ML20199F529

Insp Repts 50-373/97-20 & 50-374/97-20 on 971101-1218. Violations Noted.Major Areas Inspected:Operations,Maint, Engineering & Plant Support

ML20199F529
Person / Time
Site:	LaSalle
Issue date:	01/23/1998
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20199F484	List: IR 07100101/2012018 ML20199F480 ML20199F529
References
50-373-97-20, 50-374-97-20, NUDOCS 9802040033
Download: ML20199F529 (14)

See also: IR 05000373/1997020

Text

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U.S. NUCLEAR REGULATORY COMMISSION

REGIONlli

Docket Nos.: 50 373, 50 374

License Nos.: NPF 11, NPF 18

l Report No: 50 373/97020(DRP); 50 374/97020(DRP)

Licensee: Commonwealth Edison Company

Facility: LaSalle County Station, Units 1 and 2

Location: 2601 N. 21st Road

Marseilles, IL 61341 -

Dates: November 1 December 18,1997

Inspectors: M. Huber, Senior Resident inspector

J. Hansen, Resident inspector

R. Crane, Resident inspector

Approved by: Kenneth G. O'Brien, Acting Chief

Reactor Projects Branch 2

9902040033 900123

PDR ADOCK 05000373

G PDR

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EXECUTIVE SUMMARY

LaSalle County Station, Units 1 and 2

NRC Inspection Report No. 50 373/g7020(DRP); 50-374/g7020(DRP)

This inspection i sort included aspects of licensee operations, maintenance, engineering and

plant support. Ti. toport covers a six week period of Inspection conducted by the resident staff,

Operatigni

+ The inspectors identified a violation involving the failure of operations personnel to ensure

that all operability evaluations remained on file in the control room. The unavailability of

one operability evaluation resulted in an unnecessary entry into a Technical Specification

limiting condition for operation action statement. Engirieering personnel were removing

the evaluations in anticipation of a procedure change that had not yet been implemented.

Engineering personnel demonstrated a lack of questioning attitude regarding the

operationalimpact of removing the operability evaluations. (Section O3.1)

Maintenance

. The licensee decontaminated reactor water cleanup system work areas which increased

the licensee's ability to perform maintenance more efficiently. (Section M2.1)

+ The inspectors identified two violations which resulted in the improper installation of a test

velve assembly on an emergency diesel generator. A work planner failed to follow

administrative maintenance process requirements to ensure the associated work package

incorporated information from the vendor manual. In addition, the licensee had not taken

adequate corrective actions following previous similar problems. (Section M3.1)

E09109EiO2

+ The licensee's identification of testing deficiencies involving the emergency diesel

generators was positive and reflected an increased licensee focus on surveillance test

procedure adequacy. These issues were determined to be non-cited violations.

(Sections E1.1 and E8.2)

Elsnt Suppori

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+ The inspectors identified a violation involving the failure to properly post a contaminated

area. The lack of a questioning attitude by a radiation protection technician and by two

non licensed equipment operators resulted in contaminated liquids being vented into a

plant drain trough which was not identified as contaminated, (Section R1.1)

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Resort Datants

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Summary of Plant Stahm

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During this inspection pedod, the licensee maintained Unit 1 in cold shutdown (Operational

Condition 4) for a forced outage and Unit 2 remained shut down for a refueling outage with all

fuel removed from the reactor.  !

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j. l. Deerations j

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03- Operations Procedures and Deoumontation

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. 03.1 Failure to Adecuately Control Operabilltv Evaluations 1

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a. inspection Scope (71707)

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I' The inspectors reviewed a sample of the operability evaluations applicable to current

plant conditions, Lasslie Administrative Procedure (LAP) 220 5, Equipment Operability .

Determination," Revision 5, and the unit logs. The inspectors also interviewed operations  !

and engineering personnel.

b. Observation and Findinas

On November 19,1997, a control room supervisor (CRS) unnecessarily entered .

