IR 05000259/1990003
| ML18033B231 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 03/20/1990 |
| From: | Carpenter D, Little W, Patterson C NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML18033B230 | List: |
| References | |
| 50-259-90-03, 50-259-90-3, 50-260-90-03, 50-260-90-3, 50-296-90-03, 50-296-90-3, NUDOCS 9004020100 | |
| Download: ML18033B231 (40) | |
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UNITEDSTATES NUCLEAR REGULATORY COMMISSION
REGION II
101 MARIETTASTREET, N.W.
ATLANTA,GEORGIA 30323 Report Nos.:
50-259/90-03, 50-260/90-03, and 50-296/90-03 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 Units 1, 2, and
Inspection at Browns Ferry Site near Decatur, Alabama Inspection Conducted:
January 15 - February 16, 1990 Inspectors
.
R. Carp n
,
NRC Site anager at gne a
rson, NR Restart Coor inator Accompanied by:
E. Christnot, Resident Inspector W. Bearden, Resident Inspector K. Ivey, Resident Inspector ned Approved by:
W. S. Li t
, Section C ie
,
Inspection Programs, TVA -Projects Division SUMMARY Ago f5 a
e gne Scope:
This routine resident inspection included reviews of reportable occurrences and actions on previous inspection findings.'esults:
This inspection report is primarily.a closeout of LERs and open items.
Five LERs, two Part 21 Reports, four IFIs, four URIs, and five VIOs were closed.
This effort essentially eliminates the backlog of closure items for which the licensee has closure packages prepared.
The remaining items are under review by the resident staff or other NRC offices, or are awaiting closure packages.
9000020100 900 ~0 PDR ADOCK 0 Ovv> <<9 PE~C
,
An IFI concerning correction of the facility operating license was identified in paragraph 3.i.
The license references the Fire Recovery Plan which has been superseded by Appendix R modifications.
There are no cited violations or deviations in this inspection repor REPORT DETAILS Persons Contacted Licensee Employees:
0. Zeringue, Site Director
- G. Campbell, Plant Manager
- M. Herrell, Plant Operations Manager R. Smith, Project Engineer
- J. Hutton, Operations Superintendent A. Sorrell, Maintenance Superintendent G. Turner, Site guality Assurance Manager P'. Carier, Site Licensing Manager
- P. Salas, Compliance Supervisor J. Corey, Site Radiological Control Superintendent R. Tuttle, Site Security Manager Other licensee employees, or contractors contacted included licensed reactor operators, auxiliary operators, craftsmen, technicians, and public safety officers; and quality'ssurance, design, and engineering personnel.
NRC,Employees:
- D. Carpenter, Site Manager
~C. Patterson, Restart Coordinator
- E. Christnot; Resident Inspector
- W. Bearden, Resident Inspector
- K. Ivey, Resident Inspector
- Attended exit interview Acronyms used throughout this report are listed in the last paragraph.
Reportable Occurrences (92700)
The LERs listed below were reviewed to determine if.the information provided met the NRC requirements of
CFR 50.72, The determinations included the verification of compliance with TS and =-regulatory requirements, and addressed the adequacy of the event description, the 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 were conducted, as appropriate.
a.
(CLOSED)
LER 259/88-29, Failure to Meet Single Failure Criteria for Equipment Airlock Ventilation Dampers Because of Inadequate Design Places Plant in an Unanalyzed Conditio An inspector reviewed the LER, dated October 14, 1988, and verified that it met the requirements for timeliness and content.
This item is identical to IFI 259, 260, 296/88-05-06.
The IFI is discussed in detail in paragraph three of this report.
Any corrective actions or licensee commitments associated with this issue will be followed up during the review of the IFI.
This L'ER is closed.
(CLOSED)
LER 260/89-14, Unplanned Safety Feature Actuation Due to Trip of Transformer Feeder Breaker.
A timed overcurrent signal trip occurred on the 4KV feeder breaker to the 2A 480V shutdown board transformer.
This caused deenergization of the 2A 480V Reactor MOV Board feeding the 2A RPS Bus.
The timed overcurrent relay were tested and found normal.
Visual inspection of the connecting cable, busses, and transformer did not reveal any signs of overcurrent, excessive heating or arcing.
High pot testing was performed on the electrical cables, transformer, and breakers associated with the trip.
Results were satisfactory.
Functional tests were performed on the relays.
Test results were normal.
Instrumentation was installed to monitor current when the transformer was energized.
All indications were normal.
The investi'gation by the licensee has not been able to determine a cause.
The inspector reviewed the followup actions and tests and considers this LER closed.
(CLOSED)
LER 259/89-17, Unplanned Engineered Safety Feature Actuation Due to Potential Transformer Door Falling Open.
An undervoltage trip was caused when the compartment door to the PT on 4KV Shutdown Board B fell open, disconnecting the PT fuse contacts.
The B "diesel generator started and the normal feeder breaker tripped.
This event was attributed to a
stuck latch mechanism on the compartment allowing the door to fall open.
The licensee has implemented corrective actions including verification of operability of all safety related PT compartment doors, amplifications in the operations procedure to verify secure closure, and upgrading the
.
periodic maintenance performed on the compartments.
This LER is closed.
(CLOSED)
LER 260/89-18, Technical Specification Violation Due to Inoperable Diesel Generator Caused By Air Start Motors Failure to Disengage.
During performance of a
DG SI, the air start motors did not disengage.
The diesel was declared inoperable.
The cause was determined to be the air start solenoid valve leaking air and not properly closing.
The air start motor and the air start relay valve
were replaced; The air start solenoid valve was disassembled and repair was attempted.
After reassembly, the valve still leaked air.
The solenoid valve was replaced and the applicable sections of the SI were successfully completed.
Due to this event and a similar problem in January 1988, a note was added to NNI-6, Scheduled Naintenance Standby Diesel Generators, in which an engineer and foreman are to be present during solenoid valve inspections.
The LER is closed.
(CLOSED Unit
Only)
LER 259/89-25, Design Errors In 250V DC Electrical System Results In Unanalyzed Condition.
