Text

Insp Rept 50-298/96-31 on 961201-970219.Violations Noted. Major Areas Inspected:Operations,Maint,Engineering & Plant Support
	ML20135D999
Person / Time
Site:	Cooper
Issue date:	02/28/1997
From:	; NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION IV)
To:	;
Shared Package
ML20135D994	List: IR 05000298/1996031; ML20135D993; ML20135D995;
References
	50-298-96-31, NUDOCS 9703060190
	Download: ML20135D999 (31)
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ENCLOSURE 2 U.S. NUCLEAR REGULATORY COMMISSION

REGION IV

Docket No.: 50-298 License No.: DPR-46 Report No.: 50-298/96-31 Licensee: Nebraska Public Power District -

Facility: Cooper Nuciaar Station Location: P.O. Box 98 Brownville, Nebraska Dates: December 1,1996, through February 19,1997 Inspectors: Mary Miller, Senior Resident inspector Chris Skinner, Resident inspector Haymond Azua, Reactor inspector, . Technical Support Branch Gail Good, Reactor inspector, Plant Support Branch Dave Wiggington, Project Manager, NRR Approved By: Larry Yandell, Acting Chief, Project Branch C Division of Reactor Projects 1 ATTACHMENT: Supplementalinformation 9703060190 970228 PDR ADOCK 05000298 G PDR ,

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i EXECUTIVE SUMMARY COOPER NUCLEAR STATION NRC Inspection Report 50-298/96-31 Operations

The Technical Specifications (TS) indicated unidentified leakage rate was exceeded as a result of postulated intermittent water slug drainage from the drywell unit cooler drain trays (Section 01.1).

Due to operations' insistence, engineering provided a detailed evetuation on the likely cause of the intermittent water slugs of unidentified leakage (Section 01.1).

The shift supervisors failed to enter the TS action statement twice, based on an engineering evaluation, after the plant engineering manager had assured the inspectors that the evaluation would not be used as a basic; for operators not entering the TS action statement. Plant management's subsequent emphasis on the need to enter TS action statements appeared to have addressed the concern (Section 01.1).

Inspectors identified that the interim guidance to operators was weak and incomplete regarding operations' compensatory measures to address both drywell sump problems (identified and unidentified leakage sumps). After inspectors questioned the guidance given to operators, uncontrolled written instructions to pump the identified leakage sump at 2-hour intervals were provided to the control room (Section O3.1).

Inspectors identified a weakness in that, upon increased drywell temperature, radiation, humidity, or pressure, compensatory actions did not require operators to pump the equipment drain sump to determine if equipment drain leakage had increased (Section 03.1).

The licensee performed a critical assessment of the overall quality of past operability evaluations anJ identified numerous weaknesse: (Section 07.1).

Quality Assurance identified additional weaknesses in the licensee's response to the l flow bias scram reset prob!em (Section 07.2). l

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Inspectors identified a violation in that operations failed to provide procedures which j describe operator actions to mitigate slush buildup in circulating water bays caused by entrainment from river ice. Some corrective actions were compreherd.e, but !

procedure changes were not completed after 2 months and interim instructions to f operators were documented only after questioning by inspectors (Section 08.1). ]

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Maintenance >

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The inspectors identified a violation with regard to the failure by maintenance i personnel to write a Problem Identification Report (PIR) for a known valve leakage problem (Section M1.1).

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The inspectors identified a weakness in the maintenance technicians' performance when they failed to question why leakage through a new valve was worse than'the !

(leaking) replaced valve (Section M1.1). )

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Inspectors identified that licensee staff was unaware of the requirement to perform evaluations in accordance with plant procedures for changes to tolerances, scaling, calibration processes, and calibration data sheets for instrumentation without automatic setpoints (Section M1.2).

Enoineerina

Inspectors identified that the licensee has been slow to resolve the safety significance and potential generic scope of a reactor protection system vulnerability which occurred on November 22,1996,in which a failure of a voltage input for an average power range monitor (APRM)_ flow bias unit resulted in a nonconservative reset of the flow bias scram setpoint for that division (Section E1.1).

The inspectors identified an apparent violation in that no evaluation in accordance with 10 CFR 50.59 was performed when an ice deflector was not installed as described in the Updated Safety Analysis Report (USAR) (Section E2.1).

The inspectors concluded, in agreement with a licensee assessment, that prior to 1995 the licensee's 10 CFR 50.59 evaluations were poor. The licensee has implemented corrective actions that improved performance to a generally good level. However, the screens and 10 CFR 50.59 evaluations warrant continued observation due to the recent implementation of improvements and training and changes in ownership of the overall program (Section E2.2).

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-Inspectors identified an apparent violation in that the licensee failed to perform an i evaluation in accordance with 10 CFR 50.59 to address the function performed by

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the fill rate timer as described in the USAR (Section E2.3). ;

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A 10 CFR 50.59 evaluation to address operation of the plant with failed high level )

switches on the identified leakage sump was not approved until 15 days after the :

sump level switches f ailed (Section E2.3), j l

The inspectors identified an apparent violation in that information concerning l primary containment isolation valves was missing from the latest USAR update (Section E2.4).

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The inspectors reviewed a licensee Quality Assurance audit of the design control program and determined that the audit concluded that the USAR had not been properly maintained and that systematic correction of the condition was necessary (Section E2.5).

Plant Support

The inspectors identified an apparent violation in that combustible materials were observed unattended in the service water pump room, which was in conflict with the USAR, and the licensee did not perform an evaluation of this inconsistency in accordance with 10 CFR 50.59 (Section M1.3).

The inspectors identified that, after a fire alarm, an individual who was not a fire responder tailgated through a security door entering the potential fire area. The i event was logged, but no PIR was initiated nor was the control room informed until j the inspectors raised the issue (Section S4.1).

