identified in the Restart Test Program area.
7.
Followup of NRC Bulletins (92703)
(OPEN) IE Bulletin 83-08:
Electrical Circuit Breakers With An Under-voltage Trip Feature In use In Safety-Related Applications Other Than The Reactor Trip System The NRC inspector reviewed and evaluated the correspondence.between TVA and NRC and determined that TVA had made four commitments.
The four commitments and their status, as of December 21, 1988, are listed below:
a.
Information regarding the RPS power mr.itoring system design modifi-cations will be sent to NRC.
COMPLETE - TVA submitted this to the NRC in December 1988 (A27 84 0809 009)
b.
TVA will provide tests for each unit prior to the cycle 6 refueling outage and determine the replacement requirements for the RPS components.
OPEN - (TS Change) - TVA will perform a voltage verification test with the unit in normal operation and with the RPS components in
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their normal operating configuration.
c.
A setpoint of five seconds plus or minus 1 second for the alternate f
supply relays will be added in a future TS submittal.
COMPLETE - TS submittal was made in December 1988.
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TVA will correct the setpoint inequality sign in TS 4.1.B.2 in a future TS submittal.
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l Based on the above, the NRC inspector concluded that the status of this item was acceptable for fuel load, but the item will remain open pending completion of the licensee commitments.
i No violations or deviations were identified during the Followup of NRC
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Licensee Action on Previous Enforcement Matters (92702)
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(OPEN) VIO 260/83-46-03:
Failures To Install And Test An Adequate Design i
This item involved a slow response time for instrument LT-85-45A, installed on the west scram discharge instrument volume.
The NRC inspector reviewed this item during this inspection period and noted that ECN P0392, which applies to Unit 2 only, was completed to change the Rosemont Differential Pressure Switches, Type 1153DP, to Fluid-Components Incorporated Level Switches.
The inspector verified that this type of switch was installed and that all welds and connections appeared to be intact.
However, further review of PMT-110, " Scram Discharge System", indicated that Section 5.11, " Design Verification Test", could not be performed until the reactor pressure vessel is brought up to full pressure.
This section of PMT-110 is scheduled to be performed during the vessel hydrostatic test.
The inspector determined that the modification and post modification testing were adequate for fuel load.
However, this item will remain open for final review and closure prior to Unit 2 restart.
b.
(CLOSED) URI 259, 260, 296/85-13-02:
Final Determination Concerning A Possible Violation Of TS 3.1 (RPS) and/or 3.2 (Protective Instrumentation)
This unresolved item resulted from the identification of two separate instances when the TS requirement for two operable reactor trip channels for reactor water level were not met.
Escalated enforcement action was subsequently taken and TVA admitted the violations in a response dated August 28, 1985.
No further enforcement action is appropriate and this unresolved item is closed.
In March 1986, in response to Generic Letter (GL) 84-23, TVA committed to modifications to eliminate the possibility of level measurement errors such as those which caused the violations.
The licensee's actions will be monitored in future inspections to verify that the commitments in response to GL 84-23 are carried out.
In this inspection, the NRC inspector reviewed the installation of the level transmitters in the Unit 2 reactor building and noted that all reactor water level transmitters were equipped with quick disconnect devices on the reference legs as well as the variable legs.
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(CLOSED) VIO 259, 260, 296/85-57-06 (A):
Normal and Alternate DC l
Control Power to Shutdown Board A j
It had been identified that the normal and alternate 250 V DC control j
power supply to the Unit 1/2 A Shutdown Board were reversed.
The NRC l
inspector reviewed the corrective action taken by the licensee, which consisted of determining the correct normal and alternate DC power feeds to terminal boards WE and WF, reterminating the power feed from j
the DG batteries (normal DC supply) to the correct termination
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points, and splicing the power feed from the unit batteries (ultimate supply) and reterminating them to the correct termination points.
The inspector observed the proper operation of the shutdown board during the LOP /LOCA series of tests, and during RTP-082, " Diesel Generators".
