IR 05000259/1986036
| ML20211A779 | |
| Person / Time | |
|---|---|
| Site: | Browns Ferry  |
| Issue date: | 01/30/1987 |
| From: | Brooks C, Ignatonis A, Patterson C, Paulk G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II) |
| To: | |
| Shared Package | |
| ML20211A561 | List: |
| References | |
| 50-259-86-36, 50-260-86-36, 50-296-86-36, NUDOCS 8702190234 | |
| Download: ML20211A779 (12) | |
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natto UNITE 3 STATES
[a NUCLEAR REGULATORY COMMISSION
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REGION 11 I *j 101 MARIETTA STREET, N.W.
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ATLANTA, GEORGI A 30323
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Report Nos. 50-259/86-36, 50-260/86-36 and 50-296/86-36 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 Conducted: October 1-31, 1986 Inspectors:
[ h. r #ML 1/30//f G. L. Paulk,/SeniorfResident Inspector Date Signed b0 mi h, If30lD hC.A.Patterton,RespntIn~spector Date Signed
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pC. R. Brooks,CResideqt/ Inspector Date Signed -
Approved by:
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cN 1/3MJ'7 A. J. Igngtonis,'gSection Chief Date Signed Division of Reactor Projects SUMMARY ~
Scope: This routine inspection was in the areas of operational safety, maintenance observation, surveillance testing observation, reportable occurrences, emergency preparedness public information program, and design changes and modifications.
l Results:
One violation 'was identified for failure to annually report changes to
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procedure per 10 CFR 50.59(b).
I G702190234 070205 PDR ADOCK 050002S9 PDR G
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REPORT DETAILS 1.
Licensee Employee Contacted H. G. Pomrehn, Site Director J. G. Walker, Deputy Site Director J. P. Stapleton, Project Engineer R. L. Lewis, Plant Manager
- E. A. Grimm, Assistant to the Plant Manager J. E. Swindell, Superintendent - Unit Three
- R. M. McKeon, Superintendent - Unit Two
- T. D. Cosby, Superintendent - Unit One T. F. Ziegler, Superintendent - Maintenance D. C. Mims, Technical Services Supervisor J. G. Turner, Manager - Site Quality Assurance
- M. J. May, Manager - Site Licensing R. D. Schulz, Compliance Supervisor A. W. Sorrell, Health Physics Supervisor R. E. Jackson, Chief Public Safety
- D. L. Miller, Quality Surveillance
- R. D. Erickson, Plant Operations Review Staff
- Attended exit interview Other licensee employees contacted included licensed reactor operators, auxiliary operators, craftsmen, technicians, public safety officers, quality assurance, design and engineering personnel.
2.
Exit Interview (30703)
The inspection scope and findings were summarized on November 7,1986, with the Plant Manager and/or 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 (259,260,296/85-06-03)
To establish control of thermo-meters all existing power stores stock have been removed.
Future thermo-meters Jurcnases will only be through power stores in Chattanooga.
All thermometers will be certified prior to issue at Browns Ferry.
Standard Practice 17.19, " Program to Establish and Maintain Certifiably Accurate Thermometers" was revised to reflect these changes.
This item is close.
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(0 pen) Violation (259,260,296/86-25-06)
Along with the violation, correc-tion of an inadequate safety evaluation was included in this tracking number.
The safety evaluation was inadequate since it failed to address the basis of technical specification 3.2 which states that plant flood protection does not depend in any way on advanced warning.
The Division of Nuclear Engineering (DNE) performed an independent safety evaluation on October 4,1986, which took the same approach as the original evaluation, namely that flood levels of large magnitudes can be predicted in advance of their occurrence.
The DNE evaluation concluded that operators would have 3 hours3.472222e-5 days <br />8.333333e-4 hours <br />4.960317e-6 weeks <br />1.1415e-6 months <br /> available to shut the flood doors and since it takes about 5 minutes to shut the door, flood protection would not be degraded by keeping the doors normally open.