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The inspectors confirmed that an onshift dose assessment capability existed. The i commitment was described in the emergency plan implementing procedures but not I in the emergency plan (Section P3.1). :

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Summarv of Plant Status The plant was maintained at 100 percent power throughout the inspection period, except for routine turbine valve testing, during which power was reduced to 70 percen l. Operations 01 Conduct of Operations 01.1 Increased Drvwell Leakaae Rate Indication and Failure to Enter TS Action Statement Inspection Scoce (71707)

The inspectors reviewed several instances during which unidentified drywell leakage was calculated to exceed TS 3.6.c.1 limits. The inspectors reviewed operator responses and operations guidelines and procedures and held discussions concerning these occurrences with plant management, operations, and engineering personne j Observations and Findinas On December 9,1996, Sump F pumped automatically 51 minutes after the scheduled TS surveillance at midnight which had emptied the sumps. As a result, the calculated unidentified leak rate was found to be 3.33 gallons per minute (gpm),

which exceeded TS 3.6.c.1 requirements of unidentified leakage of less than a 2 gpm change in 24 hours2.777778e-4 days 0.00667 hours 3.968254e-5 weeks 9.132e-6 months . The licensee entered the appropriate TS 24-hour i shutdown action statement. The operations crew also verified drywellindications l of pressure, temperature, humidity, and radiation which indicated no increases in leakage. At 1:40 a.m.,2:51 a.m., and 3:51 a.m. the licensee unsuccessfully attempted to manually start the pump, indicating that the sump liquid level was below the low level pump start permissive switch and that no significant leakage had occurred over that period. At 4:30 a.m. engineering determined that the lack of liquid f!ow into the sump indicated an actualleak rate change of less than 2 gpm during the prior 24-hour period and the plant exited TS 3.6.c.1. A similar leak rate j increase above TS limits occurred on December 14 and the licensee issued PIRs for these occurrence In response to operation's concern regarding the cause of the increased indicated leakage, engineering performed an evaluation of the unidentified leak rate. The engineering evaluation concluded that the steady collection of liquid in drywell unit l cooler drain trays, followed by batch dumping to the sump, caused the unexplained pumping intervals and the apparent excessive unidentified leakage. It appeared that, except for tho e periods of batch flow from the cooler drain trays, only minor amounts of liquid were collecting in the sump.

l On December 19, inspectors expressed concern that the engineering evaluation I would cause operators to become less likely to believe a high leak rate indicatio l l

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2-The plant engineering manager responded that operators were expected to believe leakage rate indication and enter the appropriate TS shutdown action statement However, on December 20, a similar sump pumpdown occurred which resulted in a calculated leak rate of greater than 2 gpm increase in a 24-hour period, thus exceeding the TS 3.6.c.1 limit. The shutdown action statement was not entered because the operating crew considered the engineering evaluation an acceptable explanation of the increased indicated leak rate, which precluded the need to enter the TS action statement. Continued monitoring of the sumps demonstrated that actual leakage was below TS limits before the TS action statement shutdown would have been initiated. The same situation occurred later that day on the second shift I and, again, the licensee did not enter the TS action statemen On the evening of December 21, operations management met with shift supervisors and issued night orders to require entry into the TS shutdown action statement any time indicated leakage exceeded TS limits. Licensee management confirmed its expectations that TS action statements should be entered when there was indication that TS limits were being exceeded. Although multiple causes for this leak rate anomaly had been proposed, no explanation was provided which precluded the requirement to enter the TS action statement when reactor coolant leakage limits were exceeded. Operators responded to a subsequent high leak rate ,

indication by promptly entering the appropriate TS action statement. Continued !

monitoring again showed that actualleakage rate was below TS limit I Conclusions I

During two of five cases of an apparent unidentified leak rate in excess of TS limits, operations did not meet management's expectation by entering the required TS shutdown action statement. Operations management addressed this concern appropriately and reconfirmed that TS action statements must be entered when data indicate that TS leak rate limits are being exceede O3 Operations Procedures and Documentation l 03.1 Pooriv Documented interim Guidance to Operators Concernina Unidentified and Identified Drvwell Leakaae Indication and Sumo Operations Inspection Scope (71707)

Inspectors reviewed instructions given to operators concerning anomalous temporary increases in the drywell unidentified leakage rate, as well as instructions for compensating for f ailure of the drywell equipment drain (identified leakage) sump pump auto start function, high level indicator, and high fill rate indicatio l

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b. Observations and Findinas !

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In reviewing the drywellleakage rate anomalies discussed in Section 01.1 above, !

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l inspectors noted that no written compensatory measures were provided to l operators to ensure a consistent and conservative response to the temporary high l leakage indication I l

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When questioned by the inspectors, licensee management stated that verbal briefings and discussions. as well as current operator training, ensured conservative response to the :l1uation. Inspectors noted that, after the failures to enter the TS action statement, minimal guidance was published in the night orders to provide direction to the operators when in high leak rate situation Concerning the drywell identified leakage, on December 5,1996, the licensee noted that the high level and high-high level switches for the drywell equipment drain sump had actuated and would not clear, although the sump level had been lowere The level switch failures caused the pump to start each time the liquid level rose above the low level setpoint. To preclude excessive pump cycling, the licensee placed the pumps in pull-to-lock and attempted to start the pumps every 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months .

l The sump discharge integrator information was recorded as troubleshooting l_ information for engineering. The inspectors found that the licensee completed the l routine 12-hour TS surveillance procedures for identified drywellleakage and t included all the inventory pumped at 2-hour intervals. The control room staff l- calculated TS leakage rates, then provided the inventory information to engineering l' for further analysis.

i l The inspectors noted that pumping down the sump at 2-hour intervals was not controlled by procedure or instruction but by a comment of " pump every 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months " on caution tags on the pump switches located at the main control boards. After discussions with the inspectors on December 9, the licensee provided informal ;

l written instructions to the control room staff directing the crew to pump the j j equipment drain sump at 2-hour intervals and to review other control room indicators, such as drywell pressure, temperature, humidity, and radiation as well as

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annunciators such as recirculation seal trouble and vessel flange seal lev a i provide additionalindication of leakage between the scheduled pumpd a s. The j licensee concluded that the volumetric inventory logging and the leak icte ;

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determination at the 2-hour intervals along with other control room indicttors '

( provided adequate compensatory measures for monitoring the drywell drain pump l l activity with the level switches out of service. The evaluation of this activity in i accordance with 10 CFR 50.59 is discussed in the engineering section of this report.