An original design error in the 250V DC electrical system was identified that would have resulted in the plant not meeting the single failure criteria.
A failure of a
250V DC Battery Board concurrent with a
LOP/LOCA could have resulted in a loss of ECCS components in both divisions.
An additional single failure identified was that the failure of the 250V DC Battery Board would have resulted in a
loss of the shutdown cooling isolation valves capability to perform their function.
These errors in the original plant design were corrected for Unit 2 only.
The power supply for shutdown cooling isolation valve FCV-74-47 was changed from the Division I, 250V RNOV Board to the Division II Board.
In addition, 250V DC control power for the 480V AC shutdown boards has been moved from the, plant unit batteries to the 4160V AC Shutdown Board batteries.
The inspector reviewed portions of the closure package and discussed the battery sizing and voltage drop calculations with the system engineer.
New adequacy calculations, ED(200087042, Revision I have been performed for the effected batteries since the closure package was prepared.
Based on this review and discussion, this item is closed for Unit 2 only.
Concerns with Unit 1 and Unit 2 response to a Unit 1 accident signal still remain.
The Unit 1 accident signal is currently disabled.
This LER remains open for the other units, but should not impact restart of Unit 2.
(CLOSED) Part
259, 260, 296/P21 89-05, Susceptibility of Weld Between Core Spray Line and Thermal Sleeve to IGSCC.
In SIL 289, Revision 1,
GE recommended that additional weld areas in the CS system be added to inspections performed by NRC.
The inspector reviewed the licensee's closure package and applicable procedure revision concerning this item.
NNI 182, Reactor Vessel Internals and Ultrasonic Inspection, was revised on September 25, 1989, to add the recommendation of the GE SIL and included a
Procedure Step 8.2.2 stated to inspect the CS spargers in the region of T-Box junction welds for
signs of cracking and to inspect the weld joining the T-Box to its front cover plate and the creviced weld joint which connects the T-Box to the CS thermal sleeve.
This Part 21 is cl'osed.
(CLOSED) Part 21 259, 260, 296/P21 89-14, Foxboro Model N-Ell and N-E13 Pressure Transmitters Containing 10-50 MA Type Amplifiers May Experience Current Output Osci llations Due to Workmanship Deficiencies.
This Part
notified the licensee that certain amplifiers manufactured from January 1,
1988 through September 1,
1989, had workmanship deficiencies.
Similar model transmitters and amplifiers manufactured prior to January 1988, were not affected.
The licensee reviewed their records and the equipment at Browns Ferry with the report-identified model numbers and none were received subsequent to January 1,
1988.
This issue was not found to exist at Browns Ferry.
The inspector reviewed the licensee closure package for this item and considers the action taken appropriate.,
This Part 21 is closed.
3.
Action on Previous Inspection Findings (92701, 92702)
(CLOSED)
IFI 259, 260, 296/86-40-07, Complexities in Meeting TS Surveillance Requirements of 4.7.D. l.a, PCIS.
This item involved the TS requirements for the PCIS and the surveillance instructions written and performed to meet the requirements.
This item was originally identified during a review of surveillance activities in the November to December 1986 time frame.
The item addressed the following TS tables:
Table 3.7.A - Primary Containment Isolation Valves Table 3.7.D - Air Tested Isolation Valves Table 3.7.E - Primary Containment Isolation. Valves which Terminate Below the Suppression Pool Water Level.
Table 3.7.F - Primary Containment Isolation Valves Located in Water Sealed Seismic Class 1 Lines.
The item also addressed two tables located in other sections of the FSAR.
The concern was with inconsistencies in the tables and that the SIs at the time of the inspections did not adequately test all the isolation valves.
The inspector reviewed the current TS tables and found that the tables in the current TS are consistent about the type of valve and the category.
However, the new SI has not been validated.
The licensee indicated that a
new TS change is in preparation to change some of the valves and categories.
Based on the current TS, the
licensee has adequately addressed this issue.
If additional difficulties arise from the new TS change or the upgraded SI, it will be a new issue.
This item is closed.
- (CLOSED)
IFI 259, 260, 296/87-33-09, LOCA Re-analysis After Eg Modifications.
The licensee initiated a
LOCA re-analysis as a result of Eg modifications of 15 limitorque motor-operated valve actuators.
The Eg modifications consisted of removal of the non-qualified motor, brakes which in turn required that valve stroke times be increased.
The valves involved included injection valves for CS, LPCI, containment spray, torus spray, and HPCI.
A comprehensive LOCA analysis was performed by GE and documented in report NEDC-31580P, Safety Evaluation in Support of Extended Valve Stroke Times for Browns Ferry Nuclear Plant Units 1, 2, and 3.
The inspector reviewed the licensee closure package for this item.
The safety evaluation concluded that the extended valve stroke times will have an
'nsignificant impact on plant safety.
TS Table 3.7.A was revised to change the maximum operating time for the inboard LPCI valves from 30 to 40 seconds.
TS amendments 152, 148, and 123, for Units 1, 2, and 3, respectively, were issued on August 8, 1988.
The LOCA analysis performed by GE was the basis for, the evaluation used to approve the amendments.
Since this issue has been reviewed by the NRC and TS changes have been completed, this item is closed.
(OPEN) IFI 259, 260, 296/88-05-06, Potential Single Failure-Two Sets
. of Two Dampers from Two Trains are Actuated by One Relay.
This item was previously reviewed in IR 89-35 and involved a design deficiency in the secondary containment system.
Four ventilation dampers located in the equipment bay between the inner and outer doors are required to close on an initiation signal from either of two trains of the SBGT system.
Each of the two. signals actuate a
single relay which closes all four dampers.
The single failure of this relay would prevent operaton of all four dampers.
The licensee has committed to modify the logic to meet the single failure criteria during the Unit 2, cycle 6 refueling outage.
Until then the licensee will implement'ompensatory measures to manually close the dampers prior to opening the inside air lock doors.
The previous inspection (IR 89-35) left this item open pending the performance of a safety evaluation for the compensatory measures to be taken until the logic is modified.