The NRC inspector also observed that the alternate supply cable (2B95-1E) was spliced and landed on terminals AAP WE1 and AAN WE2 of terminal board WE and that the normal supply cable 1895-1E was landed on terminals ANP WF1 and ANN WF2 of terminal board WF inside shutdown board A as required.
This item is closed.
d.
(CLOSED) VIO 259, 260, 296/85-57-06 (B):
Low Low DG Lube Oil Pressure Indication in Control Room This item was part of Escalated Enforcement EA 86-56, dated September 8, 1986, and was referred to as part of Violation I.C.
The escalated enforcement involved the Unit 1 and 2 DG low, low oil pressure switches PS-82.29-A, B, C and D.
These pressure switches are used to indicate a low lube oil pressure condition to the operator, by means of white light indicators located next to the DG emergency shutdown buttons on the DG control panel in the control room.
The pressure switches were not installed properly during initial construction, and they gave a continuous white light indication although the local DG oil pressure gage indicated adequate oil pressure.
The operators originally removed the bulbs from the white light indicators in the control room because they remained continuously lighted all the time while the DGs were running.
The NRC inspector reviewed the corrective action taken by the licensee which involved correcting the installation of the low low lube oil pressure switch located on each Unit 1/2 DG local panel.
The NRC inspector also observed the starting and stopping of the Unit 1/2 DGs during the performance of the Common Accident Signal Surveillance Instruction (0-SI-4.9 A.3.a) which required that each DG be automatically started both singularly and together in response to a simulated common accident signal.
When each Unit 1/2 DG received an auto start signal, the white indicator for each DG illuminated in the control room momentarily and then went out, indicating that a low low lube condition existed up until the DG's came up to speed, and normal oil pressure was then established.
This item is closed.
e.
(OPEN) URI 259, 260, 296/86-06-03:
Design Control of the As-Constructed Configuration of Auxiliary Electrical System This item was originally identified by the licensee and was
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s documented in Significant Condition Report (SCR) BFNEEB8511 which dealt with the status of the DG electrical loads.
The original.
concern stated that additional loads were placed on the DGs with inadequate analysis as required by General Design Criterion 17.
Although all planned DG testing has been completed, this item is not considered a fuel load issue and will remain open pending additional review of all DG test results in conjunction with NRC Headquarters staff.
This item must be closed prior to Unit 2 restart.
f.
(OPEN) URI 259, 260, 296/87-33-05:
Simulated Loss of Offsite Power Testing and Compliance with General Design Criterion 18 This followup concerned the adequacy of sis performed on the DGs,.in light of additional loads being added to the DGs with 'nadequate analysis.
Although all planned DG testing is completed, including the performance of the upgraded surveillance, this item is not considered a fuel load issue and will remain open pending additional review of all DG test results in conjunction with NRC Headquarters staff.
This item must be closed prior to Unit 2 restart.
t g.
(CLOSED) VIO 260/88-24-04, Failure to Comply with 10 CFR 50.72 Example 1 of the violation involved the failure to promptly evaluate and report an unreviewed safety question regarding flood protection of the RHRSW pump building.
On June 17, 1988, the licensee dis-covered that the RHRSW Pump Building was not protected against flooding as required by the design basis and FSAR.
The condition existed since the original construction and was attributed to a design error.
An evaluation of the situation led the licensee to conclude that a flood could have resulted in a loss of the ability to transfer heat to the ultimate heat sink.
The licensee's design organization made a determination on July 25, 1988, that the deficiency constituted an Unreviewed Safety Question per 10 CFR 50.59.
The CAQR that documented this problem was deficient in that it did not consider the deficiency to be either significant or reportable to the NRC.
After the NRC inspector took issue with the CAQR, the licensee made a 4-hour ENS phone call on August 19, 1988, and submitted an LER on the subject.
TVA's response to this violation describing their corrective action it described in their letter to the NRC dated January 20, 1989.