! Inspectors identified that, although instructions had been given to pump the sump i every 2 hours2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months , no compensatory action had been provided to pump the sump if

temperature, pressure, humidity, or radiation in the drywell increased. Further, as of the end of the inspection period, the licensee had not provided detailed l

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-4-compensatory action instructions in the night orders or issued other controlled documents to review these indications and to pump the sump if indications warranted that actio I Conclusion i

The inspectors concluded that the instructions to operators for responding to apparent high unidentified leakage and the instructions measuring TS identified leakage after a sump levelindicator failure were weak, in that contingency steps put l in place were poorly documented and incomplet Operations Quality Assurance l 07.1 Ooerations identification of Weaknesses in Operability Evaluations i Insoection Scoce (71707)

i inspectors reviewed a PIR initiated by an operations crew shift test engineer on 1 operability assessment weaknesse ! Observations and Findinas i

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j operability assessments. An operations crew shift test engineer reviewed 35 operability evaluations performed over the past 2 years and noted that the quality of some of the evaluations was weak and that the evaluations had not been adequately examined for trends and lessons learne The licensee concluded that, although weaknesses in individual evaluations did not - i

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result in inoperable equipment, operability assessments could be improve Conclusion l

The licensee performed a critical assessment of the overall quality of past operability evaluations and identified numerous weaknesse O?.2 Quality Assurance Findinas Concernina Nonconservative Reset of the APRM Flow l Bias Scram I Inspection Scone (37551)

The inspectors reviewed Quality Assurance's findings concerning the actions taken during the failure of the APRM flow bias troubleshooting.

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l Observations and Findinas After discussions with the inspector on December 20,1996, Quality Assurance reviewed the licensee's actions to address the flow bias scram reset discussed in Section E1.1. Quality Assurance determined that no procedure was used to perform the troubleshooting, that the guidance given to operators in the form of a memo was weak, and that the condition review group disposition of the original PlR was weak. The licensee initiated PIRs to address these problem Conclusions The Quality Assurance group performed a thorough review of the activities associated with the troubleshooting of the APRM flow bias failure and identified significant weaknesses associated with the licensee's action Miscellaneous Operations issues 08.1 (Closed) Unresolved item (URI) 50-298/96026-02: Slush buildup il circulating :

water intake bays. On November 26,1996, as a result of ice slush floating on the J river surface and the absence of an ice deflector, slush buildup occurred in the circulating water intake bays. Operator's immediate actions precluded plant operationalimpact. As corrective actions, the licensee stated that procedures would be changed to instruct operators to align condensers in the backwash mode, fully open deicing gates, and use fire hoses to break up slush packs if this situation recurre The licensee initiated changes to Procedure 2.4.9.3.2, " Flow Blockage in intake Bays," Revision 13.1, and alarm response Procedure 2.3.2.3, " Panel A -

Annunciator A-3," Revision 22, for high and high-high differential pressure across traveling screens to provide guidance to the operators in this situation. The licensee expected to have the revised procedures in place by the end of January 1997. As an interim corrective measure, a night order log entry was made to inform the operating crews of these contingency plans. As additional corrective action, operations solicited lessons learned concerning other infrequent past events to ensure appropriate response actions were recorded in procedures for control room operator The safety consequence of this failure of this event was low since operators mitigated the effects of slush without reduction of reactor power. The safety consequence was also low in that the service water bay was not vulnerable to this f ailure due to the low flow velocity which did not entrain slush into the service water bay and the deicirn flow to the service water bay. However, inspectors i noted that operations faibd to provide procedures to address this issue although the l occurrence had been noted several years ago and recollection of individual l employees provided adequate information to operators during the event. The issue l s considered a violation of the requirements of 10 CFR Part 50, Appendix B, I

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Criterion V, which requires that procedures be provided which are appropriate to the l circumstances (298/96031-01). l t

The licensee's action to provide :ontrolled interim guidance and to revise the !

procedures appeared to be slow in that the procedure revision had not occurred !

after 2 months and the night order log entry for interim guidance was not made until j questioned by the inspectors on January 6,199 i 08.2 (Closed) Licensee Event Reoort (LER) 50-298/95-019: Control room emergency ;

filter system inoperability due to unavailability of emergency diesel generator. The !

control room emergency filter system was declared inoperable after recognizing that j the system was not aligned to the operable emergency onsite power sourc l TS 3.12.A.1 requires that the control room emergency filter system and its >

associated diesel generator be operable whenever secondary containment integrity is required. At the time of discovery, activities were in progress on the refueling i floor that required secondary containment integrity. The diesel generator was taken j out of service on November 22,1995, and the condition was recognized on l November 24. The system was then declared inoperable as of November 22,1996, l and operability restored within the 7 days allowed by TS Action Statement 3.1 !

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The inspectors verified that the following corrective actions were completed: !

(1) a root cause evaluation was performed by the shift supervisor, documented I in Condition Report 95-1293, and critiqued by the plant management during i the corrective action review board meetin >

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(2) training was conducted with operations personnel on this event in Lesson '

Plan INTRO 23 99-30," Industry Events," Revision (3) the control room TS tracking system was modified to identify that the removal of a diesel generator from service would impact the operability of the control room emergency filter system. The inspector observed a demonstration of the computer program modificatio .3 (Closed) Violation 50-298/95004-01: Two examples where licensee personnel failed to initiate condition reports as directed by Administrative Procedure 0.5, l- " Condition Reporting," Revision 3. The inspector verified the corrective actions

! described in the licensee's response letter, dated June 15,1995, to be reasonable

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and complete. Independent inspector observations, discussions with operators, and review of operations initiated PIRs indicated a low threshold for report initiation in the NRC's acknowledgement letter dated June 30,1995, the NRC disagreed with the licensee's position on the first example. The inspector reviewed the corrective actions implemented for the second example and concluded that the corrective actions encompassed both examples. In addition, routine observations by the

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inspectors have confirmed that a low threshold for PIR initiation exists in operations.

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7-08.4 Review of TS Interoretations (92901}

The inspectors conducted a survey of the licensee's TS interpretations and determined that none of the documents contained informal references to NRC review and approval without formal NRC documentation. The inspectors emphasized to the licensee that any informal reference to NRC review and approval in a TS interpretation is not recognized by the Commission and is not an acceptable practic . Maintenance M1 Conduct of Maintenance M 1.1 General Comments insoection Scope (62703.61726)

l The inspectors observed all or portions of the following work activities: '

Procedure liite MWR 95-4099 Service Water Booster Pump B Discharge Check Valve Maintenance MWR 95-2749: Replace Pressure Switches NBI-PS 51 A/51B MWR 96-1521: Replace Pressure Switch (NBI-PS-51 A) Drain Valve NBI-V-287 MWR 94-0111: Replace Service Water Booster Pump Drain Valve SW-862 PM-03944: Change Service Water Booster Pump 1C Bearing Oil PM-01119: Inspect Motor for Service Water Booster Pump 1C SP 6.1 ADS.301: Automatic Depressurization System Reactor Pressure Permissive Calibration and Functional and Logic Tests SP 6.1RHR.303: Residual Heat Removal Loop A Reactor Vessel Shroud Level Calibration and Functional Test (Division l}

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1 Observations and Findinas

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The inspectors observed that all the procedures and work packages used in the activities witnessed had been reviewed and approved as noted by the appropriate signatures and were the correct revision. The inspectors verified that all of the approoriate TS limiting conditions for operation had been entered prior to the i

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l initiation of the work activities, as noted in the control room shift supervisor's lo Licensee personnel were found to be knowledgeable of their responsibilities and ,

procedural compliance was noted with regard to the performance of the work !

activities. When applicable, appropriate radiation control measures were in plac !