During this reporting period, an inspector reviewed the licensee'.s closure package for this item, plant procedures, and the FSAR.
The inspector determined that operating the dampers was not described in the FSAR and the compensatory measures did riot constitute a change to the facility, a test, or an experiment.
The inspector determined that a
safety evaluation was not required to implement the
compensatory measures.
The inspector verified that procedure'-0I-64, Primary Containment System Operating Instructions, included steps to isolate and tag the ventilation dampers prior to opening the
- interior doors.
These measures should ensure that the secondary containment boundary is intact when the interior doors are open.
This issue is resolved for the restart of Unit 2; however, this item will remain open pending completion of the modifications.
(CLOSED) IFI 259, 260, 296/88-24-05, Clarification of Bypassing Hain Steam Line High Temperature Channels.
TS Table 3.2.A, note 12, allows bypassing the main steam line tunnel high" temperature channels under certain circumstances.
During a
previous operator licensing examination, an examiner ra'ised a concern that Browns Ferry personnel appeared to have been trained to inappropriately use the AOI's or EOI's to bypass the high temperature channels.
The basis for the concern was a caution in AOI-99-1, Loss of Power to One RPS Bus, which allowed inappropriate jumpering of the high temperature channels, and discussions with licensed operator candidates and training personnel.
This item was opened pending clarification and training on the basis and use of the TS allowed bypass function.
During this reporting period, an inspector reviewed the TS, the licensee's closure package for this item, applicable procedures, and operator training records.
The inspector verified that the licensee had revised AOI-99-1 to delete the reference to jumpering the high temperature channels and the operating crews had been trained on the revised AOI.
In addition, this item was reviewed by NRC license examiners during an examination given February 5-9, 1990.
No further questions were identified.
This item is closed.
(CLOSED) IFI 260/89-11-03, Deteriorated GE Cables.
During the performance of surveillance testing on Unit 2 NI, a short circuit occurred resulting in the power supply fuse blowing for the associated drawer.
The 'licensee determined that various cables in the NI System were deteriorated in that the rubber insulation had become brittle, cracked and had exposed portions of the conductor.
The NRC inspectors opened this item to document a concern about the proper identification and replacement of defective cables in all applications of this type cable in the control room and in local panels throughout the plant.
The licensee performed an additional evaluation based on information supplied by GE and determined that the use of the cables is limited to NIs, r'adiation monitors, process instrumentation, three local panels, and various control room recorders.
The cables in question are GENIE SJO SI - 53115 type cable with black Nitrile -
PVC Jacket
~
~
(GE part number 175A7294).
The conductor insulation is Styrene Butadiene (BUTA-S).
On March 29, 1989, GE issued a Rapid Information Communication Services Information Letter to all BWRs detailing the condition and the licensee issued LER 260/89-007 reporting the condition to the NRC.
In the LER the licensee attributed the cause of the failure to vendor design error in selection of the wire type.
The LER corrective actions included walkdowns of all identified panels for inspection and evaluation of the cables with replacement as necessary.
The walkdowns are to be performed prior to restart of the respective 'units.
The inspector determined that the licensee had adequately addressed the concern as stated in the original inspection report.
Any additional technical concerns will be closed as part of the followup of LER 260/89-007.
This item is closed.
(CLOSED)
URI 259, 260, 296/86-16-01, Possible Failure to Have Adequate Configuration and Status Control of High Pressure Fire Suppression System.
This item involved two incidents.
The first incident occurred on April 30, 1986, when hydrant 820 and fire protection isolation valve HCV-0-26-1021 separated from the fire header causing serious leakage from the FPHP water system.
Several fixed spray systems initiated when pressure was restored to the FPHP water system by the RSW Charging System, wetting down areas in the Unit
and 3 reactor buildings.
The second incident occurred on May 10, 1986, when a fire completely destroyed the number four cooling tower.
Following this occurrence the plant experienced additional fire protection problems when on May 11, 1986, a spurious actuation of the High Pressure Fire Suppression System occurred in the Unit 3 reactor building.
This same area was sprayed with water during the actuation on April 30, 1986.
The actuation occurred while restoring the fire protection system to normal after completion of fire fighting control efforts at the cooling towers.
Each'ire zone is actuated by operation of a deluge valve.
The valve is opened by system pressure whenever the pilot chamber pressure is diminished by any significant amount below the header pressure.
The check valves in the line charging the pilot chamber leaked when the system pressure was low.
Consequently, when the system was repressurized,,
the pressure beneath the deluge clapper exceeded the pressure in the pilot chamber and the valve actuated.
The resolution of the deficiencies in the system was identified as an URI.
The inspector reviewed the scope of fire protectio'n modifications and maintenance to prevent the spurious actuations and water spray, including the following:
Modification DCN H7887 which will blank off and disable the cable tray spray system for Unit 2 is scheduled to be completed prior to Unit 2 restart; Modifications ECN P7169, P7170, and P7772 extended the Unit 2 pre-action fire protection system to
all the Unit 2 areas including the cable trays.
This system does not use the star valve type deluge system.
The Unit 1 and 3 cable tray spray systems, have been overhauled by maintenance.
The system in these units will be modified prior to their individual startups.
The licensee initiated several MRs for the cooling tower hydrants to upgrade the valves and hydrants to include rods on the slip joints and thrust blocks to prevent separation of components.
The inspector observed that the addition of rods and thrust blocks were being performed at several other locations.
Based on the completed activities and the commitments to perform additional activities on the fire protection system, this item can be closed.
After all modifications and upgrades are finished any deficiencies subsequently identified will be handled as new findings.
The cause of the spurious actuations was mainly related to equipment problems with the deluge valves, and as such did not constitute a violation.
This item is closed.
(CLOSED)
URI 260/88-24-11, Tracking of NPP Commitments.
During a previous NRC inspection, an inspector raised the concern that commitments made by the licensee in the text of the NPP were not being tracked as licensing commitments on the CCTS.
