Text

Insp Repts 50-413/87-25 & 50-414/87-25 on 870709-0825. Violations Noted.Major Areas Inspected:Plant Operations, Containment Integrity Verification,Surveillance Operation, Maint Observation & LERs
	ML20234C624
Person / Time
Site:	Catawba
Issue date:	09/09/1987
From:	Lesser M, Peebles T, Van Doorn P; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION II)
To:	;
Shared Package
ML20234C552	List: ML20234C549; ML20234C571; ML20234C624;
References
	50-413-87-25, 50-414-87-25, NUDOCS 8709210357
	Download: ML20234C624 (13)
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k# Urt'lTED STATES

Mooq*'

/ o ' WUCLEAR REGULATORY COMMISSIOM

REGION 11

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' ATLANTA, GEORGI A 30323 ,

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Report Nos.: 50-413/87-25 and 50-414/87-25

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Licensee: Duke Power Company

422 South Church Street

Charlotte,:NC 28242

Docket Nos.: 50-413 and 50-414 License Nos.: NPF-35 and NPF-52

Facili ty ' Name: Catawba 1 and 2

Inspection Conducted: July 9 - August 25, 1987

Inspectors: // 8/ 4$/ 9 7'

Cite' Signed

P'i K. Van D orn

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M.'S?'L6sser

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Efat4 Signed

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Approved by: . a f -f '/ -

T. A( Peeblesi Section Chief. Date Signed.

Projects Branch 2

Division of Reactor Projects

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SUMMARY

Scope: This routine, unannounced. inspection was conducted onsite, inspecting .

in the areas of. review of plant operations; containment integrity verification;. q

L surveillance observation; maintenance observation and . review of. licensee -

nonroutine event reports. -- '

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Results: Of the five (5) areas inspected, two apparent violations' were - 1

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identified in two areas (Channel Check on Auxiliary Feed Flow . Instrumentation .I

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Inadequate to Meet Specified Acceptance Criteria paragraph 3.b, and- Failure

i to Demonstrate Operability of Offsite Power Sources Within Required Time Frames

L paragraph 9.b).

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REPORT DETAILS

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'1. Persons Contacted

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Licensee Employees-

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J.'W. Hampton, Station Manager '

H. B. Barron,' Operations Superintendent

W. F. Beaver, Performance Engineer-

W. H. Bradley, QA Surveillance

S. Brown, Reactor Engineer . l

'B..F. Caldwell, Station Services Superintendent _ 'l

R. N. Casler, Operating Engineer i

R. H. Charest, Station Chemistry-Supervisor !

M. A. Cote, Licensing Specialist q

T. E. Crawford, Integrated Scheduling Superinte* Jent ;

W. P. Deal, Health Physics Supervisor- i

B. East, I. & E. .Er.gineer 1

C. S. Gregory, I. & E. Support Engineer j

C. L. Hartzell, Compliance Engineer

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J. Knuti, Operating Engineer

F. N. Mack, Project Services Engineer. '

W. W. McCollough, Mechanical Maintenance Supervisor .,

C. E. Muse, Operating Engineer i

F. P. Schiffley, II, Licensing Engineer -

G. T. Smith, Maintenance Superintendent ,

J. Stackley, I. & E. Engineer i

D. Tower, Shift Operating Engineer

R. F. Wardell, Technical Services Superintendent

J. W. Willis, Senior QA Engineer, Operations

Other licensee employees contacted included technicians, operators, 9

mechanics, security force members, and office personnel.

Attended exit interview

2. Exit Interview

The inspection scope and findings were summarized, on August 25,.1987,

with those persons indicated in paragraph 1 above. 'The inspector

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described- the areas inspected and discussed in detail. the- inspectica

findings. No dissenting . comments were received from the licensee. The- ,

licensee did not identify as proprietary any of the materials provided ,to i

or reviewed by the inspectors during ;this inspection. The,following new-

items were identified at the exit:

Violation 414/87-25-01: Channel Check on Auxiliary Feed Flow Instruments-

tion Inadequate to Meet Specified Acceptance Criteria-Required by Procedure

(paragraph.3.b).