Finally, communications between field and control room personnel were found to be ,

excellen l On January 9,1997, the inspectors observed licensee maintenance personnel replace Service Water Booster Pump 1C Gland Seal Flow Indicator Switch (FIS-2703C) Drain Valve SW-862 using Maintenance Work Request 94-0111. This was l the second of four valves being replaced under this maintenance work reques When the replacement was completed, the inspector noted that the new valve was leaking by as much as had been noted prior to the valve having been installed. The inspector visually verified that the valve was in the closed position, then questioned i the maintenar ce personnel as to the acceptability of this condition. After '

discussions with the maintenance planner, maintenance personnel removed the new valve and reinstalled the old one until the problem could be determine in discussions with the maintenance planner, the inspector determined that another new valve would be used to replace the one that leaked by. The inspector determined also that the licensee had not raised the question as to why the original replacement valve had leaked as it did. The planner stated that they did not know at the time, but planned to determine the caus i During a followup discussion 2 weeks later, the inspector questioned the licensee as to the cause of the leaking valve. The licensee responded that no determination had been made thus far. Also, the inspector determined that a PIR had not been written regarding this issue. The licensee stated that a PIR might be considered for this specific issu As further followup to the issue, the inspector questioned the maintenance planner about the first valve (SW-861) that had been installed under the work request. From comments written by maintenance personnel on the work request, the inspector learned that this valve also showed signs of leakage. The inspector asked if the leakage identified in Valve SW-862 could have been caused by the same problem Valve SW-861 experienced. The maintenance planner did not believe so because the problem identified with Valve SW-861 had been corrected. The inspector found that, prior to the installation of the new valves, maintenance personnel had replaced the original vendor installed gaskets with similarly sized gaskets found in the Cooper Nuc! ear Station warehouse. This was done when the valves were disassembled for inspection and deformed gaskets had to be replaced. When the first valve installed was found to leak by, the cause of the leakage was determined to be that incorrect gaskets were used for the valve reassembl After the licensee contacted the vendor for gasket size information, the

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correctly sized gaskets were installed and the valve leakage stoppe Maintenance personnel concluded at this time that the problem had been correcte The inspector questinned if a PIR had been written with regard to this issue and the licensee responded that one did not need to be written because the cause of the problem had been identified, i.e., that the valve leaked because of the incorrectly sized gaskets. The licensee stated that it was common practice for maintenance personnel, upon identifying a deficiency while working on a maintenance work request, to list it in the work package. Efforts would then be initiated to attempt to address the deficiency prior to closing out the work package. If the deficiency could not be resolved prior to the work package being closed out, then a PIR would most likely be written. The inspector noted to the licensee that this was contrary to ,

the rcquirements set forth in Administrative Procedure 0.5, " Problem Identification I anc. Resolution," Revision 8; the intent of which was to make sure that problems i wore pc. ply documented and appropriately dispositioned. The inspector noted thot, by having ebsed out the leakage !ssue in the maintenance work request r,ackage, the root cause as to why the incorrect gasket sizes had been used was not appropriately adCressed and there were no records of this issue for trending )

purposes. The licensee agreed that the maintenance practices of addressing identified deficiences through an open work package appeared to create a confusion fe.: tor for maintenance personnel with regard to when a PIR should be 1 writte r The failure of the licensee to issue a PIR in regard to the leakage of Valve SW-861 is a violation of 10 CFR Part 50, Appendix B, Criterion V, and Administrative Procedure 0.5(298/96031-02),

in regard to Valve SW-862, the licensee later stated that a PIR would have i eventually been written against the identified leakage problem. It is unclear that this action would have been taken had the inspectors not raised the question, due to the time that had elapsed and the maintenanco personnel policy of addressing deficiencies through open maintenance work requests, c. Conclusions Good procedural compliance by licensee personnel during the performance of maintenance and surveillance activities and excellent communication between plant personnel were note The inspectors identified a violation related to the failure by the licensee to write a PIR for an observed valve leakage problern. This violation is of significance because the NRC and the licensee's corrective action program personnel have identified that some maintenance organizations lagged behind other site organizations in identifying problems through the PIR report process. Although some recent data by the licensee would indicate an improvement in this area, this violation demonstrates

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- 10-a lack of understanding by the maintenance personnel regarding the purpose of the PIR process, i.e., for trending purposes and guaranteeing that problems identified :

are properly addressed. This violation also provides an example of how established maintenance practices have contributed to the confusion regarding when a PIR q should be writte M1.2 (Open) Inspection Followun item (IFI) 50-298/96026-06: Control of Scaling, j Calibration Procedures, and Tolerances for Instruments Used During Routine, l Abnormal, Emergency and Emergency Operating Procedures inspectors identified that calibration processes, scaling, and tolerances maintained on instrument data sheets were not controlled for instrumentation without automatic setpoints and that no evaluations were performed for changes to those documents and processes. This included instruments both inside and outside the control room which were used in emergency operating procedures, emergency procedures, abnormal procedures, and routine procedures. The inspectors determined that the setpoint analysis process only reviews changes to instrumentation with automatic setpoints, j interviews with maintenance, operations, and engineering personnel determined that maintenance was allowed to change instrument scaling, tolerances, and calibration processes without evaluating for consistency with operations procedures end acsign calculations. Further review by the inspectors identified that, for those instruments without automatic setpoints, changes to processes, instrument data sheets, tolerances, and scaling were not subject to design control review in accordance with 10 CFR 50.59 or station operation review committee (SORC) review. The inspectors also determined that engineering was not aware that maintenance personnel did not evaluate these changes to the instrument scaling, data sheets, tolerances, or calibration procedures in accordance with 10 CFR 50.5 The licensee indicated the intent to review past changes to these items to determine if scaling, tolerances, data sheets, and calibration processes were within design values and operating procedure assumptions. The licensee also agreed to review the procedures governing this activity to ensure that future changes to instruments without automatic setpoints received proper evaluatio The inspectors initially concluded that no procedure requirements existed that governed changes to the scaling, tolerances, data sheets, or calibration processes of these particular instruments. Following the inspection, the licensee determined that Engineering Procedure 3.26, " Instrument Setpoint and Meter Banding Control,"