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Technical Specification (TS) limiting condition for operation (LCO) action

Statement 3.3.7.3 a., following the identifloation of an upscale meteorological differential  ;

temperature indication. The CRS was aware of an operability evaluation, No. 97016, that  ;

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had been completed to support operability determination of the meteorological differential j

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temperaturs instrument during certain atmospheric conditions. - Guidance and analysis in

the operability evaluation would have prevented the need to enter the LCO action -

statement. However, the CRS could not locate the operability evaluation in the control

room file.  :

On December 4,1997, the inspectors discussed the LCO entry with the CRS. Following

further review, the inspectors identified a broader conoom regarding the control of

operability evaluations. The inspectors determined that engineering personnel had

removed operability evaluation No.- 97106 from the control room because they anticipated ,

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the approval of a procedure change which would have required storage of operability

evaluations by engineering personnel outside of the control room. However, the

inspectors determined that the procedure change had not been approved. Also,

engineering personnel did not question the operational impact of removing the operability

evaluations from the control room or verify that the procedure had not yet been

implemented.; Step F.2.n.10 of LAP-220 5 required that operations management file the

- completed operability evaluations in the control room.- The failure of operations

management to maintain on file the completed operability evaluations in the control room

as required by LAP 220 5 is an example of a violation of 10 CFR Part 50, Appendix B, ,

- Criterion V, as described in the attached Notice of Violation (Notice) (50 373/97020-01a;-

50 374/97020-01a).

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On December 8,1997, the operations manager reiterated the procedural requirements by

informing the shift managers to maintain the operability evaluations in the control room.

Also, as a corrective action, operations and engineering personnel performed audits to

determine if any other operability evaluations were missing from the control room files.

They identified that 15 operability evaluations were not being maintained in the control

room. Enseering personnel subsequently filed these operability evaluations in the

control rot. - Operations personnel did not identify any additional instances of

unnecessary TS LCO action statement entries. The inspectors determined the corrective

actions to be appropriate,

c. Conclus19m

Operations personnel did not ensure operability evaluations were proper 1y controlled to

support plant operations. Also, engineering personnel demonstrated a lack of

questioning attitude regarding the operationalimpact of removal of the operability

evaluations from the control room.

II. Maintenance

M2 Maintenance and Material Condition of Facilities and Equipment

M2.1 General Comments

a. Inspection Scope f62707)

The inspectors observed portions of the maintenance associated with the reactor water

cleanup (RWCU) system and reviewed the associated work request (WR),

No. 970052068, Task 7, " Demolition RT (Reactor Water Cleanup System)

Suction / Discharge Piping and Supports.*

b. Observations and Findinas

The licensee was modifying the RWCU system to address material condition deficiencies

in the system including fluid flashing / voiding at the RWCU heat exchangers and pump

vibrations. Maintenance personnel were using the work proceduren at the work site.

Radiation protection personnel were observed by the inspectors at the job site to support

the work. Once the workers completed demolition of the system, the work areas were

decontaminated to allow workers access to the work areas without the constraints of

anti-contamination clothing,

c. Qonclusions

The licensee decontaminated the RWCU work areas which increased the licensee's

ability to perform maintenance more efficiently. No problems were identified by the

inspectors.

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MS Maintenause Procedures and Documentation

M3.1 Deficiencies Related To Emernanov Diesel Generator fEDG) Test Valve Assembiv Failure

a. inaceallon Soone (62707)

The inspectors evaluated the licensee's response to a loose test valve casombly on the

0 EDG and interviewed maintenance and engineering personnel, in addition, the

inspootors observed maintenance of the test valve assernblies and reviewed

documentation and procedures which included:

.- WR 970119536, " Cylinder 15 Test Valve Unattached to Diesel Cylinder"

. Engineering Request (ER) 9707312. " Evaluate Extent of Condition for Test Valve

Cylinder Threads"

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. LAP 13001, " Action / Work Request Processing," Revision 67

. LaSalle Mechanical Maintenance Surveillance (LMS) DG-01, " Main Emergency

Diesel Unit Surveillances," Revision 18

. Maintenance Memorandum 200-02, " Instructions to Work Package Preparers

(Wott Analysts)," Revision 11

. LAP 10015 " Vendor Information Technical information Program," Revision 7

b. Observations and Findinns

on November 11,1997, an equipment operator found that the test valve adapter for .