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Unresolved Item 413/87-25-01: Inoperable Motor Driven Auxiliary Feedwater

Pump 18 Due to Isolated Pressure Transmitter (paragraph 5.g.). Violation

413/87-25-02, 414/87-25-02: Failure to Demonstrate Operability of Offsite ,

Power Sources Within Required Time Frames (paragraph 9.b.). 1

Licensee Identified Violation 413/87-25-03, 414/87-25-03: Missed Fire

Detection Instruments Surveillance (paragraph 9.c.). j

Unresolved Item 414/87-25-04: Deletion of Incident Investigation Report

Conclusions from LER Without Station Manager Approval (paragraph 9.d.).

3. Licensee Action on Previous Enforcement Matters (92702)

a. (CLOSED) Unresolved Item 413/86-50-02, 414/86-53-02: Determination l

by Licensee as to Whether the Computer Point High Temperature Alarm ;

for the SNSWP Temperature Meets the Intent of the FSAR. The licensee

has determined that a computer point alarm for SNSWP temperature is l

acceptable. The inspector reviewed a January 29, 1987 letter from

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R.F. Wardell to W. A. Haller of proposed revisions to the FSAR. Item

(8) proposes a clarification concerning the SNSWP temperature alarm.

Based on this letter, this item is closed. l

b. (CLOSED) Unresolved Item 414/87-05-04: Channel Check on Off Scale.

High Auxiliary Feed Flow Gauges. The inspector reviewed Duke Design

Engineering's evaluation of off scale high Auxiliary Feed Flow gauges

when at 100% and the inability to perform a reliable channel check.

As documented in Problem Investigation Report (PIR 2-C87-0031) Duke

Design Engineering concluded that in order to compare control room

and remote indications, the flowrate "must be within the gauge

capability or the loop must be isolated and a test differential

pressure put on the flow transmitter". The inspector also received a j

memorandum from NRC:RII dated May 14, 1987 from Alan R. Herdt to

Thomas A. Peebles concluding that the use of pegged gauges to perform I

channel checks is unacceptable "because a gauge could fail high and

not be detected".

The memorandum further stated that " pegged gauges cannot be checked

unless some action is taken to place the indication on scale". On

three different occasions (8/26/86,1/4/87, and 3/1/87) the channel

check performed by the licensee was inadequate to meet the acceptance

criteria of 30 gallons per minute as specified in PT/2/A/4600/03A,

Monthly Surveillance Items, because the gauges were off scule high.

This is identified as Violation 414/87-25-01 Channel Check on I

Auxiliary Feed Flow Instruments Inadequate to Meet Specified

Acceptance Criteria required by Procedure,

c. (CLO3ED) Unresolved Item 414/86-15-04: Review of Training and

Guidance for Performing Reactor Trip Reviews. The licensee has l

completed the committed upgrading of the trip review program and no 1

violation of NRC requirements was identified. In addition a recent i

NRC team inspection reviewed the trip review program in detail . i

(See Report No. 50-413,414/87-23). !

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d. (CLOSED) Unresolved Item 414/86-16-04: Resolve Questions Associated

with Removal of Auto Close Signal Associated with Feedwater Pump .

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Discharge Valves. The inspector ~ reviewed the justification for this .2

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modification which appeared acceptable and also' verified that' the

change was incorporated into the-FSAR, Rev. 15.

e. (CLOSED) Unresolved Item 413/84-87-03: Review of Operations -l

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Corrective Action Program. The inspector verified that the . final j

action to. implement the new program had been completed. This action j

was a change to Maintenance Management Procedure 1.0, Work Request

Preparation.

