ML20209G634

Insp Repts 50-259/87-02,50-260/87-02 & 50-296/87-02 on 870101-31.Deviations Noted:Failure to Have Battery Room Curbs Installed as Described in FSAR & No Neutron Dosimeters Provided in New Fuel Storage Vaults Per FSAR

ML20209G634
Person / Time
Site:	Browns Ferry
Issue date:	03/17/1987
From:	Bearden W, Brooks C, Ignatonis A, Patterson C, Paulk G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II), NRC OFFICE OF SPECIAL PROJECTS
To:
Shared Package
ML20209G571	List: ML20209G566 ML20209G598 ML20209G634
References
50-259-87-02, 50-259-87-2, 50-260-87-02, 50-260-87-2, 50-296-87-02, 50-296-87-2, NUDOCS 8705010029
Download: ML20209G634 (23)

See also: IR 05000259/1987002

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, p ter . UNITED STATES

ug*o - NUCLEAR REGULATORY COMMISSION

"

[' o REGION 11

3 ,j 101 MARIETTA STREET, N.W.

• ATLANTA, GEORGI A 30323

's

%+.../

Report Nos.: 50-259/87-02, 50-260/87-02, and 50-296/87-02

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Licensee: Tennessee Valley Authority

6N 38A Lookout Place

1101 Market Street

Chattanooga, TN 37402-2801

Docket Nos.: 50-259, 50-260 and License Nos. DPR-33, DPR-52, &

50-296 DPR-68

Facility Name: Browns Ferry Nuclear Plant

Inspection Conducted: January 1 - 31, 1987

Inspectors: kMd% k /ei

G.L.Paulk,SeniorRes{ den $ Inspector

22

' Date Signed

O h n , W ,.hdr r YN

C. A. Patterson, Residylt yspector Date/ Signed

we b Sb7

Date' Signed

C.R.Brgoks,Residentyn(pector

Gld9J 1 jn

W. C. Bearden, Inspector

3 /5/S?7

Vate /5igned

g

Approved by: dd- -r,h/i u

A. Ignhtonis, 5.etti6n Chief

7//7/87

DAte'5igne'd

Division of TVA" Projects

SUMMARY

Scope: This routine inspection was performed in the areas of operational

safety, maintenance observation, surveillance testing observation, reportable

occurrences, action on previous enforcement matters, Configuration Management,

Quality Assurance, refueling preparations, and Plant Operations Review Committee

activity.

1

Results: Two deviations were identified - failure to have battery room curbs I

installed as described in the FSAR; and no neutron dosimeters provided in the I

new fuel storage vaults as described in the FSAR. '

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870501o029 870417 i

PDR ADOCK 05000259

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REPORT DETAILS

1. Licensee Employees Contacted

H. G. Porrehn, Site Director

J. G. Walker, Deputy Site Director

G. T. Chapman, Project Engineer

-*R. L. Lewis, Plant Manager

• J. D. Martin, Assistant to the Plant Manager

• J. E. Swindell, Superintenient - Unit Three

• R. M. McKeon, Superintendent - Unit Two

T. D. Cosby, Superintendent - Unit One

T. F. Ziegler, Superintendent - Maintenance

D. C. Mims, Technical Services Supervisor

J. G. Turner, Manager - Site Quality Assurance

• M. J. May, Manager - Site Licensing

• J. D. Savage, Acting Compliance Supervisor

A. W. Sorrell, Health Physics Supervisor

R. E. Jackson, Chief Public Safety

Other licensee employees contacted included licensed reactor operators,

auxiliary operators, craftsmen, technicians, public. safety officers,

quality assurance, design and engineering personnel.

• Attended exit interview

2. ExitInterview(30703)

The inspection scope and findings were summarized on February 5,1987,

with those persons indicated by an asterisk in paragraph one above. 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.

A meeting was held on January 20, 1987, between the Design Verification /

Baseline Project Manager and his staff and the resident inspection staff.

Discussions focused on program progress and a description of problems in

the baseline program as evidenced by resident inspectors concerns is

identified in paragraph 8 of this report. The inspector concerns high-

lighted during this meeting are delineated in paragraph 8 of this report.

3. Licensee Action On Previous Enforcement Matters (92702)

(Closed) IE Bulletin (86-03) This bulletin was titled " Potential Failure

of Multiple ECCS Pumps Due To Single Failure of Air-0perated Valve In

Minimum Flow Recirculation Line. The inspector reviewed the bulletin and

the licensee's response dated November 14, 1986. The licensee stated that

the single failure vulnerability discussed in the bulletin does not exist

at Browns Ferry. The bulletin describes four recent cases concerning

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safety injection pumps in Westinghouse designed reactors and IE Bulletin

86-01 with a problem at Pilgrim. The safety injection pump problem is not

applicable to Browns Ferry and IE Bulletin 86-01 is closed in this report.

The ECCS Systems at Browns Ferry consist of two high pressure systems:

the high pressure coolant injection (HPCI) and the automatic depressuri-

zation (ADS) system, and two low pressure systems; the low pressure

coolant injection (LPCI) mode of the residual heat removal (RHR) systen

and the core spray (CS) system. The reactor core isolation cooling (RCIC)

system is similar to the HPCI system design but is not an ECCS.

HPCI has a single minimum flow recirculation valve and path back to the

torus. The valve is powered by 250 volt direct current and fails "as-is".

The HPCI system uses only a single train minimum flow design because its

backup system, ADS, allows use of LPCI if HPCI fails.

Both RHR and CS systems are similar. Each system has a total of four

pumps which are divided into two loops with two pumps each. Each loop has

a separate minimum flow valve and path the suppression pool. Each valve

is powered by 480 volt alternating current. Failure of one power supply

will only affect one loop of each system.

With the exception of HPCI, the single failure vulnerability is not

applicable to Browns Ferry. HPCI system is single train using ADS as a

backup. This bulletin is closed.

(Closed) IE Bulletin (86-01). This bulletin concerned a problem with the

minimum flow logic for which a single failure could disable all residual

heat removal (RHR) pumps. The inspector reviewed the bulletin and the

licensee's response to the bulletin dated June 2,1986. The licensee

stated that due to the system configuration and logic design a single

failure was precluded. The bulletin stated that at Pilgrim the minimum

flow bypass lines for each pair of RHR pumps are connected to a single

line and controlled by a single minimum flow bypass valve. The valve

would close upon sensing flow in either of the RHR loops.