The inspector identified two problems with the safety evaluation.
First, Reference 1 in the safety evaluation clearly assumes 35 minutes are required to shut the flood doors, not 5 minutes as stated.
Secondly, the one safety evaluation failed to address the TVA response to the Atomic Energy Commission (AEC) licensing question 2.2., Probable Maximum Precipitation.
In this response, TVA stated that the reactor and radwaste buildings are protected from floeding by the reactor and radwaste building watertight doors for the condition resulting from a 1-hour storm with probable maximum precipitation.
Since TVA had not established compensatory operator action to guard against flooding in this situation, plant manage-ment closed the flood doors on October 21, 1986, pending further evaluation of the normal position of the doors.
This item remains open pending further study by the licensee.
(0 pen)
Inspector Followup Item (259,260,296/86-25-07) While following up on this open item, the inspector identified another deficiency with the flood doors.
The upper, west wing hinge pin on door 182 had dropped about two inches downward due to a missing cotter pin. With the missing cotter pin, nothing prevented the hinge pin from dropping totally out of the hinge.
Damage to the door would be almost certain should the door have been operated with a missing hinge pin.
This item remains open pending licensee's corrective action to the deficiency identified above.
The Mechanical Maintenance Instruction 19, Inspection and Testing of Flood Protection Devices needs to be reviewed for adequacy by the licensee.
(Closed) Violation (260,296/84-48-01, Clearance Problems)
The licensee conducted training to emphasize requirements of the clearance procedure.
Standard Practice, BF-14.25, was revised to provide sufficient detail to ensure consistency in tagging control switches and motor operated value handwheels.
This item is closed.
(Closed) Open Item (259,260,296/82-23-02, Secondary containment PMS) The licensee has completed various design change requests and maintenance requests associated with the documented deficiencies.
The inspector observed operation of the standby gas treatment system and noted no further deficiencies.
This item is closed.
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(Closed) Violation (296/85-06-07, HPCI inoperable)
The licensee has corrected the drawing error which resulted in the incorrect limit switch setting for the HPCI steam supply valve.
SI-4.5.E has been revised to clarify conditions under which the SI is performed and to include require-ments to perform post-maintenance testing by accomplishing SI-4.5.E.1.d and e any time maintenance is performed that could affect the 25 second time requirement to deliver rated flow.
This item is closed.
4.
Unresolved Items * (92701)
There is one unresolved item described in paragraph five related to compliance with Corrective Action Program procedures.
5.
Operational Safety (71707, 71710)
The inspectors were kept informed of the overall plant status and any significant safety matters related to plant operations.
Daily discussions were held with plant management and various members of the plant operating staff.
The ins;:ectors 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.
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 l
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.
Weekly verifications of system status which included major flow path valve l
alignment, instrument alignment, and switch position alignments were performed on the electrical distribution, pressure suppression chamber and
l residual heat removal systems.
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- 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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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.
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a.
Cable Tray Supports Various cable tray supports in areas of the control bay, Diesel Generator and Reactor Buildings, station Class I structures or systems, were not adequately designed to withstand a design basis earthquake (Reference Enforcement Action Report 50-259/86-56).
TVA contracted United Engineers and Constructors (UE & C) to perform an interim seismic qualification of the unit two cable tray system.
UE &
C issued a report containing support qualification calculations and modifications required prior to Unit 2 restart.
Fourteen modifications
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were recommended.
The initial report was three volumes and contained four recommended modifications.
This was termed phase one.
During phase two the scope of the program was expanded.
A five volume report was issued with the remaining modifications.
Phase one modifications are described in volume three, appendix E, and phase two modifications are in volume five, appendix H.
A brief description of the 14 modifications (fixes) is given below:
Fix # 1 Addition of axial brace in the pump house to prevent cable tray displacement in the axial direction.
