IR 05000259/1986040

Insp Repts 50-259/86-40,50-260/86-40 & 50-296/86-40 on 861101-1231.Violations Noted:Failure to Correct Condition Adverse to Quality,Per 10CFR50,App B

ML20207U024
Person / Time
Site:	Browns Ferry
Issue date:	02/18/1987
From:	Brooks C, Ignatonis A, Patterson C, Paulk G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20207U013	List: IR 05000259/1986040 ML20207U012 ML20207U016 ML20207U024
References
50-259-86-40, 50-260-86-40, 50-296-86-40, EA-84-108, IEB-80-11, NUDOCS 8703240522
Preceding documents:	ML20215A267 (29 May 1987)
Download: ML20207U024 (19)

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UNITED STATES Do

- [m3 Ktk NUCLEAR REGULATORY COMMISSION

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-g, ,j 101 MARIETTA STREET, ,

• ^t . ATL ANTA, GEORGI A 30323 -

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f Report Nos.: 50-259/86-40, 50-260/86-40, and 50-296/86-40 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: November 1 - December 31, 1986

Inspectors: bkAu da, O Date' Signed G. L. Paulkg Senior Reside j'inspegtor'

flut% L K ahh, -

D4te Gign'ed C.A.Pattepon,ResidentIljspectQr

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Gab-k3 C.R.Broolf,ResidentInsp(ttorG A-n, -

o2/UR 7 DSte / Signed Approved by: M , c , ymc 5, 6L[/f 7 A. J. Igna'tonisig$ection Chief Oate Sfgned Division of TVA Projects l'

SUMMARY Scope: This routine inspection was in the areas of cperational safety, main-tenance observation, surveillance testing observation, reportable occurrences, licensee action on previous enforcement matters, management meetings, design changes and modifications, facility modifications, and the new employee concern program.

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Results: One violation for failure to correct a condition adverse to quality as

, required by 10 CFR 50, Appendix B, Criterion XV1.

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REPORT DETAILS n Persons Contacted Licensee Employees l *H. G. Pomrehn, Site Director i

• J. G. Walker, Deputy Site Director

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• R. L. Lewis, Plant Manager

• J. D. Martin, Plant Manager Office

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• E. A. Grimm, Assistant to the Plant Manager

• J. P. Stapleton, Project Engineer

• J. E. Swindell, Superintendent - Unit Three

• R. M. McKeon, Superintendent - Unit Two

*T. D. Cosby, Superintendent - Unit One T. F. Ziegler, Superintendent - Maintenance

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• D. C. Mims, Technical Services Supervisor

- J. G. Turner, Manager - Site Quality Assurance

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• E. Hartwig, Project Management

• M. J. May, Manager - Site Licensing

• C. Beasly Information Office

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A. W. Sorrell, Health Physics Supervisor

• J. A. Savage, Site Licensing

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R. E. Jackson, Chief Public Safety

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• C. McFall, Site Licensing

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• R. K. Golub, Technical Services i *H. E. Hodges, Technical Services Other licensee errployees contacted included licensed reactor operators,

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I auxiliary operators, craftsmen, technicians, public safety officers, quality

assurance, design and engineering personnel.

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I NRC Resident Inspector i

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• C. Bearden

, * Attended exit interview Exit Interview (30703)

. The inspection scope and findings were summarized on December 12, 1986 and i January 9,1987, with the Plant Manager and/or Superintendents and other

! members of his staff. Additionally, daily discussions were held with plant

{ management and various members of the operating staff and biweekly discus-sions were held with the Deputy Site Director.

i 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 inspectio . Licensee Action on Previous Enforcement Matters (92702)

(Closed) Follow-up Item (259,260,296/83-17-02) Two plant drawings (48W1256-1 and 48W1255-1) were revised to show the completed structural modification The two wall numbers were 24 and 92 in the control bay. The revisions close the follow-up ite IE Bulletin 80-11 concerning masonry wall, however, remains ope (Closed) Follow-up Item -(259,260,296/86-11-02) This item was to review the licensee's evaluation of a licensee vendor audit findin The licensee performed a first time torque testing of the Magnetrol scram discharge volume level switches. The nut holding the switch assemblies together did not meet the required 200 to 225 foot pound torque. All the switches torque tested between 65 and 185 foot pounds torque. These nuts were torqued at the Magnetrol factory prior to installation. The licensee reported this problem in licensee event report (LER) 50-259/86-17. Corrective action consisted of tightening all the assemblies to within specified torque values, including the torque value on vendor drawings, and revision of maintenance instructions. Stated in the LER was that Magnetrol had independently reported this to the NRC under Part 21 provisions. This item is close (Closed) Open Item (259/84-53-02) and (Closed) Unresolved Item (259/84-53-03) During surveillance testing conducted December 6, 1984, the licensee discovered tnat the computer printout frem the sequence of events recorder did not correspond to the scram discharge instrumentation being teste This information is used to conduct post-trip review The licensee committed to verify a sampling of computer output data. Sixteen other alarm points were verified to be correc The specific points in error were corrected by discrepancy reports 85-0244 and 85-0243. An additional person was assigned to the team performing this surveillance for monitoring the sequence of events recorder printout. This item is close Troubleshooting found errors in the design drawings, surveillance procedure, and post modification test. Drawings 45N621, 45W1673-3, and 45W3673-3 were corrected. The drawing errors resulted in wiring errors in the alarm circuits to detectors "F" and "G" for the sequence of events recorder for units one and thre The wiring errors were correcte Surveillance instruction SI 4.1.A-8 was corrected to show LE-85-45B in the west instru-ment volume and LE-85-45H in the east instrument volume. The methodology of conducting post modification testing was changed. Tests will stand alone in the future and not be solely dependent on surveillance instruction This item is close (Closed) Violation (296/85-36-02) This violation was against 10 CFR 50, Appendix B, Criterion V for failure to have the High Pressure Coolant Injection (HPCI) system torus suction valve, 73-27, electrically connected in accordance with plant drawings. The motor series and field windings were found connected as a differentially compounded direct current motor instead of cumulatively compounded. This configuration caused the valve to operate faster than similar valves on other units. A special test was performed to

