ML20135G039

Insp Repts 50-327/85-26 & 50-328/85-26 on 850706-0805. Violation Noted:Failure to Follow Procedures for Whole Body Frisking & to Follow Procedures to Document & Correct Individual RPI Module Deficiency

ML20135G039
Person / Time
Site:	Sequoyah
Issue date:	09/06/1985
From:	Jenison K, Linda Watson, Weise S NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:
Shared Package
ML20135G032	List: IR 05000706/2008005 ML20135G029 ML20135G039
References
50-327-85-26, 50-328-85-26, IEIN-84-31, NUDOCS 8509180072
Download: ML20135G039 (16)

See also: IR 05000327/1985026

Text

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UNITED STATES

4 #p>* MGoq'o NUCLEAR REGULATORY COMMISSION

[ -

'n REGION ll

5 101 MARIETTA STREET N.W.

• *" *j ATLANTA, GEORGI A 3o323

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\...../

i Report Nos.: 50-327/85-26 and 50-328/85-26

Licensee: Tennessee Valley Authority

6N11 B Missionary Ridge Place

Chattanooga, TN 37402

Docket Nos.: 50-327 and 50-328

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License Nos.: DPR-77 and DPR-79

Facility Name: Sequoyah.1 and 2

Inspection Conducted: July 6 - August 5, 1985

Inspectors: 6 ./) . - d w h J, 9/4/85

q K.M.Jenis66,SenfofResidentInspector Dat( Signed

G 0. n v: c/C/85 t

t

. J. Watson,LResidep ~Insp c~ tor DatV Sfgned

Accompanying Insp ct r: J. W. York, Senior Resident Inspector, Bellefonte

Approve by: h7 A 7

S. P. $1se, Section Chief Date S'igned

Division of Reactor Projects

SUMMARY

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Scope: This routine, announced inspection involved 256 resident inspector-hours

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onsite in the areas of operational safety verification including operations

performance, system lineups, radiation protection, security and housekeeping

inspections; surveillance and maintenance observations; review of previous

inspection findings; followup of events; review of licensee identified items and

. review of licensee response to NRC IE Information Notice 84-31.

Results: In the areas inspected, two violations were identified (Failure to

follow procedures for whole body frisking after exit from a contaminated zone

(paragraph 5); and Failure to follow procedures to document and correct an

Individual Rod Position. Indication module deficiency (paragraph-5)).

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8509180072 850906

i PDR ADOCK 05000327

G PDR

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REPORT DETAILS

1. Persons Contacted

Licensee Employees

• H. L. Abercrombie, Site Director

• P. R. Wallacc, Plant Manager i

• L. M. Nobles, Operations and Engineering Superintendent

• B. M. Patterson, Maintenance Supervisor

M. R. Harding, Engineering Group Supervisor

J. M. Anthony, Operations Group Supervisor

D. C. Craven, Quality Assurance Supervisor ,

D. E. Crawley, Health Physics Supervisor '

• J. L. Hamilton, Quality Engineering Supervisor

• G. B. Kirk, Compliance Supervisor ,

• D. H. Tullis, Mechanical Maintenance Group Supervisor ,

• D. L. Love, Mechanical Maintenance Engineering Section Supervisor

• J. T. Crittenden, Public Safety Supervisor

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Other licensee employees contacted included technicians, operators, shift

engineers, security force members, engineers, and maintenance personnel.

• Attended exit interview

2. Exit Interview

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The inspection scope and findings were summarized with the Plant Manager and

members of his staff on August 7, 1985. Violations described in paragraphs

5.a. and 5.d. were discussed. The licensee acknowledged the inspection

findings. The licensee did not identify as proprietary any material

reviewed by the inspectors during this inspection. During the reporting <

period, frequent discussions were held with the Site Director, Plant Manager

and his assistants concerning inspection findings. At no time during the

inspection was written material provided to the licensee by the inspector.

3. Licensee Action on Previous Enforcement Matters (92702)

(Closed) Violation 328/83-29-01. The licensee's response of February 8,

1984, was reviewed and the indicated corrective actinns were audited. The

licensee took administrative disciplinary action in this case which involved

an operator who failed to follow established system alignment procedures.

System Operating Instruction S01-14.3, Condensate Demineralizer Waste

Disposal, was amended to include independent verification. Plant personnel

have also received additional training in. procedural compliance. The

licensee's corrective actions are considered complete.

