ML20216B667

Insp Rept 70-7001/98-03 on 980120-0309.Violations Noted. Major Areas Inspected:Plant Operations,Maint,Surveillance & Engineering

ML20216B667
Person / Time
Site:	07007001
Issue date:	04/06/1998
From:	Hiland P NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20216B646	List: ML20216B641 ML20216B659 ML20216B667
References
70-7001-98-03, 70-7001-98-3, NUDOCS 9804140043
Download: ML20216B667 (15)

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l U.S. NUCLEAR REGULATORY COMMISSION i i

REGION 111 Docket No: 70-7001 Certificate No: GDP-1 Report No: 70-7001/98003(DNMS) l Facility Operator: United States Enrichment Corporation Facility Name: Paducah Gaseous Diffusion Plant j Location: 5600 Hobbs Road P.O. Box 1410 Paducah, KY 42001 I Dates: January 20 through March 9,1998 l

1 Inspectors: K. G. O'Brien, Senior Resident inspector J. M. Jacobson, Resident inspector Approved By: Patrick L. Hiland, Chief l Fuel Cycle Branch l Division of Nuclear Materials Safety l

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9004140043 900406 i PDR ADOCK 07007001 i C- PDR ,

EXECUTIVE

SUMMARY

United States Enrichment Corporation Paducah Gaseous Diffusion Plant NRC inspection Report 70-7001/98003(DNMS)

Plant Operations

. The plant staff appropriately initiated an investigation into repeat malfunctions of the P-515 pressure instrumentation channel that led to safety system actuations for the Autoclave 2 East Steam Pressure Control System. ' At the conclusion of the inspection period, the root cause and corrective actions for the malfunctions had not been determined. (Section O1.1)

. The certificatee identified that the assumptions used to develop the source term for liquid uranium hexafluoride releases from Buildings C-310 and C-315, during a potential seismic event, were non-conservative. After discussions with the NRC, the certificaise submitted a Request for Enforcement Discretion to allow continued withdrawal operations with adminis'rative controls on the liquid uranium hexafluoride inventories in the accumutr.iors. At the conclusion of the inspection period, the NF C had not completed review of the certificatee's Request for Enforcement Discretion. ;Section 01.2)

The plant staff expeditiously developed a reasonable plan to remediate a planned expeditious handling deposit discovered in the Number 1 Purge and Evacuation Pump Cooler in Building C-335. (Section 01.3)

. The inspectors identified a Technical Safety Requirement violation, in that, the plant shift superintendent failed to declare the Building C-333A Process Gas Leak Detection System inoperable, following boss of the local alarm function; failed to direct implementation of the Technical Safety Requirement Action Statements; and, failed to make a report to the -

NRC of the disabled safety-related equipment, in accordance with plant procedures. The inspectors also identified that the long-term order process did not ensure that the plant shift superintendents were periodically retrained on long-standing orders, a process weakness. (Section O1.4)

Maintenance and Surveillance

. _ The inspectors identified an ongoing recirculating cooling water valve leak that was not well contained and had the potential to negatively affect the functionality of a cascade cell

. emergency stop button. The inspectors noted that the management-by-walking-around program did not appear to have identified or evaluated the leak's impact. (Section M1.1) 2 e

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. The inspectors concluded that the approach taken by plant staff to resolve an operability Issue associated with jetting the residual steam from autoclaves after completion of a cylinder feeding or sampling cycle was reasonable. (Section E1.1)

. The inspectors concluded that the approach taken by plant staff to resolve an operability issue associated with the trouble and alarm setpoints for the criticality accident alarm system was reasonable.-(Section E1.2) i r

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Report Details

1. Operations 01 Conduct of Operations 01.1 Failures of Autoclave Steam Pressure Control System

a. Inspection Scope (88100)

The inspectors reviewed the circumstances surrounding two notifications to the NRC tha+

actuations of the autoclave steam pressure control safety system had occurred on Autoclave 2 East in Building C-337A.

b. Observations and Findinas On January 13,1998, the Plant Shift Superintendent (PSS) notified the NRC that a safety system actuation of the steam pressure control system for Autoclave 2 East in Building C-337A had occurred (Certificatee Event Report 33521). The steam pressure control system was designed to prevent overheating a cylinder by limiting the pressure of saturated steam in the autoclave to less than 8.0 pounds per square inch gauge (psig).

