ML20148F819

Insp Rept 70-7001/97-02 on 970303-0421.Violations Noted. Major Areas Inspected:Plant Operations,Maint & Surveillance, Engineering & Plant Support

ML20148F819
Person / Time
Site:	Paducah Gaseous Diffusion Plant
Issue date:	05/30/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20148F792	List: ML20148F788 ML20148F812 ML20148F819
References
70-7001-97-02, 70-7001-97-2, NUDOCS 9706050055
Download: ML20148F819 (38)

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.I U.S. NUCLEAR REGULATORY COMMISSION I

REGION 111 j'

Docket No:

70-7001

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' Inspection Report No:

70-7001/97002(DNMS)

Facility Operator:

United States Enrichment Corporation Facility Name:

Paducah Gaseous Diffusion Plant l

Location:

- 5600 Hobbs Road P. O. Box 1410 Paducah, KY 42001 1

Datos:

March 3, and April 21,1997 l

Inspectors:

K. G. O'Brien, Senior Resident inspector J. M. Jacobson, Resident inspector Approved By:

Tim Reidinger, Acting Chief i

Fuel Cycle Branch l

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9706050055 970530 PDR ADOCK 07007001 C

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EXECUTIVE

SUMMARY

l United States Enrichment Corporation Paducah Gaseous Diffusion Plant NRC Inspection Report No. 70-7001/97002(DNMS)

Plant Operations The inspectors identified that the Building 360 Technical Safety Requirement manning limits were not always met during the inspection period. As a result, a Technical Safety Requirement violation occurred. (Section 01.1)

The inspectors determined that some plant management and staff were not i

cognizant of and did not adhere to plant procedures limiting the number of hours worked without prior approval. As a result, a Technical S6fety Requirement violation occurred. (Section 01.2)

Building 360 operators failed to determine the cause of a safety system actuation,

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prior to retuming the system to service, resulting in a second actuation and a l

Technical Safety Requirement violation. (Section 01.3) t A Technical Safety Requirement violation was identified, in that, on two occasions, I

the plant did not ensure that the criticality accident alarm system was operable for l

l areas requiring system coverage. (Section 01.4) l l

Maintenance and Surveillance The Plant Shift Superintendent and engineering staff identified an inadequate surveillance procedure and responded in a timely manner. (Section M1.1)

An incomplete understanding of the process gas leak detection system by some operations and maintenance staff contributed to an incorrect operability determination. In addition, the Plant Shift Superintendent's response to the perceived inoperability was untimely. (Section M1.2) j Incorrect implementation of the work control program resulted, on two occasions, in the incorrect replacement of safety-related components (nitrogen cylinder) with non-safety-related qualified components. (Section M1.3)

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Continued criticality accident alarm system problems indicated that an increased surveillance frequency may not be adequate. (Section M1.4) l Enaineerina l

A Technical Safety Requirement violation occurred as the result of a non-conservativo interpretation of the existing nuclear criticality safety requirernents for legacy procass equipment. (Section E1.1) 2 I

e A Confirmatory Action Letter was issued in response to plant staff-identified weaknesses in the understanding and implementation of some nuclear criticality safety approvals. (Section E1.4)

A Notice of Enforcement Discretion was issued in response to a Technical Safety Requirement inaccuracy caused, in part, by past inadequate reviews of the Technical Safety Requirements and past plant configuration problems.

(Section E1.5)

The inspectors identified several examples of change control process implementation problems. (Section E1.6)

Plant Sucoort j

An untimely 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> report for an inoperable fire protection system was made as the result of an incomplete understanding, by some operations staff, of the engineering basis for the fire protection sprinkler systems. (Section F1.1)

The plant staff undertook a significant effort to perform a detailed walkdown of the entire Technical Safety Requirement-related fire protection sprinkler system. The walkdown identified both design and configuration control problems for which appropriate correction actions were initiated. Because the criteria for eriforcement discretion specified in Section Vll.B.3 of NUREG-1600, " General Statement of Policy and Procedures for NRC Enforcement Actions," were met, this matter is considered a Non-Cited Violation. (Section F1.2)

Two examples of inadequate contro! of classified materials were identified and cited as a violation of 10 CFR 95. (Section S1.1)

A 10 CFR 71 violation was identified based upon two examples of inadequate j

implementation of American National Standards Institute Standard N14.1. Each j

example occurred, in part, due to the continuation of long standing past practices without assurance of their conformance to current applicable requirements.

(Section T1.1) 3

i DETAILS

1. Operations 01.

Conduct of Operations' 01.1 Buildina 360 Minimum Staffina Imolementation a.

Insoection Scoce (88100)

The inspectors reviewed implementation of the Technical Safety Requiremmts (TSRs) for minimum staffing in Building 360 facility.

b.

Observations and Findinas On April 14,1997, during a routine tour of Building 360, the inspectors observed that the facility was in the process of heating and sampling uranium hexafluoride (UFe) filled cylinders. The conduct of these activities placed the facility in Modes 5 and 7.

During discussions with one of the operators, the inspectors observed a j

second operator enter the building office area. The operator appeared to have just completed taking a shower and was dressed in personal clothing.

j The inspectors determined that the second operator had left the work area cpproximately ten minutes earlier to take a shower. The operators indicated that it was common practice to leave the work area, in order to take a shower, prior to being relieved by the oncoming operator.

As a followup to these observations, the inspectors reviewed both the plant procedures and TSR requirements for Building 360 staffing. Several plant procedures included staffing requirements, as noted below:

Procedure Recuirement Shift Turnover Remain in work area until relieved (OPS 11)

Shift Routines Proper attention to plant conditions (OPS-8)

Required Reading Read, initial, date immediate reading prior to doing the tasks or duties Topical headings such as 01, MS, etc., are used in accordance with the NRC standardized inspection report outline contained in NRC Manual Chapter 0610. Indwidual reports are not expected to address all outline topics, and the topical headings are therefore not always sequential.

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e TSR 3.2.2.a. and TSR Table 3.2.2.1, " Minimum Staffing Requirements,"

required that Building 360 (facility) be staffed by: 1) one operator during i

Mode 5 operations, and: 2) two operators, one located in the laboratory area and the other in the facility, when in Mode 7.

The inspectors reviewed the Building 360 required reading file and control room logs for the period from March 3 through April 14. The inspectors determined that several operators were not included in the building required reading program. These operators were not routinely assigned to the building, but had stood watch during operations since March 3. As a result, i

several operators had not read, initialed, and dated the immediate required reading materials prior to assuming responsibility for performing watch station tasks or duties.

j The inspectors discussed these observations with the building and plant management. Management indicated that staff taking showers or in street clothing were previously considered able to respond to emergencies. This position was based upon historical events and actual response actions.

However, this approach did not consider the need for staff to respond to expected normal and off-normal conditions. Management acknowledged the j

deficiency in the implementation of the required reading program.

Based upon the above, the operators actions to: 1) leave to work area to take a shower; 2) return to the office area in personal clothing, and; j

3) assume watch standing responsibilities without having reviewed, initialed, and dated immediate required reading materials is a Violation of TSR 3.2.2.a (VIO 70-7001/97002-01).

At the end of the inspection period, plant management took steps to ensure j

that operators: 1) remained on watch until properly relieved; 2) were cognizant of and adhered to plant protective clothing requirements for standing watch, and: 3) reviewed, initialed, and dated immediate required i

reading materials prior to assuming a watch which would include j

responsibility for performing the referenced tasks or duties.

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Conclusions The inspectors identified that Building 360 TSR manning requirements were not always met during the inspection period. Specifically, the operators:

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1) left their work area without proper turnover; 2) returned to the area in personal clothing, thus precluding entry into the normal work area, and; 3) did not review, initial, and date immediate required reading materials prior to assuming watch standing responsibilities.