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determine if the incorrectly wired motor could develop sufficient torque to operate the valve with differential pressu'e across the valve. The valve was found able to perform it's intended function. Valve 73-27 was correctly wired and operated at the same speed as other identical valves. A review of all valves with timing criteria was conducted by the licensee. Six valves were found to time out inconsistently. The differences for four valves were due to mechanical differences such as different gear ratios. However, two valves, 2-74-30 and 2-74-57, required resetting of the limit switches for the motor operator These were reset per plant electrical instruction EMI-1 This item is close (Closed) Violation (259/81-28-09, 260/81-28-06, 296/81-28-06) This violation was against 10 CFR 50, Appendix B, Criterion XVI for failure to take correc-tive action for conditions adverse to quality concerning the emergency equipment cooling water (EECW) air release valves. The licensee submitted a licensee event report (LER), 259/84-13, concerning the valves. This LER was closed in report 86-06. The inspector reviewed the completed engineering change notice (ECN P0739) which replaced the air release valves with seismically qualified cast steel body valves rated for 185 psig design pressure. This item is close (0 pen) Violation (259,260,296/86-16-07) This violation involved failure to properly control procedures for radiological environmental monitorin In response to this finding, the licensee's Nuclear Quality Audit and Evaluation Branch performed an audit of Radiological Environmental Monitoring, Radiological Effluent Monitoring and Environmental Dose Assessment at Browns Ferry and Sequoyah during the month of September 198 Several deviations were identified and will be tracked under this violation for corrective action and recurrence control. Examples of the deviations identified were: radiation sources used for calibration of radiation monitors were not traceable to the National Bureau of Standards (NBS);

effluent monitoring samples were not representative of the material released; the accuracy of sample flow rates on gaseous effluent monitors was indeterminate; the Quality Assurance Program for radiological effluent monitoring program was not adequately establishe This item remains ope (0 pen) Inspector Follow-up Item (259,260,296/86-25-02) This item contained several deficiencies identified during a walkdown of the Control Room Emergency Ventilation System (CREVS). One of these deficiencies was a recurring problem with the backdraft damper on CREV A sticking open follow-ing shutdown of CREV A. At the time of the inspection, the function of the backdraft damper was unknown. A documentation review by the inspector has since found reference to the dampers in Amendment 40 to the original FSAR which contains the licensee's response to Atenic Energy Commission (AEC)

licensing question number Q10.2. This response Indicated that the damper prevents outflow (and possible loss of the abtlity to maintain the required pressure on the control room) in the event of a fan failure and to prevent contamination of the charcoal when the unit is inoperativ This item remains ope _ .. . _ - . _ .. . . _ _ . _ . _ _

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I (Closed) Unresolved Item (259/81-32-07) This item was to review the licen-see corrective action regarding concerns about the respiratory protection

program. An inspector had found that no quality assurance program existed I~ and that particulate filters were routinely reused witnout ett1ciency or resistance checks. Six health physics . respiratory protection section instruction letters are the implementing procedures for the respiratory protection program. Instruction Letters HP-RP SIL 2, Respiratory Protection Equipment Inventory, Control, Accountability; HP-RP SIL 3, Respiratory Protection Equipment Inspection and Repair; HP-RP SIL 5, Respirator Mask and Cannister Testing were reviewed by the inspector. NUREG-41, Chapter 10 provides guidance for a quality assurance program. The instruction letters

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. reviewed contained the guidance provided in NUREG-41. The inspector toured the respiratory protection repair facility. The licensee demonstrated the i method of checking the respirator masks for efficiency and resistance. Also

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discussed were the quality assurance measures taken for the masks and test j equipment. This item is close (0 pen) Open Item (259/85-06-09) This item concerned aluminum-electrolytic capacitor aging. In response to the concern the licensee revised Browns