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(Closed) Violation 328/83-29-03. The licensee's response of February 8,

1984, was reviewed and the indicated corrective actions were audited. The

licensee took administrative disciplinary action in this case which involved

an operator who failed to follow established system operating procedures.

Also, plant personnel have received additional training in procedural

compliance. The licensee's corrective actions are considered complete.

(Closed) Violation 327, 328/84-25-02. The licensee response of November 23,

1984, was reviewed and the indicated corrective action was audited. The

licensee submitted a revision to LER 84055 describing the events which led

to the inadequate LER and providing additional information on the breaches

of the ABSCE. LER 84055, Revision 1, was closed in Inspection Report 327,

328/85-16. In addition, the inspector noted an improvement in the quality

of LER submittals. The licensee's corrective actions are considered

complete.

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(Closed) Violation 327, 328/84-29-01. The licensee's response of

January 18, 1984, was reviewed and the indicated corrective action was

audited. The licensee revised Surveillance Instruction SI-256.2, Inspection

of Molded Case and Lower Voltage Circuit Breakers, to provide alternate

power to breaker 213 en the 125 volt Vital Battery Board III when the

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breaker is being tested and to include a signcff to ensure that the alter-

nate power is provided to the fuse column.

(Closed) Violation 327, 328/84-35-01. The licensee's response of

January 21, 1985, was reviewed and the indicated corrective action was

audited. The licensee reviewed System Operating Instruction SOI-63.1 and

verified that the procedure had been revised to include the two missing

level transmitter root valves for the cold leg accumulators. The licensee's

corrective actions are considered complete.

4. Unresolved Items

Unresolved items are matters about which more information is required to

determine whether they are acceptable or may involve violations or devia-

tions. One unresolved item identified during this inspection is discussed

in paragraph 8.

5. 0perational Safety Verification (71707)

a. Plant Tours

The inspectors observed control room operations, reviewed applicable

logs, conducted discussions with control room operators, observed shift

turnovers, and confirmed operability of instrumentation. The

inspectors verified the operability of selected emergency systems,

reviewed tagout records, verified compliance with Technical Specifica-

tion (TS) Limiting Conditions for Operations (LCO) and verified return

to service of affected components. The inspector verified that

maintenance work orders had been submitted as required and that

followup activities and prioritization of work was accomplished by the

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licensee. Tours of the diesel generator, auxiliary, turbine buildings

were conducted to observe plant equipment conditions, including

potential fire hazards, fluid leaks, and excessive vibrations and plant

housekeeping / cleanliness conditions.

During the performance of a routine Unit 2 Control Room tour the

inspectors noticed a wedge of paper pressed between two Individual Rod

Position Indication (IRPI) modules on the main control panel 2-M-4.

This same wedge of paper was' still in the panel three days later on a

succeeding inspection tour. When the assigned Reactor Operator was

questioned about the foreign object, he stated that the paper was to

keep the module from vibrating. Vibrations cause the rear contacts on

the module to become loose, resulting in a loss of indication and an

inoperable module. No Maintenance Request (MR) had been written to

repair the module, and there was no Temporary Alteration Control Form

(TACF) prescribing the use of the paper wedge. Administrative

Instruction 12 requires that personnel formally identify deficiencies

and that such deficiencies be promptly identified and corrected.

Failure to follow procedures .foi ' entification and correction of the

vibration deficiency affecting o,mrability of the subject IRPI is a

violation (328/85-26-09).

The inspectors walked down accessible portions of the following

safety-related systems on Unit I and Unit 2 to verify operability and

proper valve alignment:

Safety Injection System (U$its 1 and 2)

Turbine Driven Auxiliary Feedwater System (Units 1 and 2)

Upper Head Injection System (Units 1 and 2)

Diesel Generators (Units 1 and 2)

Emergency Gas Treatment System (Units 1 and 2)

On July 17, 1985, during a walkdown of diesel . generator 1A1, the

inspectors observed an air leak in pressure reducing valve

0-FCV-82-172A. The licensee wrote MR A-537220 to correct the

deficiency. It was determined that the body of the valve was cracked.

A new valve has been ordered and the valve will be replaced during the

first surveillance (i.e. , the next time the diesel will be _out of

service) after the valve is received. The licensee stated that the DG

was operable due to the redundant air start system for the engine. The

DG was started successfully on August 7, 1985. Repair of this valve is

identified as Inspector Followup Item (327, 328/85-26-01).

b. Verification of Ice Condenser Door Operability

The inspector's review of ice condenser door operability is documented

in Inspector Report 327, 328/85-23. During this inspection period, the

inspector discussed the methods for independent verification of removal

of door blocks from the lower ice condenser doors with the licensee.