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cylinder 15 on the EDG was loose and could be removed from the cylinder by hand. The

operator was performing pre start checks of the 0 EDG as directed by LaSalle Operating

Surveillance (LOS) DG-M1, " Diesel Generator Operability Test," Revision 35. The

licensee initiated a problem identification form (PlF) and immediately vertfled the test

valves in the four remaining E00s were tight. The licensee investigated the cause of the

loose valve and determined that the threads in the cylinder that held the test valve

adapter failed and that the test valve adapter could not be threaded back into the

housing.

On November 12,1997, a maintenance work analyst developed work instructions and

maintenance personnel reinstalled the test valve assembly. Engineering personnel

completed en operability evaluation and concluded that the loose test valve was an

isolated case and did not impact the operability of the other EDGs. On November 13,

1997, operators declared the 0 EDG operable following successful completion of an

operability surveillance test.

After the licensee declared the o EDG operable, engineering personnelidentified that

Stop E.29.21 in LMS-DG-01, a maintenance procedure used by mechanics to reinstall the

test valve assemblies following periodic maintenance performed prior to the failure -

identified on November 11,1997, was not correct for the valve assemblies currently

installed on the EDGs. Specifically, the procedure was to be used for the previous type of

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valve that was installed and appropriate valve adapter torque requirements were not

prescribed for the currently installed valves. As the operability of all EDGs was again in

question after the repairs made on the 0 EDO oylinder 15 test valve, plant personnel

identified contingent actions and test valve tortuing equipmerd was staged by the *

appropriate personnel while the operability evauustion was completed.

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As a result of the licensee's determination that the torque values in LMS DG-01 were not

applicable to the current valve assemblies, the inspector reviewed the work request,  ;

- WR 970119536, and identified that the test valve for cylinder 15 on the 0 EDG was  !

installed using superoseded vendor information. The work analyst used the original test i

valve installation work procedure from 1995, whloh was different from LMS DG-01, to

generate the WR. The WR required the test volve adapter be tightened until snug with

the valve knob pointing down. However, in August 1997, the licensee's EDG vendor

equipment technical information program (VETIP) manual had been revised to include a

vendor presortbed torque requirement for the test valve adapters of 45 to 50 foot pounds.

The l6censee's process for developing work procedures, defined in LAP 13001

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Step B.2.5.1, required the work analyst to incorporate work requirements and restrictions

as discussed in Maintenance Memorandum 200-02. Step F.1.d.4 of the maintenance

memorandum required the work analyst to apply applicable information contained in tha

VETIP manual when developing work Instructions. However, the inspectors determined >

that the work analyst did not reference the manual due to familiarity with the test valve  :

assembly and a perceived time constraint to retum the EDG to operable status. The .

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work analyst's failure to reference appropriate documents and incorporate work

requirements discussed in Maintenance Memorandum 200-02 when developing

WR 970119536 for installing a test valve on the O EDG, spoolfically torque requirements

from the VETIP manual, is an example of a violation of 10 CFR Part 50, Appendix B, -

Criterion V, as described in the attached Notice (50 373/97020-01b; 50 374/97020 01b).

Subsequently, engineering and operations personnel again determined the EDGs to be

operable. The inspectors reviewed the operability esaluation and concluded that the i

operability determination was appropriate.

- Engineering personnel determined, in both the initial oporability evaluation and the

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supporting engineering evaluation, that the loese test valve assembly was an isolated

case and that the repaired test valve assembly had been correctly installed. However,

the inspectors identified that a similar event had occurred at Dresden Station in .

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November of 1996 which resulted in Nuclear operations Notice .

