ML20204J639

Insp Repts 50-373/86-21 & 50-374/86-20 on 860510-0611. Violation Noted:Failure to Follow Procedures Re Operational Safety Verification

ML20204J639
Person / Time
Site:	LaSalle
Issue date:	08/04/1986
From:	Wright G NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20204J605	List: ML20204J602 ML20204J639
References
50-373-86-21, 50-374-86-20, NUDOCS 8608110155
Download: ML20204J639 (11)

See also: IR 05000373/1986021

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U.S. NUCLEAR REGULATORY COMISSION

. REGION III

Reports No. 50-373/86021(DRP); 50-374/86020(DRP)

. Docket Nos. 50-373; 50-374 Licenses No. NPF-11; NPF-18

Licensee: Commonwealth Edison Company

Post Office Box 767

Chicago,~IL' 60690

) Facility Name: LaSalle County Station, Units 1 and 2

Inspection At: LaSalle Site, Marseilles, IL -

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Inspection Conducted: May 10 through June 11, 1986

l Inspectors: M. J. Jordan

! J. Bjorgen

R. Kopriva

Approved By: . C. Chief

Reactor Projects Section 2C

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Inspection Summary

Inspection on May 10 through June 11, 1986 (Reports No. 50-373/86021(DRP);

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50-374/86020(DRP))

Areas Inspected: Routine, unannounced inspection conducted by resident

inspectors of operational safety; surveillance; maintenance; training; unit

trips; refueling / outage; and regional requests.

} Results: Of the seven areas inspected, no violations or deviations were

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identified in six areas; two violations were identified in the remaining area

(failure to follow procedures, Paragraph 2). During this inspection period,

. several times the operator log and shift engineer log did not adequately

reflect the shift occurrences. This is a continuing problem at LaSalle. In

addition, personnel errors and failure to follow procedures continues to be a

problem.

8608110155 860804

gDR ADOCK 05000373

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DETAILS

1. Persons Contacted

• G. J. Diederich, Managerk LaSalle Station

• R. D. Bishop, Services Superintendent

• C. E. Sargent, Production Shperintendent

D. Berkman, Assistant Superintendent, Technical Services

W. Huntington, Assistant Superintendent, Operations

J. C. Renwick, Assistant Superintendent, Work Planning

R. W. Stobert, Quality Assurance Supervisor

P. Manning, Tech Staff Supervisor

T. Hammerich, Assistant Tech Staff Supervisor

W. Sheldon, Assistant Superintendent, Maintenance

• J. Atchley, Operating Engineer

• D. Winchester, Senior Quality Assurance Inspector

The inspectors also talked with and interviewed members of the operations,

maintenance, health physics, and instrument and control sections.

• Denotes personnel attending the exit interview held on June 11, 1986.

2. Operational Safety Verification (71707)

The inspector observed control room operations, reviewed applicable logs

and conducted discussions with control room operators during the inspection

period. The inspector verified the operability of selected emergency

systems, reviewed tagout records, and verified proper return to service of

affected components. Tours of Units 1 and 2 reactor buildings and turbine

buildings were conducted to observe plant equipment conditions, including

potential fire hazards, fluid leaks and excessive vibrations and to verify

that maintenance requests had been initiated for equipment in need of

maintenance. The inspector by observation and direct interview verified

that the physical security plan was being implemented in accordance with

the station security plan.

The inspector observed plant housekeeping / cleanliness conditions and

verified implementation of radiation protection controls.

The inspectors, while touring the plant, noticed oil leakage from some of

the post tension connections for containment on Unit 2. This leakage was

brought to the attention of the station management for evaluation based

on a similar problem at the Farley Unit 2 plant addressed in IE Information

Notice 85-10, Supplement 1. The licensee reported back that no problem

existed with the post tension connections because less than one gallon

total leakage had occurred and less than one percent of that contained in

any one tension was noticed. The next grease coverage check surveillance

required by technical specifications is scheduled for late 1986. Since no

free standing water was noticed in the tension cap, no further action was

needed at this time. The inspector also reviewed the method by which IE

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Notice 85-10, Supplement 1, was implemented at the site. LTS 1000-1 had

been implemented to require a laboratory analysis of the grease samples

and identified the action to be taken if the samples indicate free water.

