ML20199H657

Insp Rept 70-7001/97-08 on 970826-1007.Violations Noted. Major Areas Inspected:Plant Operations,Maintenance & Surveillance,Engineering & Plant Support

ML20199H657
Person / Time
Site:	07007001
Issue date:	11/20/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20199H615	List: ML20199H612 ML20199H645 ML20199H657
References
70-7001-97-08, 70-7001-97-8, NUDOCS 9711260185
Download: ML20199H657 (25)

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1 U.S. NUCLEAR REGULATORY COMMISSION .

- REGION lli

.. . t Docket No: 70 7001-Report No: 70-700 l/97008(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: August 26 th ough October 7,1997 Inspectors: K. G. O'Brien, Senior Resident inspector

.J. M. Jacobson, Resident inspector J. R. Kniceley, Physical Security inspector, Region ill R. G. Krsek, Fuel Cycle inspector, Region Ill' Approved By: P. L. Hiland, Chief Fuel Cycle Branch

/11260185 971120-PDR ADOCK 07007001 C PDR w

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. EXECUTIVE

SUMMARY

United States Enrichment Corporation

' Paducah Gaseous Diffusion Plant ^

NRC Inspection Report No. 70-7001/97008(DNMS) -

'i PJLant Operations

. The inspectors determined that a Technical Safety Requirement violation occurred, in I that, operations staff failed to document the reason for and the special entry and exit points of procedures used during restart of the purge cascade. As a result, the direct causes for increased uranium releases could not be clearly determined. In addition, the  !

inspectors identified that an intcmalinvestigation of the loss of purge cascade event l failed to identify sither the procedural problem or 'he bases for (ther problems experienced during restart of the purge cascade. (Section 01.2)

. The inspecurs determined that a Technical Safety Requ!mment violation occurred, in that: 1) management used an outdated procedure during their oversight of the response to a small fire in Building C-310; and,2) numerous controlled manuals of operations procedures were not sufficiently maintained to ensure that only current approved procedures were available and used for operations activitW I.Section 01.3)

. The inspectors identified the failure to file timely problem repons for certain events and deficiencies meeting the problem reporting threshold as a violation of Technical Safety Requirement 3.9.1 and the problem reporting procedure. (Section 01.5)

Maintenance and Surveillance

. The inspectors concluded that plant staff had accomplished the actions identified in Confirmatory Action Letter Rill-97-003 for restart of the Building C-400 spray booth. No additional concems were identified as a result of the review. (Section M1.1) l Enaineerina

. The inspectors identified a violation as a result of the plant Jtaff's management of an as-found condition. Specifically, the plant staff failed to evaluate an as-found condition involving the purge cascade, as required by 10 CFR 76.68(b), until after the inspectors brought the issue to their attention. (Section E1.1)

. Inconsistent implementation of some controls or requirements relied upon for criticality safety appeared to result from a lack of clear communication between the nuclear criticality safety staff and those personnel responsible for implementation. (Section E1.2)

Plant Support

. Reponses to events observed by the inspectors were effective and in accordance with the Emergency Plan ineffective communication of the release point to the incident commander in one instance delayed identification and sampling of the area involved with the minor release. (Section P1.1)

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r The inspectors identified on inconsistency in the issuance of problem reports for.

radiological protection deviations by plant health' physics personnel. . (Section R1.1)

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The inspectors noted that the internal dosimetry program was effectively:

implemented through thorough and extensive dose investigations in addition to good employee participation in the routine urinalysis bloassay program. (Section R1.2)

Inspector review of a sampling health physics technician training and qualification records indicated that the technicians had completed applicable mandatory training ,

and were properly qualified. (Section R1.3)-

- i The results of a physical security inspection indicated that the protection being afforded low-enriched uranium was in compliance with the commitments in the Physical Security Plan.- (Section S1.1) 3 3

' DETAILS

1. Operations

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01. Conduct of Operations' 01.1 StatusmfEan1J88100)

During the inspection period the plant staff increased the power load from approximately 1100 megawatts (MW) to approximately 1450 MW. No abnormal operations were noted as a result of the gradualincrease in load.

01.2 Loss of Pume Cascade

a. Inspection Scope (88100)

The inspectors reviewed the plant staff's assessment of and corrective actions to preclude recurrence of the July 1997 purge cascade loss,

b. Observations and Findinos The inspectors reviewed the completed assessment and corrective actions plan developed for the July 1997 loss of the purge cascade event. During the review, the inspectors noted that the investigation neither identified nor evaluated the basis for several decisions, made during the restart, which appeared to contribute to increased uranium releases. Speci5cally, the investigation did not evaluate the basis for restarting of the purge cascade without: (1) a clear understanding of the cause for the trip;(2) procedural guidance for the restart activity, a complex and seldom performed entution; (3) using operators experienced in the evolution; and, (4) documentation of the initial or the final system alignments.

The inspectors discussed the restart evolution with the facility operations manager and the system engineer. These individuals were also members of the investigation team.

During the dialogue, the facility operations manager indicated that the quick purge cascade restart, following the trip, was implemented based upon a historical approach to plant operations. The historical approach did not require a completc understanding of all the event root causes before retuming the system to service, instead, plant management relied upon a follow-up investigation to determine all the event root causes and to define long-term corrective actions. The facility operations manager also indicated that management was aware of the changes in staff experience levels and the absence of procedural guidance by which to conduct the restart. As a result, facility operations management reviewed the existing procedures, to ensure that restart activities would not violate existing procedures, and involved the system engineer in structuring a restart sequence of activities. The facility operations manager then directed the operators to use Topical headinos such as 01, MS, etc., are used in accordance with the NRC standardized inspection report outline contained in NRC Manual Chapter 0610. Individual reports are nat expected to address all outline topics, and the topical headings are therefore not always sequential.

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4 the partial procedures to conduct the restart activities, concurrent with some system engineer oversight. The facility operations manger stated that neither the partial procedures, used to conduct the restart, nor a listing of the applicable sections of the procedures was maintained following the completion of the restart.

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During discussions regarding the failure to document either the initial or finalYystem alignment, the inspectors leamed that the facility staff were unaware of a system alignment change recommended by the system engineer to preclude recurrence of the purge cascade loss. The recommended system alignment change was implemented during restart activities; however, operations staff later retumed the system to the normal alignment. The latter a'.tiori was taken based upon successfulleak testing of adjacent valves. The leak testing was performed to rule out another possible initiator for the event.