Specific examples included Improved Management Practices and Implementation of the Systems Engineering Concept with respect to the surveillance program.
During this reporting period, an inspector held discussions with licensing personnel and reviewed the licensee's closure package for this item, including the NPP commitment list and tracking system printouts.
The inspector noted that at the time of the previous inspection the tracking system for NPP commitments included only those items on the NPP commitment list (Volume III, Attachment IV-2,
BFNPP Volume 3 Commitments)
and did not include statements of intent from the NPP, Volume III', text.
In response to this item, the licensee reviewed the text of the NPP, Volume III, for statements of intent and added them as additional commitments in the NPP commitment tracking system.
The inspector verified that the two examples identified during the previous inspection were included in the NPP commitment tracking system.
Under the new TYA NEAP, the program for tracking commitments to the NRC is detailed in ONP-STD-6. 1. 1, Managing and Tracking NRC Commitments, and is applicable to all TVA sites.
The licensee will further detail the requirements of this Standard for the Browns Ferry Site in SDSP 15.6, which will be issued after management review.
The Standard and associated SDSP include methods to track, schedule, and complete commitments made to the NRC in response to violations and bulletins, in LERs, during conversations with NRC personnel, and in other written correspondenc ~
~ The inspector concluded that this item did not constitute a violation or deviation in that the licensee had not failed to complete commitments to the NRC, and that the licensee had a program in place to ensure that commitments contained in the NPP, Volume III, as well as statements of intent, will be completed.
This item is closed.
(CLOSED)
URI 259, 260, 296/88-36-05, Adequacy of Corrective Actions in Response to NRC Violations At Sequoyah in the Area of Configuration Control.
During a previous NRC inspection, the inspectors cited the licensee (VIO 88-36-01)
for the failure to have an approved procedure governing the control of equipment, and noted various problems with the system configuration control process.
The inspectors also noted that violations for similar problems had been previously cited at the TVA's Sequoyah plant.
This item was opened to assess whether the licensee had reviewed the Sequoyah violations for applicability to the Browns Ferry configuration control program and whether corrective actions had been taken at Browns Ferry.
During this reporting period, an inspector held discussions with licensee personnel, reviewed the licensee's closure package for this item, and reviewed the licensee's program for the review of events and issues from other TVA nuclear plants.
From these reviews, the inspector determined that the Sequoyah inspection reports were forwarded to the Browns Ferry site for information and were later reviewed by the operations group.
Operations personnel at the time failed -to take appropriate corrective actions to upgrade the
'onfiguration control procedure, OSIL 43, to an approved plant procedure.
The inspector concluded that the licensee's program to review violations from other TVA sites for applicability to Browns Ferry was implemented correctly in this case.
The failure of the operations group to take appropriate corrective actions to upgrade the OSIL was cited in VIO 88-36-01 and the corrective actions for that violation will resolve this issue.
This item is closed.
(CLOSED)
VIO 259, 260, 296/84-20-01, Failure to Assure That Applicable Design Requirements Were Correctly Translated Into Specifications, Drawings, Procedures, and Instructions.
This VIO consisted of five examples and was originally identified during design reviews of various systems as follows:-"
(1)
Example A
On Nay 8, 1984, the licensee reported that a design deficiency was discovered during the Appendix R review as related to the DG logic control system.
SDB 3, of section 8.5 of the FSAR, Standby A-C Power and Distribution stated:
For the lo'ng term, greater than 10 minutes, three of the Units
and
DGs, paralleled with the three respective Unit
DGs, shall be II
adequate to supply all required loads for the safe shutdown and cool down of all three units in the event of a loss of offsite power and a
design basis accident in any one unit.
All long-term analyses for the DGs have been based upon paralleling the Units 1 and
The operational mode switch used to modify the function of the engine governor and the voltage regulator for paralleled operation is inhibited from working in parallel with unit or system mode if an accident signal is present.
Operation of the DGs in parallel with the mode control switch in the SINGLE UNIT mode of operation was analyzed in response to AEC question 8. 11, dated March 25, 1971.
In addition, there was no documented evidence that the DGs can handle the long-term load requirements for a loss of offsite power and design basis accident without paralleling the DGs SDB 10.
The licensee was informed that this item was a violation on May 25, 1984.
The inspector reviewed the licensee's response, dated October 15, 1984, which stated that the immediate corrective action was to revise plant operating instructions to include-temporary procedures to establish paralleling capability.
Field change requests were issued for all three units to allow immediate wiring changes.
ECN P5096 was issued on May 10, 1984, to revise all affected design drawings.
The logic wiring changes have been implemented for all units.
Paralleling of the DGs has been successfully tested as part of the restart test program.
No deficiencies were identified during the performance of the RTP test procedures.
With the operation and testing of the DGs, and with the AOI in place, the licensee has adequately addressed this issue.
This example is closed.
Example B
The licensee reported, on May 14, 1984, that design deficiencies had been noted related to operating the HVAC system for the unit shutdown board rooms.
The design deficiency was discovered during an Appendix R design review, NCR BRNMEB 8403.
The specific design deficiencies are discussed below.
(a)
The shutdown board rooms exhaust fans for Units 1 and 2 are automatically and permanently shed from the diesel power supply upon receipt of an accident signal concurrent with a
'oss of offsite power.
(b)
A design error allowed redundant cooling systems to be supplied from the same power source.
A single failure of this power source will result in loss of cooling to essential electrical equipmen (c)
Additionally, any single failure resulting in loss of 480V reactor MOV board 1A-Unit 1, 2A-Unit 2, or 3A-Unit 3 will disable redundant cooling systems in shutdown board rooms A-Unit 1, C-Unit 2, or the Elevation 621 shutdown board room for Unit 3.
The inspector reviewed the licensee's response, dated October 15, 1984, and the corrective action.
The licensee completed ECN 0956 to accomplish the modifications.
The inspector reviewed the ECN and WPs 2013-88, 2014-88, and 2016-88 which documented that the new AC units are powered from separate divisions.
The inspector noted that PMT 161 was performed to verify P0956.