The actuation was caused by a malfunction of the steam controlinstrumentation channel (P-515) which failed in the negative direction (readout of- 3.75 psig) causing a demand for steam which in tum caused the steam control valve to fully open. The opening of the control valve to the maximum limit created a steam pressure within the autoclave of 6.5 psig which actuated the safety system (set to actuate between 6.5 and 7.5 psig) through the other pressure detection channel (P-514).

On February 1 another actuation of the steam pressure control safety system occurred because of a malfunction of the P-515 pressure instrumentation channel for Autoclave 2 East in Building C-337A (Certificatee Event Report 33637). This additional malfunction occurred after corrective maintenance had been completed for the previous event and appropriate post-maintenance testing had been performed.

in both instances, operators responded in accordance with the appropriate Alarm Response Procedure and the PSS declared the safety system inoperable. Plant staff noted that this particular instrumentation had been modified during the autoclave instrument upgrade process which was conducted in accordance with the Compliance Plan. Plant staff began an investigation into the repeat malfunctions. The initial investigation, which wac continuing at the end of the inspection period, pointed to a six-pin connector or amplifier in the channel as a potential cause of the malfunctions.

Plant staff noted varying quality on the soldering which had been performed during the installation of the six-pin connectors for the instrument upgrade project. However, subsequent troubleshooting and non-destructive testinn indicated that these components were functioning as designed. Plant staff were contint.ag to evaluate the new pressure transducers utilized in the P-515 channel as another potential cause of the malfunction.

Destructive tests at the manufacturer were scheduled to further evaluate the malfunctions. The inspectors will continue to track the root cause and corrective action processes for the malfunctions as part of the followup to the certificatee event reports.

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t c. Conclusions l

l Plant staff appropriately initiated an investigation into repeat malfunctions of the P-515 l pressure instrumentation channel that led to safety system actuations for the Autoclave 2 l l

East Steam Pressure Control System. At the end of the inspection period, the root cause  ;

l and corrective actions for the malfunctions had not been determined. )

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01.2 Reauest for Enforcement Discretion for Product and Tails Withdrawal Buildinas

a. Inspection Scope (88100)

The inspectors participated in an NRC review of the compensatory actions developed to i support a Request for Enforcement Discretion for continued liquid uranium hexafluoride l (UF ) operations of Withdrawal Buildings C-310 (Product) and C-315 (Tails). The request

! was made as a result of plant staff's identification that the potential consequences of l

seismically induced equipment failures exceeded those previously analyzed based on assumptions of minimalliquid UF, inventories in the building's condensers and accumulators.

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b. Observations and Findinas l' On February 5, the NRC issued a Request for Additional Information related to the Safety l Analysis Report Update analysis of postulated seismic failures of the liquid-UF, vessels

! and piping in Buildings C-310 and C-315. The initial analysis assumed no significant liquid releases from the condensers or accumulators because of very low inventories of material during routine operations. After further discussions with the NRC, the United l States Enrichment Corporation (USEC) submitted a letter dated February 20,1998,

! under 10 CFR 76.0, indicating the assumptions for the seismic accident analysis in Buildings C-310 and C-315 were non-conservative. The letter also indicated that the consequences could be more severe than those currently reported in Chapter 4 of the Safety Analysis Report.

The February 20,1998, letter provided compensatory actions for the potential increase in the amount of liquid UF, released as a result of a seismic event. The actions included:

1) using only one accumulator in Building C-315; 2) instituting a crew briefing for operators on the seismic concems with the accumulators; 3) issuing a long-term order for l cascade coordinators on when to reduce tails withdrawal rates to minimize the amount of liquid UF, stored in the Building C-315 accumulator; 4) issuing long-term orders requiring

empty cylinders be kept available at ali operable withdrawal positions not in use;  ;

! 5) placing Buildings C-310 and C-315 on recycle in the event that accumulator capacity in either build.ing should be reached; and,6) restricting access to Buildings C-310 and i C-315 to only those individuals essential to operations or inspections.

! An NRC prelirninary review of the February 20,1998, letter indicated that the proposed

compensatory actions did not include quantified limits for accumulator use or estimates of l the potential offsite consequences for those limits. Pursuant to further discussions with the NRC, USEC submitted a Request for Enforcement Discretion on February 25,1998.