However, the testing did not resolve all of the concems which led the system engineer to request off-normal alignment. The system engineer indicated that he was unaware that the system had been retumed to the normal alignment.

Technical Safety Requirement 3.9.1 requires, in part, that procedures shall be implemented for activities described in the Safety Analysis Report, Section 6.11, Appendix A. Safety Analysis Report, Section 6.11, Appendix A identifies procedures management (use) as an activity that shall be performed in accordance witt, approved procedures. Procedure CP-2-PS-PS1038,"Use of Procedures at PGDP,' requires, in part, that the first line manager must approve and document the reason for, and the special entry and exit points for use of, a partial procedure. The failure to document the reason for and the special entry and exit points of procedures used to restart the purge cascade was a Violation of Technical Safety Requirement 3.9.1 (VIO 70-7001/97008-01),

c. _C_q1clusions The inspeciors determined that a Technical Safety Requirement violation occurred, in that, operations staff failed to document the reason for, and the special entry and exit points of, partial procedures used during restart of the purge cascade. As a result, the direct causes for increased uranium releases could not be clearly determined in addition, the inspectors identified that an intemalinvestigation of the loss of purge cascade event failed to identify either the procedure use problem or the bases for other problems experienced during restart of the purge cascade.

01.3 Pome Cascade Cell Motor Fire

a. Lnispection r Scope (IP 88100)

The inspectors observed the plant staff's response to a small motor fire which resulted in the tripping of a purge casc.ide cell.

b. Observations and Findinas On September 2 at 11:28 a.m., Building C-310 purge cascade cell 4 tripped while onstream. During investigation of the tiip, the operators noted that the stage 6B motor had grounded, starting a small fire. The operators responded to the fire per off-normal Procedure CP4-CO-OIN3016a, "C-310 Building Fire."

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'The inspectors observed the response actions taken by both the building operations staff and plant management. The inspectors noted that the area control room operators implemented the procedure-required immediate response steps and were actively tracking further evaluations of the situation by the sito emergency squad During follow-up discussions w:th the operators, the inspectors determined that theoperators were not aware if an off-normal procedure existed for the shutdown of a high speed purge casaade cell. This type of procedure could be used to respond to any off-normal condition which resulted in an unplanned perturbation of the purge cascade.

The inspectors observed that the managers, located in the main control room, closely followed building operator actions and carefully considered the need for other actions, either in Building C-310 or for the remainder of the cascade. As a part of their oversight efforts, the managers reviewed and referenced procedure CE 16, "Emercency Operations in Building Fires." The procedure directed the cascade coordinator to take specific actions depending upon the size of the fire. The inspectors concluded that the managers oversight of the activities was appropriate.

Following completion of the building operators' and management's response to the fire, the inspectors identified that the two facilities used different procedures to direct response activities. The inspectors compared the procedures and noted that the actions directed by the two procedures were consistent. However, the procedure used in the main control room, CE-16, was not a valid procedure, in that, it had been superseded in July 1997. The current valid procedure was the procedure used by the building operators. During follow-up discussions with the managers, the inspectors determined that none of the managers were aware of the proccdure differences. As a follow-up to the observations, the plant staff issued a problem report, removed the incorrect procedure from the controlled manuals in tho main control room, and initiated action to tum over maintenance of the manuals to the site document control staff.

Approximately a month after the initial finding of outdated procedures in the main control room manuals, the inspectors performed a cursory review of off-normal and emergency operations procedures maintained in the Building C-310 area control room. The inspectors noted that the controlled manuals included several canceled or outdated off-normal and emergency operation procedures. A subsequent review of the controllud manuals maintained in the main control room identified additional examples of canceled or outdated off-normal and emergency operations procedures in response to the findings, plant management initiated a detailed review of other procedure manuals to define the problem scope and necessary corrective actions. The review documented a large number of controlled manuals which included outdated or canceled operations and other procedures. Final corrective actions for the issue were being developed at the end of the inspection period.

Technical Safety Requirement 3.9.1, requires, in part, that written procedures shall be maintained, and implemented for activities described in Safety Analysis Report, Section 6.11, Appendix A. Safety Analysis Report, Section 6.11, Appendix A identifies operations, including system procedures addressing startup, shutdown, normal operations, and abnormal operations as activities requiring written procedures. The failure to use approved procedures to direct the manager's response to the motor fire in Building C-310 and to maintain current written procedures for operations, including 6

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. ' system startup, shutdown, normal operations, and abnormal operations was a Violation of Technical Safety Requirement 3.9.1 (VIO 70-7001/97008 02),

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c. Conclusionsi The inspectors determined that a Technical Safety Requirement violation occurred, in that: (1) management used an outdated procedure during their oversight of the response to a small fire in Building C-310; and, (2) numerous controlled manuals of operations procedures were not sufficiently maintained to ensure that current approved rmedures were available and used for operations activities.

01.4 Technical Safety Reautrement inconsistencies

a. Inspection Scope (88100)

While performing follow-up inspection for issues raised at the Portsmouth Gaseous Diffusion Plant (PORTS) and Paducah Gaseous Diffusion Plant (PGDP), the inspectors .

Identified two inconsistencies in the PGDP Technical Safety Requirements (TSRs).

b. Observations and Findinas Wet Air inleakaoe Technical Safety Reauirement The inspectors noted that the applicability statement for TSR 2.4.4.4, " CASCADE WET AIR INLEAKAGE," did not include MODE Cascade 3, that is, cascade equipment that wcs not in use and was at a uranium hexafluoride (UF ) negative. However, the TSR Action Statements indicated that the TSR applied for MODE Cascade 3, since a deposit of uranium-bearing material, requiring moderator control for nuclear criticality safety, could be present in equipment which was not in use. The similar PORTS TSR included the MODE equipment not in use in the applicability statement.

The inspectors were concemed that the associated surveillance requirement for quarterly surveys of cell and unit drops might not encompass those systems which were not in use, but could potentially contain an unsafe deposit. Disccssions with plant staff performing the surveys indicated that, in fact, cascade cells and units which were not in use, continued to be surveyed for deposits and no known unsafe deposits had been located during previous surveys. At the conclusion of the inspection period, plant staff indicated that a request was being prepared for submittal to Nuclear Regulatory Commission (NRC) to revise the TSR to include MODE Cascade 3 in the applicability statement.

