ML20207U024

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
Download: ML20207U024 (19)

See also: IR 05000259/1986040

Text

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UNITED STATES

Do

- [m3 Ktk NUCLEAR REGULATORY COMMISSION

[ n REGloN 11

'

-g, ,j 101 MARIETTA STREET, N.W. ,

• ^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

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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

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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

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1. Persons Contacted

Licensee Employees

l *H. G. Pomrehn, Site Director

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

assurance, design and engineering personnel.

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NRC Resident Inspector

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• W. C. Bearden

, * Attended exit interview

2. 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.

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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.

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3. 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 modifications.

The two wall numbers were 24 and 92 in the control bay. The revisions close

the follow-up item. IE Bulletin 80-11 concerning masonry wall, however,

remains open.

(Closed) Follow-up Item -(259,260,296/86-11-02) This item was to review the

licensee's evaluation of a licensee vendor audit finding. 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 closed.

(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 tested.

This information is used to conduct post-trip reviews. The licensee

committed to verify a sampling of computer output data. Sixteen other alarm

points were verified to be correct. 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 closed.

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 three. The wiring errors were corrected. 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 instructions. This

item is closed.

(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 operators. These were reset per plant electrical instruction

EMI-18. This item is closed.

(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 closed.

(0 pen) Violation (259,260,296/86-16-07) This violation involved failure

to properly control procedures for radiological environmental monitoring.

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 1986.

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 established. This item remains open.

(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 inoperative. This item

remains open.

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(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 closed.

(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 years. 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 controlled.

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 requirements. 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 tested. The ' licensee stated they would

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confirm the status of all the relief valves. 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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4. Unresolved Items * (92701)

There is an unresolved item discussed in paragraphs 7 and 10.

5. Operational Safety (71707, 71710)

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.

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.

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 systems.

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.

a. 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 lines. 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 building. 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 sump.

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 room. 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 discussed.

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 conducted.

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 resolved.

Also, the licensee is conducting a critique of the fire protection

system initiation and spray down of the Unit 2 reactor building.

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

required.

a. 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 supply. 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 plant.

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

channel.

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 update.

(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

trips.

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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l b. 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

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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 exposed. 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,

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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 length. .

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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.~ manner.

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

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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 concern. 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.

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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

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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 elements.

4) 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 turnbuckle.

5) The tunnel floor drain was missing its grating and as a result, it

was fouled with trash.

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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 tags. 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 past. 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 Building.

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

, representative following the inspection. The following MRs were

initiated by the licensee to resolve the deficiencies: 777201, 777202,

777203, 759605, 759604, 745124, 745125, 759602, and 759601.

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7. 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 ,- a. 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 CFM. The inequality sign was a typographical

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.I error. 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

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y- with the T.S. The SI allowed an acceptable setting of 2000 to 4000

CFM. The licensee is considering submitting a licensee event report

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. _ . . - - _ . _ - .

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(LER) to report this licensee identified violation. This item will

remain an inspector follow-up item for correction of the T.S.

(259,260,296/86-40-06).

b. Primary Containment Isolation Valves

'

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

SI. 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 Valves.

Table 3.7.E - Primary Containment Isolation Valves which Termi-

nate Below the Suppression Pool Water Level.

Table 3.7.F - Primary Containment Isolation Valves Located in

Water Sealed Seismic Class 1 Lines.

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.

!

. Table 7.3-1 - Pipelines Penetrating Primary Containment.

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 specifications.

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 specifications.

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 signal. In addition, the inspector learned that

the valves had not previously been stroke-timed in the closed direc-

tion. 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 testing.

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 closed. 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 violation.

c. 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 Cylinders.

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 floor.

(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 required.

(4) Acetylene-oxygen gas cylinders were found in storage (by the

definition contained in the BF-FPP) in safety-related areas.

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 corrected.

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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

259/86-17 Feb. 28, 1986 Failure to Properly Torque

Enclosing Tube Assembly Nut.

• 259/86-31 Nov. 5, 1986 Inadvertent Engineered Safety

Feature Actuation Caused by

Undervoltage Relay Tripping of

the Reactor Protection Motor

Generator Set.

259/86-30 Sept. 22, 1986 Inadvertent Isolation of Reactor

Water Cleanup.

• 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 towers.

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 required.

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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9. 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.

10. 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:

a. Population increase in excess of predicted values.

b. Major changes in transportation routes or in the movement of hazardous

cargo near the facility.

c. Change in the routing of oil or gas transmission lines near the site.

d. Changes or additions of major industrial, institutional or military

facilities near the site.

e. Erection of dikes or dams across cooling water supply.

f. Naturally occurring changes in geologic. hydrologic, or meteorologic

features in site areas.

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 date. 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 site.

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 licensee.

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 licensee.

11. Design Changes and Modifications (37700)

'

The inspection of Design Changes and Modifications which was started in

October 1986 (refer to report number 86-36) was completed during this

reporting period. 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

upgrade.

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 pressure. 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 76-24. 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 closing. 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).

12. 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 indicators.

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 harm.

13. 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 employees. 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 1,1986.

An Employee Concern Program Site Representative (ECP-SR) was designated for

each nuclear plant site and corporate office location.

Briefings were held with supervisors and all employees within the TVA

organization to orient everyone to this program.

,

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

'

its inception. 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 concerns. The ECP program at

BFNP has been a positive attribute to increasing site morale and proved

'

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 Alarm.

ECP-86-566-001 Unresolved NRC Item 259/260/296/85-07-01,

Adequacy of Actions Taken with Regard to

Allegations Concerning Category I Supports.

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 completed.

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 nature.

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 specific.

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