The
licensee response to this violation was reviewed by the NRC and i
determined to be acceptable, i
In order to address the question of whether other CAQRs had been I
improperly evaluated for significance and deportability, the licensee
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concluded on October 18, 1988, with the finding that sixteen CAQRs were originally classified improperly.
All of these were reported
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which could place the plant in an unanalyzed condition or condition I
which could have prevented the fulfillment of a safety function.
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order to prevent future occurrences, the licensee established a CAQR Management Review Committee (MRC) on September 28, 1988.
The MRC provides daily oversight on CAQR significance classification and deportability determinations.
The inspector reviewed the documen-tation associated with these licensee activities and observed the MRC during one of its daily CAQR review meetings.-
These" activities effectively addressed the evaluation and deportability aspects of.the violation.
With' regard to the.USQ involving the lack of ' RHRSW building flood L
protection,; the licensee designed and installed watertight seals on each of the piping penetrations.
The NRC inspector reviewed-the drawings. and workplans associated with Design Change Notice (DCN)
H1888A and visually inspected the seals in the pump rooms.
On the basis of.the positioning of :he seals available for visual inspection and the documentation associated with the workplan, the inspector concluded that the RHRSW pump. room flood protection ' seals had been properly installed.
Example 1 of the violation is closed.
Exemple 2-~of.the violation identified the failure by the licensee to make the required notification pursuant to 10_-CFR 50.72b.2 following an _ unplanned ESF actuation.
The actuation was caused by a spurious low reactor water level signal which occurred while an instrument technician was returning a. reactor water level transmitter to service.
The failure to report the event was attributed to a misinterpretation of the meaning of a planned versus unplanned ESF actuation.
After discussions were held with licensee management, the plant immediately changed the policy for reporting ESF actuations to bring the criteria for reporting within NRC requirements.
This is described in a Plant Manager Memorandum dated August 19, 1988 (R35 880815 935).
This memo states that an ESF actuation that is not a result of a planned part of a specific approved written procedure is reportable.
The licensee issued LER 50-260/88005 to report the event.
The corrective actions taken should be adequate to prevent future occurrences.
Example 2 of the violation is closed.
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(OPEN) VIO 259, 260, 296/88-24-08:
Fai f ure To Impleme'. Timely
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Corrective Action In Performing Preventive Maintenance Every Five Years On Safety Related 4 KV Breakers This item was initially identified by the NRC inspector when it was made known that the preventive maintenance recommended by the manufacturer to be performed every five years was not documented, and in fact had never been performed.
The licensee initiated a contract with.GE to refurbish / rebuild 72 safety-related 4 KV breakers prior to fuel load.
The inspector reviewed the list of breakers which are contained within the RHRSW, Core Spray, Fire Pump, Diesel Generator,
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and Shutdown l Board systems.
The inspector reviewed and observed field activities. involving testing and manipulating 'the rebuilt /
refurbished breakers and determined that this activity was adequate for fuel load.
The licensee has.taken steps to improve the preven-tive maintenance program, implementing a revised structured program
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Implementation of this program will be reviewed in a future inspection, and must be adequately addressed prior to Unit 2 restart.
9.
Followup of Open Inspection Items (92701)
a.
(OPEN) IFI. 259, 260, 296/84-32-02:
Torus Level Instrumental' ion Problems Between Separate Level Detectors This item identified the concern that the level difference between the L detectors for narrow range suppression chamber (torus) volume level indicators 2-LI-64-54A and 2-LI-64-66 was sometimes greater than the maximum allowed in the TS.
The NRC inspector reviewed the licensee's corrective actions which consisted of a DCR and an ECN to modifyf the sensing lines to the level transmitters, performance of post modification testing, and performance of. sis for the torus water Llevel.
TS 3/4.2.F, Table 3.2.F and Table 4.2.F require.that only the wide range indicator be operable.