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a 01.2 Hours of Work a.

Insoection Scone (88100)

The inspectors noted several problem reports which documented plant staff working hours in excess of the TSR limits. Each of these events were i

reviewed and discussed with the responsible management. The inspectors also reviewed the current procedure for authorizing staff to work hours in excess of the TSR limits and its implementation.

b.

Observations and Findinas During the inspection period, plant staff filed several problem reports to document staff working hours in excess of the procedure and TSR prescribed limits without management preauthorization. The inspectors reviewed the problem reports and noted that the exarnples were dominated

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by staff working: 1) in excess of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in a 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> period, and; 2) with less than an 8 hour9.259259e-5 days <br />0.00222 hours <br />1.322751e-5 weeks <br />3.044e-6 months <br /> break between shifts. in most of the cases, the staff were not involved in safety-related work.

However, on March 4,1997, two individuals worked 17 continuous hours on safety-related equipment, the criticality accident alarm system. The final hours of work were focused on completing repair and documentation activities. Although the work was directed by the line management, the inspectors noted that no effort was undertaken by either the staff or management to get prior approval to exceed the TSR hours of work limitations.

TSR 3.2.2.b requires, in part, that the working hours of facility staff ' e c

controlled in accordance with the TSR guidelines and that deviations from the guidelines be authorized in advance by managoment.

The failure to receive prior management approval to deviate from i

TSR 3.2.2.b guidelines is a Violation (VIO 70-7001/97002-02).

As a followup, the inspectors reviewed procedure CP2-HR-LR1030,

" Limitations on Hours of Work." The procedure provided instructions to both management and staff on the methods and tools to be used to ensure that TSR limits were met. In the cases noted above, these methods were not utilized. Through discussions with plant staff, the inspectors determined that some managers and staff were not fully cognizant of the individuals covered by the procedure; while others were not rigorously tracking the hours worked by themselves or their staff, 6

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Conclusions The inspectors determined that some plant management and staff were not cognizant of and did not adhere to plant procedures limiting the number of hours worked without prior approval. As a result, two individuals, performing safety-related activities, worked 17 continuous hours without the proper authorization. This is a TSR violation.

01.3 Suildina 360 Safety Systein Activation and Return to Servism a.

inspection Scope (881001 The inspectors reviewed the circumstances surrounding a Building 360 j

safety system actuation and the subsequent return to service of the equipment.

b.

Observations and Findinas On April 9, Building 360 operators simultaneously initiated heat cycles for UF, bearing cyliriders in autoclaves number 1 and number 2. Approximately thirty minutes into the heat cycles, a high condensate alarm and.

associated safety system actuation was received on autoclave number 1. The operators responded to the alarm and determined that all automatic actions had occurred as described in Alarm Response Procedure (ARP) CP4-CO-AR8360, Rev 0, "High Drain Level."

Step 9 of the ARP specified supplementary actions which required, in part, that: "When any condensate is drained and thc cause of alarm has been determined and corrected,.... return autoclave to service." As fulfillment of this step, the operators: 1) pulled a vacuum on the autoclave for twenty minutes; 2) opened the autoclave shell and inspected for standing water; and,3) checked the drain line for a plugged strainer. Based upon the absence of standing water and a clear drain line strainer, the operators concluded the actuation was spurious. The autoclave was returned to service approximately one hour after the actuation. The Plant Shift Superintendent (PSS) approved the decision to return the autoclave to service.

Subsequently, during the same operating cycle, a second high condensate j

alarm and safety system actuation occurred for autoclave number 1.

However, at the time of the occurrence, the autoclave was in a Mode that did not require the safety system to be operable.

On April 10, the inspectors reviewed the event and the operators actions.

On the night of the event, the operators received an alarm from only one of the two condensate probes. The alarm cleared during the course of subsequent operator actions. However, the bases for getting an alarm from l

only one of the two probes was neither pursued or determined. The

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inspectors noted that the condensate probes were located opposite each other and at the same elevation in the drain line. Therefore, absent an

. equipment failure, both should respond to a high condensate level.

The inspectors notad that the autoclave system pressures and temperatures were also traced 'oy equipment maintained in the lab area. _ The inspectors reviewed the traces with the system engineer and observed changes in system parameters which appeared to validate the alarm. However, the operators indicated that the traces were not used to either assess the validity or cause of the alarm. Instead, the operators stated that the absence of standing water or a clogged drain line strainer was the bases for classifying the alarm as spurious. This approach appeared inconsistent with j

the system design. Specifically, the autoclave was an air tight system and

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the alarm response procedure required evacuation of the autoclase. Based l

Upon the pressures and temperatures involved, this process would normally remove all water from the autoclave, even on a valid alarm.

The inspectors further discussed with other operations management and i

staff the expectations for the indications operators should use when determining alarm validity. Most individuals cited indications similar to those j

used by the Building 360 operators. The inspectors noted that the applicable alarm response procedure also included standing water and a clogged strainer as two of four possible causes for the alarm. However, the system engineers and engineering manager indicated that the absence of l

these indications would not be sufficient to determine that the alarm was spurious. As a result of these observations, management issued a long term i

order (LTO) requiring system engineering involvement in the assessment of safety system actuations, prior to returning the equipment to service.

Additional system engineer-led inspections of the autoclaves identified several other potential generic causes for the alarms. These included:

1) piping configuration (i.e. two autoclave drain lines tied together);

2) degraded drain line check valves, and; 3) the absence of drain vents.

TSR 3.9.1. requires,in part, that written procedures shall be p; spared and implerr.ented to cover activities described in Safety Analysis Report (SAR)

Section 6.11.4.1 and listed in Appendix A to SAR Section 6.11. Appendix A to SAR Section 6.11 lists Operations (including alarm responses) as activities requiring procedures. The operator's failure to determine and correct the cause of the high drain alarm prior to retuming the autoclave to j

service, as required by ARP CP4-CO-AR8360, Rev. O,is a Violation of j

TSR 3.9.1 (VIO 70 7001/97002-03).

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Conclusions L

L The inspectors determined that the Building 360 operators retumed the l

number 1 autoclave to service without adequately determining the cause for L

a safety system actuation. As a result, a second actuation occurred.

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-01.4 Criticality Accident Alarm System Ooerability issues a.

Insoection Scone (88020)

The inspectors reviewed several criticality accident alarm system (CAAS) operability issues.

b.

Observations and Findinas On March 4, plant staff identified that two of three Building 337 CAAS cluster "V" detectors (modules) had indicated background readings of less j

than 9.5 millirem per hour (mrem /hr), the minimum value required for cluster j

operability. This observation was made during a weakly walkdown of tho clusters. Management had implemented a schedule of weekly walkdowns due to engineering concerns over systam electrical stability.. Previously, 4

engineering had identified a trend of unacceptable (background too low) module readings. In most cases, the unacceptable readings involved only i

one of the three CAAS detector modules. Proper functioning of two CAAS detector modules was required for operability.

j in response to the finding,' the plant reviewed the TSRs, entered the i

applicable action statements, and took the required actions. The inspectors observed these activities and noted that implementation of some required actions was delayed. Specifically, the Cascade Coordinator staff made the initial inoperability announcement, in the building, approximately thirty minutes after the issue was identified. However, the same announcement to the general plant population was not made until almost an hour later. In i

addition, plant staff took a significant amount of time to place barriers around the affected area to preclude staff entry.