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Ferry Standard Practice BF-6.8, Aluminum Electrolytic Capacitors, to include the recommendation from General Electric Service Instruction Letter Number 290. The procedure requires that the capacitors shall not be stored for a j period greater than three year The inspector toured the power stores '

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facility on December 17, 1986, and inspected several storage drawers

containing capacitors. An estimated 10 to 30 percent of the capacitors in

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the drawers were controlled as aluminum electrolytic capacitors. A computer printout lists all the capacitors and specifies which are aluminum electro-lytic. However, numerous uncontrolled capacitors in the drawers appeared identical to the capacitors being controlle A sampling of the uncontrolled capacitors indicated a number were being stored past the three year requirement. The inspector discussed with the power stores supervisor a concern that the computer printout may not adequately state which capacitors are aluminum electrolytic. This item j remains open.

(0 pen) Violation (259,260,296/84-34-03) This violation was for failure to test the core spray system 400 pound relief valves per ASME Code Subsection IWV-3510 requirement The relief capacity is used by the licensee as a

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basis for setting valve leakage limits between high and low pressure piping

to assure the low pressure piping is not overpressurized. In response to

the violation TVA revised Surveillance Instruction, SI 3.2, Inservice Valve Testing required by ASME Section XI, to include the relief valves. This i violation was part of a civil penalty (Enforcement Action 84-108) for the i core spray system overpressurization for Unit 1. The inspector questioned whether all the relief valves have been tested for all units. Apparently,

only Unit i relief valves were teste The ' licensee stated they would

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confirm the status of all the relief valve Because the violation was for not testing the relief valves, merely revising the procedure is not adequate for closure of this item. Although this item was presented to the l inspector for closure by the plant licensing staff, this item will not be i closed until all relief valves are tested.

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5 Unresolved Items * (92701)

There is an unresolved item discussed in paragraphs 7 and 1 . Operational Safety (71707, 71710)

The inspectors were kept informed of the overall plant status and any significant safety matters related to plant operation The inspectors made routine visits to the control rooms when an inspector was on sit 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 supervisor General plant tours were conducted on at least a weekly basis. Portions of the turbine building, each reactor building and outside areas were visite 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 tour Weekly verifications of system status which included major flow path valve alignment, instrument alignment, and switch position alignments were performed on the spent fuel pool and suppression chamber system 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 post In addition, the inspectors observed protected area lighting, protected and vital areas barrier integrit Inadvertent Initiation of Fire Protection Spray System During a thunderstorm on December 23, 1986, a voltage fluctuation occurred on both the 161 and 500 kilovolt off-site power line This resulted in a Unit 2 fire panel common alarm (XA-39-111A1) and

• 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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initiation of the fixed spray system on elevation 593 of the reactor building. On December 24, panels sprayed during the initiation were inspected. dried. and resealed. Also, a clogged floor drain in the reactor building was flushed to permit water drainage. On December 25, standing water was found in the floor drain of Unit 2 battery board room located in the control bay. A possible source of the water was the floor drain flushing activities in the reactor buildin Plant drawing 47W852-1 shows that the battery room and battery board room drains located in the control bay and the reactor building floor drains share a common drain header to the reactor building floor drain sum The common drain may be a breach of secondary containment not previ-ously considered. The battery room and battery board room drains were installed in 1978 after fire protection sprinkler systems were installed in the roo Engineering Change Notice L-1978 made the modifications. The inspector reviewed the change notice and the associated Unreviewed Safety Question Determination (USQD). The issue of secondary containment integrity was not discusse Plant drawings show a loop seal in the battery board room and battery room drains. The loop seal may maintain secondary containment if

, filled with water. However, visual inspection by the licensee found the Unit 3 battery room drain completely dry. No periodic inspections of the drains are conducte Prior to the question concerning the common drain header, the licensee performed a USQD concerning a number of secondary containment penetra-tions which were not seismically qualified. This was completed on December 12, 1986, and concluded that secondary containment was operable for fuel handling. The inspector questioned whether the drain connection had previously been considered in the USQD previously completed on December 12, 1986. The inspector has asked for a list of the penetrations in question but no list has been provided to date The licensee stated the drain connection was being evaluated and that it was not known if the drain connection was previously considered. This will be an inspector follow-up item for review of the licensee's evaluation (259,260,296/86-40-01). The licensee has placed expandable plugs in the drains in the control bay battery and battery board rooms until the issue is resolve Also, the licensee is conducting a critique of the fire protection system initiation and spray down of the Unit 2 reactor buildin Spurious initiation of the fire protection system is a recurring problem. A previous event was the subject of an Advisory Committee on Reactor Safety subcommittee meeting. The licensee reported that event in Licensee Event Report 259/86-14. A review of the critique will be an inspector follow-up item (259,260,296/-86-40-02).