Fffezer/ Sublimer Coolant Ovemressure System Technical Safety Requirement 2.4.3.4, "R 114 COOLANT OVERPRESSURE SYSTEM,"

provided Limiting Control Settings for coolant systems associated with the cascade cells, boosters, and purge and evacuations pumps, but not for the freezer /sublimers (rated at

- 150 pounds per square inch gauge (psig]). Technical Safety R*quirement 2.4.3.3,

" FREEZER / SUBLIMER R 114 VENT LINE MANUAL BLOCK VALVE," required that the

. coolant block valve be sealed in the open position in order to ensure the availability of the pressure relief system in the presence of an overpressure situation. The TSR requirement that the block valve be opened to the pressure relief system without a 7

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' corresponding TSR requirement for the pressure relief system appeared inconsistent .

since putting an inappropriately sized rupture disk in the relief system could also disable

-the system.

The inspectors subsequently discovered is response to a question (TSRQ192), p.osed during the certification process, describing why the freezer / sublimer relief system was not included in the TSRs; Specifically, the answer stated that given the pressure and temperature of the recirculating cooling water system which acted as the heat source and -

sink for the R-114 coolant system, the ma timum coolant pressure (approximately 60 psig)

- could never achieve the 150-psig rupture pressure for the system. Nevertheless, the inspectors noted that it appeared illogical to have a TSR requirement on the manual block valva for the piping to the rupture disks, but no TSR requirements for the rupture disks.

The laspectors noted that the freezer / sublimer coolant overpressure systems were controlled procedurally to ensure appropriate rupture disks were used.

The inspectors will track resolution of the above noted inconsistencies in the TSRs as an Inspector Follow-up item (IFl 70-7001/97008-03).

c. Conclusions Two inconsistencies in the Paducah Gaseous Diffusion Plant Technical Safety Requirements were identified for follow-up by the inspectors.

01.5 Untimely Problem Reports

a. Inspection Scope f88100)

The inspectors reviewed the problem reporting system for items identified by the inspectors during tours of plant facilities and routine inspections as well as self-disclosing events.

b. Observations and Findinas The inspectors noted that timely problem reports were not provided to the plant shift superintendent (PSS) for a number of issues or events which occurred during the inspection period. Specifically, problem reports were not filed by the end of shift for the following:

. On September 22, the inspectors noted that the Building C-300 PSS logs included an entry indicating that a conductivity alarm and subsequent autoclave containment had occurred on September 19 in Building C-333A for autoclave 4S.

Shortly after the conductivity alarm, a high condensate alarm (a Q safety system) occurred due to excessiva water in the autoclave. The facility operators and the PSS judged the high condensate alarm to be

• invalid," thus not reportable, because the safety function (containment) had already been completed in response to the conductivity alarm (a non-safety systerr) prior to the safety-system alarm.

The cause of the conductivity alarm was cracks in the probe housing for the conductivity probes which resulted in excessive leaking of the steam condensate 8

4 from the housings. Building C-333A operators indicated this problem had occurred in the past. The inspectors noted that a problem report was not filed with the PSS as a result of the event. Section J.2.C of Table 6.9-1 of the Safety Analysis Report,

• Event notification and reporting criteria applicable to USEC,"

stated: " Invalid Q safety system actuations are documented and evaluate,d through the Problem Reporting System."

i On September 23, during a tour of the Building C-310 withdrawal area, the inspectors observed a fissile high-efficiency particulate air (HEPA) vacuum cleaner in a spacing pan with the vacuum hose loosely wrapped around the vacuum. The nuclear criticality safety (NCS) posting on the vacuum stated:

" Vacuum hoses shall not be wrapped around this vacuum." The inspectors discussed the issue with an operator in the area and were told that the condition of the hose appeared to be in violation of the posting and the associated nuclear criticality safety approval (NCSA), OEN-04. A problem report for the issue was not filed with the PPS until September 29 (Problem Report PAD-97-5515).

In subsequent discussions, NCS staff indicated that the wording on the posting was meant to convey that hoses should not be tightly wrapped around a fissile vacuum cleaner. The staff further indicated that the as-found coadition of the hose in Building C-310 was acceptable because of the calculations performed for the cvaluation supporting the NCSA. Nevertheless, the inspectors determined that the discrepancy between the restriction on the posting and the operator's understanding indicated an NCSA or procedural problem.

t On September 24, at approximately 10:00 a.m., during a tour of Building C-400, the inspectors identified that the " Originator" section of a check sheet for independent verification of assay had not been signed. At the time of the tour, the checksheet, used to confirm the enrichment of equipment to be decontaminated, already included assay data for the equipment. A problem report for the m'! sing signature svas not filed until September 25 (Problem Ruport CO-97-5433).

. During the aftemoon of Septembe,' 24, the inspectors identified that plant staff had not performed a rated load test for two NCH-35 cylinder haulers, used to transport cylinders containing solid UF, across the site, following enodifications.

Mod;fications to the cylinder hauler boom extensions were performed in September and October 1996. The NCH-35 cylinder haulers were rated at 35 tons. Step 6.1 of Procedure CP4-TO-Ol1087, Revision 0,' INSPECTION AND TESTING OF MOBILE CRANES," datea October 27,1995, required quality inspectors to perform a rated load test on all new, altered, modified, or repaired cranes. Plant staff filed a problem report (Problem Report SF 97-5429) the moming of September 25 after questions by the inspectors, but continued to use the NCH-35 cylinder haulers following identification of the problem. After receiving the prob:em report, the PSS took the NCH-35 haulers out of service.

When the first cylinder hauler was load-tested on September 26, the cylinder hauler failed the test Subsequent discussions with the manufacturer indicated the failure was caused by a hydraulics problem. Following maintenance to correct the problem the haul 6rs passed the load test.

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Me inspebtors noted that the NRC observed only a small fraction of the activities s.nd '- -l Levants which occur routinely at the plant. Therefore, the inspectors could not draw ai ,

l global conclusion about the timeliness with which problems were identified and reported =1

at the plant. Nevertheless, the number of inspector-identified late problem reports

- appeared to indicate a negatsve trend. The inspectors noted that some staff.wora- -

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. attempting to analyze or solve problems before formally identifying the issues through the ,

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J problem reporting system - even for certain issues which, upon identification, cleariy met .

- the criteria of the problem reporting procedure. In all cases, problem reports were filed after questions were raised by the inspectors.-

1 A similar concem over filing problem reports is discussed in Section R1.1. The cencem arose because health physics technicians indicated problem reports were not always written for radiation work permit or procedure violaf!ons.