Fix # 2 Longitudinal restraint (4 x-braces) for cable tray in the pump house.
i Fix # 3 Decoupling of a Unit 2 control rod drive (CRD) hydraulic unit support from a cable tray support since the cable tray deformation could adversely affect the essential CRD support.
Fix # 4 Replacement of a missing bracket / rod hanger in the unit two cable spreader room.
Fix # 5 Correction of support clip overhang in the Cable Tunnel.
Fix # 6 Remove one U-bolt from conduit supported by two U-bolts for proper conauit flexibility.
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Fix # 7 Provide lateral and longitudinal restraints.
Fix # 8 Unrestrain cable tray at elevation 021.25 in the Unit 2 reactor building.
Fix # 9 Replacement of two fabricated clips with standard clips on elevation 621.2S in the Unit 2 reactor building.
Fix # 10 Replacement of a bent clip on one support in the Unit 2 drywell.
Fix # 11 Replacement of a missing cable tray section and two support brackets in the Unit 2 drywell.
Fix # 12 Recommends that a heating ventilation and air conditioning blower bank frame in the unit two drywell be evaluated for loads imposed by a cable tray support.
Fix # 13 Included fifteen hardware-related recommendations to correct supports in the control bay of Units 1, 2, and 3.
Fix # 14 Attachment of a vertical run to the cable tray in the Unit 2 drywell which was not attached to one support.
The inspector toured the facility to determine the scope of these modifications with a licensee representative.
Most of the fixes have been completed.
Fix # 11 was not completed and is awaiting component procurement.
Fix # 12 was evaluated and no modification was determined to be necessary.
The modifications were not as extensive as expected for the magnitude of the problem.
Several of the fixes such as fix # 5 and fix # 10 required only minor hardware changes.
These types of items have been identified in the past by the resident inspectors.
A violation was issued in Inspection Report 83-33 for failure to have cable tray hold-down clips installed per plant drawings.
At that time, a survey by the licensee identified the following deficiencies:
Unit 1, 25 deficiencies; Unit 2,18 deficiencies; and Unit 3, 21 deficiencies.
NRR is evaluating the interim program.
For the long term resolution of the inadequate design, TVA has contracted with Earthquake Engineering (EQA).
Browns Ferry is planning to use the methodology of NUREG 1030 which was developed to resolve Unresolved Safety Issue A-46, Seismic Qualification of Equipment in Operating Nuclear Power Plants.
This method of resolution uses the results of damage surveys conducted in conventional power plants and industrial facilities which have experienced actual earthquake ground motions.
These surveys may show that nonnuclear grade equipment, including cable trays, are seismically
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rugged in general and do not fail under seismic loading.
Without a favorable long term resolution the plant will require extensive major modifications (if feasible) to bring the supports up to the standards of new plants presently being constructed.
b.
Corrective Action Program During the course of tracking the progress of outstanding conditions adverse to quality, the inspector identified a lack of compliance with site procedures.
Site Director Standard Practice 15.3, Potentially Significant Safety Issues, was initiated in November 1986 as a mechanism to provide for the timely identification of significant safety issues to site and corporate management and to provide for systematic analysis, reporting and tracking of the issue to resolution.
This procedure was written to correct a weakness in past management awareness and control of significant safety issues.
The inspector's concern is that even though procedure SDSP 15.3 was initiated properly and was fine tuned through a revision as late as March,1986, it has not been adhered to since this last revision date.
The procedure requires that each principal manager maintain a list of significant safety issues and forward a report containing proposed resolutions and alternatives and current status to Planning and Scheduling (P&S) by the 15th of each month.
New issues are to be reported to the Site Director on a weekly basis.
P&S is required to distribute a monthly report of the Significant Safety Issues, alternative for resolution, and current status.
The Plant Manager is tasked with ranking the significance of each issue on the list.
However, the plant licensing and P&S representatives stated that these monthly reports are no longer generated per this procedure.