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f. . (OPEN) ' Unresolved Item 413/87-10-01, 414/87-10-01: Single Failure

Vulnerability of the Nuclear Service Water System. On' August 7, 1987 ,

a letter to H. B. Tucker, Vice President Duke Nuclear Production, from '

Steven A. Varga NRC:NRR conveyed concerns of the adequacy of the

Nuclear Service Water (RN) System to respond to a certain accident

scenario. The licensee was requested under 10 CFR 50.54 (f) to

provide an analysis justifying the system adequacy and'a description

of actions planned. In response to NRC concerns Catawba ' agreed to j

swap Nuclear Service Water suction from Lake Wylie to the Standby- l

Nuclear Service Water Pond (SNSWP) when a' diesel generator is _!

inoperable greater than '12 hours1.388889e-4 days 0.00333 hours 1.984127e-5 weeks 4.566e-6 months . This- action would eliminate ..the - -j

single failure susceptibility. in question .and was implemented' by 'l

Catawba Operation. Technical Memorandum 17-18 of August 8, 1987. j

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On August 14 Duke responded to the August 7 request in a letter from 1

Hal B. Tucker to USNRC. .The letter conveyed that an' analysis had

recently been completed demonstrating the adequacy.'of a single RN . i

pump under the accident scenario. The lettercadditionally described I

compensatory measures which had been taken in the absence of the

analysis. A meeting is scheduled for further discussion in Bethesda,

Md. on August 27, 1987. This item remains open pending further

review by NRR of adequacy of the RN system. ,

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One violation was identified as described in paragraph 3.b. above. l

4. Unresolved Items *

Two new unresolved items are identified in paragraph 5 and 9. j

5. Plant Operations Review (Units 1 & 2) (71707 and 71710)

a. The inspectors reviewed plant operations throughout the reporting

period to verify conformance with regulatory requirements. Technical

Specifications (TS), and administrative controls. Control room-logs, i

danger tag logs, Technical Specification Action Item Log, and the

removal and restoration log were. routinely reviewed. Shift turnovers .

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were observed to verify that they were conducted in accordance with

approved procedures.

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

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

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The inspectors verified by observation and interviews that measures l

taken to assure physical protection of the facility met current -l

requirements. Areas inspected included the security organization, l

the establishment and maintenances of gates, doors, and isolation

zones in the proper condition, that access control and badging were

proper and procedures followed. j

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In additior. to the areas discussed above, the areas toured were - )

observed for fire prevention and protection activities. These !

included such things as combustible material control, fire protection

systems and materials, and fire protection associated with maintenance

activities. The inspectors reviewed Problem Investigation Reports to

determine if the licensee was appropriately documenting problems and

implementing appropriate corrective actions,

b. Unit 1 Summary

Unit 1 began the reporting period in Mode 3 after a manual reactor j

trip on July 6 resulting from a loss of a main feed pump. The unit j

was started up on July 8. On July 10 a turbine trip at 94% power I

caused a reactor trip. The turbine tripped during turbine valve j

testing when the disc dump valve on a combined intermediate valve l

stuck open and allowed Emergency Trip System pressure to bleed off. j

The unit returned to power on July 14. On August 23, seal leakoff s

was lost to two reactor coolant pumps and the decision was made to l

shutdown and attempt to adjust the seals by isolating the standpipes. I

Upon energizing the Source Range Nuclear Instruments, one channel

failed high causing the reactor to trip. l

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c. Unit 2 Summary

Unit 2 started this period at 100% power and on July 11 was reduced

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to 90% power for scheduled core conservation. On July 27 the unit

tripped from 90% when a condensate polisher high differential

pressure caused loss of the condensate booster pumps. This in turn

caused the main feed pumps to trip, a turbine trip, and a reactor

trip. The unit returned to service on July 28. On August 7 the unit j

required shutdown by Technical Specifications when 2CA-150 (2B Steam i

Generator Main Feed Bypass to Auxiliary Feed Containment Isolation !