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On March 18, during a routine facility tour, the inspectors identified that the

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i Building 337A "N" CAAS cluster had two modules reading less than the i

minimum allowed limit of 9.5 mrem /hr. This observation was immediately

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brought to plant management's attention. In response to the inspectors' finding, plant staff reviewed the applicable TSR and spent approximately 25 minutes attempting to determine if the facility was in a Mode that required j

the CAAS to be operable. This action appeared non-conservative. At the i

time of discovery plant staff expected the cluster to be operable. Following a PSS determination that the system should be operable, the facility entered the appropriate actica statements, and took the required actions.

Technical Safety Requirement 2.2.4.3.a (333A and 337A Facilities) and Technical Safety Requirement 2.4.4.2.a (Enrichment Cascade Facilities) require that criticality accident detection shall be operable in areas, equipment, or processes which contain greater than 700 grams of uranium-235 at an enrichment greater than or equal to 1.0 weight percent (w/%) uranium-235. The failure of the plant to maintain the Building 337A "N" and the Building 337 (an enrichment facility) "V" CAAS clusters 9

o operable in areas which contained greater than 700 grams of uranium-235 enriched to greater than one w/% is a Violation of TSRs 2.2.4.3.a and 2.4.4.2.a (VIO 70-7001/97002-04).

c.

Conclusions A TSR violation was identified, in that, on two occasions during the inspection period, plant staff did not ensure that the CAAS system was operable for areas that contained quantities of uranium-235 greater than 700 grams and enriched to greater than one w/E in addition, the plant staff's response to these inoperabilities, at times, was delayed.

08.

Miscellaneous Ooerations Matters (90712) 08.1 (Ocen) Certificatee Event Reoort 31892: Discovery of inoperable criticality accident alarm cluster V in Building 337. See discussion in Sections 01.4 and M1.4. This event report is open pending a review of the certificatee's written report and response to the associated violation (CER 70-7001/97002-05).

08.2 (Ocen) Certificatee Event Reoort 31897: Actuation of the autoclave water inventory control system (primary condensate alarm) for autoclave 1 north in Building 333A. An open circuit on the autoclave thermocouple transmitter due to corrosion caused the temperature and pressure control loop to respond to a perceived high temperature and close the steam control valve. The rapid reduction in pressure and condensation of steam, while the uranium hexafluoride in the cylinder was at the triple point, caused condensate to accumulate in the drain line and actuated the condensate alarm. The high condensate safety system isolated the steam supply and thermovent line for the autoclave, as designed. This event report is open pending a review of the certificatee's written report (CER 70-7001/97002-06).

08.3 (Ocen) Certificatee Event Reoort 31954: The plant staff made a 10 CFR 21 report to the NRC relative to Hunt alloy 636 packing nuts for uranium hexafluoride

. cylinders. A number of these packing nuts have exhibited cracks due to potential manufacturing or material defects. The plant staff reported that cracks in the packing nut were responsible for two known outgassings of uranium hexafluoride during cylinder filling operations. This event report is open pending a review of the esrtificatee's written report (CER 70-7001/97002-07).

08.4 (Ocen) Certificatee Event Reoort 31968: Discovery of inoperable ciiticality accident alarm cluster N in Building 337A. See discussion in Sections 01.4 and M1.4. This event report is open pending a review of the certificatee's written report and response to the associated violation (CER 70-7001/97002-08).

08.5 (Closed) Certificatee Event Reoort 32044: An annual surveillance of the UF, reisase detection and isolation system for the Building 310 west Normetex withdrawal pump was determined to have made the similar safety system for the east pump inoperable. The pump was in Mode 2, withdrawal, at the time. The 10

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i LCO required that the safety system be operable for this Mode. The plant staff i

later retracted this event report. ' The inspectors reviewed a subsequent engineering l

evaluation that documented that the east Normetex pump safety system was never s

inoperable and concluded that evaluation was reasonable. See Section M1.4 for i

~ further discussion'(CER 70-7001/97002-09).

I 08.6- (Onen) Certificatee Event Report 32048: Legacy G-17 valve assemblies, containing i

c unknown amounts of potentially fissile material, were not spaced in accordance I'

-with the applicable Nuclear Criticality Safety Approval (NCSA) GEN-27. The NCSA required that the valves be spaced 2 feet from other items containing potentially fissile material. The issue of legacy equipment identification and maintaining spacing controls is discussed in Section E1.1. This event report is open pending a review of the certificatee's written report (CER 70-7001/97002-10).

08.7 (Onen) Certificatee Event Report 32128: On March 9,1997, the water inventory l

control system for autoclave number 1,in Building 360 (Toll Transfer and Sampling), actuated upon a valid signal. The signal was originally thought to be spurious by the operation staff. See further discussion in Section 01.3. This event report is open pending a review of the certificatee's written report (CER 70-7001/97002-11).

II. Maintenance and Surveillance M1, Conduct of Maintenance and Surveillance M1.1 Buildina C-720 Criticality Accident Alarm Duration Test

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a.

Insoection Scoos (88102)

The inspectors reviewed the circurnstances surrounding and response to a plant staff-identified failure to rigorously perform and document a Building 720 CAAS horn battery surveillance.

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Observations and Findinas On March 3, the PSS and engineering staff identified that documentation did not exist to demonstrate that the Building 720 CAAS horn batteries had been properly tested. Surveillance Requirement (SR) 2.6.4.1.b-3 required an annual check of the batteries to ensure that the batteries cou!d power the electronic horns for at least 120 seconds. Plant staff had performed a surveillance of the CAAS cluster just prior to the NRC assuming regulatory authority on March 3,1997. However, this surveillance did not specifically address the 120-second requirement. Staff that performed the surveillance indicated that horns sounded for a period of time in excess of 120 seconds while other surveillance activities were accomplished.

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Upon ' discovery that the previous surveillance did not adequately document j

the CAAS horn battery performance, the PSS utilized TSR 1.6.3.3 to allow the system to remain operable. This TSR provides a 24-hour exception to the operability requirements in those cases where the failure to perform a

required surveillance within the maximum acceptable time interval was the l

only reason for the inoperability. TSR 1.6.3.3 also directed that within this i

24-hour window,'the plant was to either perform the required surveillance or place equipment in an operating Mode for which the CAAS would not be-F required. The PSS determined that the " maximum interval" fended with the

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NRC assumption of regulatory authority at 12:01 a.m. 'on March 3,1997.

i Therefore, staff action was required by 12:01 a.m. on March 4,1997.

J Maintenance staff completed the necessary surveillance within the allotted

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time. No problems were identified with the CAAS horri batteries.

c.

Conclusions c

I The PSS and engineering staff identified an inadequate procedure for l

conducting a CAAS surveillance that had been performed prior to NRC j

assumption of regulatory authority. Plant staff responded to the finding in a L

. timely manner by utilizing TSR 1.6.3.3 to maintain the CAAS operable while

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the SR was performed under a revised procedure to assure CAAS electronic

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horns would sound for the required 120 seconds.

M1.2 Buildina 310 West-Normetex Leak Detection Surveillance i-a.

Insoection Scone (88100. 88102) j The inspectors reviewed the initial problem report, work package, and system drawings for a failed surveillance nf the west Normetex pump

-uranium hexafluoride (UF ) release detection system.

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. b.

Observations and Findinas t

On March 27, during the performance of an annual surveillance of the west i

N*ctmetex pump UF, process gas leak detection (PGLD) system, instrument 1

mechanics (IMs) identified a faulty "B" signal conditioner for. one of the two instrument channels. At the time of discovery, the "A" channel was also i

out c'l service. After discussions between the PSS and the IMs, the PSS

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.%ntified that the surveillance procedure may have unintentionally caused i

j bcth east pump PGLD channels to be inoperable. As a result, the PSS d;rected Building 310 operators to place the east pump in Mode 3 (standby).

