ML20140C516

Insp Rept 70-7001/97-01 on 970107-0302.No Violations Noted. Major Areas Inspected:Plant Operations,Maint & Surveillance, Engineering & Plant Support

ML20140C516
Person / Time
Site:	07007001
Issue date:	04/10/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20140C504	List: ML20140C498 ML20140C516
References
70-7001-97-01, 70-7001-97-1, NUDOCS 9704170042
Download: ML20140C516 (33)

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U.S. NUCLEAR REGULATORY COMMISSION REGION III Docket No. 70-7001 Observation Report No. 70-7001/97001(DNMS)

Facility Operator: United States Enrichment Corporation Facility Name: Paducah Gaseous Diffusion Plant Location: 5600 Hobbs Road P. O. Box 1410 Paducah, KY 42001 Dates: January 7 through March 2, 1997 Inspectors: K. G. O'Brien, Senior Resident Inspector J. M. Jacobson, Resident Inspector W. J. Tobin, Senior Security Inspector Approved By: Tim Reidinger, Acting Chief Fuel Cycle Branch l

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97041'70042 970410 PDR ADOCK 07007001 C PDR o

EXECUTIVE

SUMMARY

United States Enrichment Corporation Paducah Gaseous Diffusion Plant NRC Observation Report 70-7001/97001(DNMS)

Authority Statement: The Department of Energy (DOE) and the Nuclear Regulatory Commission (NRC) have agreed to cooperate to facilitate the NRC obtaining information and knowledge regarding the gaseous diffusion plants and the United States Enrichment Corporation's (USEC) operation thereof through observation / inspection activities during the interim period before the NRC assumes regulatory responsibility. This report is a summary of NRC observations for the period stated. Each of the observations was communicated to the DOE Site Safety Representatives during and at the end of the observation period.

All items were discussed and reviewed with the DOE Site Safety Representatives to allow for their future followup and evaluation, as they deem appropriate.

The inspectors determined that the facility continued to operate in a safe manner. An Executive Summary follows:

Plant Operations e Weaknesses in oversight of ongoing activities contributed to the improper implementation of some equipment Limiting Conditions for Operations (LCOs). As a result, a 30 day LCO was exceeded, some equipment was removed from and returned to service without proper controls, and some LC0 prohibited activities were conducted.

e Weaknesses in procedures and plant staff knowledge of the Transition Requirements, Justifications for Continued Operations (JC0) requirements, and Technical Safety Requirements (TSRs) contributed to several apparent violations including: (1) feeding of incorrect or non-TSR authorized cylinders; (2) conducting withdrawal operations without the use of the cylinder valve closure system; (3) continuing occupancy of an area without the required Criticality Accident Alarm System (CAAS) or building howler audibility; and (4) operating the cascade at greater than atmospheric pressures without an operable Process Gas Leak Detection (PGLD) system for eight hours.

Maintenance and Surveillance e The unplanned actuation of a CAAS horn during maintenance activities was an unnecessary challenge of a safety system causing personnel to take emergency actions without cause. Corrective actions for previous unplanned actuations during maintenance and the pre-job briefing did not successfully address the problem of mis-identifying CAAS cluster cables.

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l e Weaknesses in staff and management's knowledge and adherence to the work ,

control process resulted in the inappropriate removal and return to j service of some safety systems. A large number of work control process- i related problem reports, filed during the Observation period, appeared to indicate minimal success with past corrective actions.

Enaineerina e Weaknesses in some Plant Operations Review Committee (PORC) members' understanding of the SAR, TSRs, and NRC position relative to Technical Safety (TS) clarifications contributed to the inappropriate approval of two TSR clarifications and a TSR-related JC0 clarification. These actions resulted in TSR violations, e Weaknesses in the design and implementation of a second cascade building CAAS survey program contributed to operability and response problems.

Corrective actions to a previous survey program were either not ,

developed or ineffective. )

e A self-assessment process identified a significant nuclear criticality I safety approval (NCSA) control violation and management took prompt  !

action to discontinue operations and conduct a thorough review to address the administrative control design or implementation  ;

deficiencies. However, the number of cylinders washed prior to '

identification of the problem and the ease with which a criticality control .was not implemented indicated a weakness in management's involvement in NCSA approval and implementation.

Plant Support e An UF, release from a valve in one of the Building 333 feed headers i indicated that a potential pathway existed for an intake of uranium without the knowledge, by sight or smell, of the personnel involved. 4 e The failure to meet a scheduled Nuclear Material Control and Accountability (NMC&A) Compliance Plan commitment appeared to result from plant staff rationalization of not adhering to requirements.

e Security protection afforded low enriched uranium (LEU) at Paducah was in accordance with the commitments with few exceptions .

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pETAILS I. Operations

01. Conduct of Operations

• 01.1 General Comments The inspectors observed selected activities to confirm that the facility was operated safely and in conformance with guiding programs and procedures. These activities were confirmed by direct observations, facility tours, interviews, discussions with management and staff, and

-reviews of facility records.

01.2 Operations Oversight

a. Insoection Scope The inspectors reviewed several events and situations which indicated inadequate control of operations. The specifics of each situation are discussed below.

'b. Observations and Findinas

1) Limiting Conditions for Operations (LCO) Management Inoperable Fire Protection System Sectional Valve On January 17, 1997, plant management and staff met to l discuss the current status of safety equipment and to assess  !

the impact on transitioning the cascade. facilities from Operational Safety Requirements (0SRs) to the TSRs. During l the discussions, the staff identified and then dismissed, as inconsequential, the current inoperable status of a fire protection system sectional valve. This action was taken based upon staff comments that the required LC0 action had been completed. The LC0 required that an alternate flow path be determined to be open. Although this was true, the staff did not consider the need to perform this LC0 compensatory action every thirty days, consistent with the normal surveillance requirement. This realization should have led to continued tracking of the system status.

' Topical headings such as 01, M8, etc., are used in accordance with the NRC standardized inspection report outline contained in NRC Manual Chapter 0610. Individual reports are not expected to address all outline topics, and i the topical headings are therefore not always sequential.

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Following these discussions, the inspectors asked the staff 1 to explain the bases for the fire protection system's j operability, given that no outstanding LCO Action Statements '

were identified. During this explanation, the staff determined that the plant was.indeed in an LC0 for the fire protection system. In addii,1on, this situation was expected to cont'inue until the . valve was repaired. Subsequent to the inspectors questioning, the valve and system fault were added to the newly created plant system for tracking LCOs.

The inspectors noted that the original decision not to include the equipment fault in the tracking system appeared to be based upon majority opinion versus a clear set of well documented and thought-out criteria.

Dearaded/Inocerable Criticality Accident Alarm System (CAAS) l Concurrent with the aforementioned transition of the cascade facilities from the OSRs to the TSRs, the DOE Site Safety Representative (SSR) identified that the plant staff had failed to ensure that some CAAS LC0 controls were maintained. Specifically, the plant staff had moved seal i exhaust pumps, tagged as " legacy" equipment containing  !

potentially fissile materials, within or through an area of degraded CAAS detectabilty. Movement of the equipment was not permitted under the CAAS LCO Action Statements. During followup reviews of this issue, the plant staff also i determined that the LC0 Action Statements required the condition to be rectified within 30 days of the initial out of service (005) condition. However, at the time of transition, the system had been 00S in excess of 37 days.

The inspectors noted that several groups appeared to be I aware of the CAAS 00S condition and its impact on daily l activities. However, none of these groups took i responsibility to ensure that either the system was returned to service or that the plant was placed in a mode for which coverage was not required. The difficulty in accomplishing this latter point may have contributed to the lack of aggressive tracking of the issue.