Technical Safety Requirement 3.9.1 requires, in part, that written procedures sNil be prepared and implemented to cover the activities described in Safety Analysis deport,

. Section 6.11.4.1, Section 6.11.4.1 of the Safety Analysis Report stated: "As a minimum,

' a procedure is required for any task that is described in, or implements a commitment--

that is described in, the SAR..." Section 6.8.2.4 of the Safety Analysis Report, " Problem Reporting," stated: "All plant employees have the responsibility to write problem reports

" . on safety, operating, and noncompliance items... Corrective actions are tracked through the plant's corrective action program." Procedure UE2 HR-Cl1030," PROBLEM REPORTING," Revision 0, dated April 10,1996, identified " false alarms or false actuations related b safety system items" and " violations of, or deviations from, programs, policies, and procedures or deficiencies which could cause safety, operability,

- or reportability concems"_ as problems requiring a problem report (PR), Step 6.1.3C ,

required that the problem report form be delivered to the PSS as soon as practical, but

- always prior to the end of the shift. The failure to file protslem reports for the deficiencies and alarms noted above by the end of the shift (within either 8 or 12 hours1.388889e-4 days <br />0.00333 hours <br />1.984127e-5 weeks <br />4.566e-6 months <br /> of the time of discovery) was a Violation of Technical Safety Requirement 3.9.1 (VIO 70-7001/97008-

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As an immediate corrective action, plant staff began a series of articles in'the plant '

newspaper ar.d employee bulletin designed to raise the staff's awareness of when a problem report was required to be filed and how to properly file a problem report.

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c. Conclusions -

The inspectors identified the failure to file timely problem reports for certain events and deficiencies meeting the problem reporting threshold as a violation of Technical Safety Requirement 3.9.1 and the problem reporting procedure.

08. Miscellaneous Matters 08.1- (Closed) Certificatee Event Report 32906: Building C-337 Unit 4, Cell 8 freezer / sublimer (F/S) three-way recirculating water (RCW) valve failure when the F/S was operated from the cold-standby to sublime mode.

After the initial report, the certificatee performed a test of the three-way valve. The valve operated properly. However, the RCW pump motor coupling had failed and the RCW 10 .[

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1 pump was not running. The certificatee identified the failed motor coupling as the cause l of the slow sublime rate for the freezer / sublimer wh9n the operator changed from the j cold standby to sublime mode. The result of either an inoperable RCW pump or an l inoperable three-way valve would be a slow sublime rate as noted by the operator at the time of the event. The F/S monitoring system did not have any alarm to indicate,that the RCW pump was not operating.

The RCW pump and three-way valve were both components of the F/S high-high weight trip safety system identified in the Safety Analysis Report and safety system bour.dary manual. As such, a discovery that either component was inoperable was reportable.

However, a subsequent Engineering Evaluation, EV-C-821-97-041, Revision 0,

" Freezer / Sublimer RCW Pump Operation," dated Sopiember 20,1997, documented that the RCW pump was not required to be operating in order for the safety function to be achieved. The evaluation documented that the function of the safety system trip, te close the inlet valve to the F/S to preclude further solidificaFon of material and open the outlet and vent valves to low-presure piping and allow slow sublimation of UF6 bas.k to the cascade, was not affected by the RCW pump not running in the tripped mode (modified hot standby). This was primarily due to the fact that the intermediate heat transfer (R 114) loop pump was off in the modified hot standby mode, so the status of the RCW pump did not affect the F/S sublimation rate in this mode.

As a result of the follow-up inspection of the three-way valve and the subsequent evaluation of the RCW pump, the certificatee revised the Safety Analysis Report to indiccte that the RCW pump was not required for the safety system to perform its intended function and retracted the event report for an inoperable safety t.ystem. The inspectors reviewed the evaluation and Safety Analysis Report change and concluded that the subsequent event report retraction appeared reasoncble. The inspectors had no further questions and this item is closed 08.2 (Closed) Certificatee Event Report 32909: Discovery of a tripped RCW pump for the Building C-333 Unit 2, Cell 2 freezer / sublimer. The certificatee subsequently retracted this event report based on the information provided in Section 08.1. The inspectors had no furthet questions and this item is closed.

08,3 (Closed) Certificatee Event Report 32947: Failure to properly space two 5.5-gallon fissile waste drums containing used cylinder valves.

The failure by opeutions personnel to follow the approved nuclear criticality safety approval for handling potentially fissile waste while splitting the conteots of a 3G-gallon accumulation drum into separate waste drums was identified by the certificatee. The root cause was the failure to follow the implementing procedure for handling waste drums which required that the drums be placed in approved physical restraints to ensure two-foot spacing for interaction control. The certificatee immediately spaced the drums in the approved physical restraints in order te reestablish interaction control. Additional training on the proper handling of waste drums was provided to the involveo personnel. This non-repetitive, certificatee-identified and corrected violation is being treated as a Non-Cited Violation, consistent with Section Vll.B.1 of the NRC Enforcement Policy. The inspectors had no further questions and this item is closed.

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08.4 (Closed) Certificatee Event Report 32967: Discovery of three tripped recirculating cooling water pumps for freezer /sublimers in Buildings C-333 and C 337. The certificatee subsequently retracted this event report based on the information provided in Section 08.1. The inspectors had no further questions and this item is closed.

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08.5 Cer1ificatee Event Reports (90712) -

The certificatee made the following cperations-related event reports during the inspection period. The inspectors reviewed any irnmediate safety concems indicated at the time of the initial verbal notification. The inspectors will evaluate the associated written mports for each of these items following submittal.

Number Status Title 32852 Open Safety System Failure of Autoclave Head-to-Stiell Gasket on Position 2 South of Building C-333A (CER 70-7001/97006 05) 33038 Open Safety System Actuatien of a Building C-333A Process Gas Leak Detector Head following Minor Release (CER 70-7001/9700846) 33039 Open Safety System Actuation of Two Building C-333A Autoclave Eteam Pressure isolation Systems following Retum to Service (CER 70-7001/97008-07)

II. Maintenance and Suiveillance M1. Conduct of Maintenance and Surveillance M1.1 Buildina C-400 Sprav Boqth

a. 10 spection Scope (88102. 88100)

The inspectors reviewed the maintenance and surveillances, nuclear criticality safety approvals, and procedurt s associated with the restart of the Building C-400 spray booth pumuant to NRC Confirmatory Action Letter (CAL) No. Rlll-97-003, dated February 28, 1997 (see Inspector Follow-up item 70-7001/97002-15). In puticular, the inspectors reviewed the following:

(1) Procedure CP4-CU-CH2108, " Operation of the C-400 Spray 500th;"

(2) Nuclear Criticality Safety Approval No. 3973-06, "C-400 Spray Booth;" and (3) Letter from United States Enrichment Corporation, "Paducah Gaseous Diffusion Plant (PGDP), Docket No. 70-7001, Restart of C-400 Spray Booth Operations,"

dated August 19,1997.