A subsequent meeting between the NRC and USEC on March 3 led to a revised letter, dated March 5,1998, which included quantified administrative limits on the UFe inventories allowed in the accumulators for Buildings C-310 and C-315, the major 5

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contributors to an offsite release, The administrative controls required notification of the PSS and NRC after one hour of accumulator use, calculating quantities of UF, accumulated, and initiating high priority actions to resolve unscheduled outages and -

reduce withdrawal rates, if the crJ niated inventory in the online accumulator reached 4,000 pounds in Building C-310 or 10.000 pounds in Building C-315, flow to the affected l accumulator would be stopped. Poteritial offsite consequences for releases of this magnitude were also provided.

As of the end of the inspection period, the NRC had not completed review of the

- certificatee's Request for Enforcement Discretion.

c. Conclusions The certificatee identified that the assumptions used to develop the source term for liquid uranium hexafluoride releases from Buildings C-310 and C-315, during a potential p' seismic event, were non-conservative. After discussions with the NRC, the certificatee l submitted a Request for Enforcement Discretion to allow continued withdrawal operations ,

l with administrative controls on the liquid uranium hexafluoride inventories in the  !

!. accumulators. As of the end of the inspection period, the NRC had not completed review l of the certificatee's Request for Enforcement Discretion. ,

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01.3 Planned Expeditious Handlina Deposit in Buildina C-335 l a. Inspection Scope (88100)

The inspectors followed up on the discovery of a p!anned expeditious handling (PEH) deposit in the Number 1 Purge and Evacuation Pump Cooler in Building C-335 on January 15,1998.

b. Observations and Findinas Upon discovery of the PEH deposit, the plant staff took actions to ensure the deposit remained L. r. fluorinating environment, as required by Technical Safety Requirement (TSR) 2.4.4.4, " Cascade Wet Air Inleakage." In addition, the pc.nt staff initiated an investigation into the cause of the deposit and developed a remediation plan within the 30-day time frame required by the TSR Action Statements. The approved plan of action,  ;

da'ed February 9,1998, documented that the cause of the deposit was most likely the accumulation of reaction products on the gas cooler heat transfer fins over years of operation. The plan initially called for performing a chemical treatment of the deposit in the cooler to try to reduce the deposit in situ. This phase was to be followed by non-destructive assay (NDA) scans to determine if the mass of the deposit was reduced below safe mass. If this treatment proved unsuccessful, plant management would evaluate other methods to reduce or remove the deposit, including cutting out the cooler.

The inspectors reviewed the plan and concluded that the plant staff had developed a reasonable approach to remediating the PEH deposit in the Number 1 Cooler.

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c. Conclusions The plant staff developed a reasonable plan to remediate a planned expeditious

' handling deposit discovered in the Number 1 Purge and Evacuation Pump Cooler in Building C-335; 01.4 Inocerable Buildina C-333A Proces_s Ga* Leak Detection System

a. Inspection Scope (88100)

The inspectors reviewed the circumstances surrounding a failed autoclave programmable L logic controller (PLC) in Building C-333A and the resultant . impact on the autoclave process gas leak detection (PGLD) system.

b. Observations and Findinas On March 5 during a routine tour of Building C-333A, the inspectors noted that both the Number 1 North and South Autoclaves were inoperable. The inspectors also observed that the autoclaves contained cylinders, indicating that the cylinder heating cycles were unexpectedly interrupted. The inspectors discussed the observations with the building operators and were informed that an equipment failure interrupted the heating cycles.

- Specifically, on March 1 at approximately 10:15 a.m., a PLC logic card for the Number i North and South Autoclaves failed causing: 1) the PLC to go into halt mode; 2) the autoclaves to go into containment, TSR Mode 3; and, 3) the heating cycle to be '

l interrupted. The inspectors reviewed the TSRs and determined that the building -

L operator's handling of the cylinders was in accordance with TSR 2.2.4.11, " Valve Clarity / Heating Cycle Interruptions."