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f At the time the PSS directed this action, the east pump UF, PGLD system -

i had been inoperable for approximately 95 minutes. This was 35 minutes in i

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- excess of the TSR 2.3.4.3 Action Statement completion time 'of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. The F

PSS also promptly filed Event Report, No. 32044, to document that the i

plant had operated for a period of time in excess of that allowed by the TSR 1

with an inoperable safety system.

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4 An annual surveillance of the Normetex pump PGLD systems was required by TSR 2.3.4.3 and SR 2.3.4.3-1. The objective of the surveillance was to ensure that the pump would trip and the discharge valve close, thereby limiting any potential release to 250 pounds of UFe, as specified in the SAR, given the activation of two adjacent PGLD system detector heads.

The inspectors subsequently reviewed the event, the electrical drawings, and the procedures with the system engineer. The review in'dicated that each pump PGLD system has two " trains." The detector heads for each pump were wired into two separate "A" and "B" channels with two separate power supplies for actuating ~the trip and isolation circuitry. - However, a

- caution note in the procedure indicated that work on one train for west pump could make the same train on the east inoperable. The inspectors j

determined that if the mechanics followed the procedure, as written; no

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inoperabilit.ies would be created. Therefore, the procedure caution ~ note appeared technically incorrect.

j The inspectors noted that although the system engineer'was knowledgeable f

of the design and its implications, relative to the TSR; some maintenance and operations staff appeared to have a weak understanding of the actual electrical configuration. This resulted in the.PSS's determination that the i

safety system for the east pump had been made inoperable when, in fact, it l

had not. Following further engineering review of the issues, the plant staff retracted the event report.

The inspectors also noted that the entry into TSR 2.3.4.3, Action Statement D, occurred well beyond the 1-hour completion time requirement.

At the time the IMs discussed the failed "B" channel signal conditioner with the PSS, the east pump was operating, in Mode 2. During these l

discussions, the PSS determined, based upon the procedural caution note, that the potential existed for the east pump to have two inoperable PGLD l

channels. The PSS also assessed that this condition had existed for

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approximately 55 minutes 5 minutes less than the maximum TSR Action i

Statement completion time of 1 hour1.157407e-5 days <br />2.777778e-4 hours <br />1.653439e-6 weeks <br />3.805e-7 months <br />. At this point, the PSS requested j

additional management guidance, but did not direct the staff to place the east pump in Mode 3. This course of action appeared inconsistent with the j

TSRs and non-conservative, given the information available at the time.

c.

Conclusions Some maintenance and operations staffs' weak understanding of the withdrawal pumps PGLD system design and function, in conjunction with an inaccurate surveillance procedure, led to an incorrect determination that the east Normetex pump PGLD system was inoperable and a subsequent event report. Irrespective of the final operability determination, the PSS's actions, J

during the event, were untimely and inconsistent with the TSRs.

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. M1.3 Use of "Non-Q" Nitroaan Cylinders in the Criticality Accident Alarm System t

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Insoection Scone (88103)

The inspectors reviewed the circumstances surrounding the use of replacement nitrogen cylinders for CAAS horns which had not been tested and accepted in accordance with the configuration control program.

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Observations and Findinas On April 8, in NRC Event 31207, the plant reported that a safety system '

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component, a compressed nitrogen cylinder for Building 333 CAAS cluster "Z," had been replaced with a component which did not meet the j'

engineering and configuration control program requirements. In accordance with the plant quality assurance program, nitrogen cylinders were designated as "O" or quality components because they provided the safety-related motive force for the air-operated CAAS horns. As such, the cylinders were.

included in the boundary definition of the CAAS. Because the installed cylinder had not been approved as acceptable for use within a "O" safety system, the PSS initially declared the CAAS cluster inoperable and reported

.the event in accordance with 10 CFR 76.120(c)(2). In addition to the incident on April 8, maintenance staff also replaced a r:itrogen cylinder for j

Building 310 CAAS cluster.H with a "non-Q" cylinder during the inspection period.

i In reviewing the events and the associated work packages, the inspectors noted that the requirement to replace the nitrogen cylinders with "O" cylinders did not appear in the work package. As a result, the maintenance

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staff performing the work were not advised that a standard nitrogen cylinder l

was not acceptable. Additionally, in discussions with engineering staff, the inspectors were informed that nitrogen cyiinders were controlled under a j-program established to ensure that cylinder nitrogen purity and pressures were acceptable for all plant uses. Both programs specified the same basic l

requirements for nitrogen pressure and purity. The inspectors also noted

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that the safety function of the nurogen cylinder was based on the pressure of the cylinder as a back-up source of " air" for CAAS horns. Based on the review, the inspectors did not identify any significant differences for supplying the required " air" pressure between "O" cylinders and "non-Q" j

cylinders controlled and filled under the standard program.

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Plant engineering staff performed an operability evaluation which documented.that the CAAS horns would be operable if "non-Q" nitrogen a

cylinders were used. After the evaluation, the PSS retracted the event E

report because the system was never actually inoperable. The inspectors reviewed the evaluation and concluded that it was reasonable.

j 4

4 8

1 14

-. - _ -. - - -. -. _ _ _, _. ~. ~.. _ - - -.... _ _.

t The inspectors noted that these issues demonstrated a weakness in implementation of the work cor:tml program. Specifically, the staff did not

-l did not ensure, on separate occasions, that safety requirements were effectively incorporated into a safety-related work package. As a result, a safety system (CAAS clusters) h6d to be removed from service to correct j

maintenance deficiencies. In addition, CAAS coverage was not available c

while the deficiencies were remedied.

i c.

Conclusions -

Weak implementation of the work control program resulted in the l

development and use of work packages for replacement of a safety system component (nitrogen cylinder) which were not detailed enough to ensure -

j maintenance staff were aware of which cylinders were acceptable for use.

l As a result, CAAS clusters had to be removed from service to re-perform j

maintenance activities.

' M1.4 Inonerable' CAAS Clusters

)

a.

Inspection Scone (88102) l l

i The inspectors reviewed a trend of inoperable CAAS clusters to assess its significance.

b.

Observations and Findinas Prior to NRC assuming regulatory authority on March 3,1997, plant engineering staff identified a problem with drift in the circuit generating an electronic background of 10 millirem per hour (mrem /hr) for the_CAAS detectors. The electronic background was used as a means to ensure each detector operated properly by providing an indication, on the meter, that the detector was sensing a signal.

The original CAAS design calculations were based on the detector alarming upon sensing 10 mrem /hr of gamma radiation above the generated background. The alarm setpoint was defined as 19.5 mrem /hr. Thus, engineering staff determined that a drift in the background signal, which caused the background to drop below 9.5 mrem /hr, would increase the amount of gamma radiation required to alarm the detector. This condition would also reduce the radius of coverage below the SAR specified 400 feet.

Each CAAS cluster contained three detectors connected in two-out-of-three voting logic. Therefore, any cluster with two or more detectors with background readings less than 9.5 mrem /hr was considered inoperable.

The two events discussed in Section 01.4 were examples of CAAS clusters that were discovered to be inoperable. Based upon past information, engineering staff assumed the cause of the inoperabilities to be drift in the background signal, in an effort to more closely monitor CAAS cluster 15

)

status, engineering initiated a weekly surveillance cf all CAAS clusters. The surveillance's focus was to ensure that the CAAS electronic background had j

i not drifted below that assumed in the coverage calculations. However, the discovery of two inoperable CAAS clusters, during this inspection period, indicated that periodicity of the drift that rendered the CAAS clusters inoperable was not fully understood.

c.