The licensee stated that the lower ice condenser doors would be

numbered and that a double visual verification of the removal of the

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door blocks would be conducted and double signoffs recorded for each

door. This is identified as Inspector Followup Item (327, 328/

85-26-02) until licensee actions are completed.

c. Security

During the course of the inspection, observations relative to protected

and vital area security were made, including access controls, boundary

integrity, search, escort, and badging. Two security concerns were

identified during the inspection period which will be reviewed by NRC

Region II security inspectors for potential violations. These concerns

involved movement of materials into the protected area and the

visibility of the protected area boundary.

d. Radiation Protection

The inspectors observed Health Physics (HP) practices and verified

implementation of radiation protection control. On a regular basis,

radiation work permits (RWPs) were reviewed and specific work

activities were monitored to assure the activities were being conducted

in accordance with applicable RWPs. Selected radiation protection

instruments were verified operable and calibration frequencies were

reviewed.

On July 16, 1985, after observing a functional test on the B train

waste gas compressor, the inspector exited the contaminated zone and

went to the personnel frisking station on EL690. After frisking, the

inspector noted that two individuals who had also exited the

contaminated zone had not arrived at the frisking station. The

inspector found the individuals outside the regulated area. The

inspector interviewed the two individuals and determined that they had

not conducted a whole body frisk when exiting the contaminated zone.

The individuals had used a hand and foot monitor at the exit of the-

regulated zone. The inspector discussed the incident with Health

Physics. The two individuals were recalled, a whole body frisk

performed, and surveys conducted of the areas the individuals had

entered after leaving the contaminated zone. The hand and foot monitor

and whole body and area surveys performed by the licensee after the

incident indicated that the individuals were not contaminated. Radio-

logical Control Instruction RCI-1, Radiological Hygience Control, which

was established to implement the requirements of Technical Specifica-

tion 6.11, states that individuals exiting a contaminated zone are

required to perform a whole body frisk to prevent the spread of

contamination to other areas. RCI-14, Radiation Work Permit (RWP)

Program, states that it is the responsibility of each employee to

adhere to the requirements of RWPs and RWP Timesheets. RWP 02-0-85663,

which was issued to control access to the waste gas compressor room on

July 16, 1985, required employees to perform a whole body frisk upon

exit from the contaminated room. The failure to perform a whole body

frisk is a violation (327, 328/85-26-03).

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On August 5, 1985, during a routine tour of the Auxiliary Building, the

inspector observed several yellow poly bags on an Instrument Mainte-

nance Section cart which were labeled only with the sections's name and

telephone number. The material was unattended in a remote location and

not marked with survey data or radiation warning tape. The inspector

requested Health Physics (HP) to survey the bags. The inspector

discussed the observation with the licensee. The licensee stated that

the bags contained instruments which had been used inside a contami-

nated zone during the morning and were to be used in another

contaminated zone during the afternoon. The technicians using the

instruments had left the regulated area for lunch. Although the bags

were unattended and not marked, the licensee felt that the technicians

were cognizant of the location and content of the bags and were

controlling the material properly. The inspector discussed the

observation with the NRC Region II office and determined that the

procedure was act.aptable in cases where work is still in progress, the

articles are contained in yellow poly bags with identification of the

responsible section on the bag, and licensee employees have been

trained to recognize that material in yellow poly bags are potentially

contaminated. Since the procedure which controls the movement of

radioactive material inside the regulated area, RCI-1, was not clear in

how this situation should be handled, the licensee stated that the

procedure would be revised to address these requirements. The licensee

was cited previously in Inspection Report 327, 328/85-20 for not

labeling a yellow poly bag containing contaminated material which had

been left unattended in the Auxiliary Building on May 22, 1985.

Corrective actions for this previous violation are still under review.

6. Engineered Safety Features Walkdown (71710)

The inspector verified operability of the centrifugal charging pump flowpath

through the boron injection tank on Units 1 and 2 by performing a complete

walkdown of the accessible portions of the systems. The following specifics

were reviewed and/or observed as appropriate:

a. that the licensee's system lineup procedures matched plant drawings and

the as-built configuration;

b. that equipment conditions were satisfactory and items that might

degrade performance were identified and evaluated (e.g. , hangers and

supports were operable, housekeeping, etc., was adequate);

c. with assistance from licensee personnel, the interior of the breakers

and electrical or instrumentation cabinets were inspected for debris,

loose material, jumpers, evidence of rodents, etc.;

d. that instrumentation was properly valved in and functioning and cali-

bration dates were appropriate;

e. that valves were in proper position, breaker alignment was correct,

power was available, and valves were locked as required; and

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f. local and remote instrumentation was compared and remote instrumenta-

tion was functional.