With the completion of ECN P0966 and PMT 161, the licensee has adequately addressed this issue..
This item is closed.
Example C
This violation involved failure to maintain the HPCI/ADS electrical separation criteria as required by FSAR section 8.9.
The inspector reviewed the licensee's response, dated October 15, 1984, which stated that the long-term corrective action would consist of separating the ADS from the HPCI Division I cables to ensure that the single failure criteria is met.
ECN P-0753 was written to accomplish this modification and separation criteria had been defined for this specific change.
Inspection activities were conducted in October 1987, and reported in IR 87-42 as, well as IR 88-02.
These inspections documented that ECN-P0753 and the following associated Work Plans were reviewed:
WP 2166-85 - Cable Relabeling WP 2005-85 - Install New Conduit WP 2084-85 - Pull/Terminate New Cables WP 2085-85 - Rework Cable/Internal Wiring WP 2100-85 - Perform Modification Testing WP 2162-85 - Install Cable Tags.
Based on the inspections documted in IRs 87-42 and 88-02, the licensee adequately"'addressed this issue.
This item is closed.
Example D
On May 9, 1984, the licensee reported that a design deficiency had been noted during the licensee's Appendix R review as related to the design separation criteria for the ADS and the manual relief valve system.
The separation criteria is identified in Part X, Section A of TVA's Browns Ferry Nuclear Plant Design Changes for the Recovery from the Fire of March 22, 1975, as included in the units'perating license The Fire Recovery Plan required that the cables associated with the six valves assigned to the ADS features be separated from the cables associated with the five manual relief valves.
ECN L9059 was written and worked to accomplish the required separation.
This ECN was closed on March 9, 1976.
TVA concluded that adequate separation was never achieved and the lack of separation was made worse later due to the addition of modifications to plant systems.
During the licensee's Appendix R evaluation involving the examination of issued cable schedules and other drawings, it was found that the cables associated with the six valves assigned to the ADS shared raceways, conduit, and/or cable trays with the cables associated with the seven manual relief valves.
In addition, two manual valves have been added since 1976.
This violated the separation criteria requirements.
The licensee was informed on May 25, 1984, that this was a violation.
The inspector reviewed the licensee's response, dated October 15, 1984, and the corrective action which stated, that as part of BFN's compliance with 10 CFR 50, Appendix R, SRV control and power cables will be separated to meet Appendix R
requirements so four SRVs (manual mode) will be available for a fire located anywhere within the reactor building.
ECN P0822 was prepared to install this modification as part of the modifications being proposed to meet.Appendix R.
The licensee stated that the separation of manual SRYs from ADS SRVs was not being committed to as the Appendix R modification.
The results of additional evaluations and reviews by the licensee determined that ECN P0822 was not required, that no field work was performed as a result of this ECN, and no additional changes were necessary.
Consequently, the ECN was cancelled in June 1988 due to an Appendix R review that concluded that a
change was not required.
A number of modifications applicable to the Browns Ferry Fire Recovery Plan were no longer required under Appendix R.
The inspector reviewed the cancelled ECN and observed that approximately twelve other ECNs were cancelled due to TVA BFN Report BFEP-E1-86019, which concerned Appendix R criteria.
The inspector noted that previous NRC inspections were performed in the Appendix R
area and the findings were documented in Inspection Reports 89-13 and 89-28.
In addition to the two inspections, the NRC also issued SERs on the Browns Ferry NPP:
NUREG 1232, Volume 3, dated April 1989; and Supplement 1 to NUREG-1232, dated October 1989, Section 3. 1, Fire Protection.
A review of the NUREG, Supplement 1 and the inspection reports indicated that once all modifications are completed, the licensee's fire protection program would be acceptable for the restart of Unit 2.
A separate inspection to close open Appendix R issues is schedule to be conducted in April 1990.
Based on this information, the. licensee a'dequately addressed this issue.
This item is close (5)
Example E
During a review of seismic boundaries for IE Bulletin 79-14, the licensee identified that there was no secondary containment isolation on the reactor building heating system steam heating lines.
The plant was made aware of this on June 4,
1984, in nonconformance report BFN-NEB-8404-Rl.
The secondary containment isolation criteria is stated in the FSAR section 5.3.3.5, Locks and Penetration.
Six building heating lines between the turbine building and the reactor buildings are neither seismically qualified nor do they contain isolation valves.
The lack of secondary containment isolation valves could reduce the ability of the SGTS to draw a negative pressure in secondary containment for post accident seismic conditions.
The possible solutions included adding isolation valves or adding a seismically qualified loop seal to each line to effect secondary containment isolation.
In March 16, 1988, the licensee proposed a penetration seal upgrade program for conduits, -cable trays, and nonseismic piping.
This program was reviewed in NUREG-1232, Supplement
and found acceptable.
The program will bring secondary containment into conformance with the FSAR original design statement.
The program will be completed prior to restart.
Any new issue will be treated as a
new item.
This item is closed.
During the review of NOV item 88-20-01, example d, the NRC inspector noted that the BFNP, Unit 2 Facility Operating License, Docket No.
50-260, License No.
DPR-52, states in Section 2.c.(5);
"The facility may be modified as described in Section X of "Plan for Evaluation, Repair, and Return to Service of Browns Ferry Units 1 and 2 (March 22, 1975 Fire)" dated April 13, 1975, and revisions thereto."
The review indicated that Section X of "Plan for Evaluation, Repair and Return to Service of Browns Ferry Unit 1 and 2 (March 22, 1975 Fire)"
may be at variance with the Appendix R requirements and modifications.
This is identified as IFI 259, 260, 296/90-03-01,
'eview of Facility Operating Licenses, DPR-33 and DRP-52 for Appendix R vsSection X Requirements.
This is a restart item.
j; (CLOSED)
VIO 259, 260, 296/85-41-02, Inadequate Corrective Actions for Safety-Related Cable Tray Systems.
On February 18, 1981, licensee representatives identified a
nonconformance and initiated CAR No.81-035 in accordance with Browns Ferry Standard Practice 10.3, Corrective Action Program.
