IR 05000259/1987033
| ML20236S727 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 11/18/1987 |
| From: | Brooks C, Christnot E, Ignatonis A, Andrea Johnson, Patterson C, Paulk G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20236S706 | List: |
| References | |
| RTR-NUREG-0737, RTR-NUREG-737, TASK-2.F.1, TASK-TM 50-259-87-33, 50-260-87-33, 50-296-87-33, IEIN-87-013, IEIN-87-13, NUDOCS 8711300062 | |
| Download: ML20236S727 (25) | |
Text
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ATL ANTA. GEORGI A 30323.
'k ,/ l ..... , Y i Report Nos.
50-259/87-33, 50-260/87-33, and 50-296/87-33 i Licensee: Tennessee Valley Authority 6N 38A Lookout Place f 1101 Market Street l Chattanooga, TN 37402-2801 i Docket Nos. 50-259, 50-260, and 50-296 License Nos. DPR-33, DPR-52, and DPR-68
- i Facility Name: Browns Ferry Nuclear Plant Inspection at Browns Ferry Site near Athens, Alabama Inspection Conducted: September 1 - 30, 1987 Inspectors: 6d dx d . /l!/hA 7 G. L. @ lki Sp ior Resident Inspecior Dath Sitri'ed Ifllklf9 d /T u C. A. ~ Pat (drson,'fesiden't Inspector Dage54g6ed A l) a,b l, h JXbY7 n C. R. 8F'coiff, R 'sid%n't ' Inspector Dat/e fiigne'd I C a - A i :- h/u/r7 E. F. Chris14fot,~ Rpident' ins;iector Date Sfgned i /7/1 d }/ 1, //!/ Pb7-A.JohnsonTIgspect6f Dat( Sif ned' Approved by: [ d.
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' SUMMARY . Scope: This routine inspection was in the areas of operational safety, maintenance observation, surveillance: testing observation, reportable-occurrences, restart test program, review of licensee's management goals and objectives specified in Nuclear Performance Plan, simulated loss of offsite power testing and compliance with general design criterion, design changes and i modification, diesel generators, administrative control of special tests, information notice review, and followup of licensee's action to the identified crack in the main stack.
Results: Three violations were identified: a.
A violation of 10 CFR 50, Appendix B, Criterion III for inadequate design control
j b.
A violation of 10 CFR 50, Appendix B, Criterion V for inadequate maintenance control of activities c.
A violation of 10 CFR 50, Appendix B, Criterion XVII for a lost QA record i l i _ - _ - - _ - - - - - _ - _ _ _ _ - . - -
. _ _ _ - - _ _ _ _ - _ _ - _ _ _ _ _ _ _ _ _ _ - ' ' pn Rrog'o,^ UNITED STATES , / NUCLEAR REGULATORY COMMISSION 'f3 9 -) D ~ REGION 11
I.
101 MARIETTA STREET, N.W.
ATLANTA, GEORGI A 30323 %...../ Report Nos.
50-259/87-33, 50-260/87-33, and 50-296/87-33 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 Nuclear Plant Inspection at Browns Ferry Site near Athens, Alabama Inspection Conducted: September 1 - 30, 1987 Inspectors: Md. - ,a,I J /l!/NA 7 G. L. yulk',~ Sphior Rbsident Inspector Date Sitried n n.,,A sd.:. n/ix/r? C. A. Tat {drson,'ybsident Inspector DafeS/igned I /l /) ~,b l, ll IXh7 - C. R. BFooky, R 'sfdf n't ' Inspector Da3/e 'S~i'g ne'd M,1Li;- h/is/r7 E.F.ChristMot,"Rpident' Inspector Dafe SYgned 77 /1 /8 _ SIC L //!/f/f7 A. JohnsonT IpspeEt6f Datd Signed' ] Approved by: [ d. -, k, m
/// ///f'-) A.J.Igna[5his','$$tTonChief, Data Sigjn'bpf Inspection Programs TVA Projects Division i i
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. SUMMARY Scope: This routine inspection was in the areas of operational safety, maintenance observation, surveillance testing observation, reportable occurrences, restart test program, review of licensee's management goals and objectives specified in Nuclear Performance Plan, simulated loss of offsite power testing and compliance with general design criterion, design changes and modification, diesel generators, administrative control of special tests,
information notice review, and followup of licensee's action to the identified crack in the main stack.
Results: Three violations were identified: a.
A violation of 10 CFR 50, Appendix B, Criterion III for inadequate design control \\ b.
A violation of 10 CFR 50, Appendix B, Criterion V for inadequate maintenance control of activities c.
A violation of 10 CFR 50, Appendix B, Criterion XVII for a lost i QA record
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- _ _ - _ - _ _ - . . . . . REPORT DETAILS 1.
Licensee Employees Contacted: H. G. Pomrehn, Site Director
- J. G. Walker, Plant Manager P. J. Speidel, Project Engineer
- J. D. Martin, Assistant to the Plant Manager R. M. McKeon, Superintendent - Unit 2 J. S. Olsen, Superintendent - Units 1 and 3 T. F. Ziegler, Superintendent - Maintenance
- D. C. Mims, Technical Services Supervisor J. G. Turner, Manager - Site Quality Assurance
- M. J. May, Manager - Site Licensing
- P. P. Carier, Compliance Supervisor A. W. Sorrell, Health Physics Supervisor R. M. Tuttle, Site Security Manager J. R. Kern, Fire Protection Supervisor
- D. A. Pullen, Office of Nuclear Power, Site Representative Other licensee employees contacted included licensed reactor operators, auxiliary operators, craftsmen, technicians, public safety officers, quality assurance, design and engineering personnel, and licensee contractors.
- Attended Exit Interview 2.
Exit Interview (30703) The inspection scope and findings were summarized on October 2,1987 with the Plant Manager and Superintendents and other members of his staff.
The licensee acknowledged the findings and took no exceptions.
The licensee did not identify as proprietary any of the materials provided to or reviewed by the inspectors during this inspection.
3.
Licensee Action on Previous Enforcement Matters (92702) (Closed) Violation (260/86-40-04).
This violation was for failure to correct conditions adverse to quality as required by 10 CFR 50 Appendix B, Criterion XVI.
Maintenance requests (MRs) were written to repair conduits i to the valve operators for the Main Steam Isolation Valves, but the MRs were closed out without correcting the deficiencies.
This was recognized as a programmatic deficiency in the dispositioning of MRs.
TVA revised Site Director Standard Practice 6.5, Maintenance Request and Tracking to require coordination of MRs with the originator when the deficiency or concern cannot be located.
MRs 777217, 770707, 777203 were compicted to correct the identified deficiency.
The inspector reviewed the completed MRs and procedure revision.
This item is closed.
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- _ - _ _ - _ - _ - - - - _ __ - . . . . . (Closed) follow-up Item (259, 260, 296/86-40-13).