This condition does not exist at Browns Ferry. There are two minimum flow

valves, one for each loop, which contains two RHR pumps. Both minimum

flow valves have their own discharge piping to the suppression pool. Each

valve has an independent power source. For example on Unit 1 the valve, '

FCV-74-7, for RHR pumps A and C is powered by 480 volt reactor motor

operated valve (RMOV) board ID. For the other loop, FCV-74-30 for pumps B  ;

and D is supplied from RMOV Board 1E. This bulletin is closed. l

(0 pen) Follow-up Item (259/260/296/86-40-01) This item addressed a concern

of possible breach of secondary containment, not previously consicered,

due to the existence of a common drain header between the control bay and

reactor building. The battery room and battery board rooms located in the

control bay have floor drains which connect to a common drain for the

reactor building floor drain leading to the reactor building floor drain

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sump. The inspector reviewed the Final Safety Analysis Report (FSAR)

section 10.11 on fire protection systems which discusses the drains.

Manual preaction sprinklers are located in the battery and battery board

rooms. Drains and curbs are supposed to be provided to remove a minimum

of 100% flow, assuming all sprinkler heads in one room are open and 200

gpm is being provided by hand hoses. In addition, the curbs and drains

were to be installed to prevent hazards to personnel and safety equipment

due to battery acid spillage.

The inspector toured the battery and battery board rooms on January 13,

1987. _ The battery rooms and Unit 2 battery board room contained floor

drains. No drains were found in Unit 1 and Unit 3 battery board rooms.

The other rooms contained an expandable plug installed in the drains

because of the secondary containment integrity question. No curbs were

installed for any of the battery or battery board rooms.

The inspector reviewed the associated Engineering Change Notice (ECN)

L-1978 which installed the sprinkler systems and drains. This ECN was for

all three units and was closed in 1978. The associated work plans were in

lifetime storage. The Unreviewed Safety Question Determination (USQD) for

the ECN discussed the concern about acid spillage but did not address .the

secondary containment integrity question. Thus, the USQD was inadequate

and the closure of the ECN as well.

The missing curbs and missing drains are a deviation from FSAR section

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10.11 (259/260/296/87-02-01). The inspector also questioned whether

an evaluation was completed for installing the plugs in the drains.

Follow-up on the safety evaluation will be done as part of the deviation

follow-up.

(0 pen) Follow-up Item (259/260/296/86-40-02) The inspector attempted to

review an Operations Critique of the inadvertent initiation of the fire

protection spray system in the Unit 2 reactor building on December 23,.

1986. However, all of the associated Maintenance Requests could not be

-located. The cause of the spurious alarm was believed due to improper

setting of the automatic transfer timer and a defective solenoid valve for

fire zone 2A. The following Maintenance Requests could not be located:

MR-A-779060 Check the timer.

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MR-A-779093 Test automatic supply voltage transfer from alter-

nating current to direct current.

MR-A-779066 Test water flow switch electrically and mechanically.

Plant personnel in Planning and Scheduling who track Maintenance Requests

(MRs) stated the three MRs did not exist. The review by the inspector of

the event could not proceed due to missing MRs.

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A meeting was held on January 23, 1987, with the fire protection super-

visor and compliance personnel to discuss the event. The status of the

three missing MRs was given. A-779093 was really A-779063 due to a typing

error. This MR was written on December 29, 1986, but only started on

January 20, 1987. A-779060 was performed on December 24, 1986, but was

not received in planning and scheduling until January 22, 1987. A-779066

to test the water flow switch electrically and mechanically was deleted

because it was believed another surveillance instruction (SI) performed

included the scope of this MR. However, discussion and review of SI found

this not to be the case.

As a result of the meeting this item will remain open pending the licensee

response on the following:

a. The results of MRs to be developed to examine the physical actuation

of the solenoid valves.

b. The revision to the Critique of the Unit 2 fire protection fix spray

initiation.

c. Revisions to sis that will address the periodic checking of the

timers.

d. Clarification of the MR tracking associated with this event and the

follow-up actions.

(0 pen) Unresolved Item (259/260/296/86-25-04) This item was opened to

track the progress and evaluate the significance of the unanalyzed

condition related to control room panel anchoring details. The inspector

learned that a licensee representative has delayed investigation of this

condition and has recommended deferral until next outage. This was to

allow the industries Seismic Qualification Utility Group (SQUG) to issue

guidelines on anchoring electrical equipment. The plant licensing group

was informed that a more detailed safety evaluation and/or justification

for continued operation would be required in order to accept deferral of

this potentially unqualified situation.

4. Unresolved Items * (92701)

There are five unresolved items identified in paragraphs five, seven and

eleven. i

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"An Unresolved Item is a matter about which more information is required to l

determine whether it is acceptable or may involve a violation or deviation.

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5. Operational Safety (71701,' 71710)

The inspectors were kept informed of the overall plant stiius and any

significant safety matters related to plant operations. Daily discussions

were held with plant management and various members of the plant operating

staff.

The 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; instru-

ment 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 fuel handling system and secondary containment 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 activi-

ties 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.

Wrong Gear Ratio in HPCI Isolation Valve

On January 13, 1987, the licensee made a four hour report to the NRC that

the Unit 2 High Pressure Coolant Injection (HPCI) System steam isolation

valve 2-FCV-73-2 may not be capable of closing against design differential

pressure of 1250 psid. This is the containment isolation valve located in

the drywell for the steam supply line to the HPCI turbine. The Limitorque

operator had a worm gear ratio of 33:1 instead of the acceptable 60:1

ratio installed on the HPCI valves in Units 1 and 3. This was initially

reported to have been found during the environmental qualification (EQ)

system walkdowns.

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The inspector reviewed the licensee's engineering report (SCRBFNMEB8615R0)

concerning the problem. The error occurred during initial installation.

Limitorque installed the incorrect worm gear in the operator due to either

erroneous information from Crane Company or an error during assembly. The

report stated that this was a generic problem that is potentially applic-

able to all plants that have Crane Company valves with Limitorque actua-

tors or all Limitorque operators that are located in a safety system.

Discussion with design engineers on site and at Knoxville revealed the

problem was discovered as part of IE Bulletin 85-03, Motor-0perated Valve

Common Mode Failures During Plant Transients Due To Improper Switch

Settings, and resolution of previous Advisory Committee on Reactor

Safeguards member concerns. A previous failure of the reactor coolant

isolation cooling inboard containment isolation valve 71-2 occurred on

March 21,1984. A detailed discussion of this failure is in IE Report

84-52, Inspector Follow-Up Item 84-52-05. Therefore, the initial report

that the error was found during EQ system walkdowns was incorrect.