The nonconformance dealt with overfilled, cable trays and cable penetrations in the cable spreading rooms.
The root cause determination and corrective action associated with this
nonconformance was delinquent and inadequate until the CAR was upgraded to a significant status on July 9, 1985.
Various corrective actions were initiated over the period from February 1981 to July 1985, none of which succeeded in correcting or evaluating the overloaded condition of the cable trays.
The inspector reviewed the licensee's response to this violation dated October 8,
1986, which indicated that increased management attention was needed.
A contributing factor was that the importance of the CAR program was not always recognized by line management.
As part of the corrective action for this violation, the response stated that management was committed to resolving CARs expeditiously.
CARs were discussed in daily guality Assurance staff meetings, and the Site Director met periodically with the Site guality Manager and quarterly with the Corporate Director of Nuclear guality Assurance to discuss CAR status and escalate corrective actions.
Managers at all levels were being held accountable for timely resolution of CARs in their area of responsibility.
The site gA organization had been expanded to better track and evaluate CARs.
The guality Assurance Technical Services Staff had been established and was responsible for tracking the issuance, verification, and closure of CARs.
The staff was assigned the responsibility of preparing monthly reports to management to keep them apprised of the status of plant-initiated CARs.
The inspector reviewed the above procedures as well as additional information supplied by the licensee.
The original CAR was superseded by three CARs:
No.
860078 for Unit 1, No.
860079 for Unit 2, and No.
860080 for Unit 3, each dated April 17, 1986.
Each of these CARs was superseded by a series of CAgRs', starting with No.
880285 for Unit 1, No.
880286 for Unit 2, and No. 880287 for Unit 3, each dated April 19, 1988., Additional CAgRs were generated as part of the licensee's program of documented CAgRs from the TVA Sequoyah, Watts Bar, and Bellefonte nuclear facilities which have generic implications.
The end result of these activities over the past eight years has been that the licensee, through the CA(R process and the applicable engineering procedures has established an adequate corrective action program.
This item is closed.
(CLOSED)
YIO 259, 260, 296/87-38-02, Ineffective Followup on Audit Findings.
During a special NRC team inspection conducted in the licensee's DNE offices in Knoxville, TN., an inspector identified that ITRs were not being maintained as gA documents as required by 10 CFR 50, Appendix B,
Criterion XVII and the licensee's NIZAM.
The inspector subsequently determined that the problem had been previously identified by the licensee during several different gA audits.
This
constituted a failure to identify and implement proper followup corrective action.
The inspector reviewed the licensee's response to the violation dated April 22, 1988.
In the response, the licensee attributed the violation to failure to provide effective long term followup on audit findings due to not identifying a trend related to training within the DNE organization.
The adverse trend was not identified because of the methodology used in specifying audit deficiency categories for trend analysis.
As part of the corrective action the licensee reviewed all deficiencies related to training identified during EA audits from fiscal years 1985, 1986, and 1987.
The adverse trend in maintaining ITRs in DNE identified during that review will be closed as part of the followup to VIO 259, 260, 296/87-38-01.
This violation was similar to two previous violations (259, 260,
.
296/83-53-04 and 259, 260, 296/85-03-01).
These violations were identified during separate NRC special gA team inspections conducted during November - December 1983 and January 1985 at TVA's Chattanooga and Knoxville Offices to evaluate the adequacy of the licensee's management and gA controls related to licensed activities.
The closure of these two previous violations is documented in IR 89-50.
During the followup review performed for closure of these violations the inspector -determined that the licensee had developed and implemented a
new standardized Corrective Action Program which the licensee stated will result in directing management priorities at reducing the number of CAgs, and established prompt response to audit deviations.
The gA manual and plant procedures had been revised for the propose of ensuring prompt attention to corrective action and provide escalation of those items that are not completed within the required time frame.
This is to be accomplished by tracking and trending of open CAgRs and providing reports to be reviewed at the appropriate managerial levels.
Escalation to higher levels of management is performed whenever required to obtain the proper amount of attention to overdue corrective action.
Another change was that audit findings are now documented as CAgRs r'ather than tracked separately as audit items.
Audit items identified prior to the use of CAgRs for that purpose were tracked on TROI with corrective action established for each example.
During this period, the inspector reviewed the status of selected CAgRs identifying audit findings associated with several recent completed licensee gA audits.
The results of that review was the determination that the audit findings received the same consideration as other licensee identified conditions adverse to quality, i.e.
late,.approval or submission of corrective action plan or completion of scheduled corrective action items had been escalated to higher management unless approval for extension of due date was authorized.
Based on this and other recent reviews of corrective action documented in IR 89-48 and 89-49, the inspector determined that the
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licensee currently has a workable program in place to insure that planned corrective action occurs within required timeframes, or that escalation to higher levels of management occurs.
Based on discussions held with members of the licensee's gA organization and reviews of outstanding late corrective action, the inspector determined that a correct amount of attention is now being given to
,
tracking and trending CAgRs.
Except for the previously identified concern about a significant backlog of corrective action items, as identified in Inspection Report 89-49, the inspector determined that the licensee had made a considerable improvement in the area of management/tracking of gA audit findings.
This item is closed.
l.
(CLOSED)
VIO 259, 260, 296/88-16-03, Six Examples of Failure to Follow Procedures.
CFR 50, Appendix B, Criterion V.
This violation encompassed six examples of licensee's failure to adequately follow approved procedures.
The six examples are individually discussed below.
The examples, identified as la thru 1d-, were originally identified as NRC concerns as a result of the Unit 2 Drywell Fire which occurred November 2, 1987.
The examples, identified as 2a and 2b, were identified in Report 88-16.
(1)
Example la, Failure to Use a
TACF for Long Term Temporary Alterations.
The inspector reviewed the licensee's response, dated October 7, 1988, which stated that the PORC considered and discussed all the procedures which affect controls for temporary alterations.
These procedures included PMI 8.1, the SDSP for MRs, and the TVA NIZAM.