Revision 8, Section 2.3.12, requires that changes to these instrument's setpoints receive a safety review per Administrative Procedure 0.8, " Safety Assessments and Unreviewed Safety Question Determination" Revision j l

After January 11,1997, the licensee performed an initial assessment to determine !

if the required safety evaluations of changes to calibration processes, data sheets, j 1 _ __

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setpoints, and tolerances for instrumentation used by operators to perform plant abnormal, emergency, and emergency operating procedures have been performe The inspectors determined that, by the end of the inspection period, only changes in the instrumentation and controls area have been reviewed and that these changes had been initiated by engineering design changes, which include the required evaluation The inspectors noted that, although maintenance and engineering were aware of this issue on December 20,1996, and plant management was informed by December 31, no PIR documented this issue until inspectors raised questions on January 9,1997, regarding its resolutio At the close of the inspection period, the licensee has not completed evaluation of past changes to instrument calibrations to determine if changes had resulted in instruments performing outside design values or procedure assumptions. Plant management has emphasized to the plant staff the need to identify immediately the safety significance of any changes to plant instrumentation processes, data sheets, tolerances, or scaling. The inspectors will review the licensee's assessment as part of the continuing followup of this IF M1.3 Qqmbustible Materialin Service Water Booster Pumo Room Area 1 inspection Scone (62707) l I

Inspectors evaluated licensee controls, USAR requirements, and design requirements for control of combustibles in the service water booster pump roo Observation and Findinas On December 2,1996,the inspector observed combustible materials (rags, papers, j and flammable chemicals)in the service water booster pump room. The materials were located for scheduled maintenance on the service water booster pump. No licensee personnel were observed in the area for 20 minutes. The inspector noted that the combustible materials were removed before the end of the shift after the maintenance activity was complete l The inspector determined that USAR Section X.8.2.8.C, " Common Mode Failure Analysis - Fue," required that no combustibles be located in the service water booster pump room area since both trains of service water were located in close proximity. The USAR noted that a common mode f ailure would be precluded by assurances that no combustibles would be located in that area. The inspectors determined that no common mode f ailure analysis had been performed to justify the presence of combustible materials in the service water booster pump area.

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Procedure 0.7.1, " Control of Combustibles," Revision 6, allowed up to 90 pounds of wood or 5 gallons of flammable liquid for this area of the plant. To address the contradiction between the USAR and the plant procedures, the licensee initiated a PI The presence of combustible materialin the service water booster pump room in nonconformance with the description in USAR Sectior. X 8.2.8.C, without a safety evaluation, is an apparent violation of the requirements of 10 CFR 50.59(b)(1)

which states, in part, that the licensee shall maintain records of changes in the f acility to the extent that these changes constitute changes in the facility as described in the safety analysis and that these records must include a written safety evaluation which provides the basis for the determination that the change did not involve an unreviewed safety question (298/96031-03). Conclusion The inspectors identified an apparent violation related to a failure by the licensee to perform a safety evaluation of the presence of combustible materialin the service water booster pump room are M8 Miscellaneous Maintenance issues (92902)

M8.1 (Closed) Violation 50-298/95003-01: Motor-operated valve failure caused by interference between the stem locking nut and stem cap. The inspectors verified the corrective actions described in the licensee's response letter, dated May 15, 1996, to be reasonable and complete. No similar problems were identifie i 111. Enaineerina E1 Conduct of Engineering E 1.1 (Open) IFl 50-298/96026-07: Potential nonconservative failure of APRM flow bias featur On November 22,1996, during a surveillance test of Division i APRM E, alarms indicated an unexpected condition in the Division I reactor protection syste Instrument and controls technicians correctly diagnosed that the low-side voltage power supply to the flow bias unit had failed. The unit provided scram signalinput l for all three APRMs in Division I, not just to APRM E which was under a surveillance j test and bypassed at that time, and no half scram occurre An operability evaluation, dated December 11, documented that the failure caused the flow bias scram setpoint to reset to a higher setpoint, a nonconservative value, and documented that the TS action statement should have been entered, as discussed in NRC Inspection Report 50-298/96-26. The licensee stated that the

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vendor had concluded that this particular failure mode was unexpected and that l other reactor plants of similar design may be vulnerable. The licensee's review of reactm protection system circuits determined that, when APRM E or B was li bypassed, this vulnerability would occur. As a compensatory measure, the licensee provided directions to the control room that, when either APRM E or B was in bypass condition, a control rod should be selected, which would result in a control rod block monitor alarm if this particular failure were to recur. The inspectors concluded that these compensatory measures were appropriate, but that further evaluation by the licensee is necessary. This issue may include a potentially nonconservative reactor protection system failure mode and/or failure to incorporate industry experienc During the NRC exit meeting, the licensee stated that the vendor had notified NPPD in mid-December that this problem was discussed in the scope of Services Information Letter 445 issued in 1988, with revisions and supplements in 1989 and 1990. The licensee provided documentation to show that the recommendations in Services Information Letter 445 had been implemented, but that this action did not include the APRM power supplie The licensee concluded that the failure was not significant, because the flow bias i scram feature was not credited in accident analysis and the high power flux tiip !

bounded the analysis. Observations by quality assurance are discussed in -

Section E7.1 of this report. The inspectors will continue to review the licensee's 1 actions in response to this issue as part of the continuing followup of this IF E2 Engineering Support of Facilities and Equipment ,

E Review of Uodated Safety Analysis Report Commitments A recent discovery of a licensee operating facility in a manner contrary to the USAR description highlighted the need for a special focused review that compares plant practices, procedures, and/or parameters to the USAR description. While performing the inspections discussed in this report, the inspectors reviewed the applicable portions of the USAR that related to the areas inspecte Failure of the licensee to document changes of the plant configuration that were consistent with the USAR are documented in Sections M1.3, E2.2, E2.4, E2.5, and E2.6 in this repor E2.2 (Closed) URI 50-298/96026-02: Slush buildup in circulating water intake bays. On November 26,1996,in response to a slush buildup in the circulating water bay the licensee identified that an ice deflector had not been installed. The NRC identified that Section XII.2.2.7.1," Intake Structure," of the Cooper Nuclear Station USAR states in part, that in order to keep ice away from the intake structure during cold weather, an ice deflector is installed during the winter months. The ice deflector