In Inspection Report 85-36 a violation (85-36-02) was issued for not

having the HPCI torus suction valve (3-FCV-73-27) wired in accordance with

plant drawings. The shunt and series windings for the direct current

motor operator were connected wrong which caused the valve to operate

faster than the same valve on other units. The licensee checked all

valves for differences in timing criteria. Six valves were found with

differences. Two were attributed to limit switch setting and were

corrected. Four had different operator gear ratios. This was documented

in a Memorandum to the Engineering Section Supervisor in May 30, 1985.

However, no action was taken to explain the differences. The four valves

in question were 69-1, 69-2, 69-12, and 73-2.

Since the 73-2 valve was one of the four valves having different gear

ratios, the inspector questioned the acceptability of the other three

valves. This will remain an Unresolved Item pending TVA's evaluation of

the valves (259/260/296/87-02-02).

6. Maintenance Observation (62703)

Plant maintenance activities of selected safety-related systems and

components were observed / reviewed to ascertain on whether they were

conducted in accordance with requirements. The following items were

considered during this review: the limiting conditions for operations

were met; activities were accomplished using approved procedures;

functional testing and/or calibrations were performed prior to returning

components or system to service; quality control records were maintained;

activities were accomplished by qualified personnel; parts and materials

used were properly certified; proper tagout clearance procedures were l

adhered to; Technical Specification adherence; and radiological controls  ;

were implemented as required. l

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Maintenance requests were reviewed to determine status of outstanding jobs

and to ' assure that priority was assigned to safety-related equipment

maintenance which might_ affect plant safety. The inspectors observed the

below listed maintenance activities during this report period:

a. Special Test 8619; Temperature and Dewpoint Traverse of the Drywell -

Unit 3 (70307)

The inspector monitored Special Test 8619 to determine if applicable

requirements were met. This special test was conducted to complete

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temperature and dewpoint area surveys within the Unit 3 drywell and

suppression chamber. Data from this test will be used for generic

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applicability to all three units. The temperature and dewpoint

surveys were taken of the drywell and suppression chamber to deter-

mine the placement of Containment Integrated Leak Rate Test (CILRT)

instrumentation for all future CILRTs. ANSI N.45.4-1972, a standard

reference of 10 CFR 50, Appendix J, requires that a temperature

survey be made of the primary reactor building containment to deter-

mine any areas of localized temperature variations so that CILRT

instrumentation can be appropriately installed. The temperature

pattern revealed by the survey shall be employed in determination of

the mean representative temperature for the absolute method of

leakage-rate testing. Observation of the test conducted indicated no

significant deficiencies.

b. Unit 3 Preparations on the Refuel Floor for Fuel Off Load.

c. Fire Protection System Initiation troubleshooting.

No violations or deviations were observed in this area.

7. Surveillance Testing Observation (61726)

The inspectors observed and/or reviewed the below listed surveillance

procedures. The inspection consisted of a review of the procedures for

technical adequacy, conformance to technical specifications, verification

of test instrument calibration, observation on the conduct of the test,

removal from service and return to service of the system, a review of test

data, fulfillment of limiting condition for operation, testing accomplished

by qualified personnel, and that the surveillance was completed at the

required frequency.

a. Continuous Air Monitor Functional Check

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On January 16, 1987, the inspector observed the performance of

Surveillance Instruction (SI) 4.8.B.4-3A, Reactor Building

Ventilation Monitoring System Functional Test. The SI procedure

had been through the SIRUS review program. This program checks the

procedure for technical accuracy. This SI was performed on Continuous

Air Monitor (CAM) RM-90-250 for Unit 1 as required by Technical

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Specification 4.8.B.4. The functional test is performed monthly.

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The CAM monitors for high particulate, gaseous, or iodine activity

from the reactor building, refuel floor, and turbine building. If

the CAM alarms, the source of the activity can be determined by

selection of a single suction path by valve manipulation.

During performance' of the SI the local flow disturbance alarm would

not function. The test requires that with the vacuum pump running

the inlet line be closed. This should cause, within 20-40 seconds

after the flow is disturbed, the local red lamp to flash and an

audible alarm to actuate. When all three suction line valves were

shut and the inlet hose crimped, the Magnehelic flow gage indicated

normal flow and no alarms actuated. The Magnehelic flow- gage

measures flow by sensing the differential pressure across an orifice.

The SI was stopped and Maintenance Request A-764139 was written to

correct the problem. A loose orifice flange and a large deadband on

the Magnehelic gage were found. These items were corrected and the

SI completed on January 16.

The inspector learned of these equipment deficiencies on January 20,

1987. Also, the licensee found a previous Maintenance Request

(A-770053) had been completed on November 18, 1986, to measure the

orifice size. This would have required loosening.of the flange for

the measurement. The inspector questioned whether the potential for

an unmonitored release had been evaluated or reported since the CAM

only appeared to be drawing in air from the loose flange or other

leakage paths. This will be an Unresolved Item until the licensee

completes its evaluation for the reportability of this item (259/260/

296/87-02-03). Also, the inspector learned that the new SIRUS

calibration SI 4.8.B.4.3 was pulled because of technical problems.

The old SI was restored to allow restoration of the CAM in a timely.

manner.

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b. RHRSW Relay Calibration

The inspector reviewed the documentation associated with the last

performance of Surveillance Instruction (SI) 4.2.B-14, RHRSW Time

Delay Relay Calibration. This SI was recently updated, revised

and reissued as Revision 0 by the procedures upgrade group on

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September 8, 1986. Three performances were required to satisfy all

of the requirements for Unit 1 (10/21/86, 10/28/86 and 10/29/86).

Three performances were required to satisfy all of the requirements

for Unit 2 (10/29/86,11/10/86 and 11/10/86). Three performances

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were - required to satisfy the requirement for Unit 3 (11/11/86,

11/12/86 and 12/18/86). Many problems were encountered with most

of the attempted performances. After each aborted . attempt, the

equipment was restored to its normal lineup and an Immediate

Temporary Change (ITC) was issued. The procedure was then re-started ;

after performing the applicable precautions, limitations and

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, prerequisites .at. the step Yhere the procedure.was previously halted.

, The ITC administrative procedure'was sometimes abused in the process.

The number of ITCs issued prior to a successful completion of the SI

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were 3,- 10 and 10 for Units 1, 2 and 3 respectively.

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The ITC process is controlled through Site Director Standard Practice

.(SDSP) 2.11, Review, Approval and Change. of Site Generated Proce-

dures/ Instructions. An-ITC (also referred to as a non-intent' change).