No violations or deviations were identified.

7. Monthly Surveillance Observation (61726)

The inspectors observed Technical Specification (TS) required surveillance

testing and verified that testing was performed in accordance with adequate

procedures; that test instrumentation was calibrated; that Limiting

Conditions for Operation were met; that test results met acceptance criteria

requirements and were reviewed by personnel other that the individual direc-

ting the test; that deficiencies identified, as appropriate, and that any

deficiencies identified during the testing were properly reviewed and

resolved by management personnel; and that system restoration was adequate.

For complete tests, the inspector verified that testing frequencies were met

and tests were performed by qualified individuals.

The inspector witnessed / reviewed portions of the following surveillance test

activities:

SI-7, Electrical Power System: Diesel Generators

SI-64, Boric Acid Flow Paths - Valve Position Verification

SI-281, Functional Tests of Radiation Effluent Monitors with Automatic

Actuations (Quarterly)

ST -'5, Channel Calibrations for Radiation Monitoring Systems

No violatict.s or deviations were identified in this area.

8. Monthly Maintenance Observations (62703)

a. Station maintenance activities of safety-related systems and components

were observed / reviewed to ascertain that they were conducted in

accordance with approved procedures, regulatory guides, industry codes

and standards, and in conformance with TS.

The following items were considered during this review: LCOs were met

while components or systems were removed from service; redundant

components were operable; approvals were obtained prior to initiating

the work; activities were accomplished using approved procedures and

were inspected as applicable; procedures used were adequate to control

the activity; troubleshooting activities were controlled and the repair

record accurately reflected what took place; functional testing and/or

calibrations were performed prior to returning components or systems to

service; quality control records were maintained; activities were

accomplished by qualified personnel; parts and materials used were

properly certified; radiological controls were implemented; QC hold

points were established where required and were observed; fire

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prevention controls were implemented; outside contractor force

activities were controlled in accordance with the approved Quality

Assurance (QA) program; and housekeeping was actively pursued.

b. Corrective maintenance was reviewed on containment isolation valve

2-67-580D, Essential Raw Cooling Water supply to upper containment

cooler 20. The following documents were reviewed:

Maintenance Requests: A-291918 A-024413

A-533609 A-112021

A-128336 A-151118

A-024411 A-024414

A-024415

Maintenance Instructions: 6.20, Configuration Control During

Maintenance Activities

11.4, Maintenance of CSSC Valves

6.21, Repairs and <: placements of

ASME Section XI Components

6.15, General Procedure, Tightening

bolts

Temporary Alteration Control Forms: 2-85-2012-67, 1-85-5007-67

Surveillance Instruction: SI-158.1, Containment Isolation

Valve Leak Rate Test

Isolation check valves 580A through 0_ are scheduled for replacement

during the next outage period. These valves normally allow ERCW flow

through the upper compartment air cooler and become crud traps as a

result of the low flow rates and the raw condition of the ERCW. Review

of MR-A151118 indicated that a waxy substance had been found during

maintenance on the disc of valve 2-67-580D. After interviews with

those technicians that performed the disc replacement maintenance on

valve 2-67-5800, it appeared that no foreigr. material was placed into

the valve. Several tests were conducted in order to get the valve to

seat correctly and to pass the leak test. This brings the repro-

ducibility of the test into question. In addition, post modification

test data was not taken and treated as a Quality Record for those tests

that failed leak testing. The licensee is committed to ANSI standard

18.7-1976, which requires that' post maintenance test failures be

treated as quality records. Such records provide maintenance history

data for trending and planning purposes. Tests that are intended only

as troubleshooting diagnostic devices and on which operability of the

valve would not be based appear not to require retention. Because of

the iterative nature of lapping valve internals, the status of the

failed tests is unclear as a quality record. The valves were turned

over to a separate group (Engineering Test) for test evaluation four

times. During interviews with the technicians that performed the

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maintenance on 2-67-580D, the inspector also learned that its internal

flapper arm was out of round, but not replaced due to lack of parts.