The inspectors reviewed the operator logs for March 1 and noted entries for:-

1) 10:15 a.m., indicating the PLC going into halt mode; 2) 10:45 a.m., indicating the  ;

operators tested the PGLD system and determined the local alarm function was not  !

working, and; 3) 12:35 p.m., indicating the PSS had deciated safety systems, associated with the Number 1 North cnd South Autoclaves, inoperable. However, the 12:35 p.m.,

i log entry did not list the PGLD system as one of the systems declared inoperable. The E inspectors discussed the apparent inconsistency between the 10:45 a.m. and 12:35 p.m.,

log entries with the building operators and operations management. The operators and operations management indicated the PGLD system was considered operable but l degraded. Specifically, the operators and operations management stated that the PGLD .i system, if actuated, would still provide an alarm in the Operations Monitoring Room l l

(OMR). Therefore, the PSS did not consider the PGLD system's inability to alarm in the  !

autoclave area as a condition that would make the system inoperable. However, both the i L operators and operations management agreed that loss of the PGLD system's local j alarm required the implementation of compensatory measures. ' As a result, operations management stated the compensatory measures described in long-term order (LTO) 333A-97-014 were implemented. l c Based upon a review of the Safety Analysis Report (SAR) description of the PGLD system and the applicable TSRs, the inspectors concluded that the operation staff's assessment of the System as operable but degraded was incorrect. Specifically, SAR Section 3.2.5.6, entitled, "UFe Detection System," stated that the PGLD system includod l

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4 i-l l i- - detection and local sounding of an alarm at each autoclave. Section 4.3.1.1.1, of the L . SAR, entitled, " Valve and Pigtail Failures," stated that PGLD system alarmed in the local area. The Basis Statement for TSR 2.2.4.1, entitled, "UF, Release Detection System,"

stated that the [PGLD) system sounded an alarm to alert operating personnel to initiate corrective and mitigative action. Finally, Table 3.2.2.1, entitled,

• Minimum Staffing l Requirements," of TSR 3.2.2, allowed, in part, the building operators to be located in l areas of the building that would preclude their awareness of a PGLD system OMR alarm.

l Therefore, PGLD system operability required functioning alarms in both the OMR and L local area.

The inspectors reviewed the LTO and concluded the required compensatory measures i- . were not implemented. Specifically, LTO ltem 3 was not implement. Item 3 required the j initiation of a smoke watch for the autoclave heated housing, piping trench, and the jet l- station. During discussions with the operators, the inspectors were informed LTO ltem 3 l was not implemented because it applied only to Building C-337A. The inspectors noted i

LTO ltem 3 included a reference to the Building C-337A; however, the reference was only l- for the Building C-337A west wall detectors. The other LTO ltem 3 listed detector areas which were common to Buildings C-333A and C-337A and should have been covered by a smoke watch. Implementation of LTO item 3 following the PLC failure would have l

fulfilled the TSR 2.2.4.1, Action Statements for an inoperable PGLD system.

During review of the LTO compensatory measures, the inspectors determined that the operations staff's initiation of LTO ltem 1 was a conservative and appropriate action.

LTO ltem 1 required continuous manning of the OMR until the PLC was repaired.

L Continuous OMR manning, without a PGLD system ostector failure, provided a positive means by which to ensure operators, outside the OMR, were informed of a PGLD system actuation. The inspectors noted, however, that the comperisatory measures described in

LTO ltem 1, though effective for the current situation, did not fulfill the TSR Action L Statements.

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The inspectors discussed the findings with the PSS and the shift operations manager.

Based upon the PSS's reevaluation of the circumstances that occurred on March 1 and the current system status, the PSS declared the PGLD system inoperable and made the l required 24-hour report to the NRC in accordance with 10 CFR 76.120 and plant j procedures on March 6. The system was repaired and retumed to service on March 7.

During followup discussions with the operations staff, the inspectors determined the PSS

, onshift during the March 1 activities was not fully aware of the LTO for a failed Position 1 l PLC. The PSS was made aware of the LTO by the building staff following the PLC i' failure. The inspectors noted that the PSS staff were required to review LTOs upon initial l issuance; however, the LTO process did not require periodic PSS retraining on the LTOs.

l The absence of a retraining requirement was a weakness in the process. In addition, the j onshift PSS informed the inspectors that a review of the PGLD TSRs was done at the I time of the initial inoperability; however, the need for a functioning local alarm was not l identified. Therefore, the system was not declared inoperable, as required.

l Technical Safety Requirement 3.9.1, requires, in part, that written procedures shall be

! implemented for activities described in Appendix A, to Safety Analysis Report, Section 6.11. Safety Analysis Report, Section 6.11, Appendix A, described administrative activities, including operations, as activities that shall be implemented using procedures.