Valve) failed to fully shut. An Unusual Event was declared based 4

upon the loss of containment integrity and the required shutdown. On

August 9, while still in Mode 3, reactor coolant pump 2B seal leakoff

showed signs of continuing degradation so the decision was made to j

replace the number one seal. The unit reached Mode 5 that same day. 1

Additional work included in the outage was: !

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- Bearing Replacement of 2A Auxiliary Feedwater Pump j

- Replacement of Auxiliary Feed Water Check Valves i

- Eddy Current Testing of 2A and 2D Steam Generators i

- Diesel Generator Load Sequencer Modification q

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d. An inconsistency was discovered by the licensee in Technical ~j

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Specification (T.S.) 3.8.1.1, ' Electrical Power Systems, AC sources.

With one offsite circuit and one diesel generator inoperable, Action

Statement 3.8.1.1.b requires the remaining diesel generator to be

demonstrated operable by starting and loading it, unless this-has 3

been successfully performed within.the last 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . . T.S. Amendment J

10 (Unit 1) and Amendment 3-(Unit 2) allowed credit to be taken for a- 1

previous te'ts within the last 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months in order to reduce unnecessary 1

starts. on the diesel engines. Similar credit- was allowed by j

T.S. 3.8.1.1.a, if only' 'an off site circuit is ' inoperable. .The l

inconsistency lies in T.S. 3.8.1.1.c, which applies if only one l

diesel generator is inoperable. -In-this case, credit'is not allowed - j

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for a previous test, although this situation is more conservative

than that encountered in T.S. 3.8.1.1.b where both an offsite circuit '

and a diesel generator are inoperable. The licensee was, at the'

time, in a situation where one diesel generator was inoperable but

.the remaining diesel generator had been tested-'within the past 24 l

hours, and posed the question to the inspector if the intent of i

T.S. 3.8.1.1.c was to allow credit for ' this previous ' test. The j

situation was discussed by telephone with Kahtan Jabbour and James

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Lazevnick of NRC:NRR.' The inspector was informed ' that the . T.S.

amendment was based upon precedent by North Anna and Perry Nuclear

Plants. It was intended to allow credit for previous diesel tests

when one offsite circuit is inoperable' but not when an offsite. ,

circuit and a diesel generator were inoperable. Apparently Action 1

Statement 3.8.1.1.b was inappropriately approved by NRR. The history

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and inconsistency of this T.S. is therefore being documented for

future reference.

e. One item was recently identified by an-' NRC Quality Verification

Inspection Team involving Urgent Modification NSM CN-11069100 and

20454 on the diesel generator load sequencer. The purpose of the' !

modification was to allow resetting of the load sequencer without

enabling the non-emergency diesel trips. The urgent modification

sent from design engineering installed a push- button to manually

enable the non-emergency trips after resetting the ' load . sequencer.

However, contacts of this push button were in series with two 1

independent emergency start ci rcuits . A signal failure (open '

circuit) of this push button would render both start ~ signals

inoperable, apparently failing to meet single failure criteria. The

referenced drawing i s CNEE .0120-01.01. The NSM package was. sent-

back to Duke design to correct.this def.iciency along with some other-

more minor problems. The. inspectors reviewed the revised NSM

drawings to verify correction of the single failure criteria. The

modification has been installed and tested on both ' Unit 2 diesel'

generators,

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f. Problem Investigation Report (PIR) 0-C87-0197 identified that a

retest was never performed on the Unit 1 Auxiliary Shutdown Panel

af ter installation of a modification to add 'a manual reset function

for an Auxiliary Feedwater auto start. 'The inspector reviewed an

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attached operability statement which justified continued operation

without the retest. The justification evaluated the consequences if

the modification had been installed ' incorrectly. If the reset switch

would not work, the auxiliary feed (CA) pumps and flow control valves

would not be controllable by the operators after an auto start, until

the low low steam generator level cleared. Auxiliary Feedwater

Isolation Valves would, however, be controllable. If the reset switch

was wired backwards, the motor driven CA pumps would be unaffected

but the turbine driven CA pump might trip. The operator would be

able to restart the pump. The licensee will perform a complete

retest of the modification during the upcoming refueling . outage in

October 1987. This item has been assigned an open item number. in IE

Report: 413/87-23, 414/87-23.