- (NON DR [Dresden} 12 9618) being issued. Also, on December 12,1996, at LaSalle

County Station, operators had identified ten loose cylinder test valve assemblies on three

EDGs. The operators initiated Problem identification Form 96 5228 in which the shift

manager indicated that the system engineer believed that torque values for the test valve

assemblies needed to be assigned due to the past history of becoming loose.

Maintenance personnel tightened the test valve assemblies but no specific torque values

- were applied, Maintenance personnel noted on the work request that actions should be

taken to keep the test valve assemblies from working loose. Neither engineering nor

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maintenance personnel performed any follow-up action to either previous occurrence.

The failure of the licensee to identify and implement adequate corrective actions for loose

test valve assemblies on the EDGs is a violation of 10 CFR Part 50, Appendix B,

Criterion XVI, as described in the attached Notice of Violation (50 373/97020-02;

60 374/97020 02).

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During the licensee's root cause investigation of the test valve failure, station personnel

determined LAP.10015 did not provide clear guidance regarding procedure revisions

which resulted in the maintenance procedure LMS DG-01 not being updated following .

revisions to the VETIP manuals. Engineering personnalinitiated changes to L.AP-10015 ,

to ensure appropriate procedures would be revised following vendor manual revisions, in ,

addition, the inspectors identified that a formal mechanism to ensure vendor information . t

specifically provided to another Commonwealth Edison facility would be reviewed for ,

applicabilNy to other sHes did not exist at the LaSalle Station. However, the Quad Chios  :

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Station EDG system engineer had informally communloated vendor information regarding

the test valve assemblies to the LaSalle system engineer during a poor group meeting in  ;

the spring of 1997. The test valve assembly vendor had provided a letter pertaining to the

torquing of the test valve assemblies to Quad Cities Station EDG system engineer in

January 1997.

c. Conclusion

While the safety significance of the loose test valve on the O EDG was minimal, the

licensee's failure to address the past valve failures resulted in the loose test valve

assembly on the o EDG and the inadequate work request to repair the valve. The work

analyst's failure to follow procedures was representative of previous human performance

problems at the LaSalle Station. However, the licensee has identified the corrective

action program and human performance deficiencies at LaSalle and continuws to address

the problems.

MS Miscellaneous Maintenance issues

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M8.3 (Closed) Licensee Event Report (LER) 50-373/97034 00: Missed TS Surveillance to

Verify Auto connected Loads on the EDGs Do Not Exceed Rating Due to

Misinterpretation of TS

The inspectors reviewed the issue described in LER 50-373/97034-00 and discussed the

results of the review in NRC Inspection Report 50 373/97016; 50 374/97016. In the

inspection report (50-373/97016; 50 374/97016), the NRC granted enforcement discretion

under Section Vll.B.2., " Violations identified Dudng Extended shutdowns or Work

- Stoppages," of the NRC Enforcement Policy (NUREG 1600) for a violation related to this

issue. The inspectors did not identify any new issues in the LER.

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111. Ennineerina

E1 Conduct of Engineering

E1.1 Inadeaunte Division til Hioh Pressure Core Sorav (HPCS) Testino

a. inspection Scope (92903. 37551)

The inspectors reviewed the licensee's actions to addrest, testing of the HPCS EDG that

was not performed in accordance with the TSs. The inspectors reviewed LaSalle

Technical Surveillance (LTS)-800109, Revision 6, "1B Diesel Generator Trips and Trip

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Bypass Logic Test," schematic diagrams of the HPCS EDG protective relays, and

discussed the issue with engineering personnel,

b. Observations and Findinas

On November 25,1997, the licensee identified that the surveillance procedure used to

demonstrate that the Unit 1 and Unit 2 HPCS EDGs could supply the normal Division lil

bus loads following a simulated loss of the offsite power from the station auxiliary

transformer (SAT) was inadequate. Technical Specification Surveillance

Requirement 4.8.1.1.2.d.11.b required the licensee to verify that 6 simulated trip of the

EDG overcurrent relays tripped the SAT feed breaker to the Division ill bus and the EDGs

continued to supply the normal bus loads. The TS required that the EDG be operating in

a test mode and connected to its bus. However, the test procedure was not performed

with the EDGs operating, nor did the procedure verify that the overcurrent relays would

operate as designed.