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Procedure UE2-TO-NS1032," Operability Evaluations and Resolution of Degraded and Nonconforming Conditions," Revision 0, dated December 31,1996, required, in part, the plant shift superintendent to: 1) declare a structure, system, or component inoperable -

that was unable to fulfill an intended safety function; 2) to initiate the Technical Safety Requirement Action Statements; and,3) to make required regulatory reports in accordance with Procedure UE2-RA-RE1030, " Nuclear Regulatory Event Reporting."

Procedure UE2-RA-RE1030 required,'in part, the plant shift superintendent to make reports to the NRC for disabled Technical Safety Requirement related equipment. The plant shift superintendent's failure to: 1) declare the Building C-333A Process Gas Leak Detection System inoperable following a loss of the local alarm capability on March 1; 2) direct implementation of the Technical Safety Requirement Action Statements; and,

3) make a report to the NRC of inoperable Technical Safety Requirement-related equipment, in accordance with plant procedures is a Violation of Technical Safety Requirement 3.9.1 (VIO 70-7001/98003 01).

c. Conclusions The inspectors identified a Technical Safety Requirement violation, in that, the plant shift superintendent failed to declare the Building C-333A Process Gas Leak Detection System inoperable, following a failure of the Position 1 Autoclave Programmable Logic Controller, failed to direct implementation of the Technical Safety Requirement Action Statements,

' and failed to make a report to the NRC of inoperable Technical Safety Requirement-related equipment. The inspectors also identified that the long-term order process did not ensure that plant shift superintendents were periodically retrained on long-standing orders.

08.1 Certificatee Event Reports (90712)

The certificatee made the following operations-related event reports during the inspection period. The inspectors reviewed any immediate safety concems indicated at the time of -

the initial vert >st notification. The inspectors will evaluate the associated written reports for each of the events following submittal.

Number Status Title l 33571 Open Loss of High-Voltage Process Gas Leak Detection System Annunciator in Building C-310 from Breaker Trip 33591 Open Actuation of Building C-333 Cell Floor Howler Rendering Criticality Accident Alarm System Audibility Function inoperable i t

'33637 Open Actuation of Autoclave 2 East Steam Pressure Control System in Building C-337A 4 33852 Open inoperable Process Gas Leak Detection System in Building C-333A l

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g 4 08.2 Bulletin 91-01 Reports (97012)

The certificatee made the following reports pursuant to Bulletin 91-01 during the inspection period. The inspectors reviewed any immediate nuclear criticality safety concems associated with the report at the time of the initial verbal notification. Any significant issues emerging from these reviews are discussed in separate sections of the report.

Number Date Title 33785 2/24/97 Potentially Fissile Sampling Buggies Stored in Surge Drum Room l

11. Maintenance and Surveillance M1. Conduct of Maintenance and Surveillance M1.1 Buildina C-331 Recirculatina Coolina Water Control Valve

a. Inspection Scope (88102 and 88103) l The inspectors reviewed the material condition of cascade equipment in Building C-331.

b. Observations and Findinas During routine tours of Building C-331, the inspectors observed an ongoing leak from l Cell 31.3.8 Recirculating Cooling Weter (RCW) System. The leak originated from an I RCW control valve located on a platform above the cell floor. A water wagon was staged  !

below the valve to catch the leaking water; however, structural components, located l between the valve and the wagon, allowed a portion of the water to migrate to other  ;

areas. The inspectors noted the intervening structural components were covered with j precipitate deposits and an emergency stop button, for the adjacent cascade cell, was potentially impacted by the leak.

Through discussions with operations personnel, the inspectors were informed that the  !

leak was first noticed in July 1997. At that time, a work order was written to fix the leak; l however, the current maintenance schedule indicated the leak would not be repaired until l the cell was shut down in August 1998. Following the inspectors observations, operations management reviewed the situation and determined that operation of the  !

emergency stop button was not immediately impacted by the leak due to the electrical l

box housing the button. However, operations management directed modifications to the ,

system used to contain the leak in order to preclude the potential for future material j' condition problems.

The inspectors noted that the area of the leak was toured as a part of the management-by-walking-around program; however, the potentialimpacts of the leak on adjoining equipment and the timeliness of scheduled corrective maintenance did not appear to have been recently identified or evaluated.

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c. Conclusions i

The inspectors identified an ongoing recirculating cooling water valve leak that was not well contained and had the potential to negatively impact the functionality of a cascade cell emergency stop button. The potentialimpacts of the leak did not appear to have been recently identified or evaluated by the management-by-walking-around program.