Conclusions The discovery of two inoperable CAAS clusters during the inspection period indicated that actions taken to address a previous trend in background signal drift may not be adequate. The inspectors will continue to track the resolution to this issue through the violation documented in Section 01.4.

M1.5. Crane Surveillances a.

Insoection Scoce (88102. 88103)

The inspectors observed TSR required crane surveillances for cranes in the Building 333A Feed Facility, Bui! ding 310 Product Withdrawal Facility, and Building 315 Tails Withdrawal Facility. In addition, the inspectors observed selected portions of an annual surveillance for a Building 333A crane and reviewed records of shiftly, monthly, and annual surveillances for cranes located in the three other buildings.

j b.

Observations and Findinos i

i The inspectors determined that the surveillances observed and records reviewed were performed in accordance with procedural and regulatory requirements. Discussions with and observations of operators and maintenance staff indicated that the plant staff were generally knowledgeable of TSR requirements relating to crane operation and surveillance. The inspectors also observed that surveillances were documented on checklists to ensure the individual performing the surveillance covered all the critical items.

f&[191131031 11 C.

Operations and maintenance staff performed crane surveillances, during the inspection period, were in accordance with TSR requirements.

M8.

Miscellaneous Maintenance issues (90712)

I M8.1 (Closed) Certificatee Event Reoort 32107: Maintenance staff replaced a CAAS safety system compressed nitrogen cylinder with a cylinder which had not been approved as a "O" cylinder in accordance with quality assurahce program. A subsequent operability evaluation demonstrated that the "non-O" cylinder was able to perform the functions required of a "O" cylinder. Thus, the CAAS cluster was 16

never inoperable. See discussion in Section M1.3 which documented that both types of cylinders could reasonably supply the required pressure to sound CAAS homs. The plant staff subsequently retracted the event report. The inspectors concluded that the retraction was reasonable (CER 70-7001/97002-12).

III. Enaineerina E1.

Conduct of Engineering E1.1 Barriers and Postina "or Leoacy Process Eauioment 1

a.

Insoection Scooe (88020,88100)

The inspectors performed tours of the major process buildings to observe the implementation of criticality safety controls. In addition, the inspectors reviewed pertinent nuclear criticality safety approvals (NCSAs) and associated irnplementing procedures.

b.

Observations and Findinas During routine tours of Buildings 331,335, and 337, the inspectors noted that equipment (compressors, valves, etc.), bearing legacy process equipment tags, was not roped off or otherwise posted with nuclear criticality safety (NCS) signs. On April 16, the inspectors observed two compressors and a G-17,30-inch block valve, which were tagged with legacy equipment tags, indicating the equipment contained unknown amounts of uranium anriched to greater than 1.0 w/%. The tags were only visible when the equipinent was approached directly from one direction. In addition, the inspectors determined that the size of the tags made it difficult for personnel to observe them and be aware that the equipment required a two-foot exclusion area on all sides.

The applicable criticality safety requirements were in a general plant approval, NCSA GEN-27, " Handling and Storage of Legacy Process Equipment," dated September 16,1996. This NCSA required that legacy process equipment, defined in part as equipment removed from the cascade but not characterized, was spaced from other fissile or potentially fissile materials by two feet. The NCSA defined three types of equipment:

(1) Planned Expeditious Handling (PEH); (2) Uncomplicated Handling (UH),

and: (3) Limited Spacing (LS) equipment. Once characterized, the NCSA directed that the legacy process equipment should be retyped into one of these three categories, or should be defined as fissile exempt. The inspectors noted that NCSA GEN-27 and an associated NCSA, GEN-10, required that PEH, UH, and LS equipment be roped off and posted with specific NCS signs. These actions were directed to ensure that other fissile or potentially fissile materials were not placed within two feet of the equipment.

17

t o'

i j.

+

!~

During discussions with operations and engineering staff, held following the April 16 observations, the inspectors were informed that nuclear criticality safety staff (NCS) did not require legacy process equipment to be roped off or posted. This assertion appeared to be non-conservative, and contrary to-the requirements of NCSAs GEN-27 and GEN-10. Specifically, this assertion implied that the NCS controls for equipment containing unknown amounts of l

enriched uranium were less stringent than the controls for equipment which

'had been properly characterized as PEH, UH, or LS.

Following these discussions, NCS staff identified that an NCSA GEN-27 implementing procedure required that equipment, bearing Legacy Process Equipment Tags, be roped off and posted to ensure that the two-foot spacing requirement was met. Specifically, Procedure CP2-TS-TS1030, Revision 0, " HANDLING AND STORAGE OF LEGACY PROCESS EQUIPMENT," dated December 10,1996, step 8.2.1, required, in part, that -

personnel: " mark / rope off equipment and tag with a Legacy Process Equipment Tag."

Technical Safety Requirement 3.11.1 required that: "A Criticality Safety Program shall be established, implemented, and maintained as described sa the SAR and shall address the following elements [ including) procedure requirements land] posting and labeling requirements.".The failure to mark (i.e, post) and rope off equipment in accordance with the procedure implemanting and maintaining the criticality safety program requirements is a Violation (VIO 70-7001/97002-13).

After identification of the procedural non-compliance, the plant staff took immediate action to rope and post the legacy process equipment in the process buildings.

)

i c.

Conclusions 4

A non-conservative interpretation of the NCS requirements for legacy process equipment resulted in a violation of tha nwa%m implementing the program required in =iss 3.11.1.

E1.2 ' Autoclave Pressure Decav Testina Modification a.

Insoection Scone (88101)

The inspectors reviewed implementation of plant design changes required by Compliance Plan (CP) !ssue Number 3. The changes were required to allow for the full performance testing of autoclave pressure boundaries.

18

~.

~

o*

4

)

b.

Observations and Findinas Compliance Plan Issue Number 3 required the United States Enrichment Corporation (USEC) to make modifications to the autoclaves in Buildings 333A,337A, and 360 to allow for the separate testing of the functionality of the inner and outer autoclave containment boundaries. The CP also required USEC to submit a revised TSR to redefine the methods for the autoclave testing process.

The inspectors reviewed the plant piping and mstrumentation drawings (PIDs) documenting the installed changes. Each autoclave penetration was reviewed to identify the relied upon containment valves and to ensure that pathways were available to preclude back pressures (i.e., from other plant systems) on the valves during the testing process. In addition, c sampling walkdown of the piping was conducted.

During the review and walkdown process, the inspectors assessed conformance of the systems in the facilities to the PIDs and the TSR changes provided to the NRC. Allinstalled changes, made within the safety boundaries, were as described and documented in the PIDs and TSR changes. Some changes had not yet been implemented and the equipment was out of service. In addition, some minor errors were noted in the plant j

drawings. These were in an area just beyond the final containment valves.

j The inspectors noted that these same errors did not exist in the materials provided as a part of the TSR changes. At the end of the inspection period, the engineering staff was reviewing the engineering drawing change process to determine how the error was incorporated into the materials.

c.

Conclusions The inspectors determined that the plant had completed, for those autoclaves in service, the required engineering changes to the autoclave piping systems to allow for pressure decay testing without back pressure concerns.

E1.3 Linuid UF. Above Cvlinder Valves Durina Heatina a.

Insoection Scoce (88100)

The inspectors reviewed a plant staff identified issue relative to the potential for liquid UF, to be above the cylinder valve during normal heating operations.

b.

Observation and Findinas i

During the inspection period, plant staff continued the implementation of a LTO developed to address an apparent inconsistency between the SAR and the TSRs. Specifically, the inconsistency involved the SAR accident 19

analysis, which assumed that autoclave pigtail / valve accident scenarios would involve only a vapor release, and the TSR required cylinder fill limits for heating. Recent engineering calculations indicated that heated cylinders, when filled to the TSR specified limits, would have liquid UF, above the cylinder valve. Therefore, the SAR assumed accident would result in a liquid release vice the assumed vapor release. Preliminary engineering calculations indicated that a liquid release could result in changes to the assumed accident progression. The changes could include both a faster accident and one that developed a peak pressure greater than pressures the autoclaves were designed to contain.