Apparently other mechanisms are used such as the Scheduling Review and Risk List.
Thus, since the monthly report generation is not being accomplished per the requirements of procedure SDSP 15.3, it is not clear, if there are other aspects of the procedure that are not being adhered to.
This matter is identified as an Unresolved Item (259,260,296/86-36-01) pending the licensee's review of procedure SDSP 15.3 to determine which aspects of the procedure are not longer applicable or required and identify any other programs that are used to assure management awareness and control of significant safety issues.
The inspector found the Corrective Action Program, described in SDSP 3.1, to be progressing well.
The monthly corrective action summary report for September 1986 was reviewed.
One hundred and fourteen Corrective Action Reports (CARS) with an average age of 6.7 months remain open.
One hundred and fifteen Discrepancy Reports (DRs)
with an average age of 3.5 months remain open.
Thirty-three open CARS are over six months old and 24 DRs are over six months old.
Although many of these CARS and DRs could be considered licensee identified violations, the shear number of items makes it impractical to document
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each one in this inspection report.
The purpose of documenting each licensee identified violation is to detect repetition of minor violations which could be indicative of a weak corrective action program.
Since the Browns Ferry Corrective Action Program has already been identified as weak and measures have been established to strengthen the program, reporting of overall numbers of CARS and DRs serves the intended purpose.
One CAR, however, deserves mention due to its age and safety significance.
CAR 83-0173 discovered in November 1983 that identical Reactor Protection System (RPS) instrumentation panels are anchored to the floor differently on each unit with no drawings to show which, if any, method is seismically acceptable.
This will be tracked as an Inspector Follow-up Item (259,260,296/86-36-02).
Another more recent CAR (number 86-0138) will be tracked as an Inspector Follow-up Item (259,260,261/86-36-03) to assure proper review in the Design Baseline Verification Program.
This CAR identifies a potential for a failure of a non-safety system (control air) to cause failure of the Drywell Control Air System.
6.
Maintenance Observation (62703)
Plant maintenance activities of selected safety related systems and compon-ents were observed / reviewed to ascertain that they were conducted in accordance with requirements. The following items were considered during this review:
the Technical Specifications were adhered to and the limiting conditions for operations were met; activities were accomplished using approved procedures; functional testing and/or calibrations were performed prior to 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 clearance procedures were adhered to; 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:
a.
Torus fill - Unit 2 b.
Control Room Emergency Ventilation motor replacement and breaker troubleshooting activities.
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No violations or deviations were observed in this area.
7.
Surveillance Testing Observation (61726)
The inspectors observed and/or reviewed the below listed surveillance procedures. The inspection consisted of a review of the procedures for technical adequacy, conformance to Technical Specifications and that the limiting condition for operation was met, verification of test instrument calibration, observation on the conduct of the test, removal from service
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8 and return to service of the system, a review of test data, testing accomplished by qualified personnel, and that the surveillance was completed at the required frequency.
a.
SI-4.2.A.10 Functional, RM 90-140,90-141, 90-142 and 90-143 b.
SI-4.7.B.10 SBGT Operability 10 Hour Run
c.
Calibration of Core Spray to Reactor Pressure Vessel Differential Instrument Surveillance The inspector selected a surveillance instruction (SI) scheduled during this report period for observation which had been revised under the SI review program for Unit 2 startup (SIRUS).
SIRUS is a program to review all sis for technical accuracy.
The inspector observed SI-4.2.8-24, Core Spray Sparger to Reactor Pressure Vessel Differential Pressure Calibration, for Unit 3 on October 23, 1986.
If differential pressure decreases below 2 pounds, an alarm gives indication of a break in core spray piping outside of the vessel shroud.
The procedure was approved for use on October 9,1986.
Prior to commencing the procedure, the instrument technicians assigned to perform the SI initiated an immediate temporary change to the procedure after reading the procedure.