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g. On July 6,1987 at 3:16 pm the Unit I reactor was manually tripped

from 70% power following a loss of one main feed pump and decreasing

steam generator levels. This incident is reported in Licensee Event

Report (LER) 413/87-26. Low low water levels in 2 out of 4 steam

generators (S/G) caused an Auxiliary Feedwater System (CA) auto

start of both the motor driven CA (MDCA) pumps and the turbine driven

CA (TDCA) pump. (Low low level in one S/G would result in only the

motor driven CA pumps starting.) All three pumps started, however

ICA-468, Motor Driven CA Pump 18 . Di scharge Isolation to IC S/G,

inappropriately closed, shutting off flow. Control room personnel

noticed that 1CA-468 was closed approximately 10 minutes later and

were unsuccessful at opening it from the control room. At 3:37 pm

ICA-46B was manually opened.

The licensee initiated a separate investigation to determine why

ICA-46B went closed and the results were reported in LER 413/87-27.

The licensee discovered the motor driven pump inoperability interlock

logic had actuated and caused ICA-46B to close. The logic is

designed to protect against' a common mode failure where the TDCA

pump and the remaining MDCA pump (following a single failure of one

MDCA pump) would feed the same depressurized S/G. Under accident

conditions the CA system functions as follows: MDCA pump 1A supplies

1A S/G via 1CA-62A and S/G IB via 1CA-58A. MDCA pump 1B supplies 1C

S/G via ICA-46B and ID S/G via 1CA-428. The TCDA pump feeds S/G's 18

and 10. Normally closed valves are provided to cross connect the

MDCA pumps or to allow the TDCA pump to supply S/G's 1A and 10,

however this requires operator action. Section 7.4.1.3.2, section

l 10.4.9 and 15.2.8 of Catawba Nuclear Station Final Safety Analysis

Report (FSAR) requires that the CA system be able to supply at least

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2 intact steam generators under accident scenarios assuming a single

failure. No operator action is allowed for 30 minutes. The motor

driven pump inoperability interlock logic actuates as follows: If

MDCA pump 1B and the TDCA pump are operating following an automatic

CA start signal, and MDCA pump 1A fails to start as sensed by

discharge pressure transmitter ICAPS 5131 with a 30 second time delay,

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1CA-468 closes to prevent MDCA pump 1B and the TDCA pump from

simultaneously supplying the IC S/G. Similar logic exists to shut

ICA-58A in the event of MDCA pump 1B failure. !

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On July 6, ICA-46B closed because pressure transmitter ICAPS 5131 was

isolated and provided a false signal to the logic that MDCA pump 1A

had failed. The licensee's investigation to date has been unable to i

determine how and when ICAPS 5131 became isolated. (It is noted that I

the investigation is continuing even though LER 413/87-27 was issued

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August 6,1987 and a revision will be issued if appropriate.) The

licensee discovered that the safety related pressure transmitter J

was scheduled for an 18 month calibration in March 1987 but the

calibration had still not been performed at the time of this event.

The licensee has been unable to locate any calibration documentation

on this instrument. No other maintenance activities documented by a

work request have ever been performed on ICAPS 5131. The licensee

has not been able to demonstrate operability of this logic since

July 17, 1986 when a reactor trip occurred in which all three CA I

pumps started and the logic responded properly (LER 413/86-40).

Hence MDCA pump 1B was apparently inoperable from July 17, 1986 to

July 7,1987 as it would have supplied only 1 of 2 required S/G's

under certain conditions.

The LER safety analysis does not address whether or not the CA system

would have functioned as designed under various accident ~ scenarios

I during the time MDCA pump 18 was inoperable. The analysis does state

that "two redundant CA pumps were available throughout the event".