2) Removal and 'leturn-to-Service of Safety Equipment During the observation period, several events occurred which involved either the inappropriate removal from or return to service of safety-related equipment. Two of these events involved the fire protection system and the tails withdrawal building liquid UF, handling crane.

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Fire Protection System Return to Service .

At approximately 10:00 p.m. on January 7,.1997, the Plant f Shift Superintendent (PSS) declared the Building 331, fire protection sprinkler system number 30 operable following

• 4 repairs and testing. During the morning meetings held on

January 8,1997, the inspectors became aware of apparent 4

deficiencies in the paperwork used to control.the system repairs and to document successful completion of required post' maintenance testing. In followup discussions with the shift engineer and the work control staff, the_ inspectors were also informed that some of the required work package

reviews had not been completed. ,

Late in the morning, the inspectors received a copy of the work control and testing paperwork. Based upon the numerous paperwork shortcomings, including the absence of approved reviews, the inspectors inquired as to the PSS's bases for the previous operability and return-to-service determinations. The inspectors were informed that the PSS declared the system operable based upon direct discussions _

with the on-duty fire protection staff. This action appeared inconsistent'with plant policies governing work 4

control and operability evaluations. This action also precluded the PSS and shift engineer's independent review of the data and tests. i Following management review of the issue, the PSS declared the system inoperable and reinstituted the required fire watch. This action was taken approximately 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> after the system was incorrectly returned to service and the fire watch terminated. The system was returned to service  !

j approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> later following completion of the  ;

4 required paperwork and technical reviews. No significant j l deficiencies were found during closeout of the paperwork.

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3 During followup reviews of this issue, the inspectors noted that the plant problem reporting system included similar recent problem reports. These problem reports described

. other instances of staff not following the work control l system requirements. Discussions with staff indicated that  !

. some personnel did not fully understand the work process I i requirements. Therefore, these staff were not conducting i their activities in accordance with the work control I policies. l Buildino 3'5 Liauid UF, Handlina Crane Slinas Replacement On February 28, 1997, maintenance staff contacted Building 315 management to inform them of the planned replacement of the liquid UF handling crane slings. At the time of the l

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Fire Protection System Return to Service ,

4 At approximately 10:00 p.m. on January 7,1997, the Plant Shift Superintendent (PSS) declared the Building 331, fire protection sprinkler system number 30 operable following repairs and testing. During the morning meetings held on January 8, 1997, the inspectors became aware of apparent deficiencies in the paperwork used to control the system repairs and to document successful completion of required  ;

post maintenance testing. In followup discussions with the shift engineer and the work control staff, the inspectors were also informed that some of the required work package reviews had not been completed.

Late in the morning, the inspectors received a copy of the work control and testing paperwork. Based upon the numerous paperwork shortcomings, including the absence of approved reviews, the inspectors inquired as to the PSS's bases for 4

the previous operability and return-to-service determinations. The inspectors were informed that the PSS declared the system operable based upon direct discussions ,

with the on-duty fire protection staff. This action I appeared inconsistent with plant policies governing work '

control and operability evaluations. This action also l precluded the PSS and shift engineer's independent review of  !

the data and tests. '

Following management review of the issue, the PSS declared the system inoperable and reinstituted the required fire watch. This action was taken approximately 14 hours1.62037e-4 days <br />0.00389 hours <br />2.314815e-5 weeks <br />5.327e-6 months <br /> after the system was incorrectly returned to service and the fire watch terminated. The system was returned to service

approximately 5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> later following completion of the required paperwork and technical reviews. No significant deficiencies were found during closeout of the paperwork.

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During followup reviews of this issue, the inspectors noted that the plant problem reporting system included similar recent problem reports. These problem reports described other instances of staff not following the work control system requirements. Discussions with staff indicated that some personnel did not fully understand the work process requirements. Therefore, these staff were not conducting their activities in accordance with the work control policies.

Buildina 315 Liouid UF. Handlina Crane Slinas Replacement On February 28, 1997, maintenance staff contacted Building 315 management to inform them of the planned replacement of the liquid UF, handling crane slings. At the time of the 6

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i discussions, building management approved the work.

However, the time period when the work would be performed was not defined. Later that same day, maintenance staff came to the facility and completed the replacements. This effort was not coordinated with the on-shift personnel. As a result, operations staff did not remove the crane from service for the work and the crane was not held out of service until technical and administrative reviews of the complete work were accomplished.

Following maintenance completion of the crane sling work and '

prior to management discovery of the uncoordinated evolution, operations staff moved several cylinders (test, empty, and liquid-filled) without the crane work having been properly reviewed and approved. Subsequent, technical and administrative reviews of the work package did not identify any anomal 19s.

3) Exit Meeting Comments At the exit meeting, management acknowledged the observations and identified several actions taken which were expected to preclude future occurrences. These included organizational changes in the PSS and Cascade Coordinator functions, and the development / implementation of a formal process for identification and tracking of LC0's.

c. Conclusions Weaknesses in oversight of ongoing activities contributed to the improper implementation of some equipment LC0's. As a result, a 30 day LC0 was exceeded, some equipment was removed from and returned to service without proper controls, and some LCO-prohibited activities were conducted.

1.3 Technical Safety Requirement (TSR) and Safety Analysis Report (SAR)

Commitment Adherence

a. Inspection Scope The inspectors reviewed several events which led to plant operation in conflict with either the TSRs or the SAR.

b. Observations and Findinas

1) Feeding of Unapproved UF, Cylinders Twice during the oisservation period, plant staff identified incidents in which incorrect or unapproved UF, cylinders were used to feed the cascade.

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On January 17, 1997, plant staff identified that an incorrect UF, customer cylinder was used as a source of feed for the cascade. The cylinder in question was identified as E0-1737; while the cylinder intended for feeding was EM-1737. Both cylinders contained " normal" assay feed materials; therefore, no safety issue was created by this specific error.

1 Through discussions with the staff and a review of procedures, the inspectors learned that the plant

, experienced similar problems in the past. In response to these examples, the plant had developed a document to assist in the identification of customer-owned cylinders and included specific requirements in procedures. The inspectors noted that the guidance document was developed in 1981, but had not been updated since.

The inspectors reviewed the involved procedures and noted that they included only minimal guidance to address this issue. The procedures appeared focused toward controlling cylinders owned by one entity, but containing material owned by a second group. It was also noted that the procedures were categorized as " general intent" procedures, meaning that the procedures would not normally be referred during the evolution. Finally, the inspectors determined that the methods used to move materiais around the site appeared to be highly focused on the cylinder number, a quantity which was not unique

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The non-uniqueness of the cylinder numbers was neither addressed by the receiving personnel nor the site personnel responsible for authorizing materials movements. Both of a these groups, through the normal course of their activities, j

should have been cognizant of the repeated numbers.

1 S The second event involved the feeding of cylinders not

- authorized in the TSRs. In September 1996, the plant transitioned the feed facilities from DOE Operational Safety

. Requirements to the NRC TSRs. As a part of this transition,  ;

TSR 2.2, Appendix A became the controlling requirement for which cylinders could be fed into the cascade. This TSR ,

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required the operators to confirm that the cylinder weight l was less than that noted in the Appendix. The purpose of I the TSR was to ensure that cylinder heating would not result ,

in a rupture of the cylinder due to a previous excessive filling.

1 During training for restart of Building 360, the Toll Transfer facility, an operator noticed that the TSRs limited  ;

the type of cylinders that could be heated and the allowable '

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starting weights for heating. In addition, the operator  !

concluded that the limitations were different than i current practices used by the feed facilities. 1 Specifically, the operator determined that procedure l CP4-C0-CM6023, Levision 1, Change C, included cylinder types that were not referenced in the TSR. Therefore, staff unfamiliar with the TSRs, but following the procedures, could heat cylinders not authorized in the TSRs. In fact, l the operator believed that this had recently occurred.