This item was a concern whether a modification to ' decrease the stroke time of Primary Containment Purge Valves cddressed the pressure drop across the valves.- This modification brought the plant into compliance with NRC Branch Technical Position CSB G-4,. Containment Purge and Venting During Normal Plant Operations.
The effects of total pressure and pressure drop were considered and-evaluated in a letter to T. A. Ippolito from L. M. Mills dated June 2,1981.
This letter was in response to D. G. Eisenhut's letter dated September 27, 1979, to all light water reactors and T. A. Ippolito's letter dated October 22, 1979, concerning containment purgo valve operability and purge operation.
The correspondence and analysis was reviewed by the inspector.
The modification changed the valve closure time from 15 seconds to less than 2.5 or 5 seconds as required.
The accident chosen for determining valve operability was the design basis loss of coolant accident.
This accident produces the highest drywell pressures and therefore maximizes the loads on the valves.
Operator torque, hydrodynamic torque, and frictional torque were considered in the analysis.
This item is closed.
(Closed) Violation (259/85-36-05).
This violation was for failure to have an adequate residual heat removal (RHR) cross-tie procedure and to adhere to a high pressure fire protection system valve lineup.
The RHR System Operating Instruction (01-74) for all units was revised on
January 23, 1986, to incorporate steps to cross-tie RHR loops.
Copies of j the procedure revisions were reviewed by the inspector.
The failure to adhere to the valve lineup was determined to be a random procedural error.
I This item was discussed with operations personnel during supplemental training.
Completion of training was achieved by December 1985.
However, , TVA provided a supplemental response on June 19, 1987, because a review of l
this item indicated a significant number of operations personnel had not l received the initial training.
The additional training was completed in i l August 1987.
The training attendance sheets were reviewed by the inspector.
This item is closed.
(Closed) Unresolved Item (259, 260, 296/85-57-04).
This item noted that l out of 40 items identified by the licensee as potential licensee event l l reports, half were past the 30 day due date for reporting.
Plant I procedures have been revised or issued to resolve this concern.
Three procedures are SDSP-15.2, fiandling of Engineering Reports (ER) from Division of Nuclear Engineering; SDSP-3.7, Corrective Action; BF-15.2, Licensee Event Report.
There are two methods by which potentially i reportable items are identified: Licensee Reportable Event Determinations I (LREDs) and Condition Adverse to Quality Reports (CAQRs).
If an item is identified as a potential Licensee Event Report (LER), it is resorted to the Shift Engineer who determines whether the item is immed'ately reportable.
After this determination is made, the original LRED is forwarded to the Plant Operations Review Staff (PORS) supervisor for 30 day deportability determination.
This determination is normally completed within three days.
During this process the item is added to the Plant Safety Issues List maintained by PORS.
Items determined to be LERs are then assigned to site groups for investigation.
When this investigation _ _ _ _ _ _ _ _ _ _ ._-
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, is completed, the report is then reviewed by the Plant Operations Review Committee (PORC).
This report is due out of the plant within 30 days of the event.
If an event or item is identified on a CAQR and potentially affects operability, the CAQR is hand carried to P0RS.
In addition, any Condition Adverse to Quality (CAQ) determined to be significant is immediately transmitted to P0RS.
PORS then reviews the items for deportability.
If the item is determined to be reportable, a report will be initiated and the item handled in accordance with BF-15.2.
Items identified in the Division of Nuclear Engineering (DNE) are handled in accordance with SDSP-15.2.
DNE may forward a CAQ in two ways.
If DNE considers the CAQ to be of immediate concern to safe operation, the DNE Project Engineer will call the Site Director.
Otherwise, DNE will hand carry or telecopy the CAQ to P0RS for disposition in accordance with SDSP-3.7.
The current Plant Safety Issues List (July 13, 1987) contains only four overdue LREDs.
All of these items were identified prior to issue and/or revision of the three procedures.
Under current procedures, these items would receive prompt evaluation and reporting, if required.
The four backlog items are being worked off by various plant programs.
This item is closed.
(Closed) Follow-up Item (259,260,296/85-57-07).
This item is similar to 259, 260, 296/85-45-07 which was closed in inspection report 87-27.
This item is closed.
(Closed) Follow-up Item (259, 260, 296/85-45-06).
This item identified that clearance tags in the RHRSW building hung on local pump control switches were found displaced and not attached.
The licensee reviewed the plant clearance procedure 'BF 14.25) and found the requirements adequate.
Protective tags are required to be securely attached with waxed string, plastic ties, or nylon ties.
Additionally, BF 14.25 is being presented to all operations personnel in the accelerated requalification training.
Copies of the training subjects and Bf 14.25 were provided to the inspector for review.
This item is closed.
(Closed) Unresolved Item (259, 260, 296/86-36-01).
This item concerned the deletion of Site Direccor Standard Practice 15.3 Potentially Significant Safety Issues.
TVA stated this was deleted to focus attention on restart issues.
Significant safety issues are now resolved by means of conditions adverse to quality.
This item is closed.
(Closed) Violation (259, 260, 296/81-37-01).
This violation noted that Operating Instruction (01)-77, Operation of Radwaste Disposal System and 01-84, Containment Atmosphere Dilution, contained several procedural errors.
01-77 was revised to incorporate system walkdown comments.
A copy of the procedure was reviewed and changes noted as NRC commitments to Report 81-37.
01-84 was deficient in regards to th9 containment venting
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. procedure.
This procedure is now included in the Emergency Operating Instructions (E0I-1 and E01-2) and has been deleted from 0I-84.
Copies of E01-1 and E0I-2 were reviewed with no comments.
This item is closed.
(Closed) Violation (259, 260, 296/85-28-10).
This violation was for failure to have the high pressure fire protection system aligned to the high pressure fire heaaer as required by Technical Specification 3.11.A.1.a.
After a portion of the system was removed from service for maintenance, the system was not properly returned to service.
The licensee determined that the root cause of the event was personnel error.
The personnel responsible for the error were disciplined.
A review of the fire header isolation incident, with emphasis on procedural controls and verification at valve positioning was, conducted for operations personnel.
The inspector reviewed the lesson plan used in this review and the training attendance sheets.
This item is closed.
(Closed) Follow-up Item (259, 260, 296/86-40-01).
This item related to whether a safety evaluation concerning a number of secondary containment penetrations included a drain connection.
The inspector reviewed the safety evaluation and the list of penetrations.
The safety evaluation assumes that the standby gas treatment system does not operate and all activity is immediately released at ground level into the environment.
The safety evaluation is all encompassing.
This item is closed.
4.
Unresolved Items * (92701) Two unresolved items were identified during this inspection.
They are addressed in paragraphs 9 and 11 of this report.
5.
Operational Safety (71707, 71710) Daily discussions were held with plant management and various members of the plant operating staff.