On May 26, 1986, at approximately 21% power on Unit 2, the unit operator

was pulling control rods during a normal power increase in accordance with

the Unit Start-up Procedure LGP-1-1. The operator incorrectly pulled

Control Rod 10-19 from Position 12 to Position 48 instead of from

Position 12 to Position 24. The Rod Worth Minimizer (RWM) was bypassed

at the time and a second operator was being used as a verifier to assure

that rod movements complied with the rod movement sequence.

The unit operator selected the next rod in the sequence, Rod 10-43, and

pulled it one notch, from Position 12 to Position 14 before the movement

error for Rod 10-19 was identified. He then reinserted Rod 10-19 to the

correct position of 24. The total elapsed time during this sequence was

approximately one minute based on a review of the process computer alarm

typer. The unit operator then contacted the nuclear engineer on-call to

verify that the actions taken were correct. The nuclear engineer approved

the operator's actions based on his knowledge of the rod sequence controls

and preferred operator actions and authorized continued rod movements.

The nuclear engineer subsequently reviewed the event on his home computer

console to confirm nothing abnormal had occurred due to the mispositioned

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rod. Reactor power was confirmed to be above the 20% technical specifi-

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cation limit thereby eliminating the need for the RWM or for a second

person to act as a verifier.

Later the nuclear engineer stated to the inspector that the RWM would allow

movement of the noted rod to position 48 before an "out-of-sequence" light

would be displayed. Furthermore, the Rod Sequence Control System (RSCS),

which also does not enforce rod blocks above 20% power, would not prevent

rod movement above notch position 12.

A thorough review of this event was performed by the inspector. This

review included the applicable operator and shift engineer logs, the

applicable alarm printouts, licensee procedures, discussions with the

personnel involved, and the licensee's corrective actions.

Technical Specification 6.2.A.1 requires detailed written procedures to be

, prepared, approved, and adhered to including the applicable procedures

recommended in Appendix "A" of Regulatory Guide 1.33, Revision 2, February

1978. Section 4 of Appendix A to this Regulatory Guide requires procedures

for plant start-up. Licensee Procedure LGP-1-1 for normal plant start-up,

Paragraph F.2.h., requires controls rods to be withdrawn in accordance with

the approved rod sequence provided by the nuclear engineer.

Contrary to the above, Unit 2 Control Rod 10-19 was withdrawn from

Position 12 to Position 48 on May 26, 1986 in lieu of the required

position of 24. This is considered to be a violation (374/86020-01(DRP)).

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In addition to the mispositioned rod, the inspectors are concerned about

three other aspects of this event: the immediate operator actions; the

timeliness of the identification of the error by the independent verifier;

and the failure of the unit operator to make a log entry describing the

error. The licensee's Procedure LOA-RD-03 provides the applicable

instructions for mispositioned control rods. For a control rod withdrawn

beyond its in-sequence position, Section D.2 of this procedure requires

the operator to demand process computer printouts 00-3 (Core Thermal

Power), OD-7 Option 2 (Control Rod Position), and 00-8 (Local Power Range

Monitor readings) and then to consult the Nuclear Engineer for the method

of returning the mispositioned rod to its correct in-sequence position.

Contrary to the procedure, on May 26, 1986, the operator failed to demand

the required process computer printouts and failed to consult the

Nuclear Engineer prior to returning the mispositioned control rod to

its correct in-sequence position. This is considered to be a violation

(374/86020-02(A)(DRP)). In this case, fortuitously, the operator action

was what the nuclear engineer would have recommended. The inspector is

concerned, however, that the operator took the action prior to consulting

with the nuclear engineer. Depending on circumstances, this practice

could complicate a problem rather than help it. In addition, during an

interview with the unit operator subsequent to the event it became

apparent that the operator was not aware that he had selected and moved

another rod prior to discovery of the mispositioned rod, a result of not

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obtaining and reviewing the required process computer printouts, 00-3,

00-7 Option 2, and 00-8.