6. Maintenance Observation (62703)

Plant maintenance activities of selected safety-related systems and compo-nents were observed / reviewed to ascertain that they were conducted in

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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 returning components or system to service; quality control records were maintained; activities were accom-plished by qualified personnel; parts and materials used were properly certi fied; proper tagout clearance procedures were adhered to; Technical Specification adherence; and radiological controls were implemented as require Follow-up review of General Electric Service Information Letter (SIL)

445, " Intermediate Range Monitor (IRM) Fuse Failure" During an outage at an operating GE/BWR, all positive and negative IRM 3/4 Amp fuses (F1 and F2) connected to the 24 Vdc bus B were blown because of a power surge resulting from a switching transient on the 480V power suppl After the positive 3/4 Amp fuses (F1) were replaced, all inoperative IRM channels appeared to be operating nor-mally. However, because of continued loss of the negative power supply, of which there was no indication on control room panels, the IRM channels were inoperable and unable to process flux signals. If this condition had remained undetected, the IRM-initiated alarms and scram may not have occurred if needed during restart of the plant. The blown negative-side fuses were detected later during subsequent IRM surveillance testing prior to restarting the plan The purposes of the Services Information Letter were to recommend the following: (1) reevaluation of procedures pertaining to replacement of blown fuses and restoration of inoperative safety related channels, (2) replacement of the 3/4 Amp IRM chassis fuses with 1.5 Amp fuses in certain IRM channel designs and (3) modification, if desired, of SRM/IRM system designs to add negative voltage sensing relays to each channe An evaluation of SIL 445 by the licensee indicated that the following actions would be required:

(1) Procedure revisions were required to require SRM, IRM, and radia-tion monitor functional checks upon loss of + 24 VDC buses (OI 90, 01 92, and annunciator response procedures for respective instru-ments affected). These procedures cannot be processed and updated until the configuration control drawing walkdown group has com-pleted its updat (2) Design change requests are required to change the chassis fuses from 3/4 amp. to 1-1/2 amp. and to install 24 VDC bus undervoltage relay sensors in series with the IN0P contact to ensure safe trip These SIL 445 follow-up items will be left as an open item until corrective actions are completed by the licensee (259.260,296/86-40-03).

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i 8 l Maintenance requests were reviewed to determine status of outstanding ,

jobs and to assure that priority was assigned to safety-related equip-

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p . ment maintenance which might affect plant safety.

! The inspectors observed the below listed maintenance activities during

this report period

(1) Unit Two Reactor Cavity Drain for Recirculation Pipe Safe-end

Replacement i

j (2) Diesel Generator Annual Maintenance

, (3) Repair of damaged electrical conduit on Main Steam Isolation Valves.

l Maintenance Request No. A-718016 -was originated on October 15, 1986, i requesting that damaged flexible conduits used to interface between

! junction boxes and three Main Steam Isolation Valves (MSIVs) be i repaired or replaced. The MR originator also requested that the

! internals of the junction boxes and position indicating limit switches be inspected for any signs of corrosion. The work instructions were

! written and approved on October 22, 1986. These instructions required

! that the flexible conduit be reconnected to the connector and that the I junction box internals be inspected for signs of corrosion. The " work performed" section of the MR documented that the flexible conduit on

one MSIV was repaired (2-FCV-1-38), nothing wrong could be found with

! the other two MSIVs (2-FCV-1-15 and 27), and that no junction boxes or

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limit switches were opened. The MR was signed off as completed on l October 24, 1986, even though several actions required by the work i instructions were not completed and no attempt was made to reconcile the MR originator's concern with regard to the other two MSIVs.

l The inspector toured the Unit 2 MSIV Room on November 10, 1986, and

[ noted the following deficiencies:

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L 1) The point at which the flexible conduit mates with rigid conduit

to the solenoid control valves for ' 2-FCV-1-27 had a wide gap through which the control wires were expose The other end of this conduit which attaches to a junction box (not labeled) is loose and not properly secured by the conduit connector.

! 2) The repair made to the solenoid control valve conduit at junction i box 2091 for 2-FCV-1-38 apparently consisted of laying a bead of

! RTV silicon glue around the conduit at the junction box. Since

the conduit was not properly attached via the conduit connector, l

it could again become detached with a gentle tug on the conduit.

} 3) The point at which the flexible and rigid conduit mate to the

solenoid control valve for 2-FCV-1-52 had a wide gap through j which the control wires were exposed.

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4) The rigid conduit running from the junction box - (not labeled)

to the 90*4 open limit switch for 2-FCV-3-52 had a dipd clamp

+ missing from its unistrut channel support and "was unsupported

, through its entire lengt .

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The inspector concluded that the maintenance ahtivities performed in

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response to MR A-718016 were not effective in correcting the i&ntified deficiencies and were apparently performed in a perfunctory.~ manne The work instructions were deficient in that no reference was made to

adherence with the requirements of General Construction Specification (G-Spec) G-38 and G-40 which defines the acceptable configuration of

, flexible conduit installation. The work performed was deficient in

that the work instructions were not carried out in regard to inspection

! of the internals of junction / boxes or limit switches and two of the .

j valves that were identified as having problems were inappropriately  !

dispositioned by a statement that nothing wrong could be found with them. There is apparently no mechanism in the maintenance program to o i

require that the originator of an MR be consulted prior to disposi-tioning an MR with no action just because the maintenance crew cannot i locate the deficiency or concer This ineffective correction of i deficient conditions is a violation of 10 CFR 50 Appendix B, Criterion XV1, Corrective Action (20]/86-40-04).