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had not been installed because river level was 6 feet higher than normal for this j time of year, which interfered with the attachment points for the deflecto In response to inspector questions, the licensee determined that the ice deflector . j modification had not received appropriate evaluation in accordance with j 10 CFR 50.59, both for the originalinstallation as well as for subsequent changes t to the ice deflector configuration. The unexpected buildup of slush in the circulating i water bays resulted in a challenge to plant operators to maintain the unit at full !

power. The licensee then issued problem identification reports to address these l

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concerns. On December 18, the inspectors identified that, although the licensee had subsequently installed a portion of the ice deflector and it appeared to be preventing ice buildup at the intake bay, the partial installation of the ice deflector also was considered a configuration which had not been evaluated in accordance ;

with 10 CFR 50.59 requirement ;

The failure to initially perform a safety evaluation of the lack of installation of the ice deflector is an apparent violation of 10 CFR 50.59(b)(1), which states, in part, that ;

the licensee shall maintain records of changes in the facility to the extent that these !

changes constitute changes in the facility as described in the safety analysis report l and that these records must include a written safety evaluation which provides the

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basis for the determination that the change did not involve an unreviewed safety question (298/96031-03). -

E2.3 Chanaes. Tests, and Experiments - 10 CFR 50.59 Insoection Scope (37551)

The inspectors reviewed the licensee's program, including training, for conducting evaluations under 10 CFR 50.5 ; Observations and Findinas !

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The licensee's program for conducting evaluations under 10 CFR 50.59," Changes, !

Tests, and Experiments," is undergoing extensive revision and internal review. In !

1994,in response to criticism of the existing ,orogram, the licensee established a Review Group to enhance the quality of 10 CFR 50.59 screens and safety evaluations reaching the Station Operations Review Committee. Other actions and ;

condition reports provided a thorough analysis of the deficiencies, root causes, and j corrective actions needed for the 10 CFR 50.59 program, including the screening :

processes required prior to imposing 10 CFR 50.59 evaluations. On February 23, !

1995, licensee management directed the facility staff to adopt a single screening [

process to replace the three procedures which implement 10 CFR 50.59. However,

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Review Group was discontinued on May 28,199 '

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I l-15-l The licensee completely revised the training program and evaluator qualification requirements; however, only one class has completed the new training. The licensing group was stillimplementing portions of the program. The training and program now refer to NSAC 125, and the documented requirements exceed the interim guidance to inspectors as published in the April 9,1996, edition to Part 9900 of the NRC Inspection Manua The ownership and implementation of the 10 CFR 50.59 program at Cooper Nuclear Station was fragmented and stillin transition. The Design Engineering Department )

was responsible for station modifications, the design or plant engineer for test i procedures, the computer applications analyst for software modifications, and the !

designated engineer for preparing temporary modifications. Nuclear Licensing was the proposed future owner of the 10 CFR 50.59 program, including training. The qualification guide of 10 CFR 50.59 evaluators was also under revision regarding evaluation, testing, and proficiency. With the state of ownership, training, and qualification stillin transition, continued attention by the licensee and the NRC is warrante l The screening of proposed activities for safety relevance prior to subjecting the l activity to the 10 CFR 50.59 process is also in the early stages of improvement by j the licensee. A number of activities were called into question by the NRC and i licensee staff about the adequacy or existence of the screens. The operations and maintenance departments appeared less aware than engineering of activities and the implications associated with implementation of the 10 CFR 50.59 proces However, an improvement in the effectiveness of the 10 CFR 50.59 program was evident at Cooper Nuclear Station. Evaluations performed prior to 1995 were judged by the licensee to be less than adequate and NRC review confirmed this ,

assessment. Evaluations performed since the Review Group and the independent {

contractors efforts have resulted in evaluations which typically range from adequate to well prepared and thorough. There has been good quality assurance review of the products. However, the licensee's plans to improve the process, training, and qualification do not appear to include a continuing, centralized quality assurance functio The Station 05.erations Review Committee (SORC) review has been mixed. Prior to 1994, reviews were very poor, without identifiable overall program improvement However, review of the minutes for the past year revealed that the SORC safety review of 10 CFR 50.59 evaluations was generally goo The inspectors reviewed a representative sample of 10 CFR 50.59 evaluations and determined that the evaluations prior to 1995 were poor by the NRC's and licensee's standards. The licensee has established an improvement program, which is raising the quality of the evaluations and the attention to detail if an activity was placed in the evaluation process, the product was typically good. The screenings for the past year were reviewed, especially those that did not rise to the level

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l-requiring a 10 CFR 50.59 evaluation. Again, prior to the licensee's development of ,

consistent and adequate screening criteria and examples, the screens were very I poor. The screens are now generally good. The inspection did address the most '

recent resident inspectors issues identified with operations and maintenance failure !

to implement screens and evaluations of their activitie :

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. Conclusions ;

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The inspectors concluded, in agreement with the licensee assessment, that prior to )

1995 the licensee's 10 CFR 50.59 eva:uations were poor. The licensee has I implemented corrective actions that improved performance to a generally good l level. However, the screens and 10 CFR 50.59 evaluations warrant continued l oversight and emphasis due to the recent implementation of improvements and !

training and changes in ownership of the overall progra I

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E2.4 Failure to Address USAR Hiah Sumo Fill Rate Alarm Function in Evaluation

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. Insoection Scooe (37551)

i inspectors reviewed the licensee's 10 CFR 50.59 evaluation of the failure of the ;

high level alarm in the drywell equipment drain (identified) leakage sum j l Observations and Findinas l Engineering determined that the volume contained between the low level pump shutoff and the high level pump str. 's approximately 300 gallons. The j inspectors determined that, since ;C 3 A.c established a maximum leak rate of