The inspector also reviewed PMI 18. 1, Temporary Alterations, which indicated that the licensee had upgraded its methodology for controlling temporary alterations.
The inspector discussed the upgraded procedure with members of the licensee's operations staff and determined that acceptable knowledge of the procedure existed.
Based on the review and discussions with applicable personnel, the licensee had adequately addressed this issue.
This item is closed.
(2)
Example 1b, Failure to Adequately Fill Out gA Records - MRs.
This item was identified after the fire when it was observed that gA records did not have all blanks filled in or marked N/A.
Many MRs were found with signatures and data missing.
The inspector reviewed the licensee's response, dated October 7, 1988, which stated that CARR BFP 871107 was generated to ensure that the MRs which lacked appropriate information on the data sheets were completed by the individuals who committed the errors.
Electrical modifications and quality control personnel
were trained on the recurrence control action which details the need to fill in or write "N/A" in all blocks as appropriate on the MR.
Concurrently, the gC inspectors were instructed to sign-off the appropriate blocks for all verification.which they have witnessed.
Ensuring personnel corrected their own errors and providing additional training should prevent recurrence of incomplete work by the affected personnel.
The inspector also reviewed CARR BFP 871107 and observed that training sessions were conducted for approximately 160 personnel.
Based on this review and the training conducted, the licensee has adequately addressed this issue.
This item is closed.
Example 1C, Failure to Perform Electrical Post-Maintenance Testing.
This item identified that no, electrical checks of any nature were performed as PNT following completion of the temporary electrical splices installed under NRs 793993 and 775468.
This was attributed to inadequate controls in MAI-45, Cable Terminating and Splicing for Insulated Cables up to 15,000 volts.
Also, ENI 7.2, Test Procedure for Initial Installation and Troubleshooting of Molded Case:Circuit Breakers, was found to be deficient in that it did not test the motor starter portion of the breakers.
The inspector reviewed the licensee's response, dated October 7, 1988, which stated that MAI-45 was revised to require that electrical continuity be verified individually. Individual wire continuity verification requires the cable which is being tested to be isolated at both ends.
This note is required for any future temporary terminations.
Additionally, the note addressed the invalidating of previous cable connections as well as eliminating electrical "cross-talk."
Electrical modification and gC personnel were trained on the revised requirements in MAI-45 for required continuity testing on any temporary terminations, and the need to ensure terminations remain validated.
ENI-7.2 was revised by an immediate temporary change to include instructions for testing motor starter overload relays to eliminate past oversights.
The DNE Branch issued drawings for four DCNs specifying overload relay heater sizes for safety-related circuits.
An ECN was generated to test each overload heater and to corr'ect drawing discrepancies.
A preventive maintenance schedule will require 20 percent of the noted relays to be tested annually to ensure that the safety related relays will be tested once every five year The inspector noted that both MAI-45 and EMI-7.2 have been cancelled.
MAI-45 was superseded by MAI 3.3, Cable Terminating and Splicing for Cables Rated up to 15,000 Volts, and EMI-7.2 was superseded by ECI-0-000-BKR -008, Electrical Corrective Instruction Testing and Troubleshooting of Molded Case Circuit Breakers and Motor Starter Overload Relays.
The new procedures were part of the procedures upgrade program and were still acceptable.
Based on this review and discussion with applicable personnel, the licensee has adequately addressed this issue.
This item is closed.
Example 1d, Failure to gualify Fire Brigade Members.
This item identified that three of the six fire brigade members who entered the drywell for fire fighting operations were not eligible'for fire brigade duty due to failure to comply with the training and. qualification requirements of FPP-1, Fire Protection Program Plan.
Altogether, 67 of 126 fire brigade members assigned to five operating crews were ineligible for fire brigade duty for the same reasons.
The inspector reviewed the licensee's response dated October 7, 1988, which stated that on June 27, 1988, five dedicated emergency response teams were assigned to BFN to respond to emergencies.
An emergency response team ERT consists of a
captain and four emergency service technicians (EST),
whose responsibility is the manual suppression of fires.
Currently, these qualified EST are assigned to rotating shifts and are available to the Incident Commander.
Fire protection procedures (FPP-1, 2,
and 3) were revised to detail manual fire suppression personnel as level I qualified and technical and support personnel as level II qualified.
This newly dedicated ERT should allow operational personnel to provide technical expertise by assessing the consequences of the fire instead of fighting the fire.
The training received by the ERT is directly applicable to their day-to-day duties, whereas, the training of the op'erational personnel qualified to fight fires was only a small portion of their qualifications.
The dedicated crews will enhance preserving essential fire information at a fire scene while fighting the fire, thereby assisting personnel conducting the fire investigation.
The NRC inspector reviewed procedures FPP-l, Fire Protection Plan, FPP-2, Fire Protection
- Attachments; and FPP-3, Fire Protection Emergency Organization and Pre-Fire Plan Based on this review and discussions with applicable personnel, the licensee has adequately addressed this issue.
This item is closed.
(5)
Item 2a - Failure to Fill Out Q-List Equipment Data Packages in Accordance with NQAM.
This item identified a
non-compliance with the QA records requirements, in the Q-List data entry, update, and input sheets contained in Tab Bl, Analysis Component Pickoff, and the Tab Bl/B2, Analyses Component Pickoff.
The following examples were observed:
a majority of the QA records were not in black ink; a majority of the corrections were not made by marking a single line through the item to be changed, marking the new entry, and entering the dated initials of the person making the correction; and there was no name or date included in the reviewer block on a majority of the input sheets, nor were these blocks marked as being not applicable.
The inspector reviewed the licensee's response, dated October 7, 1988, which stated that BFN should have identified.'QEDPs as
"information only" and should not have identified them as a
QA document.
These were personnel errors.
The letter also stated that personnel who improperly identified the QEDPs were required to read applicable procedures, NQAM part III, section 4. 1, Quality Assurance Records and NEP 1.3-Records Control.
A CAQR, BFP 880374, was generated to document and resolve this problem.
The QEDPs have been reclassified as
"for information only."