While in the Main Steam Tunnel,4 the inspector detected several other

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i discrepancies not related to MR A-718016 a; described below: ,

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' 1) Piping which supplies operating air to MSIV 2-FCV-1-27 was missing pipe clamps to its unistrut ' Channel Support near the emergency backup supply Accumulator. This includes both the large diameter emergency air pfping and the small diameter normal control air

supply piping.

i The normal control air supply piping to 2-FCV-1-52 was unsupported

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and very wobbly from the ' point at which it was attached to the i

header near isolation wive 2-32-1295 all the way to the control i solenoid valve block.

i 3) Main Steam tunnel tempera'ure element No. 17A had a broken sensing

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probe. All the tunnel temperature element junction boxes had large l patches of paint cracked away (apparently from overtenTerature conditions), missing cover fasteners and loose U-bolt i which'

support the temperature element ) Many turnbuckles which connect the pipe rupture restraint tie-rods for both the main steam and feedwater pipes had jam nuts which  !

were not snug to the turnbuckl ) The tunnel floor drain was missing its grating and as a result, it was fouled with tras a

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6) Housekeeping in the tunnel was unacceptable. Although no main-i tenance was actively in progress on the day of the inspection, a the following materials were laying about: yellow poly bags,

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rubber gloves, burlap bags, paper, tangled extension cords and temporary lighting, tools, discarded sections of rubber hose, discarded light bulbs, "information only" drawings, temporary ,

ventilation hoses and discarded hold order tag The r,ature of this material causes one to doubt the effectiveness of the

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mandatory post maintenance housekeeping inspections performed in the recent pas Graffiti was also noted on the walls in the tunnel. Housekeeping this area as well as correction of the identified deficiencies will be tracked as an Inspector Follow-up

Item (260/86-40-05) to ensure reinspection prior to Unit 2 Startup.

It was noted that the Main Steam Valve Vault is not specifically assigned to any group for housekeeping inspection and responsibil-

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ities in Standard Practice 14.2. It was not clear whether this

! room would be included in the area designated Elevation 565 of the i Reactor Buildin '

. All of the above material concerns were discussed with a licensee

, representative following the inspectio The following MRs were initiated by the licensee to resolve the deficiencies: 777201, 777202, 777203, 759605, 759604, 745124, 745125, 759602, and 75960 . Surveillance Testing Observation (61726)

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The inspectors observed and/or reviewed the below listed surveillance

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procedures. The inspection consisted of a review of the procedures for i technical adequacy, conformance to technical specifications, verification

of test instrument calibration, observation on the conduct of the test, l1 removal from service and return to service of the system, a review of test

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data, limiting condition for operation met, testing accomplished by qual-i ified personnel, and that the surveillance was completed at the required frequency.

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l ,- Incorrect Technical Specification (TS)

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The licensee identified during their surveillance procedure review program a possibic error in T.S. , Table 3.2.A. The standby gas treat-I ment (SBGT) relative humidity heaters trip on low flow conditions 1 to avoid heater burnout. The table states the trip level setting for the heaters as less than or equal to 2000 cubic feet per minute (CFM).

The licensee believes the current setting for the heaters is greater than or equal to 2000 CF The inequality sign was a typographical

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.I erro The . licensee plans to submit a T.S. change to correct this

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error and establish a band between 200 to 400 CFM for the setting.

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! However, the existing surveillance instruction SI-4.2.A-13, Calibration of Flow Switches for SGTS Train A, B, and C heaters, did not comply l

y- with the T.S. The SI allowed an acceptable setting of 2000 to 4000 CF The licensee is considering submitting a licensee event report

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(LER) to report this licensee identified violatio This item will remain an inspector follow-up item for correction of the (259,260,296/86-40-06). Primary Containment Isolation Valves

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Surveillance Requirement (SR) 4.7.D.1.a requires that isolation valves that are power operated and automatically initiated be tested for simulated automatic initiation. The licensee .does not have one comprehensive Surveillance Instruction (SI) to satisfy this require-ment. Rather, many sis exist (some overlapping) that perform other tests such as logic functionals as well as the automatic initiation  ;

s test. It is an extremely tedious task to verify the SR 4.7.D.1.a is '

satisfied by the various sis particularly since there is no single source document that matches each isolation valve with its associated S It is further complicated by the numerous tables in the technical specifications that identify the isolation valves as follows:

Table 3.7.A - Primary Containment Isolation Valves.