The inspectors also perfoemed a detailed walk-through of the area and procedures with spray booth operators, 4

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. b. Observations and Findinas I The inspectors' review of surveillances of items required by the nuclear criticality safety approval (NCSA) for nu lear criticality contr al, such as pump pit water level alarms, airflow sensors, and containment pan integrity, indicted all required surveillarices< were l current. Sightglass material deficiencies and leaks for the spray booth solution storage i tanks, identified earlier in the year by the inspectois and the plant staff, were corrected prior to restart.

1 A selected review of the goveming procedure did not identify any failures to meet the  !

double contingency principle for fissile operaticas, independent verifications of the origin ;

and assay or mass of equipment to be decontaminated in the spray booth were accomplished as required by the NCSA. Operators wert. knowledgeable of the NCSA and procedural requirements and were able to readily identify the relied upon criticality safety controls.

Based on the review, the inspectors considered that plant staff had accomplished the actions identified in the CAL for restart of the spray booth. These actions included a review of the administrative controls for the spray booth, revision of the NCSA, revision of the procedures, providing appropriate training to operators, and notification of the NRC prior to restart.

c. vonclusions The inspectors concluded that plant staff had accomplished the actions identified in CAL Rill-97-003 for restart of the Building C-400 anray booth. No additional concems were identified as a result of the review.

111. Enaineerina E1. Conduct of Engineering E1.1 As-Found Evaluations

a. Lrtspection Scope (88_103 The inspectors reviewed the identification and resolution of as-found purge cascade and recirculating cooling water configuration anomalies.

b. Observations ano Findinas i

. Purae Cascade During review of the intemalinvestigation conducted by piant staff of the July 1997 purge cascade loss event, tb3 inspectors noted that inadvertent valve operation" was identified as one of two possible causes for the event, The investigation team speculated that an unplanned opening of the large orifice recycle control valve to the recycle header, a

- common header between the small and large orifice purge lines, could result in excessive unmonitored purging through the small orifice discharge line. However, th hypothesis 13 1

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. ' required the presence of some other anomaly to modify the system design .ow characteristics, in an attempt to validate the scenario, the system engineer performed a walkdown of the involved piping and valves. On August 13, the system engineer identified 2 0 o-inch valves in a pipe run originally designed to include only three-inch valves. The engineer performed preliminary calculations which indicated that the size and location of the valves, combined with an inadvertent opening of the large orifice recycle control valve, would change the system flow characteristics. The presence of the two-inch valves was documented iri a problem reporI which included a recommendation that she system should be restored to the original design configuration.

The inspectors reviewed the problem report and noted that an operabildy determination was not performed and that a non-conforming condition form was not attached to the problem report. The problem retort writeup did not provide justification for either why an operability evaluation was not necessary or why a non-conforming condition form was not attached. Compensatory messe-as, recommended by the system engineer to preclude recurrence of the purge cascade loss, the basis for the walkdown, were documented on the problem separt.

During discussions with the system engineer and the facility operations manager, the inspectors noted that the safety impact of continued operations with the anomalous as-found condition, that is, two-inch valves in a system designed to operate with three-inch valves, had not been evaluated. The system engineer indicated that engineering planned to restore the system to the original design configuration; therefore, a safety evaluation wa2 not performed. However, the inspectors noted that br,tb ritle 10 of the Code of Federal Regulations (CFR), Part 76.68 and plant precedures aquired an evaluation of as-found conditions, in part, to determine if an unreviewed safety question existed. These evaluations were crucial when the system was not removed from service to immediately restore the system to the original design configuration. At the time of the inspector's observations, the system was inservice and a schedule for the design modification, to retum the system to the original design, had not been defined. The system engineer stated that other system valves had been closed and caution tagged to preclude a repeat of the conditions that led to the previous purge cascade loss event, pending the system design change.

Subsequent to the discussions held with the inspectors, the facility operations manager reviewed the current system alignment and the system engineer performed a safety evaluation of the as-found condition. Following a review of the current system clignment, the facility operations manger informed the system engineer that no system valves were caution-tagged closed as a result of the purge cascade loss event. The facility operations manager indicated that operations staff were not aware that the system engineer had relied on the closure of any valves to preclude recurrence of the event. The inspectors noted that the original problem report, which documented discovely of the as-found condition, clearly stated that a manual block valve in the large orifice recycle line had been closed to prevent erratic purge operations.

On September 5, the system engineer determined, based upon calculations included in a safety evaluation, that the two-inch valves could not have caused or convibuted to the purge cascade loss event. Instead, the system engineer determined that an error was made in the original calculations which indicated that the two-inch valves modified the system flow characteristics. Based upon the revised calculations, the two-inch valves 14

were determined to have better flow characteristics than the original three-inch valves and were considered adequate to perform the necessary functions relied upon in the Safety /,r:alysis Report. Therefore, one of the two scenarios proposed in the final investigativi report as the cause for the purge cascade loss event was improbable.

Title 10 of the Code of Federal Regulations, Part 76.68 (b) requires, in part, tEatihe certifdtee shall evaluate any as-found conditions that do not agree with the plant's prcgrams, plans, policies, and operations in accordance with Part 76.68 (a). The failure Q evaluate the safety impact of the as-found nonconformances in the purge cascade system configuration between August 13 and September 5,1997, is a Violation of 10 CFR 76.68(b)(VIO 70-7001/97008-08).

Recirculatina Water System Late in the inspection period, the plant staff identified anomalies in current operations of the freezer / sublimer recirculating water system pumps. Specifically, the plant staff determined that the Safety Analysis Report assumed the pumps operated in several of the Technical Safety Requirement MODES. However, operator observations indicated that the area centrol room staff were not always aware of the actual operating status of the pumps and that often tne pumps were not operating.

Following discovery of the non-conforming condition, the plant staff documented the issue using the problem reporting system and removed all operating freezer /sublimers from service. Subsequent engineering evaluations were performed which demonstrated that non-operation of the recirculating water system pumps did not impact the safcdy analysis provided in the Safety Analysis Report. Based upon management review and approval of the evaluation, changes were developed and approved to the Safety Analysis Report and the freezer /sublimers were retumed to service.

c. Conclusions The inspectors identified a violation as a result of the plant staff's management of an as-found condi'. ion. Specifically, the plant staff failed to evaluate an as-found condition involving the purge cascade, as required by 10 CFR 76.68(b), un'il after the inspectors brought the issue to their attention. A second as-found condition, ijentified late in the inspection period, was properly evaluated by the plant staff.