The inspectors were kept informed of the overall plant status and any significant safety matters related to plant operations.
The inspectors made routine visits to the control rooms when an inspector was on site. 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 limiting conditions for operations; nuclear instruments operable; temporary alterations in effect; daily journals and logs; stack monitor recorder traces; and control room manning. This inspection activity also included numerous informal discussions with operators and their supervisors.
- An Unresolved Item is a matter about which more information is required to determine whether it is acceptable or may involve a violation or deviation.
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. General plant tours were conducted on at least a weekly basis. Portions of the turbine building, each reactor building and outside areas were visited.
Observations included valve positions and system alignment; snubber and hanger conditions; containment isolation alignments; instrument readings; housekeeping; 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 selected plant personnel in their functional areas during these tours.
In the course of the monthly activities, the inspectors included a review of the licensee's physical security program.
The performance of various shifts of the security force was observed in the conduct of daily activities to. include; protected and vital areas access controls, searching of personnel, packages and vehicles, badge issuance and retrieval, escorting of visitors, patrols and compensatory posts.
In addition, the inspectors observed protected area lighting, protected and vital areas barrier integrity.
Discussions were held with the site security manager in regard to methods being considered to increase l security force morale and functional capabilities.
Inadvertent Fire Protection System Fixed Sprey Initiation On August 31, 1987, Fixed Spray Automatic Deluge Valve 3-FCV-26-75A for Zone 3A was found to be discharging onto the cable trays and equipment in the northwest area of Unit 3 Reactor Building as well as the cable trays and equipment below on elevation 565.
Surveillance Instruction (SI)-4.11. A.1.a was in progress on Fixed Spray Automatic Deluge Valve 3-FCV-26-79N.
RadCon personnel notified the Unit 3 unit operator (U0) that water was running from the overhead of the 565' elevation.
The Unit 3 assistant shift engineer (ASE) and AVO were sent to the area.
Fixed Spray Automatic Deluge Valve 3-FCV-26-75A was found to be discharging.
Valve 3-FCV-26-75A was isolated and the cystem was reset for normal operation.
Testing was stopped and maintenance requests (MR) were written to test Valve 3-FCV-26-75A and to inspect equipment for damage.
Valve 3-FCV-26-75A was inspected and the valve disc, set ring, seat surface, valve internals and latch mechanism were all found to be in good working order.
The priming water pressure was then checked for possible leakage or a leaking check valve; no abnormalities were observed.
Afterwards, valve 3-FCV-26-79N was retested in accordance with SI-4.11.a.1.a and valve 3-FCV-26-75A did not initiate this time.
In addition to the above described investigation, the licensee's instrument maintenance personnel inspected the sprayed equipment and found two instruments with viater in them.
These level instruments are on the west side of the scram discharge instrument volume (SDIV), 3-LE-85-45J and 3-LE-85-45L.
It was discovered that the conduit leading to the instruments had evidence of previous water damage.
The instrument mechanics dried and cleaned the instruments and then performed a functional check of the instruments.
The electricians then sealed the conduit with RTV and after inspecting the rest of the area found no other damage.
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The conduit leading to the scram discharge instrument volume level instruments was inspected by the resident inspector.
The particular area examined was the horizontal run of conduit leading away from the vertical cable tray.
Water had entered the cable tray after the spray initiation and run along cables into the conduit.
No sealing was applied to the area where the cables enter the conduit.
While water was removed from the ' ' Unit 3 instruments described above, no action was taken to inspect the f other units for generic applicability until questioned by the inspector.
l Plant drawing 458891-1 provides the requirements for water proofing and sealing of electrical conduits.
In reviewing the Modification / Addition Instruction 27, Installation of Electrical Conduit Systems and Junction Boxes, the inspector with an electrical foreman found no reference to the l sealing of conduit openings.
On September 29, 1987, sealing of conduits was discussed with the plant manager.
This appears to be a programmatic ' deficiency not addressed by plant instructions.
A program for field inspections may be required to be developed and implemented by the
licensee to further evaluate the problem.
Sealing problems were noted during past spray initiations (Reference Report 86-16).
Accordingly, a violation will be issued for failure to comply with 10 CFR 50, Appendix B, i Criterion V, in that activities affecting quality shall be accomplished in accordance with drawings (259, 260, 296/87-33-01). Maintenance request A-753468 was generated to seal the conduit leading to the Unit 2 SDIV level elements (LE-85-45J and LE-85-45L) in accordance with drawing 45B891-1.
The corrective action to this violation should address the programmatic concerns of sealing electrical equipment, training of main-tenance personnel, and correction of modification procedures.
Resolution , of this item will be considered a startup item for Unit 2.
Fixed spray fire protection valves in the NW quadrant, elevation 593 of the Unit 3 Reactor Building have actuated on four different occasions since 1985.
They occurred on April 30,1986, Zone B (86-019); May 11, 1986, (Reference Inspection Report 86-16) Zones B and C (86-027); October 3,1986, Zone B (86-046); and October 9,1986, Zone B (86-048).
The exact cause of the initiations was not determined.
This is the second major interaction of the fire protection fixed spray system with plant i safety protection instrumentation.
In Inspection Report 86-16, an =~ interaction with the drywell high pressure switches occurred and in the current case an interaction with the SDIV instruments.
The licensee plans to pursue the problems with the deluge valves but past efforts have been ineffective.
Resolution of this problem will be tracked as a restart item for Unit 2 under the previous unresolved item 86-16-01.
6.
Maintenance Program Area a.
Maintenance Observation (62703) Plant maintenance activities of selected safety related systems and components were observed / reviewed to ascertain that they were conducted in accordance with requirements. The following items were considered during this review: the limiting conditions for operations were met; activities were accomplished using approved procedures; functional testing and/or calibrations were performed prior to o :
L i ie suiei
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. returning components or system to service; quality control records were maintained; activities were accomplished by qualified personnel;
parts and materials used were properly certified; proper.tagout l clearance procedure were adhered to; Technical Specification ' adherence;.and radiological controls were implemented as required.
, Maintenance requests were reviewed to determine status of outstanding jobs and to assure that priority was assigned to safety-related equipment maintenance which might affect plant safety.
The inspectors observed the below listed maintenance activities during this report period: , i ) (1) Electrical Maintenance Instruction EMI-85, Maintenance and Security Check of Perim - Alert II Fence Security System.
It was noted that the 2-year review of this procedure was delirquent.
A licensee representative stated that a review was currently in process.
The last review was performed in August 1984.
(2) Scram Pilot Valve Maintenance - The inspector observed the performances of Mechanical Maintenance Instruction (MMI) - 28, Control Rod Drive Hydraulic Control Unit Module Repair, Removal and Replacement.
This maintenance involved replacement of the elastomeric seals, diaphragms and 0-rings as committed to by the licensee in response to IE Bulletin 78-14.