Management review of this information confirmed the operator's speculation. Between September 5, 1996 and February 6,1997, plant staff fed 48 cylinders that were not i authorized by the TSRs. Feeding of these cylinders was a violation of the TSR.

In response to these findings, management issued a log term order to prevent further violations of the TSRs.

2) Non-Use of Cylinder Valve Closure Systems '

During the observation period, the inspectors reviewed and evaluated the plant staff's use of the cylinder valve closure system (CVCS) in the liquid withdrawal areas. The i CVCS was relied upon in the SAR and TSRs as a means by which l to terminate some accident scenarios.

As a result of the evaluation, the inspectors determined that current plant practices and procedures did not agree with the information included in the SAR. Specifically, current plant practices and procedures allow operators to disconnect the cylinder valve from the CVCS at the end of a filling cycle but before the cylinder valve is closed.

However, the SAR stated that, "The cylinder valve closer motor is used to close the cylinder valve to keep the UF.

release detection and isolation system operable until the major sources of UF, are isolated." At the time of this observation, the plant was in the process of requesting DOE authorization to operate the withdrawal facilities under the NRC TSRs. Receipt and operation of the plant under the TSRs using the current practices and procedures would have resulted in a TSR violation.

The inspector communicated the evaluation results to plant management / staff and the DOE SSR for their information and review. During these discussions, the inspectors observed that plant management and staff were not cognizant of the requirements included in the SAR. Shortly after these discussions, plant management issued a long term order to require all withdrawal facility cylinder valve closures following filling operations to use the CVCS.

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I As of the end of the observation period, management had not revised the involved operations procedure to correct the <

instructions. The reason for the delay was not immediately determined.

3) Building 331 Criticality Accident Alarm System Inaudibility On February 3, 1997, the inspectors became aware, through l discussions with plant staff, of preliminary criticality I accident alarm system (.CAAS) and building howler audibility l survey results. The February 1 and 2, 1997 surveys were conducted for the cascade buildings and incicated that the buildings included some areas where both the CAAS and l howlers could not be heard. The surveys included the i building howhrs due to plant use of this system to alert  ;

building staff of an emergent safety issue. The building howlers were also included in a DOE-approved JC0 and the NRC-Approved Compliance Plan as a compensatory measure for CAAS audibility inadequacies.

In response to the findings, additional engineering reviews were being conducted. However, plant staff failed to take action to preclude access to the areas in accordance with the Transition Requirements. Instad, the movemeit of fissile materials within the areas was halted and the areas were posted to require that staff carry a working radio when entering them. The radio was a compensatory measure applied to ensure that individuals would be notified in the event of an inadvertent criticality in other areas of the cascade building.

During discussions with plant management, the inspectors noted that personnel use of these areas concurrent with non-audible CAAS and building howlers appeared to violate the conditions of a CAAS Justification for Continued Operations (JCO). The cascade building JC0 (JC0 96-05) identified the building howlers as a compensatory measure for the degraded CAAS system and stated that the building howlers could be heard tnroughout the cascade buildings.

The audibility surveys appeared to contradict the JC0 statements. Also, the JC0 did not include provisions for the use of radios as a compensatory measure.

In response to the inspectors' findings, on February 5, 1997, plant staff filed a problem report to document recent audibility surveys of the criticality accident alarm system (CAAS) and building howlers. The problem report also indicated that the areas had been posted per procedure CP2-SF-SF1030, " Compensatory Measures Program for CAAS Alarm 10

l Inadequacies Identified in Compliance Plan' Issue 50." The inspectors reviewed the procedure and noted that.it did not appear to permit the compensatory _ measures which had been implemented by the plant.

Based upon the information included in the problem report and plant management's initial corrective measures, the inspectors and the DOE SSR raised the issue with both NRC and DOE upper management. Subsequent conversations between the DOE Regulatory Oversight Manager and plant management resulted in the implementation of JC0 96-05 and TR-required response actio~ns.. These actions included the exclusion of all personnel .from the areas of inaudibility.

4)_ Cascade Operations at Greater Than Atmospheric Pressures j On February 10, 1997, during operator turnover briefings, i the Area Control Room (ACR) operator in Building 333 1 identified that.three cells in unit I were operating at greater than atmospheric pressures. This mode of operations  :

(greater than atmospheric pressure) was currently prohibited by the TSRs, due to the inoperable status of UF, detection systems for this building. In response to the findings, the  ;

ACR staff adjusted cascade _ operations and thereby lowered

' the pressure in these cells. This action was completed- ,

approximately one hour after the TSR violation was  :

identified.  :

During followup review of the issue, the inspectors' noted i that plant data indicated the. involved cells had operated at greater than atmospheric pressures for approximately eight hours. - The start of.this period appeared to correlate with  !

cascade changes . initiated earlier in the day.

The inspectors reviewed operations procedures and practices  !

to determine if there was a reason for the extended  ;

unobserved period of TSR prohibited operation. During the i review, the inspectors-noted that operations' issued a long- .

term order (LTO) to maintain cascade pressure below ~

14.5 pounds per square' inch (psia) for Building 333, units 1  !

and 6. This order was based upon seismic concerns. The Plant Shift Superintendent also maintained the Building 333, unit 1 and 6 process gas leak detection (PGLD) system as inoperable due to problems with PGLD configuration -

management, testing, and control.

l During a review of the involved procedures, the inspectors noted that the procedures did not incorporate the information from either the LTO or the PSS PGLD status. The i; absence of this information appeared inconsistent with plant expectations that these conditions would exist for an t

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.- . l extended period of time, potentially years. In addition, the procedures used to monitor and track cascade pressure, and associated UF, detector operability status, were not consistent in their reference to units or cells.

Finally, the inspectors noted that during normal operations, the cascade building staff and the cascade coordinator have 4

computer displays which track cell pressures. However, neither of these systems was used during this event to 4 ensure that cell pressures were maintained within the TSR 4 limits.

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c. Conclusions Weaknesses in procedures and plant staff knowledge of the TRs, JC0 requirements, and TSRs contributed to several apparent violations including: (1) feeding of incorrect or non-TSR authorized cylinders; (2) conducting withdrawal operations without use of the cylinder valve closure system; (3) continuing occupancy of an area without required CAAS or building howler audibility; and,  ;

(4) operating the cascade at greater than atmospheric pressures without an operable PGLD system for eight hours.

01.4 Cylinder Vibration During Heat-up  ;

a. Inspection Scope l

The inspectors reviewed the occurrence of and plant staff response to severe cylinder vibration during heat-up and prior to cascade feeding.

b. Observations and Findinos On January 12, 1997, while heating a full UF, cylinder, operations staff heard an extremely loud noise from the feed autoclave area and felt vibration through the Building 333A facility. In

. response to these conditions, the staff reviewed the status of cylinders heating at the time and observed a pressure spike on the cylinder heating in the 3 South autoclave. Based upon these observations, the operators shut off steam to the autoclave and secured the heating cycle.

Subsequent. to the event, a visual review of the autoclave cylinder position was performed. Building staff and the shift engineer noted that the cylinder had shifted approximately seven degrees during the aborted heat cycle. No damage to either the cylinder or the feed pigtail were noted. Based upon this information, the PSS halted the initiation of new heat cycles for all cylinders from the involved supplier.

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The inspectors noted that this event was a repeat occurrence of cylinder vibration and rotation during heating and feeding to the cascade. The most recent previous event was documented in Observation Report 70/7001-96007. In that report, the inspectors concluded that non-aggressive operator followup of noises and vibration from the autoclave area allowed significant cylinder rotation and longitudinal movement. In contrast, during this event, the operators' actions appeared prompt and appropriate to the unexpected vibration and noises.