This is apparently referring only to the trip event, not the time.

frame from July 17, 1986 to July 7,1987 in which one of the two

redundant CA pumps was inoperable on numerous occasions.

The inspector postulated an accident scenario where the CA system

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would. apparently fail to function as designed. The scenario is as

follows: Assume a feedline break depressurizes S/G 1A, all three CA

pumps start, MDCA pump 1A feeds the break, rendering it useless for

supplying 1B S/G. Thirty seconds later ICA-46B closes because

ICAPS 5131 is valved out making up the motor driven pump inoperability

interlock logic. This isolates that line to 1C S/G. The postulated-

single failure is that the TDCA pump starts but fails after 30 ,

seconds. This leaves S/G ID being fed by MDCA pump 18 and is less

than the FSAR requirement that a minimum of 491 gallons per minute

(gpm) is supplied to two intact S/G's.

This is being identified as Unresolved Item 413/87-25-01: Inoperable

Motor Driven Auxiliary Feed Water Pump 1B Due to Isolated Pressure

Transmitter, pending determination by licensee as to why the

transmitter became isolated and location of calibration documentation.

The question as to the legitimacy of the accident scenario postulated

has been forwarded to NRC:NRR for resolution.

No violations or deviations were identified.

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l 6 '. Verification of Containment-Integrity (Unit 2) 61715

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l The inspector verified through local observation the proper positioning ;-

of electrical and mechenical barriers and isolation valves. associated

with various containment penetrations. A walkdown of portions of '. the i

Containment Air Return and Hydrogen Skimmer (VX) . system was performed.

The air lock local leak rate test on upper containment performed prior 'toi

entering Mode 4 was witnessed.

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No violations or deviations were identified.

7. Surveillance Observation (Units,1 & 2) (61726)

During ' the . inspection period, the inspector verified plant operations

were in compliance with various TS requirements. . Typical of .these

requirements were confirmation. of comp 1.iance with the TS for reactor

coolant chemistry, refueling water ; tank, emergency power systems', safety.

injection,. emergency safeguards systems, ; control room ventilation, and

[ direct current electrical power sources. The inspector verified that

I surveillance testing was performed in accordance with-the approved written-

procedures, test instrumentation was calibrated, limiting conditions for

operation were. met, appropriate removal and restoration of the affected

equipment was accomplished, test results met requirements and were

reviewed by personnel other than the individual directing the test,-and

that any deficiencies identified during the. testing were properly reviewed

and resolved by appropriate management personnel.

The following surveillance were either reviewed or. witnessed wholly .or -

in part:

IP/1(2)/A/3200/02A Solid State Protection System Periodic Test

l IP/1(2)/A/3200/08A Reactor Trip Breaker Actuating Device Operational

Test

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PT/1/A/4200/09A Auxiliary Safeguards Cabinet Periodic Test

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PT/1/A/4400/06F KD Heat Exchanger 1B Heat Capacity. Test

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PT/1/A/4400/09 Safety Related Raw Water for 1B Diesel Water

Jacket Cooler Flow Verification Test

PT/2/A/4200/02B Cold Shutdown Inside Containment : Integrity

Verification

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PT/2/A/4200/01J Electrical Penetration Leak Rate Test

PT/2/A/4200/01E Upper Containment Personnel Air Lock Leak Rate !

Test

No violations or deviations were identified.

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8. Maintenance Observations (Units 1& 2) (62703) l

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Station maintenance activities of selected systems and . components were

observed / reviewed to ascertain that they were conducted in accordance

with requirements. The inspector verified licensee conformance to the

requirements in the following areas of inspection: The activities were

accomplished using approved procedures, and functional testing and/or

calibrations were performed prior to returning components or systems to !

service; quality control records were maintained; activities performed !

were accomplished 'oy qualified personnel; and materials used were properly

certified. Work requests were reviewed to determine status of outstanding

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Jobs and to assure that priority is assigned to safety-related equipment

I maintenance which may effect system performance. ;

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The inspector witnessed portions of the following maintenance activities.