(3) Residual Heat Removal Service Water (RHRSW) System. The inspector obo rved troubleshooting activities associated with a leak which developed in a weld.
This weld was recently made during installation of the new RHRSW pump air release valves.
The troubleshooting activities were attempting to identify a resonance by use ' of vibration monitoring equipment.
This resonance may have been responsible for fatigue failure of the weld.
b.
Measuring and Test Equipment (617248) The licensee's program for control of measuring and test equipment (M&TE) was inspected to assure compliance with Regulatory Requirements; TVA's Topical Report TVA-TR-75-1A, Quality Assurance Program Description for Design, Construction, and Operations, Section 17.2, Rev. 8; TVA Quality Assurance Program Applicable to Station Operations, Subsection 17.2.12; Control of Measuring and Test Equipment; and TVA's Nuclear Quality Assurance Manual (NQAM), Part III, Administrative Activities, Section 3.1, Control of Measuring and Test Equipment.
The upper-tier documents are general in nature setting requirements and policies which were delegated to the various sections of plant operations.
The detailed implementation as required by Part III, Section 3.1 of the NQAM is in Browns Ferry Standard Practice 17.5, i _ _ - _ - _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _.
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. Control of Measuring and Test Equipment.
BF-17.5 requires that when I an out-of-tolerance condition is discovered an evaluation shall be L made to determine the impact of the out-of-tolerance condition, however, the procedure does not adequately stipulate that a technical review shall be performed.
The licensee has identified this item and , incorporated it into the Browns Ferry Nuclear Plant Maintenance "" Improvement Plans as Maintenance Objective Number IV B.3, Measuring and Test Equipment, which has been assigned as a startup item.
The - inspector noted that approximately seven (7) groups onsite have M&TE which include the Instrumentation and Control (I&C) shop, Plant Tool - Room, Security, Main Tool Room, Mechanical Test Group, Post Modifications Group and Reactor Engineering.
The MLTE is stored in _ < various locations such as in a control environment (I&C shop, Plant Tool Toom, etc. ), in a hallway (Vibration Group), in a screened locked cage (Mechanical Test) and in a cabinet (Mechanical Test Results).
Additional M&TE is controlled by the leak rate test group and is also stored in a locked cabinet.
The I&C Engineering group has been tasked with writing a Site Directors Standard Practices e procedures SPDP 29.0, Tool Control / Measuring and Test Equipment.
The l procedure is currently in draft form and it is expected that this upper tier document will establish better control of M&TE and address ' other maintenance improvement objectives involving M&TE.
The ' inspector will follow-up with additional inspection activities upon issuance of the SPDP procedure.
No violations or deviations were identified in this area.
l l 7.
Reportable Occurrences (90712, 92700) The below listed licensee events reports (LERs) were reviewed to determine if the information provided met NRC requirements. The determination l included: adequacy of event description, verification of compliance with ! technical specifications and regulatory requirements, corrective action ' taken, existence of potential generic problems, reporting requirements satisfied, and the relative safety significance of each event.
Additional i I in plant reviews and discussion with plant personnel, as appropriate, were conducted.
The following licensee event reports are closed: l LER No.
Date Event 259/84-12 2-14-84 Shutdown Cooling System not Available due to Valve Failure to Open.
259/85-16 1-16-85 Automatic Reactor Scram Due to Loss of Feedwater ' 259/87-04 2-21-87 Accidental Bump of Radiation Monitor Output ~ , - _ _ _ _ _ _ _ _ _ _ _ _ _ - _ _ _ _ _ _ _ _ _ _ _
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. . LER No.
Date Event (cont'd) Cable Initiates Control Room l Emergency Ventilation l l 259/87-09 4-20-87 Personnel Error Causes ,__ l Engineered Safety Features ~ Actuation 259/87-15 7-9-87 Engineered Safety Features Actuation Due to Circuit Protector Trip Caused by Jarring of Panel 260/87-03 4-20-87 Reactor Vessel Refill Following Maintenance Interval Leads to Water Chemistry E.xcursion An investigation of the failed valve FCV-74-48 (LER 259/84-12) revealed that the "B" phase winding of the motor on the valve had failed.
The gate stuck in the valve seat and the mrtor could not generate enough torque to open the valve.
The "close" tes que switch setting was higher than recommended by the manufacturer.
Electrical Maintenance Instruction 18 was revised to improve recording and review of acceptance criteria and data recording of torque switch setting.
The root cause of the reactor scram (LER 259/85-16) was the malfunction of
the master level controller which caused the steam loss to the reactor feed pump turbines that then caused the scram on low reactor water level.
The problem was traced to a cold solder joint :
- action of a wire.
The wire joint was resolved and the event was considered a random failure.
The initiation of the control room emergency ventilation units (LER 259/87-04) was caused by an assistant shift engineer who bumped the signal output cable of the radiation monitor located in the control bay air inlet.
The licensed operator was counseled for not paying close attention while performing his dutfes.
The engineered safety features actuation (LER 259/87-09) was caused by a licensed operator who unnecessarily transferred IB 480 volt shutdown board from its normal power supply to its alternate power supply during preparation for unloading 1 B 4000 volt unit board for breaker maintenance.
The operator performing the board switching did not give sufficient forethought to his actions.
The operator was counseled for the error.
The engineered safety features actuation (LER 259/87-15) was caused by
craftsmen, working near a reactor protection system circuit protector, who
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, jarred the protector panel causing the trip of the circuit protector.
The craftsmen were counseled cn proper work methods regarding treatment of plant equipment.
The water chemistry excursion (LER 260/87-03) occurred during reactor vessel reflood up following replacement of the recirculation inlet nozzle safe ends.
The event will have no effect of the reactor vessel since nitrogen oxides, the primary impurities found in the water, do nct aggressively attack stainless steel.
Although, inevitable following extended periods of reactor water drain down, this could be minimized to some degree by a feed and bleed process or by expeditiously raising the vessel level and place the reactor water cleanup system in service.
The licensee stated that future procedures will attempt to make use of these alternatives.
The following licensee event reports were reviewed and remain open pending further review: LER No.
DATE EVENT 259/85-13 4-18-85 Temporary Startup Test Panel Installation 259/86-21 6-20-86 Reactor Building Flood Level Switches Not Qualified No violations or deviations were found in this area.
8.
Design Deficiency Identification During Restart Test Program During performance of Restart Test Procedure - 032 (Control Air /Drywell Control Air) on August 31, 1987, a test discrepancy was discovered.
This discrepancy has been evaluated by the licensee and found to be contrary to Safety Design Basis 7.3.3.7.a and 7.3.3.7.g contained in the 'SAR.
The test was intended to verify that the Drywell Control Air succion valves (FCV-32-62 and FCV-32-63) failed in the closed position upon loss of power to the solenoid valve and upon loss of control air.