This valve is scheduled to be retested in October 1985 and the results

will be evaluated by the inspector. This is an Unresolved Item (327,

328/85-26-04) until the inspector further evaluates licensee mainte-

nance controls and retention of test records associated with this and

other maintenance.

c. On July 24, 1985, the inspector observed a post maintenance functional

test on the B train waste gas compressor. The following documents were

reviewed:

Maintenance Request (MR) A526712

Hold Order 993

Maintenance Instruction MI-8.20

The compressor was identified on the maintenance request as CSSC

. equipment. The licensee stated that the equipment was QA controlled

but was not safety-related. Maintenance Instruction MI-8.20 provided

detailed steps on performance of the maintenance; however, the

procedure only stated that a functional test was to be performed after

the maintenance. The procedure did not provide any criteria for the

test. This concern was discussed with the licensee. The licensee

stated that functional test criteria would be included in tests for

CSSC equipment that is not safety-related. The licensee is also

examining assignment of responsibility for functional tests to assure

appropriate test control. This is identified as Inspector Followup

Item (327, 328/85-26-05). During the system lineup, the inspector

observed an Auxiliary Unit Operator (AU0) misalign several root valves.

The AU0 later corrected the misaligned valves prior to completion of

the test. The licensee stated that the AVO has received additional

training in the requirements for system alignment. The licensee is

evaluating the need for control of root valves on System Operating

Instructions. The licensee presently controls the position of root

valves by Hold Orders and with Instrument Maintenance procedures. This

is identified as Inspector Followup Item (327, 328/85-26-06).

9. Licensee Event Report (LER) Followup (92700)

The following LERs were reviewed and closed. The inspector verified that:

reporting requirements had been met; causes had been identified; corrective-

actions appeared appropriate; generic applicability had been considered; the

LER forms were completed; the licensee had reviewed the event; no unreviewed

safety questions were involved; and violations of regulations or Technical

Specification conditions had been identified.

a. LERs Unit 1

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327/83012 Feedwater Flow Channel Declared Inoperable

, 327/83019 Rod Position Indicator Declared Inoperable

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327/83021 Failure of Heat Tracing on BIT

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327/83025 Inoperable Flow Rate Monitor for Shield Building

Exhaust

327/83026 Failure of ERCW Valve

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327/83056 Inoperable Upper Containment Personnel Inner Airlock

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327/83065 Inoperable Train of EGTS

327/83070 Diesel Generator Declared Inoperable

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327/83080 Inoperable Rod Position Indicator

327/83110 Failure of EGTS Cooldown Valves

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327/83137 Diesel Generator 28-B Inoperable

. 327/83180 Thermal Overload Devices Failed to Trip Check

327/84013 Reactor Trip on Low-Low Steam Generator Level

! 327/84018 Fire Protection Deluge Valve Isolated

327/85008 Failure to Meet Fire Watch Requirements

i 327/85010 Containment Ventilation Isolation

327/85011 Failure to Meet Fire Watch Requirements

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327/85012 Failure to Meet Fire Watch Requirements

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i 327/85013 Failure to Meet Fire Watch Requirements

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327/85015 Failure to Meet Fire Watch Requirements

327/85017 Inadvertent Auxiliary Building Isolations

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327/85026 Inadvertent Feedwater Isolation

b. LERs Unit 2

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328/83004 RWST Low Boron Concentration

and Rev. 1

328/83051 BIT Valve Inoperable

i 328/83059 Rod Position Indication Inoperable

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328/83068 RWST Low Water Level

328/83081' Upper Containment Airlock Door Inoperable

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328/83082 RHR Remote Shutdown Channel Inoperable

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328/83093 Inoperable Flow Rate Monitor

328/83121 Failure of Lower Containment Airlock to Meet Leakage

Criteria

328/83124 Inoperable Con' denser Vacuum Exhaust Flow Rate Monitor

328/83131 Inoperable Condenser Vacuum Exhaust Flow Rate Monitor

328/83153 Containment Airlock Door Not Fully Closed

328/83161 Rod Bottom Light Bistable Inoperable

l 328/83163 Inoperable Rod Position Indicator

328/84003 Inadvertent Containment Building Ventilation Isolations

328/85005 Debris Inside Containment

c. Followup on LER 328/85001 - Unit 2 Reactor Trip on January 14, 1985

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Surveillance Instruction (SI-80), Power Range Neutron Flux Channel

Calibration and Functional Test, was conducted on Unit 2 Nuclear

Instrument Power Range Channel N-41. SI-80 referenced Instrument

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Maintenance Instruction (IMI-92-PRM-CAL), NIS Power Range, as the

required calibration procedure. Step 5.2.1.6 of IMI-92-PRM-CAL

required that the instrument power fuses be removed and independently

verified. When the instrument power fuses are removed from a Power

Range (PR) Nuclear Instrument (NI) drawer, the high voltage supply to

that drawer is removed and a negative rate flux trip bistable is

activated. The negative rate flux trip bistable does not automatically

reset when the instrument power fuses are replaced and must be manually

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reset. If two separate NI PR negative rate flux trip bistables are

activated at the same time, a reactor trip results.