5801PRF Inspection of CA-56 Cage for Blockage

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NSM 10637 High Source Range Flux Input to Containment

Evacuation Alarm

Special Investigation Reactor Trip Breaker Maintenance j

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No violations or deviations were identified. l

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9. Review of Licensee Nonroutine Event Reports (Units 1 & 2) (92700)

a. The below listed Licensee Event Reports (LER) were reviewed to

determine if the information provided met NRC requirements. The

determination included: Adequacy of description, verification of

compliance with Technical Specifications and regulatory requirements,

i corrective action taken, existence of potential generic problems,

l reporting requirements sattsfied, and the relative safety significance

of each event. Additional inplant reviews and discussion with plant

personnel, as appropriate, were conducted for those reports indicated

by an (*). The following LERs are closed:

LER 413/85-22 Rv.1 Inoperability of Diesel Generator 1B

LER 413/87-01 Rv.1 Inadvertent Waste Gas Release Due to Failure

of Waste Gas Drain Trap

LER 413/87-21 Failure to Verify Availability of Offsite

Power Sources Due to Per sonnel Error

LER 413/87-22 Technical Specification Violation Because

Leak Rate Testing Not Performed Due to

Incorrect Procedure

LER 413/87-24 Failure to Verify Operability of Offsite

Power Sources Due to Insufficient Supervision

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LER 413/87-25 Unit i Vent Flow Rate Estimate Not Performed

Within Required Time Due to Personnel Error

LER 413/87-26 Manual Reactor Trip Following Main Feedwater i

Pump Speed Decrease Due to Manufacturing

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LER 413/87-29 Auxiliary Feedwater Auto-Start from Main )

Feedwater Isolation Following Opening of

Reactor Trip Breaker

LER 414/86-38 Main Feedwater Pump Trip Due to Pressure

l Switch Out of Calibration

LER 414/86-47 Rv.1 Termination of Containment Air Release Due

To Installation Deficiency and Conservative

Radiation Gas Monitor Setpoint i

LER 414/86-51 Rv.2 Reactor Trip Due to Feedwater Pump Suction

Flow Pressure Switch Out of Calibration

LER 414/86-53 Rv.2 Reactor Trip Due to Personnel Error,

Procedural Deficiency, and Failure of Main

Feedwater Pump Automatic Controller

LER 414/87-10 Unit Shutdown Due to High Reactor Coolant

System Unidentified Leakage

LER 414/87-12 Snubber Inoperability Resulting in Technical

Specification Violation Due to a Defective

Procedure

b. Followup on LER's 413/87-21 and 24 and reportable incident of

7/21/87. On June 8,1987 at 11:00 a.m. a station engineer determined

that electrolyte level in two batteries associated with diesel

generator 1A were low and required declaring the diesel inoperable.

His actions to inform proper personnel were untimely and resulted in

a failure to demonstrate operability of offsite A.C. power sources

with the time limits required by Technical Specification (TS) 3.8.1.1.

The required available power source operability check was performed

at 12:50 p.m. approximately 50 minutes late.

On June 17, 1987, 18 diesel generator was declared inoperable at

6:00 a.m. The available power source operability check was completed

at 5:55 a.m. and TS 3.8.1.1 required the next check- be performed

by 1:55 p.m. The person responsible for assuring the check failed

to remember this until it was too late. The required check was

completed at 3:40 p.m. approximately 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months and 45 minutes late. On

July 20 the 2B diesel generator was declared inoperable. Shift

personnel were sensitive to the fact that offsite power source checks

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_ _ _ _ _ . ._.

11

l

\

had recently been missed and discussed the importance to ensure that ;

it gets accomplished. On July 21 at 3:50 a.m. the time limit was

again exceeded as shift personnel forgot. The available power source

operability check was performed at 5:00'a.m. , approximately 1 hour1.157407e-5 days 2.777778e-4 hours 1.653439e-6 weeks 3.805e-7 months

and 10 minutes late.