The LTO included limitations on the fill weight for cylinders scheduled for i

heating. These limitations were lower than those included in the TSRs.

At the end of the inspection period, plant engineering staff continued with the investigation into the accident analysis assumptions and the need for a TSR change. Based upon the implemented interim corrective actions, the inspectors had no immediate safety concerns. Engineering activities to resolve this issue will be tracked as an inspector followup item (IFl 70-7001/97002-14).

c.

Conclusions Plant staff identified an apparent inconsistency between the TSRs and the SAR. The LTO was issued to mitigate the issue while further review was conducted. An inspector followup item will be used to track resolution of this finding.

E1.4 Nuclear Criticality Safety Confirmatorv Action Letter

)

i a.

Insoection Scoce (92703)

The inspectors monitored actions taken to address plant identified weaknesses in the implementation of some nuclear criticality safety j

approvals (NCSA) and a resultant NRC Confirmatory Action Latter (CAL).

b.

Observations and Findinos 1

During February 1997 NCS staff identified some inadequacies in the implementation of an NCSA written for the Building 400 cylinder wash j

operation. Specifically, the inadequacies were related to the independent verifications required by the NCSA. The independent verifications were relied upon to ensure that the cylinders had not contained uranium enriched to greater than one percent since the last wash and hydrostatic testing.

Initialindications were that the staff had processed some 22 cylinders without properly performing the required independent verifications.

20

1 es I

I Following identification of this issue, plant management shut down the

.[

cylinder wash operation and initiated an investigation. ~ Initial results

-indicated that another Building 400 operation, the spray booth, may have been' affected by similar independent verification problems. In response to i

this finding, plant management also shut down the Building 400 spray booth operation. In addition, management initiated an independent, outside review of the NCS program. The review was focused on assessment of the j

independence and effectiveness of the controls implemented.

Because of the potential safety significance of the findings and the timing relative to the NRC's assumption of regulatory authority, the NRC, issued CAL No. Rlll-97-003 on February 28,1997. The CAL was effective March 03, concurrent with assumption of regulatory authority, it described those actions committed to by plant management to ensure a thorough investigation and resolution of issues.

At the end of the inspection period, plant management continued to implement some actions described in the CAL.- All operations previously shutdown as a result of the initial findings remained shut down. The inspectors will continue to track completion of activities described in the CAL as an inspector followup item (IFl 70-7001/97002-15).

c.

' Conclusions The NRC issued CAL No. Alli-97-003 confirming management actions taken and intended in response to plant staff-identifind weaknesses in the implementation of some NCSAs. The CAL remained in effect at the end of-the inspection period. The inspectors will track the completion of the actions described in the CAL as an inspector followup item.

E1.5 Feed Facility Crane Brakes a.

Inspection Scone (88100):

The inspectors reviewed the issues surrounding a request for, and the NRC's subsequent granting of, enforcement discretion.

b.

Observations and Findinas On February 12, plant staff discovered that the actual field configuration of two feed facility cranes was not in agreement with the TSR specified design requirements. Sp6:ifically, the cranes did not have two direct current rectified shoe brabs. Instead, the cranes were equipped with two brakes;

-only one of which was direct current rectified.

In a February 14,1997, letter to the NRC, USEC requested enforcement discretion in order to allow the plant to continue to operate following transition to NRC regulatory oversight. The enforcement discretion was 21

requested beginning March 3, and until a TSR change could be processed by the NRC. This action was expected to take one to three months. On 1

February 28, the NRC provided a Notice of Enforcement Discretion (NOED),

NOED NO. GDP97-1, documenting approval of the requested discretion. The NOED was contingent upon USEC maintaining the hoist brakes in accordance with American National Standards institute (ANSI) 830.2,

" Overhead and Gantry Cranes," dated 1967.

The inspectors reviewed the NOED request and the circumstances which contributed to the need for the NOED. The inspectors noted that the Compliance Plan included severalissues related to past procedure and configuration controls problems. Previous weaknesses in these areas, as documented in the Compliance Plan and NRC Observation Reports written before March 3, resulted in plant staff having an incomplete or undocumented understanding of the actual configuration of soma systems.

As a result, plant staff responsible for developing the TSR information appeared to have relied upon plant records and not a physical walkdown of the equipment to ensure that the installed equipment matched the TSRs.

The inspectors also determined, through discussions with plant staff, that this issue may have been generically identified during the certification process. However, plant staff could not locate documented information demonstrating that the concern was previously reviewed or dispositioned.

Finally, the inspectors noted that current TSR and SAR changes were required to follow a much more rigorous process than used prior to or during the certification proceas. However, the inspectors also noted recent examples of weak implementation of this process. One such example is i

discussed in Section E1.6.

The inspectors will track closure of the concerns, for which NOED GDP97-1 was issued, as an inspector followup item (IFl 70-7001/97002-16).

c.

Conclusions Past weaknesses in the documentation and control of the plant configuration and reviews of TSR materials, during the certification process, contributed to the need for a NOED. The NRC issued the NOED on February 28,1997; effective March 3. An inspector followup item was issued to track completion of the NOED related actions.

E1.6 Imolementation of Plant G mae Process (10 CFR 76.68) a.

Insoection Scooe (88101)

The inspectors reviewed the plant staff's implementation of the 10 CFR 76.68 process.

22

.c b.

Observations and Findinas l

Comoliance Plan Issue 45: Codes and Standards During the inspection period, the plant issued changes to the TSR Bases Statements and the SAR as a result of work required by Compliance Plan issue 45, Codes and Standards. The inspectors performed a sampling review of the TSR Bases Statement and SAR changes. As a result, an incorrect change implemented to TSR 2.3.5.6, "UFe Condenser and Accumulator Minimum Wall Thickness," Bases Statement and SAR Section 4.3.3.1.2 was identified. Specifically, the inspectors identified that I

an approved change deleted a requirement that the referenced minimum wall thickness was in accordance with the 1986 American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section Vill, Division 1.

The change replaced this requirement with a reference to the Nationai Board Inspection Code. This latter code does not define nor include minimum wall thicknesses requirements.

The inspectors discussed the change with the responsible engineering staff and determined that the error was introduced during the administrative process for making the change. The original engineering input to the process was shown to be correct. However, the information was incorrectly applied and this fact was not identified by subsequent engineering or other reviewers. The inspectors also determined that the plant staff had issued a problem report to document and correct the finding. However, the inspectors determined that the problem report was closed without a root cause being determined for the problem. This approach appeared inconsistent with the potential generic consequences of such a problem and its obvious relationship to the TSRs.

Finally, the inspectors determined that the overall TSR requirement remained j

intact despite the incorrect change to the TSR Bases Statement and the SAR. Specifically, TSR 2.3.5.6 expressly required conformance with ASME requirements. However, the implementation of this change indicated a i

weakness in the configuration change control process.

4 Nuclear Criticality Safety Evaluations and Acorovals immediately prior to NRC assumption of regulatory authority on March 3, plant staff identified that changes had been made to the operation of the plant, as described in the SAR, without the performance of some required reviews. Specifically, changes were made to the nuclear criticality safety evaluations and approvals (NCSE/As), as described in SAR, Chapter 4, Appendix A, without review and updating of the material. The plant staff indicated that these activities were done, for the most part, prior to the NRC's certification of the plant. The staff also indicated that this SAR chapter was not reviewed based upcn an intemal informal understanding that the material would be removed from the SAR prior to March 3.