The inspector observed that with wide range indicators 2-LI-64-159A and 2-XR-64-159. showing a -torus
' level of approximately 14 feet 6 inches, the narrow range, indicator-2-LI-64-59A showed a. level of approximately minus (-) 5 inches, and narrow range. indicator 2-LI-64-66 showed a level of approximately minus (-) 6 inches.
The difference between both narrow range indicators was approximately one inch.
The inspector determined that this was adequate to support fuel loading for Unit 2; however, this item will remain open for' final closure prior to Unit 2 restart.
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(OPEN)-IFI 259, 260, 296/85-53-01:
Le el Difference Between Reactor Water Level Narrow Range Instruments This item identified a water level mismatch in Unit 3.
At the time of the occurrence, Unit.3 was in cold shutdown rather than startup (the plant condition during the mismatch identified in NRC Inspection Report 259, 260, 296/85-13, see paragraph 8.b).
At the time of this event, SI 4.1. A-7, " Reactor Water Level Functional Test And Calibration", had been performed, and it was noted that a loss of water level in the reference leg could have occurred due to the
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The level would not have been restored i
due to the cold shutdown and condensing char 5ers not having a source
j of water with a high enough temperature to condense in the reference l
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leg condensing pots.
During the BFN facility outage, the licensee is i
modifying the level sensing lines, changing the SI, and installing
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quick disconnects to be used during level instrument calibrations for
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Unit 2.
This item originally covered all three units.
The NRC l
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inspector reviewed the installation of the reactor level instrumen-tation located on the panels for Unit 2, and noted that quick disconnects were installed in the sensing lines.
The NRC inspector determined that these activities were adequate for fuel load.
This item will be reviewed for final closure prior to Unit 2 restart.
c.
(OPEN) IFI 259, 260, 296/37-33-04:
Deficient Welds In EECW Piping Discovered During Microbiological Induced Corrosion (MIC) Inspection l
This IFI addressed the MIC program evaluation.
Resolution of Part B,
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concerning the corrective action necessary to evaluate weld deficiencies in the EECW stainless steel piping found by radiography, was required before fuel loading.
The licensee submitted a fracture toughness evaluation (Section XI of ASME Code) for the worst case condition of these welds.
NRC OSP Headquarters personnel evaluated the submittal and found no technical errors with the fracture toughness evaluation.
However, an inspection of the radiographic interpretations for these welds will ba performed by NRC/0SP Headquarters staff prior to Unit 2 restart.
This item is acceptable for fuel loading, but final resolution :s required before Unit 2 restart.
d.
(CLOSED) IFI 259, 260, 296/87-46-02:
Review By Material Specialist Of Integral RHR Pump Wear Rings Versus Softer Wear Rings As Long Term Cracking Solution Based on the closure of LER 260/86-06 (see paragraph 5), this item is alosed.
e.
(CLOSED) IFI 260/87-42-05:
Insulation Cut Away Around Tailpiece Valves75-646 and 647 A testing manifold installed on Core Spray Loop II is equipped with a three quarter inch test / drain tailpiece containing valves75-646 and 75-647.
The tailpiece was installed so close to a larger insulated pipe that part of the insulation had to be removed.
The clearance between the two pipes had been questioned by the NRC inspector.
During a later visual inspection of the two pipes during NRC Inspection 259, 260, 296/88-21, it was noted that the larger pipe had only dead weight supports and could be easily moved horizontally l
by hand.
The inspector questioned whether the larger non-safety-related pipe could move horizontally in a seismic event and shear off
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the smaller safety-related pipe. As a result of this inspection, CAQR no. BFP 8E0632 was written to evaluate the condition.
In the resolu-tion to this CAQR, it was noted that during a seismic event the larger 2 1/2" diameter heating line would not respond with the building response since the heating line is not rigidly supported.
Therefore, the line would not produce a severe load.
A postulated impact load of 350 LBS was applied at the point of contact with the heating line and core spray line, and the stresses calculated (using i
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ASME Code) were found to be within the allowable limits.