The inspectors documented a previous similar occurrence of failure

to perform the required available power source operability check on

March 2,1987 with a Licensee Identified Violation (LIV 413/87-10-03).

The licensee's corrective actions, however, have been ineffective in

preventing recurrence. Therefore, the incidents of June 8, June 17

and July 21, 1987 constitute three examples of a violation of

T.S. 3.8.1.1, Violation 413/87-25-02, 414/87-25-02: Failure to

! Demonstrate Operability of Offsite Power Sources Within Required Time

l Frames. l

'

l

l c. On June 16, 1987 a licensee audit discovered that the 6 month l

l surveillance on Fire Detection Instrumentation required by Technical j

Specification (TS) 4.3.3.8.3 had not been performed since August 27, i

1986. The Standing Work Request (SWR) which performed the surveil- l

lance had been deleted to make way for three new SWR's which would

divide the work up. However, lack of administrative controls '

resulted in the three new SWR's never being created, thus the

surveillance were not performed until discovered by a Test Review

Committee audit. This incident was investigated and reported in

l Licensee Event Report 413/87-23. As permitted by Appendix C of

10 CFR 2, no Notice of Violation is proposed and this incident is

l classified as a Licensee Identified Violation (LIV 413,414/87-25-03)

Missed Fire Detection Instruments Surveillance.

4

d. LER 414/87-09 reported an Engineered Safety Feature (ESF) Actuation j

l of the Containment Air Return Isolation Dampers (VX) on March 12, >

1 1987 due to a defective procedure. As a result of this, applicable

procedures were reviewed including PT/2/A/4450/05B Containment Air

Return Fan and Hydrogen Skimmer Fan Periodic Test. LER 414/87-15

reported that while performing PT/2/A/4450/05B on April 6,1987 a VX

l

damper unexpectedly actuated. Again the cause was a defective

procedure, the defect being exactly the same as that reviewed af ter

the March 12 actuation. The review of PT/2/A/4450/05B had been

ineffective and the defect had not been corrected. The Catawba

!

Safety Review Group (CSRG) investigation into the April 6 actuation

l was documented in Incident Investigation Report (IIR) C-87-032-2, and ;

i correctly concluded that a "more thorough review of procedures as a {

result of the March 12 actuation would have identified PT/2/A/4450/058 )

as being defective and would have prevented this incident". This j

conclusion, however was deleted from the LER (LER 414-87-15) which is- '

normally a reproduction of the-IIR. When the licensee was questioned, !

the inspector was told that this conclusion was not the root cause of 1

the event, was not required to be reported and was for internal use

<

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.a .

12

only. -The -inspector pointed out that this was a pertinent piece of'-

information and a system weakness which contributed to-the event that

requires both reporting and corrective action.

Technical Specification 6.5.1.6 requires that all reports evaluating

reportable events be approved by the ' Station Manager. Credit has

been given for ' approval of the IIR by the Station Manager in. lieu of

the LER provided the:LER contents'are the same as the IIR. (Deleting

Sections of the'IIR such as a Sequence of Events list for the LER is

acceptable- provided- the evaluation of the- event is an. essential <

replica.) _ In this case, the' Station Manager did not approve the !

deletion of the. previously described IIR conclusion from _ the LER, .

, effectively failing toj approve- the contents of the LER. This'is-

1 Unresolved Item - 414/87-25-04: Deletion of Incident . Investigation

Report Conclusions.From LER Without Station Manager Approval; pending

evaluation by NRC management and NRC review of'any corrective actions

relative to the inadequate corrective actions.

One NRC violation was identified as described in paragraph 9 b..~above and

.

one Licensee Identified Violation was identified as described,in paragraph'

9.c.

l

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_ . _ _ _ _ _ - . _ . . _ _ .__...-2._ . . _ - _ ___..-___m.--____.-_ _.2._._m

ML20234C624

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