Two sets of overcurrent relays are in the HPCS EDG circuitry. One set of overcurrent

relays, the K35 relays, trip the SAT supply breaker to the Division lll bus when a fault is

sensed. Following an additional one half second delay, the EDG also trips. However, the

EDG will not trip if the fault clears when the SAT breaker opens. The second set of

overcurrent relays, the K33 relays, trip the SAT breaker, providing protection to the EDG

from high current fault conditions from equipment on the Division ill bus. While the

licensee identified inadequate testing of the K35 overcurrent relays, the inspectors

subsequently identified that the K33 overcurrent relays also were not tested. However,

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the licensee had not yet completed corrective actions at the time the inspectors identified

this related concem. The licensee initiated a PlF to ensure the test procedure would be

revised to provide proper testing. The HPCS systems were inoperable for other reasons

at the time this issue was identified.

The licensee's failure to perform the testing with the EDGs operating and to verify

operation of the overcurrent relays is a violation of TS Surveillance

Requirement 4.8.1.1.2.d.11.b (50-373/97020-03; 50 374/97020-03). This non-repetitive,

licensee-identified and corrected violation is being treated as a Non-Cited Violation,

consistent with Section Vll.B.1 of the NRC Enforcement Policy.

c. Conclusions

The licensee's identification of the testing deficiency was positive. Severallicensee event

reports involving licensee identification of other inadequate test procedures were issued

in 1997. This additional example further confirmed an increased licensee focus on test

procedure adequacy.

E8 Miscellaneous Engineering issues

E8.1 Incorrect Drawina in Updated Final Safety Analysis Report (UFSAR)

a. lnJpection

n Scope (37301)

While observing maintenance related to the RWCU system, the inspectors reviewed the

LaSalle UFSAR, Amendment 11, Chapter 5.4.8, " Reactor Water Cleanup System."

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b. Observations and Findinas

The inspectors identified that RWCU drawings contained in the UFRAR differed from the

actual plant configuration. The UFSAR drawings shewed that the RWCU pump suction

was directly from the recirculation system, upstream of the RWCU system heat

exchangers. The actual plant configuration was with the pump suction downstream of

the system heat exchangers. The RWCU UFSAR draAgs had been revised by the

licensee in the last revision (Amendment 11) of the UFSAR. The previous amendment

contained RWCU drawings that correctly depicted the relationship between the RWCU

pumps and heat exchangers in the plant.

After the inspectors identified the discrepancy, the licensee entered the discrepancy into

the corrective action program by initiating a PIF. However, the initial evaluation of the PIF

by the Event Screening Committeo (ESC) did not result in any action to determine the

cause of the incorrect drawing being submitted in Amendment 11 of the UFSAR. The

inspectors discussed the ESC actions with the corrective action program manager who

then had the ESC review the PlF. Action items were subsequently assigned by the ESC

to determine the cause of RWCU UFSAR drawing aiscrepancy and to provide further

corrective actions as necessary. In addition, the licensee had previously identified the

potential for incorrect UFSAR information to have been used in performing 10 CFR 50.59

safety evaluations and had assigned a corrective action program item to evaluate the

UFSAR change process. No other cases of incorrect drawings being used to revise the

UFSAR were identified by the inspectors. This is considered an inspection Follow-up

Item (50 373/97020-04; 50-374/97020-04) pending further review of the root cause sind

corrective actions by the inspectars.

c. Conclusions

The UFSAR was incorrectly updated with RWCU drawings which were not representative

of the actual LaSalle plant configuration and the licensee was investigating the cause of

drawing discrepancy. The ESC did not initially require actions to determine the cause of

the RWCU drawing discrepancy in the UFSAR which demonstrated a lack of a

questioning attitude by the committee formed for oversight of the corrective action

screening program.

E8.2 fClosed) LER 50-373/97037-00: Diesel Generator Response Time Test Not Consistent

with TS Due to inadequate Review of Procedures.