A Senior Instrument Maintenance (IM) technician removed the instrument

power fuses from PR NI N-42 in error, replaced the fuses, but did not

reset the reset the channel. When the IM technician subsequently

deenergized PR NI drawer NI-41, by removing the instrument power fuses,

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the corresponding negative rate was activated. This satisfied a

] two-out-of-four reactor protection system logic, resulting in a Unit 2

i reactor trip from 30's power. The removal of the instrument power fuses

, was determined to be a personnel error by the licensee. The inspectors

verified this through interviews with personnel involved and a review

i of the licensee's IM training and qualification system. However,

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sev-:11 actions taken by the technician after removal of the fuses are

examples of a failure to follow procedure. These actions are explained

below.

Following the removal of the PR NI drawer N-42 instrument power fuses,

the IM technician failed to follow procedures in two instances. First,

he attempted to place PR NI N-42 back in service by replacing its

instrument power fuses. The appropriate section of IMI-92-PRM-CAL

(Section 7) was not implemented to ensure that N-42 was correctly

returned to service. Second, after the IM technician had returned PR

NI drawer N-42 to what he assumed to be an operable condition, he

commenced surveillance activities on PR NI N-41 without reporting to

the RO or SRO that he had not complied with an assigned procedure and

had affected safety-related equipment not authorized for su,rveillance

activities.

Administrative Instruction AI-12, Adverse Conditions and Corrective

Actions, states that plant personnel shall report any suspected

abnormal plant condition adverse to quality in the performance of their

regular work duties. A failure to follow procedure is defined in

section 4.0 of AI-12 as a condition adverse to quality. The inspector

is not satisfied that licensee personnel clearly understand their -

i responsibility to report to the shift operating staff any unintended

affects on safety-related equipment during planned activities. This

issue will be reviewed as Inspector Followup Item (327, 328/85-26-07).

The LER is closed.

10. Event Followup (93702, 62703, 61726)

a. Unit 1 Reactor Trip Due to Loss of Main Feedwater Pump Turbine (MFPT)

Oil Pump

On July 19, 1985, Unit I tripped from 100 percent power on low-low

steam generator level. The licensee was utilizing normal procedures to

search for a ground on a turbine unit board. In this process, 480V

Unit Board 1B was switched from the normal power supply to an alternate

supply. The operator closed the alternate supply breaker to the board,

but the breaker did not close the breaker completely. He then

deenergized the normal supply. He immediately noted that the MFPT oil

pump light on the board went out and jerked up on the alternate supply

breaker completely closing it; however, the oil pump tripped.

A backup oil pump started automatically on the trip signal; however,

the momentary drop in oil pressure caused partial completion of. logic

associated with the MFPT oil pressure switch. This logic includes

closing of the train associated MFPT stop valves, a main turbine

runback, and start of the motor driven (MDAFWP) and turbine driven

(TDAFWP) auxiliary feedwater pumps among others. The logic was

completed for the closing of the stop valves resulting in coasting down

of the MFPT; however, the turbine runback logic was not completed and

steam generator levels started to decrease. The MDAFWPs started, and

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the TDAFWP did not. The operators immediately put the control rods in

auto and attempted to manually runback the main turbine. This was

unsuccessful, and the reactor tripped on low-low stcam generator level.

All reactor protection system functions reacted normally on the reactor

trip including start of the TDAFWP.

The logic associated with the MFPT pressure switch is not a reactor

protection or engineered safety feature signal. The purpose of the

switch is to provide anticipatory features, such as turbine runback,

for events which could lead to low-low steam generator reactor trips. -

The inspector reviewed the logic with a Senior Reactor Operator and

discussed the failure of the logic circuit to completely energize with

the licensee. This problem was attributed to the fast operation of the

pressure switch resulting in only a momentary contact in the logic

circuit. The mechanical contacts eif.her did not have time to makeup or

the seal-in circuit was not energized.