[ is normally of an urgent nature such that time or plant conditions

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necessitate. implementation. prior to the time required by a' Temp ~orary

Change .or -a permanent revision. This method is ; meant to take

advantage;of the flexibility of Technical Specification 6.3.B which

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allows temporary changes to be made to- procedures without prior

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approval by the Plant Operations Review Committee (PORC) provided

that the change -does not change the intent of_ the. procedure. In

i order .to proparly evaluate whether a change is a valid ITC, an ITC

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guideline checklist-is provided and must be completed for'each ITC.

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-The first- hurdle 'on the checklist requires that the change must be

necessary to make the procedure work in situations which threaten to

1 curtail operations and not when time is available to process a

procedure change. This time is generally taken to be about two weeks

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although this is not specified in the procedure. Even.less certain

j is the meaning of the phrase " threaten to curtail operations". No

one quite knows wheth r this means-strictly plant operations (a unit

must be shutdown) or the operations of a maintenance group or the

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operation of the procedure in progress. In fact this is so. vague

that it is not a meaningful criteria and was routinely ignored.

During the' last performance of SI 4.2.B-14 no plant operations were

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threatened to be curtailed since all thee Units were shutdown and had

i been for some time. The licensee representative responsible for this

procedure indicated that a revision to the administrative procedure

was in -progress and that an attempt would be made to include some

additional clarification on this matter.

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Another area of abuse on the ITC checklist is.the question which asks  !

j if the proposed change is a change in -acceptance criteria. If the  !

i answer is yes, the ITC is not allowed to be processed. Acceptance J

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criteria is clearly marked on the SI by the letters AC in the margin

i of the step. ITC No. 10 on Unit 3 changed the contact status on some

relays from " Verify continuity does not exist" to " Verify continuity l

, exists". . This change- was processed improperly as an ITC .and was

subsequently reviewed and approved by PORC.

More problems were detected in the SI. 4.2.B-14 record in another

administrative control. In order to assure the most recent revision

, ar' all outstanding ITCs of a procedure are used, an SI Review Cover-

, Sheet is attached to all sis. The SI number, revision level and all

. ITCs are to be checked prior to the performance of the procedure.

The following discrepancies were found on'these cover sheets

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(1) The sheet attached to the Unit 1 procedure started on

October 28, 1986, said SI 4.2.B-14 Revision 2 with no temporary

changes was the latest. The. correct procedure should have been

Revision 0 with ITCs through No. 3. As of the time of this

report, Revision 0 is still the latest.

(2) The sheet attached to the Unit 2 procedure started on

October 29, 1986, said no ITCs were in effect but actually ITCs -

through No. 6 were in effect. This ' error was due to the time

interval between procedure issuance by Planning and Scheduling

(September 15, 1986) and actual performance of the SI on

October 29, 1986.

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(3) The sheet attached to the Unit 2 procedure started on

November 10, 1986, indicated that there were temporary changes

in effect for the procedure but the question which asks, "Have

the temporary changes been included in the package?" was not

answered.

(4) There was ..no sheet at all attached to the Unit 1 procedure

started on October 29, 1986.

One other problem was noted with SI 4.2.B-14 which was not related to

ITC control. Precautions and Limitations Section, Step 3.11 states

that the user is to N/A (not applicable) all steps which are ~not

applicable at the time of performance of the instruction. The

inspector expressed concern about this type of blanket approval to

delete or not perform any step or series of steps, in a procedure.

This authorization appears to negate the need for getting prior

review by a senior reactor operator and after the fact review by PORC

for changes to PORC reviewed and approved procedures. All of the

above concerns were addressed to various members of plant management

throughout the course of the review. The inspector was informed that

upper management attention was already focused in this area, that

PORC was rejecting ITCs more frequently, that several Discrepancy

Reports (DRs) had been issued on this subject, that a site re-

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training program was being formulated and that a revision to the

procedure was in progress. Concerns over procedure controls were

noted in Inspection Report 259/260/296/86-06 and a violation was

issued for abuse of non-intent changes. SDSP 2.11 was drafted and

issued in February,1986, in response to this violation. Since the

previous violation has already been closed out, a new Unresolved Item

(259/260/296/87-02-04) will be opened for the inspector to track the

effectiveness of the licensee's corrective action in this area. The

item is left unresolved as opposed to a violation since many of the

problems have been identified by the licensee and corrective action

has been initiated.

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c. Ambiguous Surveillance Intervals

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The . inspector became concerned about the effects of the extended

4- outages ~.at Browns Ferry on the surveillance . intervals. for various

4 technical specification (TS) equipment. Many surveillance require-

i. ments must be performed at an. interval of once per operating cycle

(0/0C) where the definition of an operating cycle is .the interval

between the end of one refueling outage and the .next subsequent ~

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. refueling outage. for the same unit. _As an example of what this

definition can result in -follows. The TS Table 4.2. A Calibration

! of Reactor Building Isolation . Timers (0/0C) was last performed over

- four. years ago (November 22, 1982) on Unit 2 and the TS 4.7.B.3.a

'. Automatic Initiation of'SGTS (0/0C). was last performed almost four

years ago (March 16,1983) on Unit 2.- These surveillances are .

normally performed prior to plant startup, however, the . equipment

functions are . required to be operable during almost all plant-

l conditions regardless of criticality conditions. . For Unit 3

, (currently being defueled) the surveillances were - previously-

performed on February 24, 1984 and October 17, 1984_(over two years-

. ago). 'As a result of the inspector's concerns, the licensee did

complete -these two surveillances prior to fuel . movement on Unit 3;

however, the. official position by the licensee remained that_ verbatim

. licensing requirements had been met. It was of interest to note that

1

one of the Reactor Building Isolation Timers was found to be. damaged >

1 and was replaced during the surveillance.. The inspectors noted that:

it would be impossible to apply the TS definition of Surveillance

Interval which restricts the maximum allowable extension to 25% of

the surveillance interval pro _vided that the combined for any three-

consecutive performances does not exceed 3.25 times the specified

interval unless the interval is equated to a' calendar time. Inter-

pretation guidance can be drawn from Standard Technical Specifica-

tions which have, in all cases, a calendar time associated with

surveillance intervals. As an example, the R frequency is defined as

j at least once per 18 months (550 days).in Standard TS. Additionally,

j the TVA TS contains a clarification on_- the Refueling Outage

i_ definition. This definition states ~ that for the purpose. of

j designating frequency of surveillance, a refueling outage shall mean

] a regularly scheduled outage. This would strictly prohibit these

extended outages- from affecting surveillance frequencies. The

f licensee is currently evaluating all such ambiguous _ surveillance.

requirements for the current unique situation but it is not clear

whether TS changes or clarifications will be requested in order.to

permanently fix tre problem. This will be tracked as an Unresolved

Item (259/260/296/87-02-05) pending review of the licensee's

! evaluation.