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Table 3.7.D - Air Tested Isolation Valve Table 3.7.E - Primary Containment Isolation Valves which Termi-nate Below the Suppression Pool Water Leve Table 3.7.F - Primary Containment Isolation Valves Located in Water Sealed Seismic Class 1 Line Confusion exists over whether the valves in Table 3.7.F must be tested for automatic . initiation since these valves are not contained in the i Table 3.7.A list. The FSAR does not clarify matters since it contains two additional. tables as follows:

Table 5.2-2 - Principal Penetration of Primary Containment and l Associated Isolation Valves.

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. Table 7.3-1 - Pipelines Penetrating Primary Containmen t These two tables do (At contain the same i,ist of valves nor does either one of these tables match with Table 3.7.A of technical specification During a meeting with site licensing on technical specifications, a licensee representative stated that changes were planned such that only l one table (Table 3.7.A) would list all primary containment isolation 1 - valves. This will be identified as an Inspector Follow-up Item (259, l

260,296/86-40-07) to cross check the accuracy of the table with the existing tables in the FSAR and technical specification The inspector identified a set of valves which was not adequately tested for automatic initiation. These are the Drywell and Torus spray isciation valves contained in Table 3.7.F (valves 74-57, 58, 60, 61, 71, 71, 74 and 75). There was no surveillance instruction which

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verified by test that the valves would go closed (if open) upon receipt of a valid isolation signa In addition, the inspector learned that the valves had not previously been stroke-timed in the closed direc-tio Due to an error in the previous surveillance instruction, the valves had always been timed in the open direction. Thus, two impor-tant functions of the valves (timing in the closed direction and verification of automatic isolation) were never verified by periodic surveillance testin While pursuing a resolution of this finding, the inspector learned that the same finding had been made by a contractor in the licensee's Surveillance Instru; tion Review for Unit Startup (SIRUS) Program. A draft revision to SI 4.2.B-45, LPCI System Logic-Functional Test is being prepared which will verify that an actuation signal causes the valves to go close This will be tracked as an Unresolved Item (259,260,296/86-40-08) pending completion of the revision and evalu-ation of whether credit can be given for licensee identification of this potential violatio Control of Compressed Gas Cylinders During a routine tour of the Unit 2 Reactor Building on December 9, 1986, the inspector noted an unusually large number of unattended compressed gas cylinders. The required controls on these cylinders are contained in the Browns Ferry Nuclear Plant Fire Protection Program Plan (BF-FPP), Attachment 0, Storage and Labeling Hazardous Chemicals, Flammable or Combustible Liquids, and Compressed Gas Cylinder Additional controls over welding and burning equipment are located in the BF-FPP, Attachment C, Guidelines for Control of Transient Fire Loads at Browns Ferry Nuclear Plant, and Attachment I, Torch Cutting, Welding, Open-Flame, Grinding and Spark Producing Work Requirements and Precautions. The following deficiencies were noted during the December 9, 1986 tour:

(1) Five cylinders were not properly secured by wire, chain or other means at 3/4 of the cylinder height from the floo (2) Four cylinders which were not connected for use were found without the required valve protection caps installed. (Two of the cylin-ders without valve caps were also not properly secured as described above).

(3) Acetylene-oxygen gas cylinders were found unattended with the regulator not depressurized as require (4) Acetylene-oxygen gas cylinders were found in storage (by the definition contained in the BF-FPP) in safety-related area These deficiencies were discussed during a daily management meeting with licensee representatives. A follow-up tour on the next day noted that the identified deficiencies had been correcte <

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13 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 N Date Event 259/86-17 Feb. 28, 1986 Failure to Properly Torque Enclosing Tube Assembly Nu *259/86-31 Nov. 5, 1986 Inadvertent Engineered Safety Feature Actuation Caused by Undervoltage Relay Tripping of the Reactor Protection Motor Generator Se /86-30 Sept. 22, 1986 Inadvertent Isolation of Reactor Water Cleanu *259/86-06 Rev. 1 Jan. 16, 1986 Tornado Missile Protection for C Vent Towers During a design evaluation of control bay ventilation modifications, TVA design engineers identified an unanalyzed condition involving tornado-missile protection for equipment located in the control bay vent tower These vent tower buildincs house many components utilized in the control bay ventilation system. A probabilistic risk assessment (PRA) concluded that the risk to the subject equipment was very low and no changes were require During the resident inspector's review of the analysis and PRA the source of the numbers used in the PRA could not be located. The analysis was for-warded to NRR for review. The numbers were from a seven volume PRA which was never docketed. When questioned about docketing the PRA, TVA stated the PRA was considered an unreviewed draft and no meaningful conclusions could be drawn from the PRA. Therefore, TVA has been requested in the cover letter of this report to submit a revision to the LER. The NRC will review the subject LER once the licensee determines their final position on this subject matter and submits the information on a formal basis. This item is open pending LER correction (259/86-40-09).

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14 Management Meetings (30702)

During the course of the last several years, numerous management meetings have been held between senior Region II, Headquarters, NRR management and senior licensee management. Monthly management meetings are being conducted at the plant site with the NRR licensing project manager, Region II TVA Projects staff and representatives from the NRC's Senior Management Team (SMT). The purpose of these meetings is to ensure proper coordination of NRC inspection and evaluation activities with the site restart schedule and to provide a forum for detailed information transfer related to the Browns Ferry Nuclear Performance Plan.