On November 5,1997, the licensee determined that the acceptance criteria specified in

plant surveillance procedures for the 0,1 A and 2A EDG response time test was not in

compliance with TS. Specifically, the load sequence timer acceptance criteria for the A

and B residual heat removal (RHR) pumps in LTS-500-1(2)09, " Unit 1(2) Integrated

Division i ECCS [ Emergency Core Cooling System) Response Time Surveillance,"

Revision 5(6), and LTS 500-1(2)10, " Unit 1(2) Integrated Division 11 Response Time

Surveillance," Revision 5(5), was 4 to 5 seconds. However, TS Surveillance

Requirement 4.8.1.1.2.d.12 required verification that the automatic load sequence timer

interval was within +/ 10 percent of its design Interval for the 0,1 A, and 2A EDGs. Since

the design interval specified in Section 7.3.1/2.4.3 and Table 8.3.1 of the UFSAR was

5.0 seconds, the minimum intervalin the surveillance should have been 4.5 seconds.

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Following identification of the improper acceptance critoria, ths licensee declared the

O EDG inoperable because the previous load sequence timer interval surveillance test

performed on March 15,1996, was 4.2 seconds, below the TS minimum of 4.5 seconds.

The licensee reviewed (Fe previous two years of load sequence timer interval

surveillances for the 0,1 A, and 2A EDGs and identified two additional instances of the

time interval being outside of technical spec!fication requirements. The licensee's failure

to meet EDG load response time testing within +/ 10% of 5 seconds is a violation of

TS 4.6.1.1.2.d.12. (50 373/97020-05; 50 374/97020-05). This non-repetitive, licensee-

identified and corrected violation is being treated as a Non Cited Violation, consistent with

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Section Vll.B.1 of the NRC Enforcement Policy.

Pr'or to declaring the 0 and 1 A EDG's operable, the licensee recalibrated the load-

sequence timer interval to within the TS requirements. !n addition, the licensee planned

to revise the surveillance procedures with the correct RHR pump time delay interval. The

licensee also evaluated the impact of the time delay interval for the A and B RHR pumps

being us low as 4 soconds. Previous response time testing performed on the EDGs

indicated that voltage recovered to within 10 percent of nominalin approximately 2.0

seconds following initialloading of the EDG buses. Since the sequence timer did not

load the EDG with the A and B RHR pumps until a minimum of 4 seconds, adequate bas

voltage was assured prior to RHR pump start and therefore loading the RHR pumps

would not impact EDG operability.

IV. Plant Support

R1 Radiological Protection and Chemistry Controls

R1.1 Unidentified Contaminated Area in the Reactor Buildina

a. Inspection Scope (717EO. 83750)

On November 30,1997, the inspectors observed the radiological work practices of

operations personnel performing checklist items associated with out-of service (OOS)

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No. 970004551. This work involved venting of the control rod drive (CRD) charging

header.

b. Observations and Findinas

On November 30,1997, the inspectors observed operations personnel venting the CRD

charging line to a drain trough, that was not marked as radiologically contaminated,

without contacting radiation protection (RP) personnel either prior to or following the

evolution. Operations management informed the inspectors that their expectations were

that the operations personnel performing the venting should have identified that the drain

trough was not posted as contaminated and should have informed RP personnel on shift.

The inspectors contacted the shift RP supervisor who ordered a contamination survey of

the drain through and found the accessible intcmal surfaces contaminated at

10,000 dpm/100 cm' beta gamma. Radiation protection technicians then posted the area

as contaminated.

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Radiation protection management infomied the inspectors that the drain trough was

previously posted as contaminated. However, when the surrounding area was

4 decontaminated, the area was surveyed as not radiologically contaminated and the

posting had been removed. The technician removing the posting did not consider that

operation of the contaminated CRD system vents and drains could potentially

contaminate the draln through.