During the test, the , valves were found to fail-as-is upon loss of control air.
These primary containment isolation valves therefore violated the design criteria which required that no single failure within the control system shall prevent essential isolation action.
This was also a common-mode failure in that , these containment isolation valves failed to close on loss of control air.
The licensee has traced the problem to an Engineering Change Notice (ECN) P-3130 which was to upgrade the solenoid valve for environmental , qualification.
Implementation of this modification was improperly made due to insufficient detail in the work plans.
Post-modification testing after this ECN consisted only of cycling the valve and therefore did not ' detect the problem.
The ECN was worked in November 1982 on Unit 2, May 1983 on Unit I and March 1984 on Unit 3.
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. The licensee considers this finding to be a' success story which emphasizes ' the value of. the Restart Test Program (RTP) and the Design Baseline Verification Program (DBVP).
The test was fnrmulated based upon the test requirements of the DBVP.
The purpose of these two programs is to assure that the design basis is properly documented and verified.
These programs should flush out all breakdowns in configuration control (similar to this finding) which may have resulted from poor control over plant modifica-tions over the last ten years.
The only questions or potential adjustment - to the DBVP program that this example raises relates to the timing of restart testing.
In order to meet schedule commitments, system restart tests may proceed before. all outstanding modification are completed.
Considering the possibility that the restart test may have been completed prior to this modification (or any similar modification), configuration control would depend upon proper implementation of workplans and post-modification tests.
Since outstanding work plans may not include all aspects of the new interim modification control process, similar break-downs are possible.
Further NRC inspections will be conducted in this area.
Several other miscellaneous problems noted by the inspectors during their review of this event are: a.
The new configuration control drawing (CCD) for this - system (2-47E610-32-2) depicted these valves as diaphragm valves.
In actuality, these fall into the general classification of cylinder , operated valves.
I b.
The new CCD for this system erroneously indicated on Note 8 that air , supply for these valves was from the drywell control air supply.
It ! is.actually from plant instrument air.
c.
FSAR Table 7.3-1, Pipelines Penetrating Primary Containment, does not include these valves.
They should be shown as a Group B Valve at Drywell penetration X-48.
I Since this was a licensee identified violation, no citation will be issued; however, corrective action on the drywell control air suction ' valves and the above miscellaneous problems will be tracked as an Inspector Follow-up Item (259, 260, 296-87-33-02).
9.
Nuclear Performance Plan Review of Management Goals and Objectives a.
Walk Your Space Philosophy In a request for additional information related to the Browns Ferry Nuclear Performance Plan (NPP), dated October 21, 1986, the NRC implied that improvements in effective management of the facility were not evident.
Tne NRC recommended that three specific action , items contained in the Regulatory Performance Improvement Program I (RPIP) be effectively implemented after proper retraining.
These items were: ! _ _ - _ _ _ - _ __- _ -___ _ _____ _ __ _ _ __ _ --- --- _ _ _ _ _ _ - - _- __ . _.
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, (1)' Managers must perform routine observations in the workp1' ace using checklists and guidelines. with time allocated for independent observation.
(2) _ The time available for plant supervisors to be involved in workplace activities must be increased.
l (3) Require managers, supervisors, and engineers to' observe at-least one activity per day.
' TVA responded.to. this letter in Revision 1 to the. NPP dated June 23, 1987.
They stated that proceduralized frequency of observation and prescriptive guidelines. and checklists were ultimately ineffective.
Implementation of these RPIP items using checklists and daily observations n re discontinued in' favor of implementing an' overall management chilosophy defined by the phrase " walk your space".
According to the licensee, this phrase represents an approach to l management which emphasizes management awareness, involvement, and ownership of activities in the workplaco.
" Walk your space" brings managers to the problems.
It requires managers to spend time with their people where the work is being performed.
Observing the work helps to 1) ensure the procedures are appropriate and are being fot Dwed,. 2) find and solve problems in performing the work, and 3) Improve the avera?1 quality of the final product.
Time spent in the workplace facilitates communication of corporate and site policies cnd provides feedback regarding their implementation while keeping supervisors cognizant of problems and issues.
The licensee reported that they had.made a good start in adopting the " walk your space" philosophy at Browns Ferry but that continuous emphasis and follow-up would be necessary before it would become a permanent part of the BFN culture.
The inspectnr learned that " walk your space" reports were being. documented and were available for review.
No reporting or documentation of managers walk-throughs was originally required.
On June 9,1987, the Site Director issued a memorandum to the tap site managers stating that the current implementation of the " walk your space" concept was not producing the desired results.
He, therefore, began requiring weekly reports documenting the location, time, ' findings and follow-up activities of managers walk-throughs.
The ' inspector reviewed a sampling of these walk your space reports in i order to evaluate the current status of the program.
As would be expected, some managers implement this concept better than others.
A large number of the managers appear to be merely documenting their weekly activities (e.g. meetings attended, training completed, tasks l performed, discussions held, conflicts resolved).
Not much was documented in the way of procedure problems, obstacles in the performance of work, improvements in quality or productivity in these j types of reports.
Only one individual docuented walk-throughs __
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. during backshift hours.
Examples of productive activities which were documented include: (1) Additional radiological shielding was recommended for a valve inside the Unit 3 Drywell.
(2) Housekeeping was poor in the Unit 2 Steam tunnel.
l (3) Decontamination of the work space during rebuild activities on a Residual Heat Removal (RHR) pump was ordered.
(4) Management needs to avoid the " Management by meetings" syndrome.
Many complaints were voiced that work could not be completed due to an excessive meeting schedule.
(5) Supervisors and managers are not properly delegating ret., possibility and authority.
Too often, excuses are made that "he's on vacation" or "he's in training" for not having answers or statuses on key issues.
(6) A potential radiological " spill" was observed by one manager who asked thoughtful, probing questions in order to guide the responding individuals into confirming their assumptions regarding the source of a potentially contaminated liquid.
Only one problem needing resolution was significant enough to require follow-up action by the inspector.
This item related to control of software for the Process Computer and will be tracked as an Unresolved Item (259, 260, 296/87-33-03) pending completion of the licensees evaluation.
The issue is whether the software should be controlled as prescribed by the Nuclear Quality Assurance Manual (NQAM), Part I, Section 2.2.1, Quality Assurance for Computer Software Systems.
Currently it is not.
Recent NRC guidance in this area is available in NUREG/CR-4640, Handbook of Software Quality Assurance Techniques Applicable to the Nuclear Industry.
During a discussion with operational readiness personnel, the inspectors were informed that consideration is being given to development of a chetklist for use by managers during their walkthroughs.
b.
Position Descriptions As part of the root cause analysis of the management breakdown problems, TVA identified in the Nuclear Performance Plan (NPP) that organizational and personal responsibilities have not been clearly defined for past corporate and BFN organizations.