Plant operations and engineering staff continued to assess these events to determine the root causes and ensure long-lasting corrective actions,

c. Conclusions The operators response to anomalous noises and vibrations during cylinder heating appeared conservative and safety-focused.

08. Miscellaneous 00erations Matters 08.1 (Closed) Follow-Vo item 70-7001/94001-01: Inadequate understanding and documentation of initial operability determinations.

This item was opened to track corrective measures implemented to resolve an observed lack of rigor in the process for making an informed operability determination upon discovery of a non-conformance for safety systems and establishing compensatory actions. Since the item was opened in August 1994, the plant implemented a new problem reporting system. The current problem report (PR) form required a description of the problem, immediate actions, and recommended actions. At the bottom of the form, the PSS was required to make and document the initial operability assessment as well as identify any compensatory actions. In addition, plant staff developed a procedure to provide guidance to the PSS on how to determine operability given degraded or non-conforming conditions (UE2-TO-NS1032).

Routine observation by the inspectors, during the observation period, indicated that the PSS utilized the PR system to make initial operability assessments in a more rigorous manner.

08.2 (Closed) Follow-Up Item 70-7001/94001-02: Lack of guidance on the use of alarm silencing capability.

An operator involved in a training exercise did not notice an alarm condition due to a prior silencing of some cascade unit alarms. The inspectors reviewed procedure OPS-19, " ALARM RESPONSE GUIDELINES AND STATUS CONTROL," which was revised subsequent to the exercise. The revised procedure and the " Quality of Operations Plan" developed by 13

j plant staff provided formal guidance on logging 00S alarms and any compensatory actions needed, as well as responding to alarms. The policy emphasized a " blackboard" approach, i.e., responding to and l

clearing alarms in a prompt manner. i 08.3 (Closed) Follow-Vo Item 70-7001/94001-04: Weak knowledge and use of procedures. i l

The inspectors noted an apparent low level of staff knowledge and l adherence to procedures during the conduct of routine and abnormal i activities. The facility is now operating under Technical Safety 1 Requirement (TSR) 3.9 which specifies the program for the scope, review and approval of, as well as temporary changes, to procedures. In addition, the facility implemented a new procedure on the use of ,

procedures (UE2-PS-PS1034) and developed a problem reporting system to i track procedure adherence problems. j 08.4 (Closed) Follow-Vo Item 70-7001/94001-05: Lack of a spare parts control program.

The facility had a number of spare parts bins and areas in which parts for safety-related equipment maintenance were kept in an uncontrolled manner. The Materials Management Organization has assumed ownership of ,

the spare parts program and has initiated a process to gain control of  !

all safety-related spare parts. The recent spares management program aims to establish a single point of contact for all spare equipment and to review all current field equipment and material for traceability and serviceability. The facility also developed a new procedure (CP2-MA- ,

SR1032) specifying the control of material and equipment at the site. '

08.5 (Closed) Follow-Up Item 70-7001/94001-10: Ineffective implementation of ,

the non-conformance reporting (NCR) system.

The inspectors noted that non-conformance reports were not issued for I certain program, process, or performance weaknesses. In addition, non-conformance reports were not reviewed to develop generic issues and provide broad corrective actions. Since August 1994, the plant has i replaced the previous NCR system with a new problem reporting (PR) l system. Plant staff perform monthly analyses of the quality and quantity of the problem reports. In addition, the plant instituted a process to screen prs for generic /significant issues and elevate those to Significant Conditions Adverse to Quality. Senior plant management reviews and approves corrective actions for these types of prs.

08.6 (Closed) Follow-Vo item 70-7001/94002-01: Ineffective response to degraded and non-conforming conditions requiring operability assessments.

The plant approved and implemented a new procedure for operability determinations (UE2-T0-NS1032) which provides detailed guidance to the PSS on performing operability assessments for safety structures, 14

1 e s systems, and components (SSCs). The procedure also provides guidance on when to obtain an Operability Evaluation from the engineering function to assist the PSS in determining the operability of an SSC. The inspectors noted that implementation of these procedural requirements resulted in more rigorous operability assessments.

08.7 (Closed) Follow-Up Item 70-7001/95002-01: Investigation and corrective actions for three exothermic-reaction events at the Building 310 cylinder burp station.

The plant conducted an investigation of three events during June-July 1995 in which exothermic reactions / abnormal pressure spikes occurred during burping operations. The Event Report (PAD-1995-0058) identified three different causes for the events: copper oxide on new pigtails which initiated the reaction, degraded Teflon valve seats which introduced foreign material into the burp station gas stream, and foreign material introduced during sampling evolutions performed by laboratory personnel at the burp station. To prevent recurrence, the plant instituted pre-use inspections for pigtails at the burp station, a shelf-life study and requirements to periodically change out valve seats for the burp station valves, and a more rigorous sampling procedure to ensure foreign material was not introduced into the system during sampling at the burp station positions. No exothermic reaction events have occurred at the burp station since the July 1995 time frame.

08.8 (Closed) Follow-Vo item 70-7001/95002-02: Use of unreviewed temporary operating checklists (TOCs) to conduct activities.

Plant quality assurance staff identified a generic issue in that required TOC reviews were not being completed prior to the issuance of f

the TOC. TSR 3.9 " PROCEDURES," defines the certificant's requirements for use of procedures under the NRC. The program requires written procedures for safety-related activities, operator actions to prevent or mitigate an accident, and to implement programs included in the SAR.

The TSR also includes requirements for review and approval of procedures and temporary changes to procedures. Based on the TSR on procedures approved by NRC, safety-related activities are required to be conducted in accordance with reviewed and approved procedures, and checklists used to facilitate procedure implementation would be included.

08.9 (Closed) Follow-Up Item 70-7001/95005-01: Lack of formal process for making operability determinations.

This item is closed based on the new procedure discussed in Section 08.6.

08.10 (Closed) Follow-Up Item 70-7001/96005-01: Uncertainty over CAAS coverage of the Building 331 to Building 335 process gas tie line.

The inspectors reviewed KY/G-578, " CRITICALITY ACCIDENT ALARM COVERAGE OF THE INTERBUILDING TIE LINES AT THE PADUCAH GASE0US DIFFUSION PLANT,"

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approved February 28, 1996. The document provided the results of SCALE-4.3 calculations which demonstrated that the radius of coverage for .the CAAS modules at PGDP for a criticality external to the process buildings was at least 800 feet. With this radius of coverage, current clusters in the process buildings provided coverage for all of the inter-building tie lines which could contain UF. enriched to 1.0 w/o or more, including the Building 331 to Building 335 tie line.

II. Maintenance and Surveillance M1. Conduct of Maintenance and Surveillance M1.1 General Comments During routine tours of plant facilities, the inspectors observed the general materiel condition of plant equipment and some in-progress maintenance activities. The inspectors also reviewed some maintenance records and maintenance-related nonconformance reports. The focus of the observations was to assess the overall performance of maintenance activities relative to approved procedures, guides, and industry codes or standards.

M1.2 Criticality Accident Alarm System (CAAS) Actuation Durina Maintenance 1

a. Insoection Scope i The inspectors reviewed the circumstances surrounding the l unplanned actuation of a criticality accident alarm system (CAAS) during routine maintenance. The inadvertent CAAS alarm resulted in the evacuation of the affected facility.

b. Observations and Findinas On January 10, 1997, at approximately 1:30 a.m., instrument mechanics performed a semi-annual module replacement for the CAAS "B" cluster in Building C-335. Each CAAS cluster consisted of three detector modules tied together in a two-of-three logic controller. During post-maintenance testing for the installed modules, the mechanics disconnected the wrong cable from the control box. The lifting of the wrong lead caused the "B" cluster horn to sound. The mechanics immediately tried to reset the horn, but were not able to do so. In response to the CAAS horn, personnel in Building C-335 evacuated the building and proceeded to the designated assembly point.