The timeliness of the independent verifier noting the error was the second

concern. The inspector noted that the duties of the verifier, i.e., what

was expected of him, were not clearly defined. However, the action state-

ment for Technical Specification 3.1.4.1 (Rod Worth Minimizer) states that

with the RWM inoperable control rod movement and compliance with the pre-

scribed control rod pattern shall be verified by a second licensed operator

or other technically qualified member of the technical staff who is present

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at the reactor control console. Therefore, this individual should have

been aware of his duties even without the benefit of a prescriptive

procedure. (The inspectors must note again, however, that the referenced

technical specification does not apply above a reactor power of 20L

Therefore, there was no technical specification requirement for the second

operator to verify rod movement.) The licensee plans to revise the appro-

priate procedures to provide this guidance. Completion of this action will

be tracked as an open item (374/86020-02(B)(DRP)).

The failure of the unit operator to log the error is another item of

concern. In this case, the operator was instructed not to make an entry

until the shift engineer obtained clarification on the seriousness of the

error. The shift engineer then neglected to provide the unit operator with

the appropriate clarification. The inspector has previously expressed

concern with the adequacy of log entries. Furthermore, the inspector is

concerned that the operator did not make an entry because of instructions

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from the shift engineer. The log book in question is the unit operator

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log and is required to contain all pertinent information related to opera-

tion of the facility. The seriousness of the error has no bearing on

whether it should be recorded. The log book is used to record facts;

analyses of these facts can be done in another forum.

Secondly, the inspectors are concerned that the shift engineer instructed

the unit operator not to make the initial entry. As the senior management

representative onsite he should be aware of the procedural requirements.

If he so desires, his log book can describe clarifications of operational

events but he should not instruct operators to not record information.

Corrective actions for this example of inadequate logs will be monitored

as an open item (374/86020-02(C)(DRPP).

In summary, a control rod inadvertently was mispositioned to Position 48

instead of to Position 24. Although not required by technical specifi-

cations for the operational conditions at the time (21% power) a second

operator had been assigned to the control room as a verifier but did not

notice the mispositioned rod until the next rod in the sequence had been

moved.

During the month of May, the inspector walked down the accessible

portions of the following systems to verify operability:

Units 1 & 2 Emergency Diesel Generators

Units 1 & 2 Standby Gas Treatment Systems

3. Monthly Surveillance Observation (61726)

The inspector observed technical specifications required surveillance

testing and verified for actual activities observed that testing was

performed in accordance with adequate procedures, that test instrumentation

was calibrated, that limiting conditions for operation were met, that

removal and restoration of the affected components were accomplished, that

test results conformed with technical specifications and procedure require-

ments 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 inspector witnessed portions of the following test activities:

LIS-NB-214 Calibration of Reactor Vessel Pressure Switch 2B21N039N

LST 86-096 Unit 2 Special Test to Check Trip Point of Low Level Scram

Switches

. LIS-NR-402 Intermediate Range Monitor Rod Block and Reactor Scram

Functional Test

LIS-NB-204 Unit 2 Reactor Vessel Low-Low Water Level RCIC Initiation

and Low-Low-Low LPCS/RHR Initiation Calibration

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On June 1, 1986, the licensee was performing a surveillance test on the "A"

turbine driven feedwater pump when problems developed and water level

dropped below the scram setpoint but the reactor failed to scram. A

preliminary notification (PNO-III-86-52) was issued by Region III and

an Augmented Investigation Team was dispatched to the site to investigate

the matter. This entire issue will be addressed in Special Inspection

Report No. 50-374/86023.

On May 25, 1986 at approximately 35% power on Unit 2, the licensee

attempted to shift the reactor recirculation pumps from slow speed to fast '

speed. The "2B" pump tripped on neutral overcurrent. This left the "2B"

recirculation pump off and the "2A" pump in slow speed. A discussion was '