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25 gallons per minute, the potential to exceed this leak rate would be identified by successivo sump pumpdowns within 12 minutes. Licensee Alarm Response Procedure 2.3.2.25, " Panel 9-4 - Annunciator 9-4-2," Revision 31.3, indicated that the annunciator alarm setpoint for exceeding the TS 25 gpm leak rate would result in an alarm if less than 13 minutes occurred between pumping intervals. However, the failure of the high level alarm caused the sump fill rate timer to not functio The inspectors concluded that the licensee's compensatory action to pump the sump at 2 hour2.314815e-5 days 5.555556e-4 hours 3.306878e-6 weeks 7.61e-7 months intervals was not conservative with respect to the alarm limit of 13 minutes, but determined that no immediate safety issue existed based on the configuration of the identified and unidentified leakage sump The licensee stated that pumpdown of the equipment drain sump was required at 12-hour intervals by TS and, although more frequent pumpdown was not required, it was being performed only as a troubleshooting measure to assist in gathering i information for engineering. The inspectors responded that, apart from the 12-hour pumping interval, the USAR described an automatic function that would provide an

! alarm indication in the control room within 13 minutes of TS leakage rates being l exceeded, r

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-17-Section IV.10.3, " Nuclear System Leakage Rate Limits - Description," of the USAR states, in part, that each drywell sump has an alarm system and automatic starting sequence on rising water level. Both drywell sumps are equipped with a fill rate timer and alarm. This alarm can be set at or below the TS limits and would provide immediate indication when this preselected ; ate is reached or exceeded. On December 20,1996,15 days after the failure, the licensee completed an evaluation in accordance with 10 CFR 50.59 to address the f ailure of the sump high level alarm and the automatic pump starting system. This safety evaluation did not address the lack of the fill rate timer and alarm, nor did a separate safety evaluation exist on this issue. This f ailure to address the fill rate timer in the safety evaluation is an apparent violation of 10 CFR 50.59(b)(1), which states, in part, that the licensee shall maintain records of changes in the facility to the extent that these changes constitute changes in the facility as described in the safety analysis and that these records must include a written safety evaluation which provides the basis for the determination that the change did not involve an unreviewed safety question (298/96031-03). Conclusions inspectors identified an apparent violation related to the failure by the licensee to address, in a 10 CFR 50.59 safety evaluation, the leak rate of a fill rate timer for the drywell equipment drain sump, as discussed in the USAR. The inspectors also concluded that the licensee did not complete the safety evaluation in a timely manne E2.5 (Closed) IFl 298/96007-06: Examoles of Failures to Correct USAR Discreoancies During routine reviews to close out previous violations (94014-05 and 94014-06)

concerning failures to properly test containment penetrations, inspectors reviewed procedures, the USAR, and drawings, as documented in NRC Inspection Reports 50-298/96-03 and 50-298/96-07. During these reviews, inspectors identified that the licensee had failed to correct incomplete or incorrect USAR listings of containment penetration NRC inspection Report 50-298/96-03 documented that Table V-2-2, " Penetration Schedule" (pages V-2-9 through V-2-12), improperly characterized or did not list several penetrations. NRC inspection Report 50-298/96-07 documented that Table V-2-7,

" Testable Primary Containment Isolation Valves" (pages V-2-44 through V-2-46) had 23 penetrations and the associated valves missing from the list of penetrations. In conjunction with NRC Inspection Report 50-298/96-24,these are examples of apparent violations of the requirements of 10 CFR 50.71(e), which states, in part, that the USAR is to be updated periodically to assure that the information in the USAR contains the latest material developed (298/96031-04).

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E2.6 Quality Assurance Audit of Desian Control

! i Insoection Scope (37551)

The inspectors reviewed licensee Quality Assurance (QA) Audit 96-07, which was

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conducted to evaluate the effectiveness of the licensee in maintaining design i control at CNS.

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l Observations and Findinas !

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( in an audit of design control, dated June 17,1996, QA identified, the following )

l issues which were extracted from the licensee's report:

some pertinent operational requirements, resulting from assumptions in i calculations, and NRC commitments were identified as not being included in l CNS USAR updates.

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The methodology for the analysis was not always included in the LCR

! (License Change Request) or USAR update * For a new mode of operation (previously unanalyzed), it was not recognized that the mode of operation was outside of the licensing basis, nor was it I recognized when changes being considered were not analyzed for all conditions (e.g., service water booster pump windmilling issue and subsequent reports).

:SAR updates were not always user friendly or adequate to communicate !

l important aspects of the change (e.g., limited assessment of LCRs revealed in a contractor report dated May 10,1996).

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Procedural requirements for updating the USAR do not provide guidance for l where different types of analyses and information should bis contained within the USAR.

i The licensee's audit concluded that maintenance of the USAR was not adequate j and, where deficiencies in USAR did exist, there was no consolidated, formal, effective corrective action effort or plan in plac The audit further noted that significant reports by contractors dated May 7 and 10, 19P6, and corrective action history generally indicated that the USAR may not be accurate in all material respects and that surveillance procedures may not be compliant with the USAR in allinstances. The audit also noted 14 examples of Condition Reports since 1994 indicating USAR discrepancies, indicating lack of effective corrective action to resolve USAR discrepancies.

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in a response to the audit dated August 26,1996, the line organization noted that no safety issues were identified, the primary conclusions of the audit were in error ;

for several reasons, and the overall safety bask descriptions were adequat '

Additionally, the response included an industry position on the USAR, a legal history of the FSAR, concluding, in part, that most nonconformances with the USAR do not .

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translate into unsafe operation. The response also stated that a program for changes, additions, and/or deletions to the USAR for improvements to station ;

efficiency is to be evaluated by management in accordance with station prioritie '

On August 30,1996, QA did not accept the line organization response, escalated it to higher management, and documented that, although no major issues have been found representing a challenge to nuclear safety, corrective actions were necessary to systematically identify and resolve additional discrepancies in the USAR and related plant operational constraints, and QA reemphasized the current obligation to maintain accuracy and consistency between the USAR and plant operation, Conclusions On June 27,1996, a QA audit documented concerns that the USAR had not been properly maintained, and systematic correction of the condition was necessary. The line organization disagreed with this assessment and QA escalated the concern to higher management on August 30. In a meeting with the NRC on December 19, 1996, the licensee discussed plans to initiato a USAR update effor l IV. Plant SuDDort R8 Miscellaneous Radiation Protection and Chemistry issues R (Closed) Violation 50-298/95003-02: Radiation protection technician failure to follow procedures. The inspectors verified the corrective actions described in the licensee's response letter, dated May 15,1996, to be reasonable and complete. No i similar problems were identifie l P2 Status of Emergency Preparedness Facilities, Equipment, and Resources P Emeraency Preoaredness Staffina Insoection Scooe (71750)

The inspectors reviewed all emergency response positions to determine if they could be filled at all time _ . . . . - -. - - - . - . _ --..-.-.--.- -.~. -.-

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Observations and Findinas ;

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On December 18,1996, the inspector noted that severallicensee staff members qualified in the emergency response organization staff had left the site area for

numerous reasons. The inspector asked whether the emergency response

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organization was fully staffed and if individuals associated with those positions were aware of substitutions which may have been required due to the absence of

qualified members from the site. Further evaluation determined that the individual

assigned as the emergency director had not arranged for a substitute prior to departure from the site.