A problem report (PAD-97-0171), generated after the incident, indicated that inadvertent soundings of the CAAS horns during maintenance activities was a continuing problem. In fact, after a November 1996 review of a previous incident, plant staff recommended color-coding the involved CAAS cluster cables to ensure ready identification and to prevent the wrong cable from 16

)

l

. . - . ~. . . ._ _. . .. __. _ __ . __ _ _

being lifted. However, the recommendation had not been acted upon prior to the most recent horn actuation. The inspectors also

noted that the " Maintenance Pre-Job Briefing Checksheet" included a requirement that mechanics be briefed on the precautions to *

, prevent false alarms.

c. Conclusions The unplanned actuation of a CAAS horn during maintenance activities was an unnecessary challenge of a-safety system causing .

personnel to take emergency actions without cause. Corrective l action for previous unplanned actuations during maintenance and the pre-job briefing did not successfully address the problem of 4

mis-identifying CAAS cluster cables.

M1.3 Maintenance Process Implementation and Rigor J

a. Inspection Scope The inspectors reviewed a number of maintenance work packages associated with operability issues and problem reports which appeared to indicate maintenance process implementation problems.

b. Observations and Findinas t

, During the review of several issues discussed in Section 0.1, the inspectors determined that each issue appeared to include some problems with the associated work packages. These problems included: 1) inadequate staff documentation of field activities;

2) non-performed, incomplete, or untimely supervisory review of work packages; 3) staff, supervisory, and management weak  :

l understanding of and adherence to the work control process; and 1

4) paperwork movement problems associated with the site size.

i Significant among the identified problems was staff and j management's weak understanding of and adherence to the work control process. During the observation period several examples were identified in which safety-related equipment was inappropriately removed or returned to service because of an inadequate understanding of the work control requirements.

Systems involved included the liquid withdrawal cranes and fire i protection piping. None of the examples appeared to create an immediate safety concern.

During discussions with plant staff, the inspectors were also informed that some staff and management were unfamiliar with recent changes to the work control process, and others did not agree with the level of reviews required by the process. This lack of knowledge and disagreement with the process may have contributed to both the problems and their recurrence.

17 e

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l The inspectors noted that these or similar issues were also well  :

documented in the plant problem reporting system. However,.  !

corrective measures had either not yet been implemented or were not effecting the necessary changes in performance. During the observation period, the problem reporting system included several dozen work control problems with apparent root causes similar to i those events discussed in Section 01. '

Late in the Observation period, the inspectors noted that increased sensitivity by the PSS staff and operations line ,

supervision to these performance issues. As a result, several  !

pieces of equipment remained 00S while the maintenance paperwork was gathered and reviewed. Several of these reviews identified anomalous conditions requiring further evaluation and at times corrective action prior to returning the system to service. ,

c. Conclusions Weaknesses in staff and management's knowledge and adherence to the work control process resulted in the inappropriate removal and return to service of some safety systems. A large number of work control process-related problem reports, filed during the observation period, appeared to indicate minimal success with past corrective actions. ,

M8. Miscellaneous Maintenance Matters M8.1. (Closed) Follow-Vo Item 70-7001/94001-06: Weak process for evaluating documenting routine surveillance activities and resolving surveillance anomalies.

The facility implemented a new work control process which requires the plant shift superintendent and shift engineer to review and approve surveillance work packages to ensure that test results meet the appropriate acceptance criteria prior to declaring a system operable.

The packages include a history sheet to capture any deficiencies or anomalies noted during testing for follow-up and trending purposes.

Irrespectiva of the examples included in this observation report, the overall process was improved and its implementation quality increased.

M8.2 (Closed) Follow-Vo Item 70-7001/94001-08: Inconsistencies in installation of safety system rupture disks.

Plant staff identified numerous rupture disks which did not conform to design specifications because of weak configuration control. Rupture

, disk settings for safety systems were provided as a part of the I application and approved in the TSRs. Safety limits and limiting control settings are specified in the TSRs for safety systems pressure relief as well as associated surveillances.

18

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1 M8.3 (Closed) Follow-Vo Item 70-7001/95003-01: Spray booth pump pit level alarm system and floor pan not maintained as safety systems.

The pump pit level alarms and floor pan integrity were not part of the

, NCSA controls for the spray booth. The revised NCSA (No. 3973-06, '

approved April 26, 1996) included the level alarms and floor pan and specified monthly surveillances for the level alarms and annual surveillances to assure floor pan integrity.

, III. Enaineerina El. Conduct of Engineering i El.1 General Comments Throughout the observation period, the inspectors observed facility engineering activities', particularly the engineering organization i

.. performance of routine and reactive site activities, including I

identification and resolution of technical issues and problems. l El.2 Technical Safety Requirement Clarifications j

a. Inspection Scqp_q  ;

The inspectors attended several Plant Operations Review Committee (PORC) meetings and reviewed the results of some of these l meetings. '

b. Observations and Findinas During the observation period, the PORC met for both routine and i special meetings. The inspectors focused their review of PORC j 3

meetings on actions taken relative to the TSRs. Many of these '

meetings resulted in TSR clarifications. The requested

, clarifications were for the purpose of resolving staff questions i on how to implement the new or soon-to-be enacted TSRs. Three of

, these clarifications involved: 1) the liquid withdrawal area emergency valve closure system; 2) the cascade cell trip system; and 3) the criticality accident alarm system (CAAS).

The inspectors observed the discussions and reviewed the materials presented at the PORC meetings for each of the above noted clarifications. The inspectors observed that the staff presentations to the PORC often did not fully discuss the concerns or issues with the particular TSR. Additionally, the materials presented to the PORC often did not completely explore the safety 4- or regulatory basis for the TSR requirements. In spite of these shortcomings, PORC approved each of the requested clarifications, some with modifications.

19 4 )

1 Following the PORC approvals, the inspectors assessed the safety and regulatory appropriateness of these actions. Based upon these assessments, the inspectors determined that each of the PORC-approved TSR clarifications either contradicted or changed the TSR wording, the meaning, or the intent of the TSR requirements.

Specifically, the withdrawal area TSR clarification had the effect of redefining operating modes. This redefinition permitted safety- related equipment to be taken 00S prior to the point previously assumed during the certification process.

Additionally, the change contradicted statements made by the plant in response to specific questions posed during the certification process. These statements were included in the SAR, but were not identified during the review and evaluation process for the change.

The cell trip TSR clarification excluded some cascade cells from the applicability statement of the TSR. The exclusion of these cells from the TSR requirements would allow the plant to restart cells previously shutdown in a manner other than that assumed in the TSR and the DOE SAR. The shutdown manner assumed in the TSRs and the DOE SAR was intended to ensure the operability of the cell trip button, should it be needed in an emergency. The final version of this TSR clarification deemed acceptable a method for testing the cell trip function that was not allowed by the TSR.

The CAAS clarification was to a current DOE OSR and JC0; however, the same JC0 was included in the NRC-approved Compliance Plan for a soon-to-be effective TSR. The clarification extended the application of one JC0 to the areas covered by another JCO. It also allowed the use of compensatory measures previously not allowed under the applicable JCO. In essence, it changed the wording and meaning of JC0s approved for use in conjunction with TSRs.

The inspectors communicated each of these findings to plant management. Following these discussions, the plant took differing actions. In the case of the withdrawal area TSR clarification, the plant immediately rescinded the clarification.