E1.2 Clarity of Nuclear Criticality Safety Guidina Documents

a. Inspection Scope (88100)

The inspectors followed up on the continued discovery of lmplementation problems with nuclear criticality safety (NCS) controls and requirements through discussions with operators and NCS staff and a review of associated nuclear criticaiity safety approvals (NCSAs) and documentation.

b. Observatione and Findinos A number of inconsistencies in the implementation of NCS controls were identified by the inspectors oc plant NCS staff during the inspection period. The inconsistencies included 15

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storage of fissile vacuum hoses, installation of fire proof covers on removed process equipment for mouerator control, and tagging of legacy equipment containing potentially fissile material. In discussions with operators, the inspectors noted that a cernplete understanding of how the NCS controls were to be implemented did not exist because of;

1) a lack of clarity in the original NCSA; and,2) weak communication of a require. ment to the staff when a new NCSA was approved. As a result, implementation of NCSA controls varied from building to building across the plant.

An example of an unclear requirement was the control specified fer HEPA vacuum hoses.

Building operators and NCS staff held differing opinions of how to implement the requirement in Building C-310. Review of NCSA and associated evaluation indicated that wrapping a hose around a vacuum was a potent!al upset condition, but the NCSE did not include any further discussion. To prevent the upset condition, the NCSA required a posting which stated:

• Vacuum hoses shall not be wrapped around this vacuum.' The NCS staff stated that this meant not to physically wrap a hose around a vacuum, but that storage of a hose in a spacing pan with a vacuum was acceptable because safe volume was maintained. The inspectors could not obtain this understanding from the NCSA (which simply identified wrapping a hose around a vacuum as a problem). Discussions with operators indicated that they did not understand what the term " wrapping" meant. As a result, operators were faced with interpreting a requirement which had not been clearly defined as a part of the NCS review and approval of the operation.

In folle'N-up discussions, the NCS staff indicated that part of the corrective actions for the NCS audit performed pursuant to CAL Rill-97-003 was an ongoing review of the format and " useability" of the NCSAs. The NCS staff indicated that this process would continue as the NCSAs were upgraded over time.

c. Conclusion

inconsistent implementation of some controls or requirements relied upon for criticality safety appeared to result from a lack of clear communication of the requirements between the NCS staff and those personnel responsible for implementation.

IV. Plant Support P1. Conduct of Emeraency Preparedness Activities P1.1 Event Responses

a. Inspection Scope (88100)

The inspectors observed three event responses during the inspection period.

b. Observations and Findinas The inspectors observed the event responses for minor releases in Buildings C-710 and C-331 and a smoking motor winding in Building C-310. In each case, the event response by the PSS, acting as the incident commander, was conducted in accordance with the appropriate emergency response procedure and the Emergency Plan. However, during the response to the C-331 release (involving a release of hydrofluoric acid through a relief 16

V ,

valve on one of the seal exhaust pumps), the actual location of the release was not provided to the incident commander until 30-40 minutes into the response. The location of the release was known to a het!th physics technician and the control room operator at the time the emergency response was initiated. This lack of communication meant that response teams were sent into the building to take samples without clear knowle.dge of the potential release point, in contrast, the response to a minor release of uranium hexafluoride in the C-710 laboratory was aided by good communication of the potential release point and the status of affected equipment to the incident commander by the personnel reporting the event. This communication aided in the timely identification of the location and initial response actions for the responders.

c. Conclusion

Reponses to events observed by the inspectors were effective and in accordance with the Emergency Plan. Inef' ctive communication of the release point to the incident commander in one instance delayed identification and sampling of the nrria involved with the minor release.

R1, Radiation Protection R1.1 Radiation Protection Routir.e Operations

a. Inspection Scope (83822)

The inspectors interviewed health physics teAnicians (HPT) and accompanied HPTs on routine radiological surveys. During facility tours, the inspectors checked the operability of radiologicalinstrumentation.

b. Observations and Findinas The inspectors accompanied HPTs on numerous routine radiological contamination surveys in the cascade process buildings and observed radiological surveys of emptied product cylinders. Survey techniques were consistent with good health physics practices and plant procedure CP4-HP-RP2101, " Performance of Radiological Surveys." A review of routine survey frequencies and randomly selected survey records indicated that the requirements of Table 5.3-5, " Removable Contamination Survey Frequency," were met.

During tours, the inspectors checked calibration intervals and daily radiological instrumentation source check records. Radiological instrumentation were either in calibration, with a daily source check performt.d, or properly tagged out of service, per the requirements of Procedure UE2-HP RP1033,"RadiologicalInstrumentation."

During intervie*.vs with HPTs onsite, the inspectors noted discrepancies between plant policies and HPTs issuanco of problem reports. Interviews with HPTs consistently highlighted that it veas unclear to the HPTs woen problem reports shoCd be initiated; as a result, in certain situations prob!em reports may not have been generated. The inspectors highlighted to HPTs, that Procedure UE2-HR-Cl1030, " Problem Reporting,"

identified what types of incidents were required to be reported, and that the procedure did include violations of radiation protection requirements as an example. Although the inspectors did not observe any actualinstances of problem reports not being generated for radiation protection violations, the consistent results of interviews with HPTs 17 L

I i.

highlighted this issue as a deficiency in the radiation protection program. In response to the ideafication of this deficiency, the Radiation Protection Mannger (RPM) initiated a number of immediate corrective actions. The corrective actions included health physics supervisor interviews with all HPTs to ensure there was a clear and consistant understanding by the HPT of when to initiate problem reports for radiation protection issues. The RPM also indicated that the next monthly HPT meeting would include a discussion on filing problem reports. Deficiencies in problem reporting is also discussed in Section 01.6.

The inspectors noted good communication between the radiation orotection management and staff at both the Portsmouth and Paducah Gaseous Diffusion Plaats. This was highlighted during the inspection through plant staff follow-up of a potentially generic

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issue conceming the flammability of anti-contamination clothing used at both plants and by the continued communication, which took place between the two plant staffs, until the issue was resolved.

c. Conclusions The inspectors noted that routine radiation survey requirements were met, radiological instrumentation was property maintained, and survey techniques were in accordance with the appropriate procedures. However, the inspectors identified a deficiency conceming general knowledge regarding the issuance of problem reports for radiation protection deviations.