This had resulted in overlap of responsibility and, in some cases, responsibility for important items was not assumed by anyone.
Mistakes had often been blamed on "the system" with no one being held accountable.
Substandard personnel performance had often been accepted due to . - _ _ - _ _ - _ - - - ,
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, management's inability or lack of desire to hold employees accountable.
TVA stated in the NPP that this was no longer the case.
Each BFN manager and employee is held accountable in his area of responsibility for achieving excellence of performance.
TVA has committed to provide the authority and resources necessary to carry out employees and managers' responsibilities.
The basis for employee performance accountability is as follows: Position descriptions have been issued for each management position at BFN.
These descriptions have been written to avoid overlapping responsibilities and to ensure that necessary. functions are assigned.
They have been reviewed by a team of top level managers and outside contractors and have now been distributed.
As new positions are established, position descriptions are prepared and issued.
Accurate position descriptions help to ensure that BFN managers know their areas of responsibility and will be accountable for the safety, technical adequacy, and quality of their work.
The resident inspector reviewed position descriptions for the plant manager and his subordinates.
Contrary to the statements in the NPP, no approved position descriptions are yet available.
Only draft versions are complete pending final approval by the Site Director.
l The following concerns were identified during the review.
(1) The position description (PD) states that the Site Security Manager reports to the Site Director.
The current organization chart has the Security Manager reporting to the Plant Manager.
In practice he reports to the Site Director.
(2) The PD states that the Public Safety Chief reports to the Dnager, Nuclear Operations.
The organization chart has him reporting to the Site Security Manager.
(3) 7he Plant Manager's PD states that he is responsible for i submitting to Nuclear Engineering all operating, maintenance, emergency and other procedures for review of technical adequacy.
In practice no plant procedures are reviewed by Nuclear engineering.
(4) The PD states that the Plant Operations Review Staff (PORS) supervisor directs preparation and approval of all safety evaluations related to plant conditions adverse te quality.
He l also ensures proper coordination nd preparation of technically l involved Unreviewed Safety Question Determinations (l!SQD) when ' requests are made by plant sections.
The Education and Training requirements for this position are deficient in that no training l 1s required in the area of 10 CFR 50.59.
In fact, the current ! P0RS supervisor has not had 10 CFR 50.59 training and is not on j the list of approved personnel for preparing and reviewing _ ___ - - ____-_
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. l 10 CFR 50.59 safety evaluations contained in Standard Practice 12.24.
(5) The Supervisor, Plant Reporting Unit is responsible for providing day-to-day interpretations of plant Technical Specifications and maintains the Technical Specification Interpretation Manual.
No special training or qualifications are required of the person filling this position.
In practice, the person filling this position is currently performing other tasks.
(6) The Supervisor, Plant Assessment Unit is responsible for investigating problems, assessing root causes, recommending corrective action and assessing effectiveness of ongoing corrective actions.
He advises the Plant Operations Review Committee (PORC), performs detailed investigations of. equipment failures, personnel errors, procedural inadequacies and. system performance.
He also directs the site SCRAM reduction program and ensures that plant operation is within the bounds of the safety analysis provided in the FSAR.
No special training or qualifications are required of the person filling this position.
' In practice, the person filling this position is currently performing other duties.
The SCRAM reduction program is coordinated by the Reactor Engineering Section.
(7) The PDs for personnel in Technical Support Services were reviewed.
No one in this unit was responsible for the materials license, byproduct material contrel, control of special tests or containment integrity testing.
(8) The Shift Technical Advisor PD does not require him to be STA qualified.
(9) The PD for the Chemistry Unit Supervisor did not include his responsibility for the Semi-Annual Effluent Release report to the NRC.
(10) The Shift Engineer PD did not include his responsibility for complying with the one hour and four hour reporting requirement of 10 CFR 50.72.
(11) Only one PD was specific in the requirement for years of nuclear power plant experience.
The Assistant to the Technical Support Services Supervisor is required to have at least one of his two years nuclear experience to be BWR experience.
No other PD distinguished between BWR or PWR experience for satisfying the educational and training requirements.
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.. 10. Microbiological 1y Induced Corrosion (MIC) Program The resident staff reviewed activities related to the site MIC program to ascertain follow-up. actions were being taken and program implementation was-ongoing in this program area.
This topic of concern was previously discussed in inspection report 87-14.
Topics.of interest raised during the previous inspection included: a.
Overall program control and implementation; b.
Corrective action for MIC. identified; c.
Long tern MIC program plan; and d.
Corrective action for EECW weld deficiencies found during radiograph of stainless steel piping in system The current review of the NIC program indicates the following specific concerns: a.
The original corporate program guidance has been negated and current program policy and direction do not exist.
b.
No action has been taken on the known deficiencies to date.
No CAQR or action plan had been initiated to date.
After identification by the inspector, a CAQR was written by the licensee to track the concerns.
c.
Generic implications at other licensee facilities msy not be fully realized due to lack of corporate direction and involvement.
The inspector will identify this item as an inspector follow-up item (259, 260, 296/87-33-04) to verify the corrective actions implemented by the licensee and improvements in the management control process.
11.
Simulated Loss Of Offsite Power Testing At Browns Ferry And Compliance With General Design Criterion (GDC)-18 In light of the finding that Browns Ferry had failed to properly control Emergency Diesel Generator load changes resulting from plant modifications, the resident inspectors reviewed surveillance testing requirements in this area.
Since the licensee's preliminary diesel load re-analysis showed a potential for overloading, the residents researched . past surveillance to determining how this condition could go undetected.
l This research indicated that the simulated loss of offsite power l surveillance does not appear to meet the intent of General Design Criteria
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, , { (GDC)-18 or the technical specifications.
The following paragraphs detail the concerns: a.
General Design Criterion 18 - Inspection and testing of electric , power systems - requires that the systems shall be designed with a capability to test periodically the operability of the systems as a whole and the full operation sequence that brings the systems into {, operation including the transfer of power among the nuclear power , unit, the offsite power system, and the onsite power system.
- Surveillance Requirement 4.9. A.1.b implements this GDC and required ~ that "once per operating cycle, a test will be conducted by simulating a loss of off-site power and similar conditions that would exist with the presence of an actual safety-injection signal to demonstrate the following: ".".". - (1) Deenergization of the emergency buses and load shedding from the emergency buses.
== (2) The diesel starts from ambient condition on the auto-start ' signal, energizes the emergency buses with permanently connected i loads, energizes the auto-connected emergency loads through the load sequcncer, and operates for greater than or equal to five minutes while its generator is loaded with the emergency loads.
(3) On diesel generator breaker trip, the loads are shed from the emergency buses and the diesel restarts on the auto-start signal, the emergency buses are energized with permanently l ! connected loads, the auto-connected emergency loads are , energized through the load sequencer, and the diesel operated = for greater than or equal to five minutes, while its generator l 1s loaded with the emergency loads.