In the example of the cell trip TSR clarification, the plant requested, by letter dated January 24, 1997, a formal written opinion from the NRC. The NRC provided this interpretation by letter dated February 20, 1997. The NRC letter indicated that the clarification changed the applicability of the TSR and therefore was not permissible. During the time between PORC approval of the TSR clarification and the date of the NRC's formal opinion, the plant used the clarification to startup several cells. The plant's actions in this regard were therefore a violation of the TSR, 20

l l

1 Finally, relative to the CAAS JCO, the plant requested a revision j to the JCO.from the DOE to approve the actions they currently 1 implemented. The DOE rejected these actions as being inconsistent with the JC0 and the intent of the forthcoming TSR. As a result, the plant implemented the OSR-required actions for the involved system anomalies.

The inspectors also interviewed several of the PORC members.

During these discussions, the inspectors determined that some of the members did not have a full. understanding of the information

. included in either the SAR or TSRs. In addition, several of the members relied heavily on the presenters to provide all of the information pertinent to the proposed questions,

c. Conclusions Weaknesses in some PORC members' understanding of the SAR, TSRs, and NRC position relative to TSR clarifications contributed to the inappropriate approval of two TSR clarifications and a TSR-related JC0 clarification. These actions resulted in TSR violations.

i El.3 Criticality Accident Alarm System (CAAS) Audibility Survey

a. Insoection Scope The inspectors reviewed the preliminary results of an ongoing CAAS audibility survey and its use to disposition operability issues.

{,

b. Observations and Findinos During the observation period, the plant staff conducted CAAS  !

,~

audibility surveys which were planned and coordinated by the  ;

system engineering group. The surveys were conducted to verify l the audibility of the CAAS alarms and building howlers throughout  ;

, the cascade buildings. Based upon past problems with the CAAS alarms in the cascade buildings, the plant was operating under a JC0 which required use of the building howlers. The JC0 stated that the howlers could be heard throughout the cascade buildings.

l I As a result of these surveys, the staff identified areas in the  ;

j cascade buildings where the howlers could not be heard. These  !

j findings are discussed in Sections 01.3 and E1.2. ,

4 As a followup to these findings, the inspectors reviewed recently completed surveys and the survey program. The inspectors '

determined that the program was almost completely qualitative and was of questionable repeatability. Qualitative aspects included:

1) the non-identification and limited documentation of the equipment operating at the time of survey, and 2) the lack of

t 1

sound-level measuring equipment. The program acceptance criteria were also neither rigorously defined nor tied to operations a requirements. These shortcomings appeared to contribute to the

. operability and response problems discussed in this report.

Finally, the inspectors noted that this was the second CAAS survey '

program conducted in the past two years. The previous survey was also primarily qualitative and limited in its repeatability.

Inspector comments on this past survey were provided to the plant; however, either ineffective or no corrective actions were implemented.

c. Conclusions Weaknesses in the design and implementation of a second cascade building CAAS survey program contributed to operability and i response problems. Corrective actions to a previous survey ,

program were either not developed or ineffective. '

l E3.4 Nuclear Criticality Safety Control Violations

a. Inspection Scope The inspectors reviewed the circumstances surrounding two

. reportabic events for nuclear criticality safety approval (NCSA)  ;

control design and/or implementation inadequacies for activities e conducted in the Building C-400 decontamination facility.

b. Observations and Findinas i

On February 18, 1997, plant criticality safety staff identified i that an NCS control for the operation of the cylinder wash station had not been properly implemented (NRC Event Report No. 31810).

The discovery occurred during an NCS review of NCSA controls implementation for Building C-400. In particular, the approval governing operation of the cylinder wash station (NCSA 3973-04) a required two chemical operators to independently verify that a j cylinder to be washed had never contained UF, enriched to 1.0 w/o or greater. The NCSA required that the operators review the '

nuclear material control and accounting (NHC&A) database records

. for enrichment back to the cylinder's last hydrostatic test date. l l In practice, chemical operators had only been reviewing the NMC&A i release for wash letter because they did not have access to the l database records specified in the NCSA. l As a result of the implementation problem, 22 cylinders were washed for which the independent review of records was not ]

i accomplished. Thus, UF, enriched to 1.0 w/o or greater could .

potentially have been in a cylinder. (The certificant's i application defines uranium enriched to less than 1.0 w/o as not  ;

representing a criticality hazard.) The failure to implement the J 22 d

1

,=

• independent records review was identified as a loss of double contingency for this particular scenario (washing cylinders with material enriched to 1.0 w/o or greater from fillings previous to the last before wash). The current NCSA did not provide controls for handling enriched material at the cylinder wash station.

Samples taken from the wash solution storage tanks, drain pan, hydrostatic station wet vacuum and water tank after the event yielded results which. indicated that only material with enrichments near normal enrichment (approximately 0.71 w/0) had been introduced at the station.

Following the event, the staff undertook a broad review of NCSAs containing administrative controls in an attempt to identify any other operations with similar weaknesses. On February 22, this review identified that the NCSA for operation of the Building C-400 decontamination spray booth (NCSA 3973-06) specified a control for independent verification of records for parts to be decontaminated which was not truly independent (NRC Event Report No. 31834). The administrative control required that two people independently verify the assay and location of equipment removal with the originator of the request-for-decontamination tag used to track equipment. The assay for the spray booth was limited to 2.75 w/o. Plant NCS staff identified that the reliance on one person (the originator) for the historical assay determination was a single-point failure mode and thus represented single contingency control of enrichment for equipment introduced into the spray booth. TSR 3.11.5 requires that the plant maintain double contingency control for all operations which do not have a specific TSR governing them.

Although the spray booth was not in operation at the time of discovery, had the plant been under TSR 3.11.5, the reliance upon a single contingency for control of equipment historical assay would have been a TSR violation.

The inspectors noted that previous observation reports had identified continuing concerns with implementation of NCSA controls. Althogh the deviations from the NCSA have been minor, the number of recurring failures appeared to be a precursor to the significant problem identified for the cylinder wash station. In addition, the rigor of the reviews of selected NCS controls appeared suspect given the relative ease with which the assay control was violated and the identification that assay control for the spray booth was singly contingent. The number of cylinders washed before the NCS staff identified the violation, and the failure to provide any mechanism by which chemical operators could retrieve NMC&A records, indicated a weakness in Operations' management involvement in the cylinder wash process. Finally, the documentation of independent verifications required by NCSA administrative controls appeared inadequate to ensure that thorough, independent reviews had been accomplished.

I 23 l l

l

g In response to these events, the Department of Energy (DOE) issued a confirmatory action letter (CAL) to confirm the NCS review of all NCSAs with administrative controls and to confirm that restart of operations be approved by the Operations Manager and Enrichment Plant Manager. As operations were not restarted prior to NRC assuming regulatory authority for USEC operations at PGDP on March 3, the NRC issued a CAL on February 28 (CAL No. Rill 003). The CAL confirmed that a review of NCSAs would take place; that the certificant would evaluate the root cause of the events and revise NCSAs, as needed, to ensure double-contingency control of all operations; to revise procedures and conduct training for the revised controls; to perform an outside review of the NCS program focused on the independence and effectiveness of NCSA controls; and, to notify the Senior Resident Inspector prior to restart of the cylinder wash and spray booth operations.

c. Conclusions The plant's self-assessment process identified a significant NCSA control violation for which plant management took prompt action to discontinue operations and conduct a thorough review for administrative control design or implenientation deficiencies.

However, the number of cylinders washed prior to identification of the problem, and the ease with which a criticality control was not implemented, indicated a weakness in management's involvement in NCSA approval and implementation.

E8. Miscellaneous Enaineerina Matters E8.1 (Closed) Follow-Vo Item 70-7001/94001-07: Lack of a rigorous modification control process for a seal exhaust pump modification.

Development of the modification package was not sufficiently detailed to ensure all possible operating modes were considered. As a result, a section of the piping associated with the pump imploded.

Since the event, the engineering department developed a new procedure:

UE2-T0-EG1031, " MODIFICATION DESIGN CONTROL." The inspectors noted that this procedure provided the engineering staff with detailed guidance on performing the design, verification, testing, and walkdown of a plant modification.