R1.2 Intemal Dosimetry Prooram

a. Inspection Scope (83822)

The inspeders reviewed the plant's intemal dosimetry program and current bloassay data for plant personiel. Several intemal dose investigations for bioassay results, in exceedance of adrWnistrati';e plant limits, were also reviewed.

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b. Observations and Findings The inspectors reviewed the intemal dosimetry program procedures and noted that the procedures implemented the intemal dosimetry program as described in Safety Analysis Report Section 5.3.2.3. Routine employee bioassays were conducted monthly, and employee delinquencies for late participation were low. The inspectors verified that employces, who did not submit samples within 10 days of the scheduled test date, were placed on radiological work restrictions until the routine bicassay was performed.

Placement of employe3s on work restrictions was required by Procedure CP2-HP-DS1030, "PGDP Urinalysis Program for the Detection of Intakes of Radionuclides."

Procedure CP4-HP-DS7600, " Routine and Special Bioassay," listed scheduling criteria, frequencies and protocols for the bioassay program. The inspectors reviewed the current list of plant personnel participating in the intemal dosimetry program and found the practices to be in accordance with the procedural requirements. Intemal dosimetry logs indicated special bioassays were conducted according to the criteria described in Section 6.4.2, of Procedure CP4-HP-DS7600.

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l Appendix C of Procedure CP2-HP-DS1030 had three administrative action levels for uranium bioassay exposure results. The limits were 5,20 and 40 micrograms per liter of uranium (pg/L). A bioassay result of S g/L required re-sampling, and bioassay results in excess of 20 pg/L required an investigation, in addition to re-samples until the bloassay results were below 5 ug/L. Bioassay results in excess of 40 ug/L required em.ployees to be restricted from radiological areas until re-samples were below 5 pg/L and required management to conduct an investigation of the cause of the uptake. Since March 3, only one bioassay result hdicated a levelin excess of 40 pg/L. Although the actual causal factor for the uptaks was not determined, the inspectors determined the investigation was thorough and extensive in trying to determine the cause of the uptake. Several other ,

investigations for bioassay results above 20 ug/L, but below 40 g/L were also reviewed and determined to be equally thorough and extensive. All re-samples for bioassays above administrative limits were conducted until a final result below 5 ug/L was observed.

The inspectors verified that the worker whose bioassay was in excess of 40 pg/L was placed on a work restriction until bioassay re-samples indicated results below 5 g/L A review of 1996 dosimetry data for the plant illustrated that the maximum Committed Effective Dose Equivalent for the year was 8 millirem. The maximum Total Effective Dose Equivalent for 1996 was 367 millirem. These maxima were below ten percent of the 10 CFR 20 maximum dose for radiation workers.

c. Conclusions The certificatee effectively implemented an intemal dosimetry program, consisting of thorough and extensive dose investigations. The intemal dosimetry program was implemented in accordance with Safety Analysis Report Section 5.3.2.3.

R 1.3 Health Physics Technician Trainina

a. Inspection Scope (63822)

The inspectors reviewed the training and qualification program for the plant health physics technicians (HPTs).

b. Observations and Findina Procedure UE2-HP-RP1036," Radiological Protection Training and Qualification,"

addressed the implementation of the health physics training program. During the course of the inspection, a random seiection of HPT training and qualification logs were reviewed. The inspector noted that one new HPT was hired since March 3. All other HPTs were requalified prior to March 3 under the previous HPT training program.

During review of qualification records, the inspectors noted that the new HPT was issued

, a letter entitled, " Qualification as a Health Physics Technician," on September 18.

Procedure UE2-HP-RP1036 Section 6.4.4 stated, in part, that HPTs shall coi,ipMte both a final comprehensive written and oral board examination prior to qealification. However, a ieview of the HPT's training file indicated that no oral board examination was conducted prior to the HPT's qualification. Upon furtner review of the September 18 letter, the inspectors determined that the HPT was not a fully quahfied technician, but rather a trained-to-task HPT. A trained-to-task HPT was qualified only for a limited number of 19

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health physics tasks, Discussions with Radiation Protection Management indicated that  !

the oral board examinations were conducted at the completion of training prior to full qualification as an HPT, and during the two year requalification of HPTs. On October 2, the technician was subsequentiy roissued a Qualification letter as a trained-to-task HPT.

The Radiation Protection Manager stated that a clearer delineation between the two types of HPT qualification and the training interval at which the orsi board was conducted would bo eddressed in a revision to the implementing procedures for HPT training and qualification. A review of the HPTs training records for qualified tasks indicated no discropancies,

c. o Qop_clmions The inspectors performed a sampling review of the HPT training and qualification records which indicated that all HPTs had completed the appropriate training and were property qualified. The training program was implemented in accordance with plant procedures and Safety Analysis Report Section 5.3.1.8.

81. Conduct of S1cu_nilyJr1Gaf_gnuards r Activities St.1 gafoggards Proaram ImpfsmentatioD

a. hspLctipaScope (8.1.401. 81402, 81431)

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

The inspectors also reviewed implementation of site security proce:fures. In particular, the following procedures were reviewed:

(1) Procedure CP4-SS SP2200, " Access Controly Revision 1, dated July 7,1997; (2) Proceduro CP4 SS-SP2201, " Patrol Operations," Revision 1, dated June 30, 1997; and (3) Procedure CP4-SS-SP27 ~ 1, " Post Operations," Revision 1, dated May 20,1997.

b. Observahons and Findings To dotormine if adoquato protection was afforded activities involving low enriched uranium (LEU), the inspectors toured the controlled access area (CAA) and observed the integrity of the fonco, gatos, and the vehicle barriet Fences, gates, and the vehicle barrier were intact and adequately maintained. Por:onnel were identified, registered, badged, and escorted as required. Clearances and the need for access were verified.

Packa00s were visually inspected by security officers at the entrance to the CAA, as required by the Physical Security Plan (PSP). Officers at the CAA vehicle gate adequately performed random entry and exit searches of vehicles.

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All the officers were armed with a handgun and equipped with a radio. The inspectors witnessed radio tests and concluded that appropriate communication capability existed within the CAA.

The use of locks and seats was adequate. The inspectors interviewed officerspgsted and on patrol and found them to be knowledgeable of their duties and responsibilities.

Security procedures were located at the appropriate locations, provided adequate guidance for security office.- duties, and were reviewed ti the required frequency.

The inspectors toured the CAA for the purpose of observing the storage of LEU and determined that the LEU was stored or used only within the CAA.