Several technical corrections were necessary.
Procedure steps 4.2 A.38 and 4.2.B.38 required the drain lines to be replaced; however, these were never removed.
Test tee caps were replaced instead.
Also inconsistent was instrument A step 4.2.A.38 which stated "if not connected to", but instrument B step 4.2.B.38 stated "if connected to".
Steps 4.2.A.40 and 4.2.B.40 required venting of the instrument lines as necessary.
However, there was no way to vent the lines.
The test tee caps were previously reinstalled and no air should have leaked into the pressurized system.
Other typographical errors and procedure enhancements were noted by the technicians performing the SI.
These are listed below:
I Page 3 Figure 1 was of such poor reproduction quality as to be useless.
Page 10 Table A-3 the column headed " Contact Open, As-Found" should have read " Contact Open, As-Left".
l Page 11 Step A.32:
The annunciator name plate contained the word
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Heater instead of the correct word Header.
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Page 16 Step B.30 and B.31:
The initial column should have read 2nd Person,1st Repeat and 1st Person, 2nd Repeat instead of 1st Person, 1st Repeat and 2nd Person 2nd Repeat.
Normally, on the second instrument checked in an SI the people performing the checks reverse roles.
This methodology was explained on page five of the procedure.
The SI was performed without difficulty.
Except for the items listed above no other problems were noted.
Second person verification was required for valve manipulations.
Separate data sheets were provided for each instrument being tested.
One criticism for this SI would be that personnel closer to the work need to be in the review process prior to approval of the procedure.
d.
Jet Pump Operability Surveillance The inspector reviewed the upgraded revision of Surveillance Instruc-tion (SI) 4.6.E, Jet Pumps (and Recirculation Pump Speed Log).
This SI satisfies the requirement of Technical Specification 4.6.E.
This specification requires that the jet pumps be demonstrated operable since a failure could increase the cross-sectional blowdown area in the event of a design basis LOCA as well as eliminate the capability to reflood the core to two-thirds height following the LOCA.
As discussed in the General Electric Service Information Letter (SIL) 330, Jet Pump Beam Cracks, this SI detects failures after they occur.
An improvement was recommended to detect the significant degradation in jet pump performance that would precede failure of jet pumps by displacement following hold-down beam cracking.
Browns Ferry had previously implemented this recommendation through Technical Instruction (TI) 52.
Since the hold down beams have been replaced at Browns Ferry, TI 52 has been abandoned.
It is therefore not expected that advance warning of an impending jet pump failure would be detected by SI-4.6.E.
IE Bulletin No. 80-07 required that licensees implement these predictive techniques until the cause of jet pump beam failures has been identi-fied and corrected.
Since these conditions have been ratisfied, no regulatory basis exists for requiring continuance of TI 52.
One of the enhancements offered in SIL 330 dealt with detection of plugged jet pumps.
Since plugged jet pumps add flow resistance to the recirculation loop, Low Pressure Coolant Injection (LPCI) flow rate may be reduced.
The SI states that a resistance change of greater than 10%
would be needed to significantly affect LPCI flow.
A change in flow resistance is detected by comparison of the ratio of Recirculation Pump Speed to flow.
The inspector reviewed data accumulated by GE contractor personnel during their review of the Recirculation System at Browns Ferry.
Although the raw data was not available, some of the reduced data indicated that a jet pump may be plugged at Browns Ferry.
A Licensee representative accumulated the raw data to determine if a problem was indicated.
The inspector plotted the data on Figure 7 of SIL 330.
Since all the data points plotted within the normal operating range of the figure, normal operation of the jet pumps was indicated.
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As stated in the summary of SIL 330, performance monitoring is needed to verify that plant operation is within the licensing basis and for early recognition of problems.
Although the immediate problem with jet pump hold down beams has been solved, elimination of the performance monitoring program also eliminates detection of degraded LPCI flow from jet pump plugging.