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E1. Conduct of Engineering j E1.1 Autoclave Hiah Pressure isolation System Operability Evaluation

a. Inspection Scope (88101) l The inspectors reviewed the resolution of an operability issue identified by plant staff associated with one of the autoclave safety systems. To this end, the inspectors reviewed Problem Reports EN-97-7638 and EN-98-0722 and Operability Evaluation OE-C-821-98-003, Revision 0,

• Operability of the Autoclave High Pressure isolation system During the Jetting of the Autoclave." In addition, the inspectors discussed the issues involved with plant operations and engineering staff.

b. Observations and Findinas ,

l As documented in the problem reports, plant staff identified a previously unanalyzed scenario in which the autoclave high pressur9 isolation system might not function to prevent a release to the environment as designed. The scenario involved the jetting process which was used by operators to rernove the residual steam from the autoclave before opening it. The jetting process basically pdled a vacuum on the autoclave to remove the steam from the autoclave and exhaust 't to the atmosphere above the roof.

The autoclave high pressure isolation (AHPI) systen' was designed to prevent a significant release to the atmosphere or condensate Crain by closing isolation valves associated with the lines penetrating the autoclave upon a sensed pressure of 15 psig.

Because the jetting process pulled a vacuum on the autoclave, if there was a release  !

(and thus a rise in pressure form the exothermic reaction of uranium hexafluoride with ,

steam), the release might not be detected by the AHPl system. As a result, uranium or  !

hydrofluoric acid vapors could be released to the atmosphere or to the autoclave  !

condensate drain.

l Plant engineering staff developed an operability evaluation for the feed facility autoclaves (Buildings C-333A and C-337A). The jetting process in these facilities was normally performed after the cylinder had been fed to the cascade and heeled (i.e., evacuated l to 5 pounds per square inch absolute or less). Under these conditions, the cylinder, valve, and pigtail (the possible points for a release) would be at a negative pressure loading. This was the design pressure loading for the cylinder and cylinder valve. The pigtail was designed for positive pressure loading, but was negatively and positively pressure-tested prior to initiating the feed cycle. The evaluation documented that the most likely failure point was the valve, and based on failure rates for valves documented in American National Standard 500-1984 and experience at the gaseous diffusion plants, the failure of the valve under negative pressure loading conditions was incredible i

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l Was reasonable assurance of AHPl system operability in the feed facilities insofar as a l l release was incredible during the five- to ten-minute jetting process. }

The jetting process at the sampling facility (Building C-360) was not covered by the operability evaluation because the situation was significantly different. The autoclaves in Building C-360 were normallyjetted after a cylirider had been heated (approximately 70 psig) and there was a significant amount of liquid UFe in the cylinder and potentially above the valve. As a result, plant staff concluded that the operability of the AHPI system for the sampling autoclaves could not be reasonably assured, so the PSS removed the autoclaves from service. The plant staff subsequently revised the sampling procedures

! to ensure that the Building C-360 autoclaves were not jetted after completion of the l

heating or sampling cycles. The Building C-360 autoclaves were then retumed to service. The inspectors reviewed the revised procedures and observed sampling operations. The operators observed were knowledgeable of the changes and performed the sampling operations in accordance with the revised procedures. The building manager indicated the removal of the jetting process added a small amount of time to the cylinder sampling process (less than a half-hour per cylinder), but did not otherwise impact operations in Building C-360.

c. Conclusions l

The inspectors concluded that the approach taken by plant staff to resolve an operability issue associated with jetting the residual steam from autoclaves after completion of a cylinder feeding or sampling cycle was reasonable.

E1.2 Criticality Accident Alarm System Setooints I

a. Inspection Scope (88100) i The inspectors reviewed the resolution of an operability issue identified by plant staff involving the criticality accident alarm system setpoints. The inspectors reviewed Operability Eva!uation OE-C-822-98-010, Revision 0, " Criticality Accident Alarm System (CAAS) Setpoint Channel Uncertainties," and discussed the issue with plant engineering staff.

b. Observations and Findinas While performing setpoint calculations, plant engineering staff identified a scenario in which the uncertainties associated with the current trip settings for the CAAS units could potentially drift to levels at which the CAAS would alarm at 10.27 milliroentgen per hour (mR/hr) above background instead of the 10.0 mR/hr required by the Safety Analysis Report. The additional 0.27 mR/hr would be the result of the trouble trip point drifting to its lower uncertainty limit while the alarm trip point concurrently drifted to its upper uncertainty limit. The range between the trouble and alarm trip points defined the amount of gamma radiation required to put a CAAS detectorinto the alarm state.