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! The inspector found that the process for ensuring that emergency response I

!- positions were staffed required only that the assigned individuals determine whether j

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a substitute was needed and obtain concurrence from their substitute to assume i emergency response duties. The lack of an assigned substitute in the case above i

was not significant because qualified individuals were available in the site area to perform emergency director duties. Additionally, all emergency response i organization members (not just the duty team) are contacted in the event the I organization is activate l l Conclusion i

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Inspectors identified a minor weakness in that the licensee failed to identify a - )

substitute emergency response organization member before the assigned team

member left the site. The inspect'rs considered this a minor iss'ue since a qualified j individual was available local to the site and would have been contacted in the j

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event of an emergenc P3 Emergency Preparedness Procedures and Documentation

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P3.1 (!censee Onshift Dose Assessment Caoabilities (Tl 2515/134) h,30ection Scoce

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Using Temporary Instruction 2515/134,the inspectors gathered information regarding:

Dose assessment commitment in the emergency plan

Onshift dose assessment emergency plan implementing procedures

Onshift dose assessment training

, Observations and Findinas On December 16,1996, the inspectors conducted an in-office review of the

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emergency plan and implementing procedures to obtain the information requested

by the temporary instruction. The inspectors conducted a telephone interview with

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-21-the licensee on December 17,1996, to verify the results of the review. The licensee provided additionalinformation on January 9,1997. Based on the documentation review and information provided by the licensee, the inspectors determined that the licensee had the capability to perform onshift dose assessments using real-time effluent monitor and meteorological data. The commitment was clearly described in the emergency plan implementing procedures but not in the emergency plan. Further evaluation of the information obtained using the temporary instruction will be conducted by NRC Headquarters personne Conclusion The inspector confirmed that an onshift dose assessment capability existed. Th commitment was clearly described in the emergency plan implementing procedures but not in the emergency pla S4 Security and Safeguard Staff Knowledge and Performance S Fire Watch Tailoated into Vital Area Inspection Scope The inspectors observed a fire watch enter a vital area after a fire alarm in the are Observationt and Findinos On January 8,1997, after a fire alarm,the inspectors identified that an individual who was not a fire responder tailgated through the service water area security door entering the potential fire area. The inspector immediately informed the individual of the concern regarding tailgating, and informed security. .The event was logged, but no PIR was initiated nor was the control room informed until the inspectors raised the issu Conclusion A weakness was identified by the inspectors in that an individual did not card into a vital area as required and that security personnel, when informed of the event, did

. not initiate a PIR on the event or inform the control room until after the inspectors raised the issu . . . . _ . . - . _ _ . _ _ _ m... .., . _ . _ . _ - . . _ . _ _ . . . ~ . . ._ . . _ . ._. . . . - _ . . _ . . . _ . _ . . _. . . _ _ _

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I 22-l VI. Manaaement Meetinas

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X1 Exit Meeting Summary

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The inspectors presented the inspection results to members of licensee management at exit meetings on January 21,1997, and telephonically on February.19,1997. The licensee ,

acknowledged the findings presente '

The inspectors asked the licensee whether any materials examined during the inspection should be considered proprietary. No proprietary information was identifie f

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I SUPPLEMENTAL INFORMATION l

PARTIAL LIST OF PERSONS CONTACTED

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l Licensee i Dan Buman, Design Engineering Manager -

Jack Dillich, Maintenance Manager Jerry Dorn, Containment Engineering Supervisor Lisa Freeman, Licensing Rick Gardner, Operations Manager Phil Graham, Vice President of Nuclear Energy Bradford Houston, Licensing Manager Mike Peckham, Plant Manager Jim Pelletier, Senior Manager, Engineering l LIST OF DOCUMENTS REVIEWED _

Emeraency Plan Implementina Procedures i

5. Shift Supervisor Emergency Plan implementing j Procedure Revision 10 ;

5,7.17 Dose Assessment Revision 2 Other Documents Cooper Nuclear Station Emergency Plan Revision 31 i

INSPECTION PROCEDURES USED IP 37751: Onsite Engineering

! IP 37828: Engineering - Install and Test Modifications l IP 61726: Surveillance Observation ,

IP 62707: Maintenance Observation !

IP 71707: Plant Operations IP 71750: Plant Support Activities IP 92901: Followup - Plant Operations IP 92902: Followup - Maintenance IP 92903: Followup - Engineering IP 92904: Followup - Plant Support ;

IP 92700: Onsite Followup of Written Reports of Non-Routine Events at Power Reactor Facilities l

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lP 93702: Prompt Onsite Response to Events at Operating Power Reactors Tl 2515/134 Licensee Onshift Dose Assessment Capabilities f

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i LIST OF DOCUMENTS REVIEWED l Emergency Plan implementing Procedures:

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5. Shift Supervisor Emergency Plan Implementing l Procedure, Revision 10 I b.7.17 Dose Assessment, Revision 2 I l

v t e Inspection Report - Cooper - 1996031
80 \| 81 \| 82 \| 83 \| 84 \| 85 \| 86 \| 87 \| 88 \| 89 \| 90 \| 91 \| 92 \| 93 \| 94 \| 95 \| 96 \| 97 \| 98 \| 99 00 \| 01 \| 02 \| 03 \| 04 \| 05 \| 06 \| 07 \| 08 \| 09 \| 10 \| 11 \| 12 \| 13 \| 14 \| 15 \| 16 \| 17 \| 18 \| 19 \| 20 \| 21 \| 22 \| 23 \| 24 \|
1996 Quarterly Integrated 1Q (0) 2Q (0) 3Q (0) 4Q (0) Assessments Mid Cycle End of Cycle 12(0) 13(0) 15(0) 16(0) 17(0) 18(0) 19(0) 21(0) 22(0) 23(0) 24(0) 25(0) 26(0) 27(0) 28(0) 29(0) 31(0) Security Operator Exams Emergency Planning Other

IR 05000298/1996031

Text

REGION IV

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