It was strongly recommended to the plant manager that restoration of some _ form of performance monitoring be made.
Resolution of this recommendation will be tracked as an Inspector Follow-up Item (259,260,296/86-36-04).
The licensee is evaluating the recommendation.
One error was found in the newly revised SI 4.6.E.
Step 1 requires that both recirculation pump speeds be matched within 4% which is equivalent to 345 RPM.
The correct value should be 69 RPM to be equivalent to 4%.
This error was discussed with the cognizant engineer for the SI.
8.
Reportable Occurrences (90712, 92700)
The below listed licensee events reports (LERs) were reviewed to determine if the information provided met NRC requirements. The determination 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 in plant reviews and discussion with plant personnel, as appropriate, were conducted for those reports indicated by an asterisk. The following licensee event reports are closed:
LER No.
Date Event 50-259, 50-260/86-28 August 28, 1986 Momentary loss of Secondary Containment Due to Personnel Airlock Failure 50-259/86-25 July 20, 1986 Technical Specification Violation for Smoke Detectors Operability
- 50-259/86-26 July 24, 1986 Diesel Generator Inoperability that leads to a Prohibited Configuration
- 259/86-27 August 29, 1986 Faulty Starter Coil Causes Control Room Emergency
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Ventilation Fan Trips
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LER 86-27 will require a revision to accurately reflect the root cause of the spurious overloads of the Control Room Emergency Ventilation Systam (CREV).
The LER indicated that following replacement of a suspected faulty i
starter, the CREV completed its operability test.
The suspected starter was bench tested but the suspected failure could not be duplicated.
This was
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attributed to inaccurate simulatio1 of the normal operating environment during the bench tests.
In fact, the inspector learned during a routine daily tour that CREV overloads were still occurring and trouble shooting efforts are continuing. The component failure section of the LER is thus inaccurate and will require revision.
The licensee is reevaluating the submitted LER for appropriateness of root cause determination.
9.
Emergency Preparedness Public Information Program (82209)
The objective of this inspection was to verify that basic emergency information is properly disseminated to the public in the plume exposure pathway Emergency Planning Zone (EPZ).
The three sources of information reviewed for this inspection included the emergency information distributed annually by the licensee, the fact sheet entitled " Citizens Emergency Action Plan for the Browns Ferry Nuclear Power Plant" distributed by the county emergency management offices, and the detailed information on evacuation sectors and routes contained in the Law Enforcement Annex to the Alabama Radiological Emergency Response Plan for Nuclear Power Plants. A number of discrepancies were noted with the potential for hampering the evacuation of the population within the 10-mile EPZ.
These discrepancies were discussed with the TVA State Emergency Planning Coordinator and were also provided to the Region II Director of the state and Government Staff for transmittal to the Federal Emergency Management Agency (FEMA) for corrective action and followup as appropriate.
FEMA's response to these discrepancies and appropriate corrective action will be tracked as an Inspector Followup Item (259,260,296/86-36-06).
10.
Design Changes and Modifications (37700)
This inspection was performed to ascertain that design changes and modifica-tions that are determined by the licensee to not require prior NRC approval are performed in conformance with 10 CFR 50.59 and the plant technical specifications. A sampling of eight modifications were selected from the licensee's annual operating report for 1985 and 1984. This inspection was still ongoing at the end of this inspection period and will be further addressed in next months report.
While reviewing the annual operating report, the inspector noted that the section containing procedure changes was incomplete.
requires the licensee to furnish to the NRC annually or at shorter inter-vals, a report containing a brief description of changes to procedures which are described in the FSAR. The report is to contain a brief summary of the safety evaluation made for each change as well. The licensee's reports for both 1984 and 1985 simply contained a statement that various changes to procedures were made and that safety evaluations are filed at the site.
This was identified as a violation for failure to comply with 10 CFR 50.59(b) (259,260,296-86-36-05).