E8.2 (Closed) Follow-Up Item 70-7001/94003-01: Track details of revised scope and structure of the Plant Operations Review Committee.

The former plant safety committee did not include senior plant management, minimum qualifications requirements, a quorum requirement, or requirements for subjects needing a safety committee review. As part 24 1

of the certification process, the certificant submitted TSR 3.10 " PLANT

. OPERATIONS REVIEW COMMITTEE," which was reviewed and approved by the NRC. TSR 3.10 identifies the requirements for PORC membership, qualifications, meeting frequency and quorum, functions, responsibilities, and records.

E8.3 (Closed) Follow-Vo Item 70-7001/95005-03: Lack of procedural guidance for performing engineering evaluations in support of operability determinations. 1 The certificant developed procedure (UE2-T0-NS1032, " Operability Evaluations and Resolution of Degraded and Nonconforming Conditions,"

effective April 1996) which provides instructions for the PSS on when a '

formal operability evaluation from engineering should be obtained to support an operability determination. In addition, the procedure i outlines the format and content of operability evaluations for the engineering department.

• IV. Plant Suonort P4.1 Release From Building C-333 feed Header

a. Inspection Scope The inspectors reviewed the circumstances surrounding a release from the "F" feed header in Building C-333 whict, resulted in unusual intakes of uranium by two plant personnel.

ts. Observation and Finding 1 On January 26, 1997, a first line manager (FLM) in Building C-333 noted smoke near the ceiling of the feed headers on the cell floor at the south side of the building. The headers transfer UF, from the feed manifold in the Building 333A feed facility to the lower  ;

cascade feed point. As a result of the visible smoke, the FLM i initiated an evacuation of all personnel into the area control t room (ACR) and notified the PSS that a hazardous materials (HAZMAT) response to a release was needed. The PSS directed the HAZMAT team to measure the hydrogen fluoride (HF) concentrations when operators alternately valved in the two feed headers in operation: the "B" and "F" header. Based on the results, the PSS concluded that the leak was in the "B" header. After the "F" header was valved in and negative samples were obtained by a '

HAZMAT team, the PSS sounded the all clear.

i Subsequent to the HAZMAT response, approximately 1.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> later, j the FLM notified the PSS that a vague haze could be seen near the '

ceiling lights of the cell floor. However, the FLM could not l smell (a generally reliable means of detecting HF) anything and, in fact, thought there might be equipment smoldering somewhere. .

The PSS responded to the cell floor with the FLM and began taking 25

)

i samples for HF as they worked their way back toward the feed headers. All samples were negative. The PSS and FLH spent l approximately 0.5 hours5.787037e-5 days <br />0.00139 hours <br />8.267196e-6 weeks <br />1.9025e-6 months <br /> on the cell floor trying to identify the source of the haze without respiratory protection (which was not j required by procedure since a release had not been confirmed and l 4

samples were negative). The PSS finally identified a minor l 1

release was occurring from a valve on the "F" feed header by '

visually identifying an " updraft" from the valve. .However, the samples taken were still negative. The PSS then had all personnel i leave the floor and had the ACR operator valve off all the  ;

building feed headers. -

Subsequent analyses of urine samples left by the personnel involved in the event indicated that the PSS and FLM received

minor intakes of uranium which were unusual for plant staff, but {

well below any action levels (urine concentrations on the order of 200 micrograms per liter or less). While the plant staff ,

4 responded in accordance with approved procedures and no <

significant exposures to HF or uranium resulted, the inspectors '

noted that the intake appeared contrary to the common " rule of thumb" that the olfactory irritation level for HF was low enough to prevent anyone from staying in an area with a UF, release and j potentially being exposed. A potential explanation for the  ;

intakes was that the hydrolysis products (HF and uranyl i oxyfluoride) of the UF, and air moisture were produced above the heads of the personnel involved. The particulate uranyl oxifluoride then fell out of the haze while the HF gas rose upward,. creating a scenario that personnel could be receiving a continuous intake, albeit at very low concentration::, without smelling or detecting any HF. j i

c. Conclusions  !

A release from a valve in one of the Building C-333 feed headers indicated that a potential pathway existed for an intake of uranium without the knowledge, by sight or smell, of the personnel involved.

P4.2 Compliance Plan Issue Not Met

a. Insoection Scoce i I

The inspectors reviewed a Department of Energy finding that an action scheduled for completion under Issue A.2 of the Compliance l Plan had not been met by the date specified: December 31, 1996.

b. Observations and Findinas  :

A DOE team, performing closecut inspections of Compliance Plan  !

issues with scheduled completion dates before transition to NRC regulatory authority (March 3,1997), identified that a commitment 26

in Issue A.2 for certain calibrations had not been performed as of the end of February. The plant nuclear matertal control and accounting (NMC&A) staff had apparently believed that the commitment could be satisf!ed by using calculations instead of calibrations. This was another of a series of NMC&A issues (documented in previous Observation Reports) which appeared to result from a rationalization of not adhering to requirements at the NMC&A staff and first-linn manager level. The inspectors noted that as of March 3,199/, the systems requiring calibration were not in compliance with the Compliance Plan requirement. The  ;

senior manager responsible for NMC&A indicated the affected systems would remain 00S until the issue was resolved.

c. Conclusions The failure to meet a scheduled NMC&A Compliance Plan commitment ,

appeared to result from plant staff rationalization of not  !

adhering to requirements.

S1.1 Safeguards Program Implementation

a. Scope The inspector reviewed the Paducah Safeguards Program to determine whether physical security requirements were implemented in l accordance with the requirements of the Physical Security Plan (PSP), Chapter 5, " Fixed Site Requirements for Special Nuclear Material of Low Strategic Significance," and Chapter 9, " Reporting Safeguards Events."

The inspector also reviewed implementation of site security procedures.

b. Observations and Findinas To determine if adequate protection was being afforded the low enriched uranium (LEU), the inspector toured the Controlled Access Area (CAA) and observed the integrity of the fence, gates, and the vehicle barrier. Fences, gates, and the vehicle barrier were intact and not affected by erosion or disrepair. Personnel were identified, registered, badged, and escorted as required.

Clearances and the need for access were being verified. Packages were visually inspected by security officers at the entrance to the CAA as required by the PSP. Officers at the CAA vehicle gate performed random entry / exit search of vehicles.

All the officers were armed with a handgun and equipped with a radio providing communication to the Security Station and to each other. The inspector witnessed radio tests and concluded that there were no " dead spots" in communications within the CAA. The communications capabilities of the operator in the Security 21

e . . ,

1 Station were witnessed and found to be performing the intended redundant function.

The use of locks and seals was inspected. The inspector observed that additional security as well as accountability was being i provided through the locks and seals. The inspector also I interviewed officers posted and on patrol and found them to be knowledgeable of their duties and responsibilities. Security l procedures were located at the appropriate locations.  !

The inspector reviewed security procedures (SP): SP 2200, " Access i Control," SP 2201, " Patrol Operations," SP 2211. " Post l Operations," and SP 2221, " Unusual Event Response" to determine ,

whether they provided adequate guidance for security officer )

duties. During the review of procedure SP 2200, the inspector l noted that the PSS was authorized to waive certain security l measures in the event of an emergency. Measures that could be waived included: the unimpeded entry of an ambulance, access or exit of "public affairs personnel" such as the Joint Public i Information Center personnel during an emergency, and issuance of Emergency Thermoluninescent Dosimeters to "... VIPS..." to facilitate their accesa at any time. The inspector informed facility management that authority to waive NRC requirements must be limited to only those emergency situations where health and  ;

safety of personnel is involved. Facility management informed the inspector and the NRC Senior Resident Inspector that the issue would be corrected.