Project Instruction (PI)
87-52, Development and Control of BFN Unit
Phase I Q-list, was revised to state that the QEDPs are not QA documents.
The inspector reviewed PI 87-52'.
It was observed that NQAM Part III, Section 4. 1, Quality Assurance Records, had been cancelled and superseded by STD - 5.9.80, Revision 0 - Interim.
Based on this review and discussions with applicable personnel, the licensee has adequately addressed this issue.
This item is closed.
(6)
Example 2b - Failure to Identify a Non-compliance to General Design Criterion 2 of 10 CFR 50, Appendix A as a Significant CAQR.
This item identified that during, the review and closure of URI 259, 260, 296/87-27-03, it was noted that CAQR 87-180 had been issued in October 1987,. relative to the concern that the SBGT building seismic response spectra in the original plant design basis was underpredicted.
The inspector noted that during the TVA review process this significant concern had not
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been reported to the NRC or adequately evaluated by the licensee.
The inspector reviewed CARR 87-180 to determine why no report was received, and several deficiencies were noted that violated the requirement of TVA's SDSP 3.7, Corrective Action.
The inspector reviewed the licensee's response, dated October 7, 1988, which stated that the initial call on whether a condition potentially affects operability is made by the management reviewer of the CARR.
Additionally, the management reviewer has the responsibility to ensure that the information contained on the CARR accurately reflects the condition.
Since the issuance of the subject CA(R, specific individuals in BFEP have been designated as management reviewers of CAgRs and have received appropriate training.
SDSP 3.13, Corrective Actions, was issued and requires that a
management reviewer consult with an experienced source familiar with the potential affected site when performing an operability determination.
The current BFEP management reviewers reviewed a
critique on this violation.
The critique stressed the importance of verifying that the CARR accurately reflects the condition and provides guidance to ensure that any potential affect on operability is adequately addressed.
The NRC inspector reviewed.SDSP 3. 13.
Based on this review and discussions with applicable personnel, the licensee has adequately addressed this issue.
This item is closed.
(CLOSED) VIO 259, 260, 296/89-38-01, Failure to Conduct PMT.
This item included three examples in which equipment was returned to service following maintenance without performing PMT as required by plant procedures.
The inspector reviewed the licensee's response to the. violation and the associated corrective actions, and held discussions with licensee personnel.
The licensee determined that all three cases were due to personnel error in that craftsmen incorrectly signed as having completed the PMTs, and foremen signed the work packages without reviewing PMT documentation.
The inspector noted that the licensee took disciplinary action against the individuals involved and held discussions of the events with maintenance and operations personnel.
The licensee also implemented a policy requiring general foremen to review work packages prior to assignment to crews; personal field verification of activities and task completion; and personal review of documentation.
The licensee had experienced several problems with the'eturn to service of components and decided to incorporate this function with the LCO tracking system.
To implement this goal, the licensee revised SDSP 7.9, Integrated Schedule and Work Control; and
'
PMI 15.10, Tracking of Limiting Conditions for Operations, to include a unique tracking number for each LCO on the Impact Evaluation Sheet for activities having the potential of affecting operations which may affect the safe and reliable operation of the unit.
This process will ensure that the SOS and STA review all packages assigned to a
specific LCO number for completion prior to declaring a component or system operable.
An Impact Evaluation is required for all PMT or functional testing on plant equipment.
The inspector concluded that the revised LCO tracking system should ensure that PMTs related to components covered by TS LCOs are completed.
The general foreman review of all work activities should provide assurance that PMTs which do not affect TS components are completed.
No further questions or concerns were identified.
This item is closed.
4.
Exit Interview (30703)
The inspection scope and findings were summarized on February 16, 1990 with those persons indicated in paragraph
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. 'issenting comments were not received from the licensee.
'tem 259, 260, 296/90-03-01 5.
Acronyms Descri tion IFI, Review of Facility Operating License, paragraph 3.i.
ADS AOI ATF BFEP CAQR CAR CCTS CS DC DCN DG DNE EA ECCS ECN EMI EOI Automatic Depressurization System Abnormal Operating Instruction Alcohol Tobacco and Firearms Browns Ferry Engineering Project Condition Adverse to Quality Report Corrective Action Report Corporate Commitment Tracking System Core Spray Direct Current Design Change Notice Diesel Generator Design Nuclear Engineering Engineering Assurance Emergency Core Cooling System
'Engineering Change Notice Electrical Maintenance Instruction Emergency Operating Instruction
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ENS EQ FCV FPHP FSAR GE HPCI HVAC IFI IGSCC IR ITR KV LER LOCA LOSP/
LPCI MAI MMI MOV MR NEP NI NPP NQAM NQAP NRC OI ONP PCIS PMI PMT
- PORC PT QA QEDP RMOV RPS RSW SBGT SDB SDSP SI SIL STD TS TVA URI VIO V
Emergency Notification System Environmental Qualification, Flow Control Valve Fire Protection High Pressure Final Safety Analysis Report General Electric High Pressure Coolant Injection Heating, Ventilating, and Air Conditioning Inspection Followup Item Intergranular Stress Corrosion Cracking Inspection Report Individual Training Records Kilovolt Licensee Event Report Loss of Coolant Accident LOCA Loss of Offsite Power/Loss of Coolant Accident Low Pressure Coolant Injection Modification/Additions Instruction Mechanical Maintenance Instruction Motor Operated Valve Maintenance Request Nuclear Engineering Procedure Nuclear Instrumentation Nuclear Performance Plan Nuclear Quality Assurance Manual Nuclear Quality Assurance Plan Nuclear Regulatory Commission Operating Instruction Office of Nuclear Power Primary Containment Isolation System Plant Managers Instruction Post Modification Test Plant Operation Review Committee Potential Transformer Quality Assurance Q-List Equipment Data Package Reactor Motor Operated Valve Reactor Protection System Raw Service Water Standby Gas Treatment Safety Design Basis Site Director Standard Practice Surveillance Instructions Service Information Letter Standard Technical Specification Tennessee Valley Authority Unresolved Item Violation Volt
0