LaSalle Radiation Protection Procedure (LRP) 50101, Revision 5, required areas with

smearable contamination levels at levels greater than or equal to 1000 dpm/100cm' to be

posted as a contaminated area. The licenroe's failure to post the drain tsough as a

contaminated area is a violation of TS 6.2.0, as described in the attached Notice

(50 373/97020-06; 50-374/97020-06).

c. Conclusions

Radiological protection personnelinappropriately removed postings from a drain trough

which received input from the contaminated CRD charging water vent line. Operations

personnel did not meet their management's expectation by ialling to notify RP personnel

when venting the contaminated system to the drain trough which was not marked as

radiologically contaminated.

VI. Mananoment Meetinos

X1 Exit Meeting Summary

The inspectors presented the results of these inspections to licensee management listed

below at an exit meeting on December 18,1997. The licensee acknowledged the

findings presented. The inspectors asked the licensee if any materials examined during

the inspection should be considered proprietary. The licensee identified none.

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PARTIAL LIST OF PERSONS CONTACTED

Comed

'W Subalusky, Site Vice President

• F. Decimo Plant General Manager

• S. Smith, Plant Manager

• W. Riffer, Site Quality Verification / Safety Assessment Manager

G. Holstorman, Maintenance Manager i

• J. Bailey, Restart Manager  ;

R. Palmieri, System Engineering Supervisor

N. Hightower, Health Physics Supervisor

P. Bames, Regulatory Assurance Supervisor

G. Poletto, Engineering Manager

• Present at exit meeting on December 18,1997.

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INSPECTION PROCEDURES USED

IP 37551 Onsite Engineering

IP 62707 Maintenance Observation

IP 71707 Plant Operations

IP 71750 Plant Support Activities

IP 83750 Occupational Radiation Exposure

IP 92903 Followup Engineering

ITEMS OPENED, CLOSED, AND DISCUGSED

QDAD

50 373/374 97020-01a VIO Operations failed to maintain operability evaluations on file

in the control room

50-373/374 97020-01b VIO Work analyst failed to incorporate VETIP information in a

work procedure

50-373/374 97020-02 VIO Failure to take adequate correctives to preclude the

recurrence of loose test valves

50-373/374 97020-03 NCV Inadequate HPCS EDG surveillance test

50-373/374 97020-04 IFl incorre :t UFSAR drawing

50-373/374-97020-05 NCV inadequate EDG surveillance test

50 373/374 97020-06 VIO Failure to post contaminated drain trough

Discussed or Closed

50-373/97034-00 LER Missed TS surveillance to verify auto-connected loads on

the EDGs do not exceed rating due to misinterpretation of

technical specifica; ions

50 3/3/97037-00 LER Diesel generator response time test not consistent with TS

due to inadequate review of procedures

50-373/374 97020-03 NCV Inadequate HPCS EDG surveillance test

50-373/374 97020-05 NCV inadequate EDG surveillance test

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> LtST OF ACRONYMS USED

CRD Control Rod Drive

CRS Control Room Supervisor '

DRP Division of Reactor Projects

ECCS Emergency Core Cooling System

EDG Emergency Diesel Generator 1

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-ER Engineering Request '

EHC Events Screening Committee

HSCS H6gh Pressure Core Spray

IR inspection Report

IFl - Inspection Follow up item >

LAP LaSalle Administrative Procedure

LCO Limiting Condition for Operation .

LER Licensee Event Report

t.MP LaSalle Maintenance Procedure

LMS LaSalle Mechanical Maintenance Surveillance

. LOP LaSalle Operating Procedure

LRP- LaSalle Radiation Protection Procedure

LTS LaSalle Technical Surveillance

NON . Nuclear Operations Notice

NRC Nuclear Regulatory Commission

NTS Nuclear Tracking System

008 Out Of Service

PIF Problem identification Form

PDR NRC Public Document Room

RHR Residual Heat Removal

RP Radiation Protection

RT Reactor Water Cleanup System

RWCU Reactor Water Cleanup

SAT Station Auxillery Transformer

TS Technical Specification

URI Unresolved item

UFSAR Updated Final Safety Analysis Report

VETIP. Vendor Equipment Technical Information Program

WR Work Request

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