The additional loads on core spray pipe supports due to the postulated impact load were evaluated by licensee DNE personnel and found to be acceptable.
This item is closed.
f.
(CLOSED) IFI 260/87-42-02:
Failure Investigation Reports Dealing With Resolution In A Timely Manner This item involved a followup on the process for resolving failures and implementing. corrective actions in a timely manner.
The inspector reviewed PMI-6.13, " Failure Investigation of Safety-Related Items". revision 2,
and discussed the program with several maintenance managers and engineers.
The inspector examined a computerized index of all failure investigations opened since June 1988 and noted that 23 had been opened and 12 had been closed.
l Four failure investigations that were still open (88-19, 88-23, 88-26, 88-31) were discussed with a maintenance manager.
Two of these investigations had unapproved reports that were reviewed by the inspector.
The inspector concluded that the licensee had an organized failure investigation progrr-that appeared to implement corrective actions in a timely manner.
This item is closed.
g.
(CLOSED) IFI 260/88-18-01:
Undocumented Cable Conductor Splice In Core Spray Logic Wiring This item was identified as part of the Restart Test Program and involved a portion of test RTP-2-BFN-075, " Core Spray."
The NRC
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inspector reviewed the test exception (RTP-075-TE09), review, the work request' (MR A-860183), and observed the documented newly installed splice.
The inspector observed the satisfactory operation of the core spray system during LOP /LOCA testing as well as sis performed prior to fuel load.
Based on the successful completion of the retest and SI performance, the NRC inspector determined that this item was adequately addressed.
Additional undocumented splices have not been discovered during restart test activities.
The item is closed.
h.
(OPEN) IFI 259, 260, 296/88-18-05: Major Discrepancies Identified During LOP /LOCA Test C This item was initially identified during LOP /LOCA testing and
involved the locking out of the DG 3A, 3C and 3D output breakers due to the anti pumping circuitry.
Under actual test conditions, it was discovered that the DG breakers would lock out and not respond to i
the LOCA condition after first responding to the LOP condition.
The inspector reviewed the corrective actions taken by the licensee, which consisted of the installation of a time delay relay (set at 3 seconds) for all eight of the BFN DG output breakers to their I
respective 4160 volt shutdown boards.
This time delay relay was installed in order to allow the charging motors inside the DG output l
breakers to recharge the breakers during a loss of power followed by l
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a loss of co'lant accident condition.
The NRC inspector observed the satisfactory retest of LOP /LOCA C.
All systems performed according to test requirements.
However, during the review of this item, the NRC inspector could not verify that the licensee has taken adequate steps to insure that the charging motors in the DG output breakers will recharge the breakers in two and one-half seconds or less.
If this is not adequately verified and the breakers are still charging when the newly installed time dela/ relays time out, then the breaker's anti pump circuitty will lock out the breakers.
The inspector determined that the modification and followup testing were adequate for fuel load.
However, this item will remain open for final review and closure prior to Unit 2 restart.
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(CLOSED) IFI 260/88-19-01:
Torus Temperature %nitoring System Installation Completion This item remained open pending completion of the installation of an improved suppression pool temperature monitoring system (Mark I Containment Long Term Modification Program).
Engineering Change. Notice (ECN) P0533, which was issued for installation of the new system, was closed on November 30, 1988.
The inspector reviewed the c'
.ieted ECN package and discussed the system with cognizant modifications personnel.
The installaticr. consisted of sixteen temperature detectors located in thermowell penetrations into the torus shell, one at each quencher location, associated I
electronic signal processing equipment, and a new sequence of events recorder.
The system was designed to provide a more accurate indication of torus water bulk temperature and provide temperature indication for each torus quencher The inspector reviewed the following associated information and identified no deficiencies:
l (1) Changes to drawing 2-47E610-64-3, Primary Containment system Mechanical Control Diagram.
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(2) Procedures 2-SI-4.2.F-22(A), Suppression Pool Bulk Temperature j
Calibration (Division I), and 2-SI-4.2.F-22(B) for Division II.