23

y l

Based upon the initial finding, an expedited review of the SAR was i

conducted to assure that the NCSE/A implemented changes did not result in j

prohibited operations. The review also resulted in changes to the SAR chapter necessary to bring it into conformance with the regulatory requirements prior to March 3. The plant staff did not identify any changes that would have been prohibited under the change control process or 10 CFR 76.68.

c.

' Conclusions Several examples of weak implementation of the change control process, 10 CFR 76.68, were identified. In each case, inadequate reviews allowed changes to occur which were either incorrect or not properly developed.

E8.

Miscellaneous Enaineerina lasues (90712)

E8.1 (Closed) Certificatee Event Report 31985: On March 19,1997, the plant staff reported that an NCS walkdown of the Building 710 Metallurgy Laboratory identified operations: (1) which could involve fissile material; and (2) which were conducted without an approved NCSA. Specifically, cutting and grinding operations were conducted in the building which created the potential for the accumulation of fissile material in equipment reservoirs. Subsequent analyses of '

i materials deposited on the equipment and in the reservoirs demonstrated that none of the operations involved uranium enriched to greater than 1.0 w/% or greater than 15 grams of uranium-235. Based on these results, an NCSA for the operations was not required by TSR 3.11.2 (CER 70-7001/97002-17).

The inspectors reviewad the sampling results and had no further questions.

E8.2 (Ocen) Certificatee Event Report 31997: On March 21,1997, the plant staff reported that an NCS walkdown of the Building 400 Chemical Operations Facility identified an air capture system and a negative air machine (NAM), used for the maintenance operations in the area, which were not covered by an approved NCSA.

Following discovery, the NAM was removed from service and the air capture system was posted and secured until an approved NCSA was developed. The plant staff determined that the air capture system was originally designed with safe geometry at the current authorized maximum enrichment. Therefore, there

)

appeared to be no immediate safety hazard. This event report is open pending a review of the certificatee's written report and re-start of the NAM unit (CER 70-7001/97002-18).

l 24

a IV. Plant Suonort F1.

Conduct of Fire Protection Activities F1.1 Process Buildino Sorinkler System Recortability a.

Insoection Scoos (88100)

The inspectors reviewed reportability of a plant staff-identified deficiency and declaration of inoperability for a Building 331 sprinkler system.

b.

Observations and Findinos On March 12, at 9:30 a.m., plant fire department personnel performing a fire protection system walkdown discovered a disconnected sprinkler pipe in system 33 (ground floor) of Building 331 between columns V-33 and W-33.

Upon discovery, the PSS declared the system inoperable and Building 331 personnel initiated a fire patrol for the affected area as required by the TSR 2.4.4.5, Limiting Condition for Operations Action Statement, " FIRE PROTECTION SYSTEM - Building SPRINKLER SYSTEM." The applicability statement for the TSR required that the sprinkler system be operable during modes Cascade 1 through Cascade 3, "except when tube oil is valved off or removed from the cells covered by a specific sprinkler system." Since Modes Cascade 1 through Cascade 3 cover all operating cells (with lube oil valved in) as well as cells not in use, the sprinkler system was required to be operable at the time of discovery.

On March 13 the inspectors were informed of the event at a routine plant status meeting. The inspectors inquired whether or not the event was reportable. The PSS informed the inspectors that the finding was not reportable based upon the systems distance from the building lube oil tanks.

The PSS and other operations staff indicated that only those sprinkler systems required to mitigate fires in building zones directly adjacent to the lube oil tanks were covered by TSR 2.4.4.5.

Subsequent to these conversations, engineering performed an evaluation of the building sprinkler systems and the relationship to the design fire (i.e., a lube oil system fire). The evaluation indicated that the buildings had no definable fire barriers and that all branch lines were part of the system that provided coverage for the entire building. This analysis was ultimately documented in an " Engineering Position on Reportability" which was provided to plant staff to clarify the extent of the system covered by the TSR.

10 CFR 76.120(c)(2) requires a 24-hour report for an event for which equipment is disabled when the equipment is required by TSR to be available and no redundant equipment is available or operable to perform the required 25

i o

t

+,

i-

~

~ safety function. As no redundant system was avalisale to provide the safety j

function of the fire protection system when it was required to be

{.

operable, the failure to report the inoperable status of system 33 between l

9:30 a.m., on March 12 and March 27, over 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> after the system was t

j discovered to be inoperable, is a Violation of 10 CFR 76.120(c)(2)

(VIO '70-7001/97002-19).

L The violation appeared to result from the PSS's incomplete understanding of the engineering basis for defining reportability requirements under l

2 TSR 2.4.4.5. In addition, the inspectors noted that NRC had neither reviewed nor approved any " zones of coverage" for the cascade building sprinkler systems. In addition, the plant staff had never demonstrated that a l

fire in one part of a process building would not spread to other areas, including those containing significant inventories of lube oil.

j l '

The PSS reported the event to the NRC on March 27 (NRC Event Re' port 1

No. 32039), after completion of the engineering evaluation. Subsequently, the plant staff reported all sprinkler system inoperabilities.

L c.

Conclusions.

t An incomplete understanding of the engineering basis for which sprinkler l

systems were required to be operable under TSR 2.4.4.5, and hence l

reportable when discovered to be inoperable, led to a violation for failing to report an event required by 10 CFR 76.120(c)(2).

F1.2 Inonerable Fire Protection Systems a.

Inspection Scone (88100)

Throughout the inspection period, the inspectors reviewed the progress of a L

plant staff-initiated walkdown of the entire TSR required process building fire -

l protection sprinkler systems.

4' b.

Observations and Findinas l

l The fire protection manager initiated a walkdown of the entire fire protection F

sprinkler system in the procoss buildings after repeat problems were identified early in 1997. The walkdown was conducted by plant fire department personnel and consisted of floor level observation and documentation of anomalies associated with each sprinkler system and branch line. The documented anomalies were later reviewed and dispositioned by a fire protection engineer. Systems identified with design or National Fire Protection Association (NFPA) code nonconformances were i

declared inoperable and the LCO Action Statement for TSR 2.4.4.5, " FIRE PROTECTION SYSTEM - Building SPRINKLER SYSTEM," was entered. The TSR Action Statement required hourly fire patrols, initiated within two hours 1

of discovery, as a compensatory measure.

26

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

k '

l a

J During the inspection period, plant staff notified the NRC of a number of events which were identified through the walkdown process. The specific event reports are referenced in Section F.8. The eight fire protection events reported to the NRC documented the deficiencies discovered that were not j

in conformance with either plant design documents or NFPA code. The five i

basic type of deficiencies reported included: 1) obstructed sprinkler heads; I

2) sprinkler heads which were not aligned vertically; 3) missing sprinkler 4

system branch lines and missing heads; 4) branch piping not connected to-

)

i the high pressure fire water system header, and; 5) sprinkler systems

)

installed above exhaust fans during a modification project in 1975 and 1976 l

l that were never connec'ted to the building systems.

Plant staff speculated that the deficiencies resulted from a general lack of

]

i.

ownership of the fire protection system over a period of years in addition, the plant staff's historical lack of rigorous configuration control and work j

control programs created a situation in which sprinkler systems, removed from service or altered to support other modification projects or maintenance l

work, were not necessarily checked to ensure proper return of the systems j

to the design configuration.

' The overall immediate safety impact of the fire protection system L

. deficiencies did not appear to be significant. The deficiencies were distributed over numerous systems; no individual systems' deficiencies involved a major portion of the piping; immediate compensatory measures were promptly implemented; and, long-term corrective actions were i

implemented or scheduled for all the deficiencies.