, , l b.
The licensee's Surveillance Instruction for this complex requirement (SI 4.9.A.1.b, Diesel Generator Load Acceptance Test) is very simple.
Various relay contacts are booted and jumpers are installed such that only three auto-connected loads are carried by the diesels.
These are: one RHR pump, one Core Spray pump, and one RHRSW pump.
These ,' loads amount to as little as 74% of the short-term (less than 10 minutes) loads described in the FSAR analysis, Table 8.5-2 (1950 kw actual load as compared to 2625 kw accident load on the Units 1 and 2 "D" diesel).
This neither meets the surveillance requirement for permanently connected loads nor, the GDC-18 requirement for testing - the systems as a whnle. Failure to have an adequate surveillance to meet technical specification requirements has led to the potential of . s
having the diesel generators overloaded due to loss of design control over the period from initial operation.
General Electric Preliminary . Report, Standby Diesel Generator Loading Study, dated August 12, 1987, identifies Diesel Generator ID as overloaded and concludes that with the diesel generator supplying 4KV Shutdown Board "D" the Residual Heat Removal Pumps, (30 and 20), the Core Spray Pumps (10 and 20) and the Residual Heat Removal Service, water pumps (D2 and - - k i W . .. . _ _ _ _ _ _ _ _ _ _ _ _. _ _. _ _ _ _ _ _ _. _ _ _ _ _ _.. _. _ _ _ _ _ _ _ _ _ _ _
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, / D3) on that board would sequentially be lost on an overcurrent condition.
With "D" diesel generator analyzed to be in a degraded state coupled with the loss of one other Unit 1 and 2 diesel generator (i.e. single failure event), the effectiveness of the .. Emergency Core Cooling System is jeopardized.
This is contrary to the Final Safety Analysis Report Section 8.5.2.1, Safety Design Basis for Standby A-C Power Supply and Distribution Systems.
Resolution of I this concern will be tracked as an Unresolved Item pending review of the upgraded Surveillance Instruction, and further system evaluation , and modifications as required prior to startup (259, 260, 296/87-33-05).
12.
Design Changes and Modifications (37700) The inspector reviewed the design and modification activities associated with the Containment High Range Radiation Monitor (CHRRM).
This item is intended to satisfy NUREG-0737 Item II.F.1-3 and is a restart commitment.
The review consisted of verification of the NUREG-0737 requirement as . discussed in the following paragraphs.
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a.
Accuracy , Requirement - Instrument must be accurate enough to provide usable ,, information.
. Compliance - No acceptance criteria related to accuracy was included in the Test Scoping document for the post-modification testing.
There is no justification or calculation associated with the required . accuracy established for this equipment.
- b.
Location - Requirement - Monitors shall be located in containment to view a large segment of the containment atmosphere which will more accurately reficct and monitor accident conditions.
The monitors shall be widely separated as as te provide independent measurements and shall " view" a large fractior, of the containment volume.
Compliance - No engineering calculations exist nor was documentation available for other judgment used in the selection of the detector locations.
c.
Purpose And Function Discussion - The CHRRM has a two-fold purpose: (1) to detect and/or verify a breach of the Reactor Coolant Pressure Boundary and (2) to n . aid in determining the magnitude of the release of radioactive materials and continually assessing such releases. It should implement these areas by detection of significant releases, release assessment, long term surveillance and emergency plan actuation.
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s . Compliance * The Engineering Change Notice (ECN) and workplans were reviewed to ascertain.the changes that would implement these purposes.
No changes were required to be implemented to the emergency plan which would assign an emergency action level (EAL) to this monitored : parameter.
No changes were required to Accident Assessment portion of the Radiological. Emergency Plan (REP).
No calculations -were available nor was any objective or subjective ! guidance promulgated on what value of containment radiation would detect or verify a breach of the Reactor Coolant Pressure Boundary.
i Other problems noted during this review include: (1) The detector mounting-installation specified in the workplan was horizontal.
This conflicts with the manufacturers installation instructions which require a vertical configuration with the connector and downward.
No justification was provided for this discrepancy.
(2) No calculations were available to substantiate that the power supplies selected for this instrumentation can carry the additional load.
(3) The Retest Scoping Document did not require a containment integrity leak rate test associated with the new electrical penetration.
(4) The post modification test did not require a functional or operability verification of the Source Range Monitors (SRM) and Intermediate Range Monitor (IRH) cables which were re-routed using the new electrical penetration.
The licensee initiated action to develop the required justification or calculations as necessary.
These deficiencies will be collectively considered as a violation of 10 CFR 50, Appendix B, Criterion III for inadequate design control (259, 260, 296/87-33-06).
13.
Diesel Generator Overload The inspector reviewed a preliminary report concerning the Standby Diesel Generator Load Study for Unit 2.
This report was performed by General Electric, System Development Engineering Department and dated August 12, l 1987.
The subject of the report was the dynamic loading of the diesel generator.
This was a follow-up to an earlier Bechtel static load study whic.h indicated overload problems with the diesels due to addition of loads without adequate modification control.
The report was performed > using a computer model with the model for the diesel generator exciter obtained from test results at Browns Ferry in June'1987.
Special test 87-23 was performed to obtain the exciter data.
The report identifies that the Units 1 and 2 "D" diesel generator is over-loaded. (Reference paragraph 10. B) This information was reported to the NRC in a 4-hour report on September 22, 1987.
Special Tests 87-30 is scheduled in October j l i
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, 1987, to confirm previcus tests and provide operational data of the diesel generator excitation components to be used in the final General Electric report.
This test will be closely monitored by the resident inspector until the issue is resolved.
Discussion with the system engineer indicated that the addition of loads to the diesels without adequate design control was the probable cause of the overloading.
In particular was the addition of the ' 480 volt low pressure coolant injection (LPCI) motor generator (MG) sets.
! The LPCI MG sets have a large flywheel and provide a dedicated source of power for the LPCI motor operated valves.
This example may be a significant indicator of the breakdown in the design control process which could have prevented the plant systems from mitigating a design basis accident.
The LPCI MG set design changes were made in the early 1980's as part of licensee amendment number 51 for DPR-33, 45 for DPR-52, and 23 for DPR-68.
14.
Special Tests (ST) a.
Administrative Control During the review of ST 87-23, diesel generator excitation system test, a lack of administrative contt01 of special tests was noted.
Each test after completion is approved by the Plant Operations Review Committee (PORC) and the plant manager.
The copy of ST 87-23 that had been approved by PORC could not be located.
The test was completed June 26, 1987, and sent to PORC two weeks later.
Special tests are quality asserance records requiring a lifetime retention period.
Plant managers instruction 17.1, Conduct of Testing, provides the administrative controls and requirements for the handling of testing.