In response to the Secunty Plan for the Protection of Classified Matter inspection, conducted by the Division of Facilities and Security in May 1997, the certificatee was in the process of revising their security plans. After the review of the proposed revisions the ii.spectors indicated that due to the extensive changes pmposed, a meeting should be scheduled with the Headquarters security plen reviewers to formally present and discuss the proposed plan changes. The certificatee agreed that a meeting was necessary and would formally request a meeting to discuss the numerous changes.

c. Conclusions The protection afforded LEU was in compliance with the Physical Security Plan commitments.

S8. Miscellaneous Security lesues S8.1 Certificatee Security Reports (90712)

The certificatee made the following security-related one-hour reports pursuant to 10 CFR 95.57(b) during the inspection period. The inspectors reviewed any immediate security concerns at the time of the initial verbal notification, Number Status Date Title 32952 Closed 9/17/97 Possible Compromise of Classified Matter 32953 Closed 9/17/97 Possible Compremise of Classified Matter 32958 Closed 9/18/97 Possible Compromise of Classified Matter 70-7001/97008-09 Closed 9/19/97 Possible Compromise of Classified Matter 70-7001/97008-10 Closed S,19/97 Possible Compromise of Classified Matter 21

. . ~ - . . . .

p. . ...

f. .,

l.'

. 70 7001/97008-1 t_ Closed ' 9/22/97 Possible Compromise of Classified:

Matter 70 7001/97008-12 Closed 9/22/97 Possible Compromise of Classified Matter s .-

70 7001/97008-13 Closed = 9/24/97- Possible Comprotnise cf Classified

~ Matter.

70 7001/97008 14- Closed 9/25/97 Possible Compromise of Classified Matter 70-7001/97008-15 Open 10/7/97 Classified Repositories Unlocked and Unattended .

The s( curity reports identified as closed were discussed during a Predecisional Enforcement Conference held with the cc "' .atet, on October 9,1997. The reports were additional aspects of deficiencies with the control of classified matter and an apparent violation discussed in Section S1,1 of NRC Inspection Report 70-7001/97007(DNMS).

The inspectors will review the certificatee's corrective actions as a part of the response to the yparent violation.

V. Manaaement Meetina X. Exit Meetino Summary The inspectors presented the inspection results to members of the plant staff and management at the conclusion of the inspection on October 7,1997. The plant staff acknowledged the findings presented. The inspectors asked the plant staff whether any materials examined during the inspection should be considered proprietary. No proprietary infomiation was identified.

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. PARTIAL LIST OF PERSONS CONTACTED United States Enrichment Corporatio3.

J. H. Miller, Vice President - Production - . ..-

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

- Lockheed Martin Utility Services (LMUS)-

'S. A. Polston, General Manager

'H. Pulley, Enrichment Plant Manager

• W, E. Sykes, Nuclear Regulatory Affairs Manager

• S. R. Penrod, Operations Manager United States Department of Enerav (DOE)

G. A. Bazzell, Site Safety Representative

/ Nuclear Reaulatory Commission (NRC)

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

.* J. M. Jacobson, Resident inspector -

R. G. Krsek, Fue! Cycle inspector J. R. Kniceley, Physical Security inspector

• Denotes those present at the October 7,1997 exit meeting.

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

INSPECTION PROCEDURES USED IP 83822 Radiation Protection IP 88100 Plant Operations

. IP 88102 Surveillance Observations IP 88103 Maintenance Observations IP 88105 Management Oversight and Controls IP 81401 Plans, Procedures and Reviews IP 81402 Report of Safeguards Events IP 81431 Fixed Site Physical Protection of Special Nucicar Material of Low Strategic Significance IP 90712 in-office Review of Events

. IP 92702 _ Follow-up of Events 4

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. ITf!MS OPENED, CLOSED, AND D!SCUSSED Opened 70 7001/97008 01 vio failure to document tha reason for and the special entry and exit points of procedures used to restart the purge cascade 70 7001/97008-02 viu failure to use approved procedures and to maintain current written procedures for operations 70-7001/97008 03 ifi inconsistencies in the TSRs 70 7051/97008-04 vio failure to f;ie problem reports for the deficiencies and alarms 70 7001/97008-05 cer safety system failure of 2s head-to-shell gasket 70 7001/97008-06 ce,' sifety system actuation of pgid following minor release 70 7001/97008-07 cer safety system actuations (pressure isolation systems) 70-7001/97008-08 vio failure to evaluate an as-found purge cascade system anomaly 70 7001/97008-15 cor possible compromise of classified matter on 10/7/97 Closed 70 7001/97008-09 cor possible compromise of classified matter on 9/17/97 70 7001/97008,10 cor possib:e compromise of classified matter on 9/19/97 70-7001/97008 11 cor possible compromise of classified matter on 9/22/97 70-7001/97008 12 cor possible compromise of classified matter on 9/22/97 70 7001/9700a-13 cer possihte compromise of classified matter on 9/24/97 70 7001/9,4 < 14 cor possibe.' compromise of classified matter on 9/25/97 pjs_ cussed None 24

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LIST OF ACRONYMS USED ACR Area Control Room . . -

ANSI American National Standards Institute APSS Assistant Plant Shift Superintendent CAA Controlled Access Area CAAS Criticality Accident Alarm System CER Certificatee Event Report CFR Code of Federal Regulations CP Compliance Plan CRD Classified Restricted Data DOE Department of Energy HF Hydrogen Fluoride HPFW High Pressure Fire Water HPT_ Health Physics Technician IFl Inspector Follow up item LCO Limiting Condition for Operation LEU Low Enriched Uranium MW Megawatt

-MWP Maintenance Wo* Package NAM Negative Air Machine-NCS- Nuclear Criticality Safety NCSA Nuclear Criticality safety Approval NCSE Nuclear Criticality Safety Evaluation  ;

NOV Notice of Violation NQA National Quality Association -

NRC Nuclear Regulatery Commission +

PCR Plant Change Request - i PMT Post Maintenance Test PORC Plant Operations Review Committee PSIA Pounds Per Square Inch Absolute PSIG Pounds Per Square Inch Gage PSP Physical Security Plan PSS Plant Shift Supervisor QAP Quality Assurance Procram <

RAC Request for Authorization Change SAR Safety Analysis Report -

TSR Technical Safety Requirement U 235 Uranium 235 UF6 Uranium Hexafluoride URI Unresolved item USEC- United States Enrichment Corporation VIO Violation v 25

,