1 FSAR Review (30702)

One of the objectives of this inspection was to ascertain whether signifi-cant changes have occurred in the general environs of the facility. These changes may include: Population increase in excess of predicted value Major changes in transportation routes or in the movement of hazardous cargo near the facilit Change in the routing of oil or gas transmission lines near the sit Changes or additions of major industrial, institutional or military facilities near the sit Erection of dikes or dams across cooling water suppl Naturally occurring changes in geologic. hydrologic, or meteorologic features in site area The inspector has reviewed the available information for the above items and determined that the sections of the FSAR which describe the site environs and hazardous materials (Sections 2.2 and 10.12.5.3) have not been kept up to dat This is evidenced by Table 2.2.8, Li sting of Institutions (Industrial, Educational and Medical) Within the Ten-Mile Emergency Planning Zone of Browns Ferry Nuclear Plant. This Table fails to list the Westlawn Elementary School in Decatur, Alabama; it gives the peak capacity of the General Motors Plant as about one-third of the current number of employees and; it lists the peak number of employees at Browns Ferry as about 3,000 less than current number of employees on site. In addition, the Toxic Gas Hazards evaluation in Section 10.12.5.2 of the FSAR was based on 1979 data for Tennessee River barge traffic passing the Browns Ferry Nuclear Plant. A major change in barge traffic was made in 1985 when the Tennessee-Tombigbee Waterway opened. The licensee has not evaluated this change for the impact on previous assumptions used in the Control Room Habitability Analysis Following Postulated Hazardous Chemical Release. The previous study con-cluded that of the chemicals stored onsite, offsite within a 5-mile radius,

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or transported near the site by barge, rail, or road within a 5-mile radius, only chlorine traveling by barge could present a hazard to control room personnel. The study further concluded that the impact from this source is negligible since Army Corps of Engineers data (1979) does not indicate that chlorine is barged past the sit The inspector contacted the Army Corps of Engineers and obtained current data from personnel in the Lock Performance Monitoring System (PMS). This information indicates that up to 4 barge loads per month of chlorine is shipped past the Browns Ferry Site. Each load is about 1100 tons of liquid chlorine under pressure in specially fabricated chlorine barges. The major shipments of chlorine originate from the Olin Chemical plant near Charleston, Tennessee and travel downstream to various destinations. This will be tracked as an Unresolved Item (259,260,296/86-40-10) pending a reevaluation of the control room habitability following a postulated chlo-rine release to be performed by the license The licensee was unable to determine what (if any) organization within TVA was responsible for maintaining cognizance over the site environs to assure changes have been adequately evaluated and original licensing assumptions remain valid. This will be tracked as an Inspector Follow-up Item (259,260, 296/86-40-11) pending resolution by the license . Design Changes and Modifications (37700)

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The inspection of Design Changes and Modifications which was started in October 1986 (refer to report number 86-36) was completed during this reporting perio This inspection confirmed that modifications and design changes which do not require prior NRC approval were performed in conform-ance with 10 CFR 50.59 and the plant technical specifications. No viola-tions or deviations were identified; however, some follow-up activity will be necessary to properly resolve some of the concerns identified below:

Engineering Change Notice (ECN) P0369 - Installed a Motor Operated bypass valve (FCV-73-81) around the outboard containment isolation valve of the High Pressure Coolant Injection (HPCI) system for system warm-up. Workplan No. 9883 which implemented this ECN contained a memo to the file which documented a telephone conversation and indicated a written memo would follow. The phone conversation gave verbal approval for upgrading ASME Section III Class II material to Class I for use in the HPCI system. No written justification for the material upgrade was included in the completed and closed-out documentation package. A licensee representative was able to locate the material upgrade documentation which provided the basis for the upgrad ECN P0602 - Installed orifice plates on the Residual Heat Removal System (RHR) pump test return line to the torus. This orifice plate increased the backpressure on the piping downstream of valve FCV-74-73 (pump test return valve). This reduced the pressure drop across the valve and reduced the excessive vibration the valve had been experiencing. No documents could be

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located that would confirm that the piping downstream of valve FCV-74-73 had been analyzed for this increased pressur The piping is designed for 150 psi; however, the post modification tests indicated that the pressure increase may be up to 300 psi. The licensee is investigating this defi-ciency and this will be tracked as Inspector Follow-up Item (259,260,296/

86-40-12).