The inspector toured the CAA production facilities, including )

Buildings 310, 315, 333A and 360 for the purpose of observing the processing and storage of '.EU. In Building 360, the inspector i noted that kilogram quantities of LEU could be in use or in 4 storage within sample containers, but that the building was not within the CAA during day shift. This occurred when the security force unlocked the electronically controlled gate (CAA barrier) in front of the building and secured another CAA barrier behind the building to facilitate day shift truck traffic. In an attempt to compensate for this configuration, armed officers remotely operated the electronic truck traffic gate from a security position located approximately 50 yards from the gate. The inspector observed that all incoming and outgoing truck drivers routinely checked in or checked out at this security post. The inspector also observed that the electronic gate did not completely close. The opening appeared wide enough for an unauthorized individual to enter and, potentially continue unchallenged into Building 360.

The PSP contains contradictory language relative to this situation. Paragraph 5.1 of the PSP states that all LEU is located within the CAA; however, another portion of that paragraph states that Building 360 is returned to the CAA barrier at night which connotes that it is treated otherwise during the day shift.

28

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i The inspector referred facility management to 10 CFR Part 73.67(f) '

which requires all LEU to be stored or used within a CAA.

Facility management stated that it was their intent to adhere to

both the PSP and 10 CFR Part 73. Building 360 was brought within j the CAA by March 3, 1997.  !

4 During the tour of the facility, the inspector often referred to i 4

Figure 2:1, Revision 1, of the PSP which displayed the CAA barrier i location. The inspector observed that, contrary to Figure 2:1, the CAA barrier incorporated parts of the shipping and receiving  ;

area of Building 720. Additionally, Building 743 and the a

adjoining CAA barrier were not accurately reflected in Figure 2:1. .

The inspector referred facility management to paragraph 5.2.1 of I

the PSP which specifies that only Building 100 is to be part of i the CAA barrier. Facility management stated that it was their i intent to submit a revision to the PSP with accurate drawings and

better descriptions of the CAA barrier location.

The inspector also observed that as part of corrective actions to an event that occurred on November 14, 1996, involving an individual driving through an open CAA gate, facility management installed 11 large STOP signs on the gate facing outward.

Additional training of the officers was conducted to alert them to be more sensitive to manually opening gates that are rarely in use.

To determine if security officers were adequately trained and qualified, the inspector reviewed lesson plans, test results and three individual training records. The inspector noted that one of the three individuals had not received the number of hours of annual training required by the PSP. Facility management ,

completed a review of all individual training records and 2 identified several other officers who were lacking the required hours. Facility management informed the inspector of their intent to provide all required PSP training prior to March 3, 1997.

During a review of Security Event Reports, facility management's.

proposed method of reporting safeguards events was reviewed.

Chapter 9 of the PSP contains the commitment to 10 CFR 73.71

" Reporting Safeguards Events." The inspector determined that the comrnitments in Chapter 9 would not guarantee compliance with Appendix G of Part 73, section 1(a)(1) which applies to theft or unlawful diversion, and requires a one hour notification to the NRC followed by a 30 day report. The inspector informed facility management that additional attention to the requirements in Appendix G was warranted to assure compliance.

29

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j. c. Conclusions i Through observation, interview, independent verification and I records review the inspector verified that the protection being

afforded the LEU at this' facility, with the exception of the CAA j for Building 360, was complying with PSP commitments. ,

i P8. Miscellaneous Plant Support Matters I P8.1- (Closed) Follow-up Item 70-7001/94001-09: Weak iciementation of radiological controls program.

1 Facility staff were not aware of the full scope of the radiological i

' controls within the process buildings or the bases of the program.

Since August 1994, the plant has significantly restructured its 2 radiological controls program including its radiation work permit (RWP)

program and its radiological training program. . Radiologically F controlled areas of the process buildings have been surveyed, roped off, i and posted. Although isolated incidents of failure to adhere to RWPs or

! postings have occurred, the inspectors noted an increased general awareness and understanding of the radiological controls program on the part of plant personnel, including using the problem reporting system

for deficiencies identified in the field.

P8.2 (Closed) Follow-Op Item 70-7001/95002-03: Unexpectedly high potential radiation exposure during the first quarter of 1995.

{ A thermoluminescent dosimeter (TLD) reading of 19.0 rem for shallow dose i equivalent and 8.8 rem for deep dose equivalent was obtained for a plant i

employee (machinist) for a routine quarterly TLD analysis. The

. Radiation Protection Manager initiated a detailed investigation and i series of surveys, the results of which were documented in a report

! '(KY/A-573). The results indicated that the employee could not have ,

!- received an occupational exposure of the magnitude read from the TLD. - '

The TLD exposure most likely resulted from low-energy x-rays or

thallium-204, for which the employee did not have an opportunity for

exposure onsite during the quarter. As a' result, the plant appropriately calculated a dose for the employee based on his past dose history and the doses measured for employees in his work area who ,

performed like work.

4 V. Manaaement Meetinas  !

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XI. Exit Meetina Summary j 4

The inspectors met with facility management representatives and the DOE  ;

Site Safety Representatives throughout the observation period, on '

February 13,-and on March 3, 1997. The likely informational content of l 1 the observation report with regard to documents or processes reviewed

, was discussed. Information highlighted during these meetings is

contained in the Executive Summary.  !

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.--.. . .. -. . . . . . - . - - - . - = . - . . - . . - . - . - - . - .-. -.- - .. - .. .- .

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i j During the February 13,'1997 meeting', facility management responded- i I

affirmatively to each safeguards finding identified, and advised the  :

inspectors that corrective actions would be taken or were in progress.  !

, No dissenting comments were voiced. No classified or propietary l information was identified.

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PARTIAL LIST OF PERSONS CONTACTED 4

l 1

Lockheed Martin Utility Services (LMUS) i i

• S. A. Polston, General Manager

• H. Pulley, Enrichment Plant Manager 3

• W. E. Sykes, Nuclear Regulatory Affairs Manager i

• S. R. Penrod, Operations Manager  ;

• C. Hicks, Site and Facility Support Manager United States Enrichment Corporation

]

• J. H. Miller, Vice President - Production

• J. M. Brown, Engineering Manager

• J. A. Labarraque, Safety, Safeguards and Quality Manager  !

f United States Department of Enerav (D0E)

• G. A. Bazzell, Site Safety Representative l'

Nuclear Reaulatory Commission (NEL1  ;

i

• K. G. O'Brien, Senior Resident Inspector  !

• J. M. Jacobson, Resident Inspector 2

• W. J. Tobin, Senior Security Inspector i

• Denotes those present at either the February 13 or the March 3, 1997 exit ,

meetings. l Other members of the plant staff were also contacted during the observation  ;

period.

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ITEMS UDENED. CLOSED. AND DISCUSSED ,

Opened None

]

Closed

• Follow-Up Item 70-7001/94001-01 l

• Follow-Up' Item 70-7001/94001-02  ;

• Follow-Up Item 70-7001/94001-04  :

• Follow-Up Item 70-7001/94001-05

• - Follow-Up Item 70-7001/94001-07

• Follow-Up Item 70-7001/94001-08 ,

• Follow-Up Item 70-7001/94001-09 .

• Follow-Up Item 70-7001/94001-10 l

• Follow-Up Item 70-7001/94002-01  :

• Follow-Up Item 70-7001/94003-01 l

• Follow-Up Item 70-7001/95002-01 l

• Follow-Up Item 70-7001/95002-02

• Follow-Up Item 70-7001/95002-03

• Follow-up Item 70-7001/95003-01

• Follow-Up Item 70-7001/95005-01

• Follow-Up Item 70-7001/95005-03 3

• Follow-Up Item 70-7001/96005-01 i Discussed None Certification Issues None I

33

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