L 8. Configuration Management Program

!

An inspector reviewed ongoing work as part of.the licensee's-program to

'

improve the configuration management system at Browns Ferry. As discussed

in Volume 3 of the Nuclear Performance Plan,Section III, Paragraph 2.0,

i

.

-_.-_x ._ _ . , . - . _ , , . . . . _ - _.,...-.a.._- - . _ , , _ _ ~ -

. - ., - . ,.-

. _ . _ _ .. - . _ _____ _ _ _ _ _ ___ . _ _ _ _ . -._ _ _ _ _ _ _ _ _

.

4

,

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12

,

Configuration Management Program, the licensee has initiated actions to

ensure that the actual plant configuration is reflected on plant documents:

l and conforms to the design requirements (Design Baseline Program). This

'

program was established to assess the adequacy of post modification work

and to correct deficiencies.

The program is divided into five major areas: 1) Development of design

basis documents for 33 systems or portions of systems covered by the -

pre-restart phase of the program as listed in Table III-3 of the Nuclear

.

Performance Plan. These systems were selected from descriptions of

'

systems required to mitigate FSAR Chapter'14 design basis accidents and

provide for safe shutdown;- 2) System walkdowns by Field Engineering

Services personnel (FES) of as-constructed plant- flow diagrams, control

, diagrams, single line electrical drawings, valve tabulations and instru-

, ment tabulations, deviations reconciled and drawings updated to match the

actual plant functional configuration; 3) Issuance of Configuration

Control Drawings (CCDs); 4) Evaluation of ECNs and other potential systems

, changes (NCRs, SCRs, TACFs, FCRs, and Local Design Change Requests)

initiated since operating license issuance against the design basis; and

5) Establishment of design baseline for these systems prior to restart of

the associated unit. Evaluations will be performed to determine whether

! these systems, as installed, fulfill their functional design requirements.

'

The design baseline will be established for balance of plant systems but

j this effort is not required to be completed prior to restart by TVA.

! In particular the inspector reviewed the current status of system walk-

downs and determined that only one system, Diesel Air Start System

(System 86), had a completed final verification walkdown in accordance

j with SDSP 9.6, Rev. 2, Mechanical and Instrument and Control System

Walkdown. The licensee estimates that the effort associated with the

33 systems component nameplate data walkdown is greater than 90% complete

and the actual effort associated with the piping configuration walkdowns

is approximately 40% complete.

!

l Selected qualification records were reviewed <to verify experience

, requirements of SDSP 9.6, Section 4.16. This requires two years of design

i

or related experience.

As part of the inspection, interim system sketches, ~ drawing change

1 requests and the drawing discrepancy package for System 86 were reviewed.

The inspector obtained an up-to-date copy of TVA drawing 4-47E861-2

Rev. O, Flow Diagram Diesel Starting Air System, Diesel Generator B,

! Units 1 and 2. According to DCC personnel this drawing is currently

4

considered a CCD but not a validated CCD in accordance with SDSP 9.2,

i

~

Rev.1, Configuration Control Drawings. A later revision of the drawing

will be stamped as a validated CCD after establishment of design baseline.

The inspector compared the CCD with the interim system sketch with no

problems noted. Additionally the inspector used the CCD to perform an

independent walkdown of selected portions of System 86. Various concerns

!

were identified as stated below:

.

l

!

!

- . , - . . - - , . . _ , - , . . , . . . - - . , - , _ , , ,.___-__.,.-,._,_c . , . , - _ , ,~.<_-r._.-.m. 4m e.,, , , , . , ,

, . - _ - __ __ __ . .. _ _ _ _ __ .

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- -

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! 1) The . inspector.was informed .by-.FES personnel that the System 86

!~

boundary ended just downstream:of the air'Y-strainers (located.before

-the air start valves on 'the diesel engine- skid) and that those

i- portions of;the airstart system were scoped along with System 82,.

' Diesel Generator. This.is not consistent with.the system drawing.as 4

4

shown in_the FSAR, Figure 8.5-2, which shows all downstream piping _

i up to and including the air start motors to be part of the air start

system. None of the components located in this portion were included

i in the scope of _the System 86 walkdown. This portion of the drawing

i should 'also be ~ redrawn.with better defined boundariesTand to more

. clearly show actual. . configuration of the air. start valves,.

F Y-strainers, and air start motors. Additionally, the pressure. .

I

indication gauges, associated instrument lines,' and taps were omitted '

from.the drawing. . Personnel attempting to use the CCD.are unable to

i readily determine that this portion of the system is not scoped as

e part of System 86 and is not verified as being the correct configura-

tion. The-inspector noted that FSAR figure 8,5-2 was not correct and

i

required major revision' work.

a

2) The .two air compressor discharge check valves, 86-521B and 86-501B,

are marked on the CCD as not verified. The inspector determined from

discussions with~ FES personnel that these check valves are-difficult

i to recognize due to their. physical construction; however, they have

i been actually located in the field. The. inspector.was able to find

what appeared to_ be- the check valves but no effort has been-made to

-

locally label these components.

3) The inspector noted that the actual component labels installed on

{ the two automatic pressure operated control switches, 4-PS-86-36B'

and 4-PS-86-378, did not agree with. the designation shown on the

-

'

The pressure switches were . designated as 86-501B and 86-5028.

'

CCD.

The inspector learned from discussions with FES personnel-that other

, _ discrepancies of this type have been identified and the licensee

,

produced a copy of Maintenance Request 718105 which called for-the

,

correct labeling of the pressure switches and check valves.

!

j The above concerns are identified as . _ Inspector _ Follow-up Item

(259/260/296/87-02-06) System Walkdowns.

On January 20, 1987, the resident inspectors met with the licensee's

Design Verification / Baseline Project Manager and his staff to further

t

discuss inspector observations and concerns of the licensee's Design

.

Baseline and Verification program.

!

1-

The inspectors expressed the following project concerns during the meeting:

"

a. Volume III of the Performance Plan may need to be updated to reflect

the licensee usage of " Preliminary Configuration Control Drawings"

(CCD) vice permanent CCDs.'

'

l

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l

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1

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!