held among the plant management concerning meeting the technical specifi-

cations for operation with one recirculation pump and it was determined

that resetting of the Average Power Range Monitors (APRM) and the Rod Block

Monitor (RBM) could not be accomplished within the four hours allowed by

technical specifications so a reactor shutdown was commenced. The reactor

power was then reduced to less than 20% power by inserting rods. Technical

Specifications 4.1.4.1 and 4.1.4.2 require that surveillance testing of the

Rod Worth Minimizer (RWM) and the Rod Sequence Control System (RSCS) be

performed prior to reducing power to less than 20%. However, this was not

recognized by the licensee and therefore these surveillances were not

performed during this shutdown (the surveillance had been performed on

May 23, 1986 prior to commencing the startup). Subsequent to the event a

discussion between the inspector and shift engineer revealed the RSCS and

RWM were enforcing during the power reduction from 20% to 15%. At approxi-

mately 15% power the "2B" recirculation pump was restarted (successfully)

at slow speed and reactor shutdown was stopped. The RSCS and RWM then were

calibrated prior to power ascension.

The missed surveillances during reactor shutdown were identified by

the next oncoming shift control room engineer during shift turnover.

Further evaluation by the licensee and by the inspectors revealed that

Procedure LGP 3-1, " Power Changes," which the shift was using to change

power did not have any precautions or warnings that these surveillances

had to be performed before reducing power below 20%. A change to this

procedure was initiated. No violation will be issued at this time

concerning the missed surveillance because the licensee notified the NRC,

took prompt corrective action, and this issue meets the requirements for

not issuing a violation specified in 10 CFR Part 2, Appendix C.

The inspectors reviewed the log books for this period of time and determined

that neither the unit operator log nor the shift engineer log contained any

information relative to any problem occurring with missed surveillances.

There was no reference to any technical specification violation covering

missed surveillance. The procedure for both of these logs (LAP 220-1 and

LAP 220-2) requires the logging of any technical specification violation,

corrective actions, and time of return to technical specifications.

Therefore, this is a violation of the regulations and a citation normally

would be issued by the NRC. Although the licensee did not identify this

particular instance of failure to log the required information, the

licensee has acknowledged the generic problem of failure to record

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significant information in log books at LaSalle and has identified

corrective actions to improve the quality of the log books. At the time

of this latest violation, the corrective action programs had been started

but not yet fully implemented. Therefore, the inspectors believe that

this issue meets the intent of the requirements for not issuing a Notice

of Violation as set forth in 10 CFR Part 2, Appendix C and that issuing a

violation at this time would serve no purpose. However the licensee has

been informed that appropriate enforcement action will be taken for future

violations.

4. Monthly Maintenance Observation (62703)

Station maintenance activities of safety related systems and components

listed below were observed / reviewed to ascertain that they were conducted

in accordance with approved procedures, regulatory guides and industry

codes or standards and in conformance with technical specifications.

The following items were considered during this review: the limiting

conditions for operation were met while components or systems were removed

from service; approvals were obtained prior to initiating the work, activi-

ties were accomplished using approved procedures and were inspected as

applicable; 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; and fire prevention controls were implemented.

Work requests were reviewed to determine status of outstanding jobs and

to assure that priority is assigned to safety related equipment maintenance

which may affect system performance.

The following maintenance activities were observed / reviewed:

The inspector observed replacement of the Unit 1 Control Rod 34-23

Hydraulic Control Unit 111 valve (Work Request L 57655), using

Procedure LMP-RD-07.

The inspector also observed the partial disassembly and inspection of the

Unit 2 Residual Heat Removal (RHR) heat exchanger discharge valve,

2E12F003A, to investigate the reason for its failure to operate on June 3,

1986. The valve is a normally open valve which is required to be closed

during warming of the shutdown cooling piping in preparations for entering

cold shutdown. After warming is completed, the valve must then be reopened

to place the RHR heat exchanger in service for shutdown cooling. The valve

had failed to reop'en either remotely or locally when trying to enter cold

shutdown. The "A RHR shutdown cooling loop was declared inoperable and

the limitorque operator was removed for repairs. Upon disassembly, the

valve operator was found to have a broken drive sleeve and damaged drive

gear teeth.

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Discussions with licensee personnel indicate that this valve has a history

of nydraulic lock, a condition that allows leakage past the valve seating

surface into the bonnet area. This high pressure water then provides a

hydraulic lock between the valve bonnet and the wedge, preventing valve

movement. To release the hydraulic lock, the licensee routinely loosens

the valve stem packing to vent the bonnet area. The licensee suspects that

attempts to open the valve against the hydraulic lock may have resulted in

overtorquing the operator and causing the failure. The drive sleeve

failure also suggests the possibility of a fatigue failure due to the

presence of rust and signs of aging in portions of the break area.