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.The licensee attempted to duplicate the initiating conditions through

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testing to ascertain if any problems existed within the logic

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circuitry; however, the condition could not be duplicated. This

testing and performance of Surveillance Instruction SI-110.1, TDAFWP

and Valve Automatic Actuation, resulted in completion of the full

logic. The licensee adjusted the pressure switch which starts the

backup oil pump to a higher setting to preclude a significant drop in

oil pressure on trip of one of the oil pumps. This should prevent a

similar event.

No violations or deviations were identified.

b. Unit 2 Notification of Unusual Event Due to Excessive Reactor Coolant

System (RCS) Leakage

On July 29, 1985, operators noted a decrease in volume control tank

level. Subsequently, a high radiation alarm on the Auxiliary Building

Vent particulate monitor caused an Auxiliary Building isolation. The

operators followed the requirements of Abnormal Operating Instruction

AOI-31, Abnormal Release of Radioactive Materials, and AOI-6, Small

Reactor Coolant System Leak. The Auxiliary Building was evacuated and

i posted as an airborne radiation area. The operators placed excess

letdown in service and secured the normal charging and letdown flow-

paths.

. A few minutes later, a fire watch monitoring temperatures in various

pipe chases reported a high temperature in the Unit 2 EL690 pipe chase.

The Unit 2 Assistant Shift Engineer (ASE) responded to the area and

encountered two electricians exiting the area. The electricians

reported that there was a steam leak in the pipe chase. (The

electricians were subsequently decontaminated as discussed below.) The

ASE investigated the report and determined that weld upstream of sample

valve 62-674 was leaking. The weld was located at the point where the

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one inch sample line to the hot sample room joined the three inch

normal CVCS letdown line.

The leak rate was determined to be about 15 gpm. The licensee

estimated that approximately 600 gallons of RCS water was released

during the event. In addition, due to backleakage across a check valve

between the Volume Control Tank and the break, a small amount of

hydrogen was released from the VCT to the pipe chase. This leak was

promptly isolated. The licensee repaired the line and returned the

normal charging and letdown flowpaths to service on August 1, 1985.

, The licensee removed the segment of the sample line affected and sent

the weld to a metallurgical lab for analysis. It was determined that

the break which was in the heat affected zone above the weld was due to

fatigue from high cycle vibration. Two initiation points approximately

'

180 degrees apart were observed. The licensee stated that these points

were due to high cycle, low stress fatigue. The break propagated about

300 degrees. The licensee has evaluated the configuration and has

installed a hanger on the line. The installation will be evaluated to

assure that the vibration is corrected. The weld on the Unit I sample

line was visually examined, and no defects were identified. An

appropriate hanger configuration will be installed on Unit I when the

Unit 2 analysis is complete. Completion of the harger work will be

followed by the inspector and is identified as Inspector Followup Item

(327,328/85-26-10).

Eleven personnel suffered skin contamination as a result of tne primary

leak described above. The most severe case was 11,000 dpm on one

person's face. All personnel were decontaminated using normal methods

(showers) and had no detectable contaminatica after they were

decontaminated.

On July 30, 1985, the licensee posted the control building and an area

adjacent to the entrance to the Auxiliary Building on EL690 as airborne

areas. The air samples from these areas had been analyzed in the plant

chemistry lab. The readings were later determined to be erroneous due

to the high background in the lab by sending samples to the Power

Operations Training Center laboratory for analysis. These samples

indicated no airborne radiation was present.

While placing the Auxiliary Building general supply fans back in

service, three Auxiliary Building isolations occurred due to spiking on

0-RM-90-1018. The licensee stated that the setpoint on this monitor

was reset to a higher value which was still within the range required

to meet Technical Specification 3.11.2.1. Verification of the setpoint

change by the inspector is identified as Inspector Followup Item 327,

328/85-26-08. The Auxiliary Building Vent System was returned to

service with the Auxiliary Building Gas Treatment System operating to

reduce airborne levels in the Auxiliary Building.

No violations or deviations were identified.

i

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11. Review of Licensee Actions on NRC Information Notice 84-31 (92703)

The inspectors examined the licensee's handling of. NRC Information Notice

-

84-31, Increased Stroking Time of Bettis Actuators Because of Swollen

Ethylene - Proplyene Rubber Seals and Seal Set, dated April 30, 1984. The

information notice described the problem as follows:

The G. H. Bettis Company is a supplier of actuators used principally in

heating, ventilating and air-conditioning (HVAC) safety-related systems.