.A . ,; - , . - . . . - - _ , . , . - , - . ..- ......,a_, ,,_ m..._.. . -. , . .- . , - - _ , - , . . . ~- ,. ,

-

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14

,

s

b. A concern was expressed that the new valve / component numbering system

appears not to be human factors or operationally oriented.

c. In reference to the Safe . Shutdown Analysis, Systems Requirements

Calculations, Test Scopes, Design Criteria, and System Safety

Evaluations several concerns were expressed:

(1) Since the Test Scopes .are being produced in advance of the

corresponding Design Criteria Documents, the potential exists

for the Test Scope document to be inconclusive with regard to

the incorporation of certain features of the Design Criteria

Documents.

(2) A number of boundaries have been established in scoping the

,

component projects and there appears to be a complex interplay

among the boundaries. System interface drawings are complex in

that more than one drawing is used to describe the system due

to boundary interface selection criteria. This is not " user

friendly" to the operator or to other licensee personnel for

walkdown of the system. TVA is aware of this concern and is

addresing the problem.

(3) The inspectors further emphasized that there needs to be appro-

priate incorporation into licensing documents, notably the FSAR

and Technical Specifications, information - derived from the

Baseline efforts.

(4) The inspector questioned whether there were problems occurring

between the QC oversight and the project personnel. The Project

Manager said that the problems that had arisen had been' resolved

withinthemanagementstructureoftheproject.

(5) The Project Manager discussed the creation of a Punch List by

which open items such as valve tagging MRs, partial ECNs, etc.,

may be tracked to establish operability of systems for startup.

The meeting proved beneficial in establishing an effective communication

channel between the NRC and the baseline project group.

9. Reportable Occurrences (90712, 92700)

The below listed licensee events reports (LERs) were reviewed to determine

if the information provided met NRC requirements. The determination

includad: adequacy of event description, verification of compliance with

technical specifications and regulatory requirements, corrective action

taken, existence of potential generic problems, reporting requirements .

satisfied, and the relative safety significance of each-event. Additional I

in plant reviews and discussion with plant personnel, as appropriate, were j

conducted for those reports indicated by an asterisk. The following

licensee event reports are closed:

I

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o - - . . . . - . . .

.T g

"

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.. K

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,

d'

o LER No. Date Event

! *259/86-36 12-08-86 Inoperable ~ diesel generators cause

, Technical Specification non-compliance.

• 259/86-34 12-09-86 CREV system initiation from high (Remains

Open) radiation. trip.

259/86-29 9-11-86 Engineered safeguards actuatica due. to

(Remains Open) personnel error.

~

• 259/86-23 8-15-86 . Switchyard Problems Lead to - Reactor

(Remains Open)- - Scrams.

) 296/86-03 . 1-23-86 Personnel error in voltage measurement

] (Remains Open) results in inadvertent- containment

isolation.

j A review of the above ' listed LERs' root causes -indicates a' recurrent -

failure of personnel to strictly follow approved procedures during.1986.

I All - the above LERs- involved failure of personnel to strictly follow-

'

procedures. Some. resident reviews of surveillances and work in progress

3 also ' indicate this trend particularly for~ certain -plant! sections.

. Continued licensee management attention must be focused-in this area.

'

No violation or deviations were noted.

4 10. Quality Assurance (90713)

The resident inspection staff conducted a meeting with the site QA super-

visor to ascertain improvements in the QA organization and. evaluate

actions being taken by the licensee to address the overall QA program-

[ improvement. The inspector reviewed the Monthly QA Trend Report for s

November 1986, prepared in compliance with commitments -in Volumes I and

-

1

'

III of the Performance Improvement Plan. The trend report provides

performance indicator tracking for use by site managers in assessing each

,

manager's quality performance areas of interest and responsibility. This

report should provide a useful tool in evaluating .the emphasis by-the

! licensee in returning quality accountability to the line manager function.

i Trending data included such categories as: '

.

j a. Condition Adverse to Quality (CAQ) and Corrective Action Report

Summary (CARS).

I

b. Discrepancy Reports (DRs)

c. Average age of DRs and CARS indicated corrective action closure

1:

rates.

i i

d. Licensee Event Reports to NRC. l

,

l l

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e. Document /Workplan/ Instructions Review rate.

f. Maintenance Request reviews in process.

g. QA inspections completed / conducted.

h. QA inspection rejection rate by group.

Additionally, the residents kept appraised of the QA Monthly event status

involving projected workload and problem areas. The QA staff monthly

report was reviewed with the QA supervisor to better assist on-site

NRC-TVA communications efforts in the Unit 2 restart effort. During the

month of December the Quality Control Section performed 5,376 inspections

with an overall reject rate of 7.2 percent. Modifications welding /NDE

ingection reject rate was 21.1 percent. The majority of these were

har>ge inspections.

The Qm d) Engineering Section reviewed 650 plant instruction changes,

854 percM n documents, and 533 maintenance requests. The major problem

area was mai in purchase documents not meeting standard practice

criteria.

The Workplan Review Group reviewed 96 workplans in December. The majority

of the problems identified was due to inadequate design information and

insufficient steps to delineate work requirements. .

The Quality Surleillance Section completed three surveillances. In

addition to the surveys, 66 man-hours were expended on plant tours.

Problems and recommendations were coordinated with line supervisors.

There were 19 corrective action reports (CARS) issued and seven closed.

There were 128 CARS open with 14 delinquent. A total of six CAR correc-

tive action due dates have been extended for three or more times. Eight

significant CARS were issued. There were 81 discrepancy reports (DRs)

issued and 62 closed. There were 123 open with 22 delinquent.

In the Quality Improvement (QI) area the following actions were noted:

a. A Qaality Circle (Performance Action Team) on socket weld fit-up

acceptance criteria was set up and chaired by the QI section.

Participants included QA Surveillance, Quality Control, Modifica-

tions, DNE Codes & Standards, and Mechanical Maintenance.

b. Follow-up was conducted on the configuration problem identified with

the fuel pool cooling pump seismic supports. The proposed method of

utilizing a sketch from the IE 79-14 walkdown effort to resolve the

problem has indicated this method could be of considerable value for

similar problems on other systems.

. -- . - - -

, - - - . . - . .

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c. The November trend analysis report was issued. It included 29 charts -

containing 56 indicators involving CAQs and related QA tasks. Work

is . continuing toward full' implementation of the revised Quality -

~

Assurance Manual Part I, Section 2.16-R1 and the restart commitments

in the trending area.

. ' d. Routine walk-throughs of critical plant areasi were performed.