ECN P0384 - Decreased the stroke time of Primary Containment Purge Valves 64-17, 18, 19, 29, 30, 32, 33 and 76-2 This was accomplished by the replacement of solenoid control valves and increasing the air supply tubing size for the dampers. This modification brought the plant into compliance with NRC Branch Technical Position CSB 6-4, Containment Purging and Venting During Normal Plant Operations. The pcst-modification test performed after the work verified that the dampers cycled in less than 2.5 seconds or 5 seconds as required. Since the test was performed with no differential pressure across the dampers, the inspector questioned whether the configura-tion of the dampers was such that an increased pressure drop across the dampers would lengthen the damper closure time or whether the pressure drop would assist the dampers in closin It could not be determined from the documentation whether the pressure drop was addressed or not. A licensee representative is investigating this and it will be tracked as Inspector Follow-up Item (259,260,296/86-40-13).

1 Facility Modifications (37701)

On December 8, 1986, the licensee began cutting out the Recirculation System inlet safe ends in which cracks had previously been discovered. In order to support the cutting operations, reactor vessel water level was drained to about 160 inches above the bottom of the vessel. This is about 16 inches below the recirculation inlet nozzle. Temporary water level instrumentation was previously installed per Instrument Maintenance Special Instruction (IMSI)-3023, Reactor Vessel Level Instrumentation to Support Vessel Drain Down. A narrow range and a wide range level transmitter was connected to existing instrument tubing and wired to circuits normally used for jet pump differential pressure indication in the control room. Thus the operator had temporary level indication on control room panels and a temporary alarm box set at 150 inches and 170 inches (normal level control at 160 inches). In order to control plant chemistry, water level, and keep the control rod drive mechanisms clean, a feed and bleed operation was maintained at about 6 gpm. Makeup was supplied via the rod drives and letdown was via the shut-down cooling alignment of RHR. Two additional temporary sightglasses made of tygon tubing standpipes were also connected for use as backup and cross reference level indicator On December 10, 1986, the inspector noted that a Maintenance Request (MR)

sticker was attached to both the temporary narrow range and wide range level instruments. MR No. 755318 was issued on November 28, 1986, to correct a downscale reading that appeared during the water level drain down opera-tions. The failure was corrected by replacing a blown fuse; however, the operations section requested that the MR remain open. A licensee represent-ative stated that this would allow for rapid correction of any potential failure since the MR contained all the necessary administrative approvals.

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This mechanism (maintaining an open MR) was only to be used on critical equipment according to the licensee and is used to avoid any delays in approving critical maintenance activities. The inspector noted that this method was not fully 4 described in the administrative procedures controlling maintenance and was additionally subject to abuse. Furthermore, section 4.3.3 on Plant Managers Instruction (PMI)-6.2, Conduct of Maintenance, contains the necessary controls over emergency maintenance and allows the senior reactor operator (SRO) to authorize immediate approval of an MR or even to authorize immediate work without an MR in order to prevent imminent damage to major equipment or to protect personnel from an imminent threat of bodily har . Employee Concern Program In order to establish high standards of quality and safety in TVA nuclear activities, it was absolutely essential that TVA establish and maintain a high degree of trust . between TVA line managers and employee Licensee management took an important first step in that direction by establishing a new Employee Concern Program (ECP). All previous employee concern programs were consolidated into a single TVA-wide effort to refocus primary responsi-bility for problem communication and resolution back into the line organiza-tions. The new ECP was implemented on February 1,198 An Employee Concern Program Site Representative (ECP-SR) was designated for each nuclear plant site and corporate office locatio Briefings were held with supervisors and all employees within the TVA organization to orient everyone to this progra ,

The. resident inspector has been tracking and reviewing the ECP program since

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its inceptio Program concerns and categorizations were reviewed and commented on by the resident and the ECP-SR and staff at Browns Ferry are i typically performing thorough follow-ups and investigations on assigned reports. Program direction has been positive and has assisted in minimizing BFNP concerns with expeditious feedback and an independent assessment of I corrective actions required to address the concern The ECP program at BFNP has been a positive attribute to increasing site morale and proved

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beneficial in assuring NRC related concerns as well as TVA generated con-cerns are addressed. The inspector requested TVA to provide explanation of how NRC-forwarded allegations are tracked and closed out. This concern will be left as an open item for follow-up (259/86-40-14).

The resident reviewed the following ECPs during this report period for i background investigation, scope of findings, detailed evaluation, safety-j related categorization correctness, and recommendations:

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Concern Topic ECP-86-BF-074-001 Intimidation & Harassment ECP-86-BF-532-01 Management Discipline for Unapproved Absence ECP-86-BF-567-001 Personnel Action During Area CAM Alar ECP-86-566-001 Unresolved NRC Item 259/260/296/85-07-01, Adequacy of Actions Taken with Regard to Allegations Concerning Category I Support ECP-86-BF-419-001 Temporary Assignment of BFNP AU0s to Fossil Plants and Bellefonte AU0s to Browns Ferry (NRC Allegation RII-85-A0208)

Current program status under the new ECP indications 108 ECPs received with 78 investigations complete Overall, the resident -reviews indicate that ECP investigations and reports are more thorough and make more beneficial root cause corrective action recommendations than line management investigations of a similar natur The ECP reports should be beneficial to TVA management if used to address the recommended corrective actions of employee concerns in general and safety-related concerns in specifi l

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