The G. H. Bettis Company notified the. NRC via a Part 21 report that their

NCB series, N52X, N72X, N73 series, and the NT310-SR4 and 5 and NT312-SR5

actuators had potential stroking times of greater than the required 15

seconds because the EP elastomers in contact with the Mobil 28 grease

l lubricant could swell. (The 15-second stroking time was used by Bettis as

typical of customer requirements.) The actuator seals swell when in contact

with the Mobile 28 grease currently used in the manufacture of "N" series

actuators. Where it is necessary to replace swollen seals, Bettis

,

'

recommended replacing them with new seals and using Dow-Corning Molykote 44

grease, a silicon based lubricant which Bettis states has been shown to

7

cause no seal degradation and adequate lubrication.

.

The G. H. Bettis Company also identified another problem that could ,

adversely affect stroking time. Their report states that ". . .the magnitude  ;

of stroking time degradation is related to the elapsed time between actuator

i

cycles. The longer the actuator remains stationary the more " set" the seals

, take. The set characteristic causes the seal to form an intimate contact

1

with the sealing surfaces, further increasing the time required to

initialize stroke. Once the actuator begins to stroke, the seals begin to

recover their original shape, thus freeing the unit up. Stroking the

actuator three or more complete cycles using pressurized gas will cause the

seals to recover sufficiently to reduce stroking time to a minimum. No seal

degradation has been traced to periodic actuator stroking, quite the

'

opposite has been experienced. Frequent stroking tends to extend seal life ,

resulting in longer actuator cycle life."

The inspectors reviewed a memorandum from H. A. Abercrombie, Director of

.

Nuclear Services, to Sequoyah and other sites, dated August 22,. 1984

! indicating that all safety-related valves and dampers of the model series

! described in the notice should have the seals and lubricant replaced at the

j first available outage. The memorandum requested that a search be made of

site documentation in order to identify the valves and dampers involved. .t

! Twenty seven (27) valves and dampers were identified at Sequoyah and three

l Nonconformance Reports (NCRs) were . written. The NCRs with pertinent

, information are as follows:

a. NCR EEB 84-12 dated December 1984 involved valves 1-FCV-77-420 and

2-FCV-77-421. Both of these valves were still in the warehouse and

have been tagged as not to be used.

l b. NCR MEB 84-07 dated September 1984 involved 17 valves used in three

j separate areas.

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15

1. Three valves are used to actuate the vacuum breaker system for

containment and come under ASME Section XI rules for testing.

These valves have a maximum allowed stroking time of 25 seconds

and must be tested quarterly. The latest test (June 26, 1985)

indicated stroking times for the three valves ranged from 6.4 to

6.8 seconds. The three valve numbers are 2-FCV-30-46,

2-FCV-30-41, and 2-FCV-30-40,

2. Ten valves are used to actuate the Emergency Gas Treatment System.

Each train is tested every other month but each of the valves is

not tested individually to determine stroke time. The fastest the

system has to actuate is 38 seconds and no problems have been

roted during the testing.

3. Four valves are Control Room HVAC isolation valves. This system

is required to be tested every 18 months and was last tested

August 17, 1984 with no problems. The four valve numbers are

0-FCV-31A-105A, 0-FCV-31A-105B, 0-FCV-31A-106A and 0-FCV-31A-106B.

The failure evaluation / engineering report (FE/ER) classified the

deficient condition to be a Category 1 (acceptable for all modes of

operation and design condition) based upon a sampling of the valves for

stroking time,

c. NCR MEB 84-08, Rev. 1, dated October 1984 involved 8 valves installed

in the fifth vital battery room for protection from tornado depressuri-

zation. The eight valve numbers- are FCO-51-485, FCO-31-486,

FC0-31-488, FCO-31-489, FCO-31-493, FCO-31-494 and FC0-31-501. The

FE/ER classified the deficient condition to be a Category I based on

the fact that the valves were not installed in the system, and the

seals and lubricant could be changed before the valves would experience

operational use. This was not done and this violates 10 CFR 50

LAppendix B Criterion XVI in that effective corrective action dictated

by the FE/ER was not taken. However, no violation will be issued since

programmatic corrective actions are currently in progress at TVA due to

,

an Order Modifying Licenses issued June 14, 1985 (EA 85-49).

Correction of this specific deficiency is an -Inspector Followup Item

(327,328/85-26-09).

!

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