Current walk-throughs now have broad criteria to enter into informal

discussions with other plant personnel to facilitate good working

relationships and a sense of teamwork.

i e. Trending activities were coordinated with the Maintenance Section to

evaluate quality performance deficiencies.

f. A lesson plan was developed for a class to be presented to main- ,

tenance supervisors and engineers. The class training includes a

, composite of a variety of violations involving personnel failure to

adhere M procedures. Case studies involving actual events are

utilize and will be discussed with emphasis on'how the violations

-

could'have been prevented. The class will be presented in January as

frequently as required to allow all designated maintenance personnel

to receive the training.

g. On December 1, a " Maintenance Activities Surveillance" conference

i was held with the plant manager and maintenance top management.

i The meeting, which received a good response from plant maintenance

-

management, set the stage for the Maintenance Activities Surveillance

>

Program. The plant manager has requested that all maintenance

management provide line participation with QS (Quality Surveillance

Group)'during the' performance of maintenance activity surveys; during

this month, acceptance and participation by the maintenance personnel )

j with QS has been reported well received.

9

,

The Quality Assurance re-organization should benefit the--licensee site

l activities. More emphasis on'"line quality"~ is a positive attribute and l

should assist in increasing the QA performance overall. Effective imple-

mentation of the QA program changes will be evaluated at a future date.

.

11. Preparation for Refueling (60705)

,

The licensee began making preparations' for disassembly of the Unit 3

reactor pressure vessel- in early January 1987, in order to support fuel

off-load. These activities consisted of' Technical Specification ~ Surveil-

4

lance requirements, General Operating Instruction (GOI) performance, i

,

Mechanical Maintenance Instruction (MMI), Special Tests and radiological . '

control surveys. The resident inspectors followed . the progress- from

shield plug removal through drywell head removal, special temperature j

p profile test, reactor pressure vessel head removal, steam dryer removal

<

and steam separator removal. The activities progressed without incident.

. except as noted in the following paragraphs

i

n

. - - -- - - - _ - _ _ . _ _ _ _ . _ _ _ _ . _ _ _ _ . _ _ _ _ - _ - - . _ _ . - _ - . . _ _ . . - - - . _ _ - _ - . . . _ . . . _ - _ _ _ , _ _ _ . - _ . _ - . _ . - _ _ . . _ - - - - _ -

- . . .-- - . . . . - ,. - - .. ..

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4

a. Surveillance Activities

<

.

The inspector observed Surveillance Instr'uction (5 4.7.C,~ Secondary

<

Containment, performed ;on January 13, 1987. This SI is . required

i

prior to each refueling in order to demonstrate the capability of-

I

secondary containment to maintain -0.25 inch water pressure with.

I system -inleakage .less than 12,000 cfm. This SI had been updated and

re-issued as Revision 0 on October 29, 1986, by.the-SI Review Project.

'

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The-'SI _ was coordinated .by a Reactor Engineer from the. Technical -

.

'

Support Division ' who. provided the sign-off. initials' on the.SI and

directed an AssistantLUnit Operator (A00) in equipment manipulations. ,

i This created some . concern on the part of the ' inspector regarding

i qualifications and training of the engineer for- certain sign-offs.

, As an example, SI prerequisite steps 3.4 andL3.5~ require that the

Reactor Building Ventilation System be.in a normal operationalfstatus

j per Operating Instruction -(0I)-30 and that the Standby Gas Treatment

System (SGTS) be in standby readiness per 0I-65.'.-These prerequisites-

were signed off as being satisfied without consultation with licensed.

operators or ~ review ~ of the pertinent Operating ~ Instructions. Step _

'

3.7 was also signed off which required that Switch 16A-S34 be verified

,

to be not in DRYWELL BYPASS or TORUSLBYPASS. This switch was not

i labeled yet the engineer did not consult.the? licensed operator until

a problem developed.much later in the procedure at which time he

{ confirmed with the operator that the proper switch had been verified.

Although many of the upgraded procedures address this concern by

'

annotating on the procedure sign-off step whether a Shift Engineer

, (SE), Assistant Shift Engineer (ASE), or Unit Operator (U0) is i

! required to verify equipment line-ups, this 'SI ' contained no. such-

} notations.

I

As the SI progressed, the engineer made several judgements as .to

whether certain procedura1 ' requirements were necessary and marked N/A

I (not applicable) for certain steps. The engineer reasoned'that the

-

procedure was written for the limiting case of either one or two of

j the three unit plant being in operation. When this is'the case; some

steps have to be performed quickly to prevent an operating unit. trip

,

from high main steam tunnel temperatures which result ~from .the loss

of normal ventilation. Since no units were' operating during this

performance, the engineer did not have an AVO stationed at the

reactor building 480 V Vent Board as required-by Step 3.3.7 nor

i

was an AVO maintained at the SGTS discharge-dampers as required by

,

Step 6.6.2. The engineer additionally ' exercised judgement on Step

6.6.5 which requires that reactor building normal ventilation be.

'

4

reestablished for all zones being tested. This step.goes on to~say

i

that normal ventilation must be in service in order to' test the

isolation condition. The engineer failed to have ventilation

restored in the refuel zone even though this was one of the zones

'

being tested. He annotated this on the SI data sheet as RFZ (refuel

.,' zone) fans left in "0FF" position. This switch alignment ' created

some confusion on Step 6.7.2 when after restoring the Keylock test

l'

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, _ u. _ _ _ - _ ._ _ _ _ . . _ . ,a .-,_.__,_u..--,_ .-_a--._,,-

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19

switches for the refuel zone to the normal position, the white

indicating lamp which shows that the refuel zone was isolated went

off.

One additional point of confusion occurred during the surveillance.

Step 6.4 requires that reactor building indoor air temperature be

recorded. No temperature instrument was referenced for retrieval of

this data and after ' searching for several instruments, the engineer

~

finally gave up and entered "less than 100 degrees F." The engineer

stated that this number was not used in any calculations and was only

recorded to satisfy a reporting requirement to the NRC.

Two equipment performance questions arose during the SI and were to

be investigated by the licensee. The relative humidity (RH) heaters

for SGTS train B did not energize. This was possibly due to the low

flow cutout switch which may have been activated by the throttled

discharge dampers. The RH indicating controller was also giving

suspiciously low readings. Trains A and B readings were also

identical 'even though train A was maintained at over 125 degrees F. ,

while train B was 69 degrees F.

l

Some minor errors were noted in the SI. These were provided to the

licensee, tne most notable being Step 6.9 which states, " Verify

indicating lights on panel 9-?5 for FC0 64-36, the Drywell/ Torus

bypass damper". This indication is not provided on panel 9-25.