The inspector continues to be concerned about the possible generic problem

of this failure for other limitorque valve operators as to the root cause

of equipment failure and corrective actions as well as how to resolve

hydraulic lock problems of valves in this system and other systems. Until

licensee evaluations and corrective action plans are completed, this will

remain as an open item (374/86020-02(DRP)).

5. Training (41400)

The inspector, through discussions with personnel and a review of training

records, evaluated the licensee's training program for operations and

maintenance personnel to determine whether the general knowledge of the

individuals was sufficient for their assigned tasks.

Specific areas reviewed are identified in Paragraphs 2, 3, and 4. The

adequacy of training to prevent personnel from over torquing limitorque

valve operators was identified as a concern. Personnel have not been

provided with specific instructions that would limit the amount of force

applied to valve operators. This contributed to the failure of a valve

operator as noted in Paragraph 4. This concern was identified to the

licensee for evaluation.

6. Unit Trips (93702)

On May 9,1985 at 9:10 a.m. CDT, LaSalle Unit 2 experienced a reactor scram

from 85% reactor power. The unit scrammed on low reactor water level due

to the loss of power to the feedwater control system. With a loss of the

feedwater control system, the "B" Turbine Driven Reactor Feedwater (TDRFP)

Pump " locked up" and maintained a constant speed. The "2A" TDRFP did not

lock up and coasted down. The motor driven feedwater pump was started but

its feedwater control valve locked up at 20% open. With a decrease in

feedwater flow, vessel level decreased to the low reactor water level scram

setpoint and the unit scrammed. The loss of power to the feedwater control

system was caused by a worker accidentally bumping and tripping a 120V

power supply disconnect breaker to the feedwater control system. The

licensee determined the lock up of the flow control valve and the coast

down of the ATOFWP was due to the loss of power to the feedwater

controller.

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The inspector attended a meeting with licensee personnel on the scram and

a discussion was held on possible delay in receiving a low level scram when

it should have occurred. Discussions with the shift engineer indicated

that he was watching the wide range recorder and saw.the recorder reading

eight inches. However, based on previous experience with the indication on

the recorder, a six inch difference between indicated recorder level and

actual level was knovn to exist. The reason for the difference is that

these recorders are calibrated with no flow in the recirculation system.

Flow in the recirculation system causes the recorder to read low.

The licensee then checked the calibration on the level switches and the

control room narrow range indicator in the control room. No problems were

identified. Due to a similar problem with these level switches during a

feedwater transient on June 1, 1986, this event will be reevaluated and

docurented in a special report (374/86023).

On May 11, 1986-at 4.:42 a.m. while returning Unit 2 to power from the May 9

scram, the unit received a Group I isolation on indicated high main steam

line flow and a reactor scram. The Group I isolation was received while

synchronizing the main turbine generator to the grid and increasing load.

All systems functioned as expected. The licensee investigated the cause

of the high main steam line flow signal and was unable to determine the

cause. A walkdown of all steam lines indicated no problem. Testing of

the EHC system indicated no problem. A review was made of the electrical

systems and no cause could be determined. The unit was returned to power l

on May 24, 1986 after a short maintenance outage. The licensee replaced

the seals on the "B" recirculation pump and worked on leaking valve

packing seals. The licensee connected recorders to the high steam flow

sensors and switches during the startup to see if they could determine

the cause of the May 11 scram. The inspector was present in the control

room during the turbine roll and synchronizing of the generator to the

grid. No problems were identified at that time. See Paragraph 3 for

events occurring later on in the startup.

7. Refueling / Outage (61701)

During a May 21, 1986, Unit 1 cold hydrostatic test of the primary piping,

two small leaks were found in a weld on a plugged hole which had been

welded up during construction. The leak rate was less than a drop a

minute. The leaks were on the B and C low pressure injection line between

the reactor vessel and the inboard manual isolation valve, and as such,

were unisolatable. The licensee has determined the corrective action for

repair is to weld a one inch pipe coupling over the existing welded plug.