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(3) Completed PHT-140, Torus Temperature Monitoring, and PMT-143, l
Sequence of Events Recording System.
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NRC inspectors performed walkdowns of the installation in the field,
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including: 1) control room indicators, recorders, and annunciators; (
2) associated electronic signal processing equipment located in the j'
auxiliary instrument rooms; and 3) the as-installed arrangement of l
the 16 temperature detectors.
All equipment was installed as l
indicated on approved drawing:, and no deficiencies were identified.
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Based on the reviews and walkdowns performed, the inspectors concluded that the licensee had completed the installation of the new torus temperature monitoring system and this item is closed.
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'(CLOSED) IFI 259, 260, 296/88-24-06:
Clarification of Shutdown Condition In May 1988, an Operator Licensing Examination was conducted by inspectors from the Region II Office.
This item was identified
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because reactor " shutdown" was not defined in any text and the l
licensed operator candidates varied in their definitions of reactor shutdown.
The NRC inspector reviewed and evaluated the licensee's corrective action of defining reactor shutdown as reactor sub-criticality in Browns Ferry Operations Training Lesson Plan OPL 174.812 and OPL 171.057 which states:
Reactor Subcritical:
means reactor power is decreasing (i.e.,
negative period).
When this term is used in the E0I's, the reactor power level is not as important as the power trend and condition of the reactor.
APRM's, IRM's, and/or period indication should be used to make the determination of Reactor Subcriticality.
The Browns Ferry operators were also briefed on the definition of reactor subcriticality through operations daily turnover meetings.
This was verified by discussions with licensed operators.
Based on the above, this item is closed.
No violations or deviations were identified during the Followup of Open Inspection Items.
10.
Exit Interview (30703)
The inspection scope and findings were summarized on December 30, 1988, 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 licensee.
Item Description 260/88-35-01 Unresolved Item:
Surveillance Testing Concerns (Restart Item) (paragraph 3)
260/86-35-02 Unresolved Item:
Missed Surveillance on SGTS (paragraph 3)
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Acronyms ASME American Society of Mechanical Engineers BFNP'
Browns Ferry Nuclear Power Plant CAQR Condition Adverse to Quality r,eport CFR-Code of Federal Regulations CS Core Spray DC.
Direct Current DCN Design Change Notice DCR Design Change Request DG Diesel Generator
.DNE Department of Nuclear Engineering EA Escalated (Enforcement) Action ECN Engineering Change Notice EECW Emergency Equipment Cooling Water ENS Emergency Notification System ESF Engineered Safety Feature FSAR Final Safety Analysis Report GE General-Electric GL Generic Letter.
HPFP High Pressure Fire Protection IE Inspection and Enforcement IFI.
Inspector Followup Item IRM Intermediate Range Monitor KV Kilovolt KW Kilowatt LER Licensee Event Report LRED Licensee Reportable Event Determination LOCA Loss of Coolant Accident LOP Loss of Power MIC Microbiological 1y Induced Corrosion MMI Mechanical Maintenance Instruction MR Maintenance Request MRC Management Review Committee NOV Notice of Violation NRC Nuclear Regulatory Commission OI Operating Instruction PMI Plant Manager Instruction PMT Post Maintenance / Modification Test RHR Residual Heat Removal RHRSW Residual Heat Removal Service Water i
RPS Reactor Protection System l
RTP Restart Test Program l
SBGT Standby Gas Treatment SCR Significant Condition Report SDSP Site Director Standard Practice SGTS Standby Gas Treatment System SI Surveillance Instruction SP0C System Pre-Operation Checklist SRN Specification Revision Notice I
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.x0 Senior Reactor Operator TACF Temporary Alteration Change Form TE Test Exception TI Technical Instruction TS Technical Specifications TVA Tennessee Valley Authority USQ Unreviewed Safety Question V
Volt VIO Violation URI Unresolved Item USQD Unreviewed Safety Question Determination
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