Section 6.1 of the instruction requires that special tests be filed as Quality Assurance (QA) records at the plant in accordance with Site Directors Standard Practice 2.5, QA Records which implements Nuclear Quality Assurance Hanaul Part III, Section 4.1, Quality Assurance Records.
As a result of the lost record, PORC plans to review the engineers data again and reproduce a PORC approved record of the special test.
The inspector indicated to the plant manager on September 25, 1987, that the control of special tests is divided between the technical services section for 1986 and 1987 tests and document control for older tests.
The problem appeared to be a new concern when partial control of special tests was taken away from document control.
When questioned if other records may be lost, the technical services supervisor indicated a number of other tests were lost and a condition adverse to quality report was being initiated.
Accordingly, a violation will be issued against 10 CFR 50 Appendix B, Criterion XVII, Quality Assurance Records for failure to maintain quality assurance records as required (259, 260, 296/87-33-07).
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, b.
Special Test Instruction - - While following up on an issue related to control of special test instructions, the inspector note 6 a problem with Plant Managers Instruction (PMI) - 17.1, Conduct of Testing.
This procedure - i l contains the administrative cor.trols over plant testing including Surveillance Instructions and Special Tests (as defined in \\ 10 CFR 50.59).
The 1.nstruction defines the format and requirements ! of special test instructions as well as outlining activities required upon test failures, deficiencies and deferrals.
The classification of this instruction (PMI-17.1) was noted to be Non Safety-Related.
The inspector questioned the justification for this and noted that ' the procedure which controls all Plant Manager Instructions (PMI 2.1, - ' Plant Manager Instruction Preparation and Format) nad no disctssion i criteria or any other guidance to the procedure writer on how to select the proper classification of a procedure.
This deficiency , ! will be tracked as an Inspector Follow-up Item pending correction by , the licensee (259, 260, 296-87-33-08).
l l 15.
Information Notice Review (92717) The inspector reviewed the licensee's processing of Information Notice l (IEN) 87-13 " Potential for High Radiation Fields Following Loss of Water . l From Fuel Pool."
In conjunction with a physical inspection of the Unit 1 l to Unit 2 spent fuel transfer canal expansion joint.
The Unit 2 cooling " water for the spent fuel pool was placed out of service due to 10 CFR 50, Appendix "J" modifications.
In early July, the licensee cross-connected , the Unit 1 and Unit 2 pools using the transfer canal in order to utilize ' the Unit I cooling water system to provide cooling to both the Unit 1 and Unit 2 pools.
The transfer canal was flooded, however, no specific action was taken involving the mechanical transfer canal expansion joint s integrity, in the light of the failure at Plant Hatch as discussed in IEN 87-13.
The inspector reviewed the following plant drawings: , Number Date Title - 67-M-4-47W481-13 1/17/69 Mechanical Drains and Embedded Piping Stage - III & IV
67-M-1-47E855-1 7/24/70 Flow Diagram Fuel Pool ' Cooling System 67-S-4-48N965 5/6/68 Miscellaneous Steel Feel , Transfer Canal Liner
D-0698-460-00 (C.E.DWG) 2/17/79 Feel Transfer Canal - Expansion Joint for Tennessee Valley Authority, Browns Ferry Nuclear Plant
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22 , The review indicated that the expansion joint cavity is equipped with an expansion joint seal drain, which could be used for evidence of water.
The presence of water would indicate that the expansion joint had lost integrity and pool water would be leaking to other parts of the plant or to the environment.
A review of procedures indicated that no routine check of the expansion joint seal drain valves was required.
The inspector physically observed the expansion joint from the refueling floor and could not visually detect the presence of water in the expansion joint seal cavity.
The licensee added valve 1-78-564, a 1-inch test connection on the expansion joint seal drain, to the reactor building operator's log to be checked on a daily routine.
The review of the licensee's processing of IEN 87-013 was conducted to determine appropriate dissemination of the received information and its closure as stated in licensee procedure Site Directors Standard Practices 15.9, " Nuclear Experience Review Program", which superseded licensee procedure Standard Practices BF-21.17 " Review, Reporting and Feedback of Operating Experience Items."
The IEN was verified as receiving an adequate review by the licensee for applicability determination.
It also appeared to have been disseminated to the appropriate departments for preventive and corrective actions.
No violations or deviations were observed in this area.
16.
Emergency Core Cooling Systems Acceptance Criteria The licensee has initiated a re-analysis of the postulated loss-of-coolant accident in conformance with 10 CFR 50.46, Acceptance Criteria for emergency core cooling systems for light water nuclear power reactors.
This was required as a result of the 10 CFR 50.49 environmental , qualification (EQ) modifications to fifteen Limitorque motor-operated j valve actuators.
The EQ modifications consisted of removal of the non qualified motor brakes which in turn required that the valve stroke be slowed to less than 36-inches per minute.
These valves include the Low l Pressure Coolant Injection (LPCI) valvet, Core Spray (CS) valves, and High Pressure Coolant Injection (HPCI) valves located outside the drywell.
The specific re-analysis activities consist of the following: ] a.
Update the BFN specific computer inputs to reflect refinements to the licensing assumptions made since the previous licensing anal These inputs will typically tend to compensate for the " lost'ysis.
Peak Clad Temperature (PCT) margin attributable to the increased valve stroke times.
b.
Determine the BFN response to the LOCA pipe break spectrum using the latest approved versions of computer codes according to the rules of 10 CFR 50, Section 46 and Appendix K.
Only the = pipe break sizes, types, and locations and single failure combinations required to support the changes in valve stroke times will be analyzed.
It is expected that approximately eight pipe breaks will be evaluated in order to adequately cover the break spectrum.
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c.
Compare the previous licensing analysis results to the updated I . analysis results in order to qualify the effect of the valve stroke ! l time changes and updated computer model inputs.
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d.
Prepare a licensing report documenting the results of the analysis l outlined above.
This report may be used to supplement the BFNP FSAR documentation.
In addition, technical specifications charges may be J required.
l l This issue will be tracked as an Inspector Follow-up Item (259, 260, E 296-87-33-9) pending completion of the study and inspection by the NRC.
It is undetermined at this time whether the 10.CFR 50.59 analysis l associated with the-change will require prior approval and a license amendment to be submitted.
17.
Crack In The Stack During a routine inspection in February 1986, the resident inspector identified a structural crack in the main plant stack..Three representatives from DNE conducted a preliminary inspection of the stack in April 1986, to determine the extent and/or cause of the crack observed by the inspector It is their preliminary conclusion that the crack is not due to a structural defect but the most probable cause is temperature variations on the surface.
The design organization also identified the fact that no formal inspection of the. stack has been accomplished since original construction.
DNE has initiated a contract to get a ' complete inspection of the stack.
The i contract for the necessary rigging should be awarded by mid-October.
The ! stack inspection is expected to be completed by mid-November.
DNE plans { to issue a report to close out the NRC item by December 1987.
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