Prior to data manipulations, an immediate temporary change (ITC) was

made to the procedure to change the duct diameter for pitot tube air

flow calculations. The ITC takes credit for the duct wall thickness

to reduce the flow area. The effect of the ITC was to reduce the

calculated in-leakage from 11,852 cfm to 11,329 cfm. Sinc'e the

i acceptance criteria is less than 11,500 cfm, the inspector questioned

the motivation of the change and questioned whether an ITC was the

proper method for changing the procedure since it appeared to, in

effect, change the acceptance criteria. The inspector learned that i

!

an after-the-fact review by the Plant Operations Review Committee  :

focused on this topic and after a lengthy debate, the ITC was judged I

acceptable.

The above concerns related to SI 4.7.C were discussed with licensee

representatives from the operations and technical support groups on

January 16, 1987, and will be tracked as an Unresolved Item (259/260/

296/87-02-07) pending further review by the inspectors.

b. Criticality Margins for Storage of New Fuel

i

'

During a tour of the refuel floor on January 21, 1987, the inspector

noted a caution sign posted near the fire protection hose station l

2

adjacent to the Unit 1 new fuel storage vault. The sign stated, "Do

--

.

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20

not spray around an open new fuel storage vault, inadvertent criti-

cality may occur". A quick review of Section 10.2.5 of the FSAR

failed to find any reference to this condition. The safety evalua-

, tion addressed criticality .of both dry storage and inadvertent

flooding of the new fuel vault. In both cases, K f would not exceed

0.95. Following a discussion with licensee repYesentatives, the -

signs were removed. After continuing to pursue- this matter, the

inspector became aware of General Electric Service Information Letter

(SIL) No. 152, Criticality Margins for Storage of New Fuel. This SIL

discussed the new finding by GE that criticality could inadvertently

occur if low equivalent water density material were to occupy the

space in and around an array of fuel. assemblies. The SIL' recommended

additional controls be placed on fire hose stations on the refueling

'

floor to further reduce the probability of a criticality occurrence.

The ability of the licensee to control signs throughout the-facility

and to determine the basis for various caution signs is a recurring

problem. This will be tracked as an Inspector Follow-up Item (259/

260/296/87-02-08) pending a complete review of administrative

controls recommended by SIL No. 152. It should be noted that no fuel

is currently being stored in the new fuel storage vault.

While reviewing the FSAR paragraphs on new fuel storage, the

inspector noted that paragraph 10.2.4 states that each vault is also

provided with neutron dosimeters. The licensee inspected the vaults

and could not locate any neutron dosimeters. This is a deviation

!

from the FSAR (259/260/296/-87-02-09).

c. Review of Contractor Recommendations on the Fuel Handling Systems

General Electric Company (GE) was employed under the Regulatory

Performance Improvement Plan to review components and systems of

the Nuclear Steam Supply System to ensure these systems were being

operated, tested, and maintained in accordance with the design

intent. This review was also to identify any serious conflicts in

relationships between Operating Instructions (0I), Final Safety

Analysis Report (FSAR), Technical Specifications, and Surveillance

Instructions (SI) performed by control room operators.

Appendix B in the Browns Ferry Nuclear Plant Nuclear Performance Plan

identifies the actions taken and planned by TVA in response to the GE

recommendations. The inspector reviewed the Fuel Handling Systems

Report and determined that one significant deficiency remains open

almost two years after completion of the review (March 1985). The

reviewer determined that refueling platform elementary drawings were

not controlled drawings, and had not been updated following modifi-

cations. The reviewer noted that operators had complained about the

lack of refueling platform drawings for reference when problems

arise. Failure to have the revised control logic-documented for the

platform resulted in an inability of the reviewer to verify certain

platform operating characteristics as follows:

1

._,

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(1) Will the fuel grapple hook fail safe (close or remain closed)

upon loss of power to the platform?

(2) Will the fuel grapple raise motion be halted if the fuel grapple

is loaded and the hook is not completely closed?

(3) Will grapples used on the monorail or frame mounted hoists close

or remain closed upon loss of power to the platform?

The inspector determined that although the Nuclear Performance Plan

states that a project manager was assigned to coordinate and formally

track, review, and disposition the GE review recommendation and that

a cognizant engineer has 'been assigned for each recommendation with

responsibility for implementing the recommendation, this deficiency

identified in the Fuel Handling System review was not being actively

pursued.

The inspector found one further discrepancy between the. Performance

Plan, Appendix 8 and actual plant pr,actice. The performance plan

indicates that eight of the GE recommendations will not be imple-

mented and a justification for each is provided. The Fuel Handling

System reviewer recommended that GE SIL No. -222, Triangular Fuel

Grapple " Normal Up" Limit Switch Modification be implemented in order

to prevent recurrence of fuel bundles contacting the cattle shoot

during fuel movements. Although this was not one of the eight

recommendations exempted in the performance plan, Browns Ferry does

not plan to implement this modification. The implementation of GE

contractor recommendations will be tracked as an Inspector Follow-up

Item to assure consistency in the Nuclear Performance Plan

(259/260/-296/87-02-10).

12. Plant Operations Review Committee (40700)

The Plant Operations Review Committee (PORC) performed a review of the

plant activities related to the contractor evaluation of the Fuel Handling

System (see paragraph 11 for details of the contractor evaluation) on

January 27, 1986. PORC reviewed the status of seven recommendations

and concluded that fuel handling operations should not proceed until

additional actions were performed. These actions included: 1) Determine

if the recommended revision to the FSAR has been submitted and review

the impact of this revision on fuel' movement; 2) Process the drawing

discrepancy related to the control logic elementary drawings formally

with walkdown verification and full design evaluation to be completed;

and 3) Further evaluate the need to perform GE SIL 222 by conducting

discussions with experienced plant personnel. These PORC activities were

conducted professionally and appeared to comprehensively address the

concern. On a more general aspect, however, PORC failed to address in

this meeting the possibility that contractor recommendations in general

,

may not have been diligently pursued. Subsequently however, PORC did

decide to confirm the current status and closecut of contractor recom-

mendations. 'PORC action was questionable when a dispute arose over the

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22

failure history of the fuel grapple " normal' up" limit switch. Rather

than order a rigorous maintenance. history be performed, PORC chose to

order verbal discussions with - specific individuals who have .a long

history of. experience with the-grapple to determine if frequent problems

had occurred. The PORC Chairman agreed to have a maintenance history -.

review performed on this component after discussing this concern with the

inspector. No violations or deviations.were identified in this area.

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