The operability evaluation stated that in order for the pr.blem to occur, the trouble and l alarm trip points would have to drift in opposite directions. The operability evaluation documented that there were two potential causes for trip circuit drift: temperature 12 l

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fluctuations and aging. The evaluation indicated that temperature-induced drifting of the alarm and trouble trip points would be in the same direction because the circuits were contained in the same housing. Drift due to aging was addressed by annual calibrations performed in accordance with the manufacturer's recommendations. Thus, the evaluation concluded that the CAAS modules in the plant were operable, but degraded.

To address the scenario identified, plant engineering staff were developing a modification package to change the CAAS trip points to ensure the maximum spread between trip j points, accounting for channel uncertainties, would afways be less than 10.0 mR/hr The 1 inspectors concluded that plant staff's approach to resolving the issue and the analysis in  !

the operability evaluation appeared reasonable.

c. Conclusions i

The inspectors concluded that the approach taken by plant staff to resolve an operability issue associated with the trouble and alarm setpoints for the criticality accident alarm j system was reasonable.

IV. Plant Support S8 Miscellaneous Security lasues S8.1 Certificatee Security Reports (90712)

The certificatee made the following security-related one-hour reports pursuant to i 10 CFR 95 during the inspection period. The inspectors reviewed any immediate security concems associated with the report at the time of the initial verbal notification.

Date Title 1/24/98 Pedestrian Gate 27A Discovered to be Unsecured V. Manaaement Meetina i l

X Exit Meeting Summary l The inspectors presented the inspection results to members of the plant staff and management at the conclusion of the inspection on March 9. The plant staff acknowledged the findings presented. The inspectors asked the plant staff whether any materials examined during the inspectica should be considered proprietary. No proprietary information was identified.

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PARTIAL LIST OF PERSONS CONTACTED United States Enrichment Corporation J. H. Miller, Vice President - Production l *J. A. L-barraque, Safety, Safeguards and Quality Manager 1.ockheed Martir, Utility Services (LMUS)

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• S. A. Polston, General Manager  !

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• H. Pulley, Enrichment Plant Manager l

! *W. E. Sykes, Nuclear Regulatory Affairs Manager

• S. R. Penrod, Operations Manager United States Department of Enerav (DOE _1 G. A. Bazzell, Site Safety Representative i i Nuclear Reaulatory Commission (NRC) i

• K. G. O'Brien, Senior Resident inspector

• J. M. Jacobson, Resident inspector l

• Denotes those present at the March 9,1998, exit meeting.

Other members of the plant staff were also contacted during the inspection period.

INSPECTION PROCEDURES USED

! IP 88100 Plant Operations IP 88102 Surveillance Observations IP 88103 Maintenance Observations IP 88105 Management Oversight and Controls l

IP 90712 in-office Review of Events l lP 92702 Follow-up of Events l

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I ITEMS OPENED, CLOSED, AND DISCUSSED Ooened 70-7001/98003-01 VIO Failure to Declare the Building C-333A Process Gas Leak Detection System inoperable, implement the Technical Safety Requirement Action Statements, and Make a Required 24-hour Report Closed None Discussed None LIST OF ACRONYMS USED ACR Area Control Room AHPI Autoclave High Pressure isolation CAAS Criticality Accident Alarm System CER Certificatee Event Report CFR Code of Federal Regulations DNMS Division of Nuclear Materials Safety DOE Department of Energy HF Hydrogen Fluoride LCO Limiting Condition for Operation LTO Long Term Order mR/hr Milliroentgen per Hour NCSA Nuclear Criticality Safety Approval NCSE Nuclear Criticality Safety Evaluation NDA Non-Destructive Analysis NOV Notice of Violation NRC Nuclear Regulatory Commission OMR Operations Monitoring Room PDR Public Cocument Room PEH Planned Expeditious Handling PGLD Process Gas Leak Detection PLC Programmable Logic Controller PSIA Pounds Per Square Inch Absolute PSIG Pounds Per Square Inch Gage PSS Plant Shift Supervisor RCW Recirculating Cooling Water SAR Safety Analysis Report TSR Technical Safety Requirement UFG Uranium Hexafluoride USEC United States Enrichment Corporation VIO Violation 15

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