This will be accomplished prior to starting the unit up.

On May 30 during the performance of LTS 300-4 (Integrated Leak Rate Test)

for Unit 1, a Primary Containment Isolation System (PCIS) Group 2 and

Group 4 isolation for Unit 1 and Group 4 isolation for Unit 2 occurred.

The cause was a high drywell pressure isolation signal due to a momentary

open circuit during the installation of a jumper. Upon receiving the PCIS

initiation, the reactor building ventilation isolated and both trains of

standby gas treatment system started.

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The loss of continuity occurred while a station electrician was installing

a jumper per procedure LTS 300-4. Per the technical staff engineer's

direction, a spade lug jumper was used. As the jumper was being installed,

the wire lugs already on the terminal block apparently lost continuity,

despite care being taken to maintain continuity. Since the isolation logic

is normally energized, the isolation occurred upon loss of the continuity.

'Upon receiving the isolation, the spade jumper was removed and all systems

reset and returned to normal. The root cause of the isolation stems from

an inadequate procedure and a personnel error. A meeting was held to

determine and implement a feasible method for installing jumpers without

disrupting the existing circuitry. The station electrician attended the

meeting and was given instructions on how to eliminate future problems.

Also, procedure LTS 300-4 has been changed to incorporate a note as to the

consequences of the loss of continuity of the circuit. Due to the fact

that there have been several other occurrences of jumpers falling off

and/or causing isolations/actuations, the inspector will monitor the

actions of the licensee with respect to installation of jumpers and lifted

leads.

8. Regional Requests (92705)

The inspector completed a request for assistance on review of the low level

radioactive waste storage facility at LaSalle. The request was from

C. E. Norelius dated May 16, 1986. The results are as follows:

a. Is the licensee building or planning to build an onsite low-level

waste storage facility? When?

Yes, the facility is scheduled to be completed by August 1986.

b. What is the general method of construction e.g. reinforced concrete,

concrete block, butler building, etc.?

The facility is constructed of reinforced concrete,

c. What is the proposed capacity of the facility in cubic feet of

waste, square feet of floor space, and/or maximum curie content?

The facility has a floor space of 5,700 ft2 and can store

153,900 fts of waste. The facility is designed with the assumption

that each barrel has a reading of 5 R/hr which is more conservative

than the average activity distribution using curie analysis,

d. Is the structure to be free standing or will it be attached to some

existing plant structure? If it is to be attached to an existing

plant structure, what is that structure?

The facility is free standing and not attached to any existing

facility.

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e. Has the licensee performed a 50.59 evaluation of the proposed

structure? If such an evaluation has been performed, what were the

conclusions of the evaluation?

Yes, the licensee performed a 50.59 review by the Station Nuclear

Engineering Department (SNED). The review is part of the modification

package. No new unresolved safety issues were generated nor were

Technical Specifications required to be changed.

f. Has the licensee estimated the contribution to off-site dose rate

from the facility? If so, what is the estimated dose rate and what

regulatory limits were used to judge the acceptability of this

estimated does rate?

The licensee has determined that the offsite dose rate from addition

of this facility has not increased above the original design of the

entire facility. The regulation used was 40 CFR 190 and EPA

regulations. This information was obtained by telecon between the

inspector and a SNED representative.

g. Will the facility house low-level waste processing equipment or will

it simply be a repository for waste?

The facility is to house low-level waste only.

Inspection of this facility was accomplished by regional based inspectors

in September 1985 and is documented in Inspection Reports No. 373/85030;

374/85031.

9. Open Items

Open items are matters which have been discussed with the licensee, which

will be reviewed further by the inspector, and which involve some action

on the part of the NRC or licensee or both. Open items disclosed during

the inspection are discussed in Paragraphs 2 and 4.

10. Exit Interview (30703)

The inspectors met with licensee representatives (denoted in Paragraph 1)

throughout the month and at the conclusion of the inspection period and

summarized the scope and findings of the inspection activities. The

licensee acknowledged these findings. The inspector also discussed the

likely informational content of the inspection report with regard to

documents or processes reviewed by the inspector during the inspection.

The licensee did not identify any such documents or processes as

proprietary.

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