The inspectors noted that the plant was neither required to perform a detailed system review nor to conform to NFPA code as a condition of i

certification. Additionally, SAR Section 5.4.1, " Fire Protection Program,"

i stated only that the fire protection program would comply with NFPA Standard 13, " Sprinkler System," for modifications to the plant following certification. The plant staff's program for identifying items which did not conform to both historical design documents and NFPA code and the action

=

to implement modifications necessary to resolve these discrepancies exceeded the SAR requirements, s

l Section Vll.B.3 of the Enforcement Policy (NUREG 1600) addresses the j-exercise of discretion for violations involving old design issues. Under that section, discretion may be considered when the violation is licensee-identified as a result of voluntary initiative; corrective action is taken i

within a reasonable time following identification; and the violation would not j

normally be identified by routine activities such as surveillances or QA. In addition, the NRC may refrain from issuing an NOV for cases that meet the i-above criteria provided the violation was caused by conduct that is not reasonably linked to present performance.

U A

27 a

8 Q

o i

The plant staff identified and evaluated anomalies documented during the process building fire protection sprinkler system walkdown. These included both old design and installation deficiencies, as well as, NFPA code inconsistencies. Modification and work packages were developed and approved to correct a number of system deficiencies and more were in progress at the end of the inspection period. As of the end of the inspection period, corrective maintenance or modifications were completed on approximately 4r % of the deficiencies identified. The plant fire protection management was continuing to address all deficiencies identified and developed a punch list to track items to closure.

In addition, the inspectors noted that the facility was required by the NRC to develop work control and configuration control programs to ensure unauthor; zed maintenance or modification betivities for safety systems did i

not occur in the future. These programs were put in place prior to NRC's assumption of regulatory authority on March 3.

Finally, the plant had conducted acceptable surveillances tests (flow alarm tests) which were adequate to test the system performance. However, these tests were not were not of the type or nature that would have identified the deficiencies found during the walkdown.

Technical Safety Requirement 2.4.4.5 requires, in part, the fire protection sprinkler systems in process Buildings C-331, C-333, C-335, C-337 to be operable during Modes Cascade 1 through Cascade 3. Based on the considerations provided above, the self-identified and corrected violations are being treated as a Non-Cited Violation, consistent with Section Vll.B.3 of the i

NRC Enforcement Poliev.

c.

Conclusions The plant fire protection management undertook a significant effort to perforrn a detailed walkdown of the entire TSR related fire protection sprinkler system based on identification of repeat problems. As a result, further discrepancies, ban from a design and from a configuration control i

standpoint, were identified. The self-identified and corrected violations are being treated as a Non-Cited Violation, consistent with Section Vll.B.3 of the NRC Enforcement Policy.

F8.

Miscellaneous Fire Protection issues (90712)

F8.1 (Ooen) Certificatee Event Reoort 31972: Discovery of two sprinkler heads obstructed in Building 337 sprinkler system C-2. The obstruction was identified as part of a sprinkler system walkdown discussed in Section F1.2. The written report dated April 18 encompassed all the sprinkler systern inoperabilities identified through April 16. The report identified the root cause of the spankler system deficiencies as a failure to have effective programs in place relative to configuration management and work control. The report identified that, in compliance with 1

28 i

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y c.-.,

l l

l l

requirements in 10 CFR Part 76 and the Compliance Plan, the plant management i

+

. developed and implemented a configuration management program, a modification i

design control program, a problem reporting system, and a work control process.

The plant management also identified planned corrective actions to include:

1 1)

. Complete walkdown of all sprinkler systems which are visible or accessible without a permit by May 15,1997; 2)

Conduct an end-point assessment within 120 days of completing work under the engineering service order (ESO Z98400) covering the repair or l-installation needed to address the sprinkler system deficiencies. The l

purpose was to ensure all sprinkler systems were capable of performing the installed function:

3).

Update drawings affected by modifications under ESO Z98400 to reflect "os built" conditions by ' July 1,1997; and, l

4)

Revise procedure CP4-SS-FS6111, "TSR Surveillance, lespection, and Testing of Wet Pipe Sprinkler Systems," to include more detailed inspection l

criteria by August 25,1997.

l-l' The event report is open pending a review of the implementation of the corrective actions ~ described in the written report (CER 70-7001/97002-20).

F8.2. (Onen) Certificatee Event Reoort 31977:- Sprinkler system pipe struck in Building 335 by maintenance personnel. While moving a large tool box in Building 335 on the cell floor, a maintenance mechanic struck a pipe on sprinkler system 27 just above the inspector test valve (ITV). The impact caused a pipe coupling to leak which initiated a water flow alarm in the Central Control Facility (CCF). The written report, submitted to the NRC and dated April 16, identified the root cause'as j

inadequate training, in that, personnel were not properly trained to maintain control of the 1000-Ib. tool box. The event report is open pending a review of the m

l implementation of the corrective actions described in the written report (CER 70-7001/97002-21).-

t F8.3 (Open) Certificates Event Reoort 320Q2: Discovery of sprinkler head obstructions l

i and sprinkler heads at 45* from verticalin Building 331 sprinkler i

systems 21 and 29. See discussion under Sections F1.2 and F8.1. The event report is open pending a review of the implementation of the corrective actions described in the written report discussed in F8.1 (CER 70-7001/97002-22).

F8.4 (Open) Certificatee Event Reoort 32012: Discovery of incorrect sprinkler head spacing, inadequate sprinkler head coverage, and obstructed heads associated with systems B16 and C10 in Building 337 and C2 in Building 333. See discussions under Sections F1.2 and F8,1. The event report is open pending a review of the t

implementation of the corrective actions described in the written report discussed in i

F8.1 (CER 70-7001/97002-23).

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l F8.5 (Open) Certificatee Event Reoort 32020.: Discovery of branch line missing l

(3 sprinkler heads) for Building 331 sprinkler system 17 under a unit bypass. See discussions under Sections F1.2 and F8.1. The event report is open pending a

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review of the implementation of the corrective actions described in the written report discussed in F8.1 (CER 70-7001/97002-24).

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F8.6 (Open) Certificatee Event Reoort 32039: Discovery of sprinkler system 33 branch pipe not connected to the high-pressure fire water header in Building 331. See discussion under Sections F1.2 and F8.1. The event report is open pending a review of the implementation of the corrective actions described in the written l

report discussed in F8.1 (CER 70-7001/97002-25).

F8.7 (Ocen) Certificatee Event Reoort 32080: Discovery of a sprinkler system branch pipe not connected to a high-pressure fire water header in Building 333 system B9.

See discussion under Sections F1.2 and F8.1. The event report is open pending a review of the implementation of the corrective actions described in the written l

report discussed in F8.1 (CER 70-7001/97002-26).

F8.8 (Ocen) Certificatee Event Reoort 32121: Discovery of ITVs without flow restricting orifices for Building 331, systems 13 and 17, and Building 333, systems A2, B1, B13, and C5. See decussion under Sections F1.2 and F8.1. The event report is open pending a review of the implementation of the corrective actions described in the written report discussed in F8.1 (CER 70-7001/97002-27).

F8.9 (Open) Certificatee Event Reoort 32144: Discovery of branch line not connected to the high pressure fire water header in Building 337 system A5. See discussion in Sections F1.2 and F8.1. The event report is open pending a review of the implementation of the correct;ve actions described in the written report discussed in F8.1 (CER 70-7001/97002-28),

f F8.10 (Ocen) Certificatee Event Reoort 32162: Discovery of branch lines not connected to the high pressure fire water header for sprinkler systems B1, B4, B5,89, B13, C1, C5, C9, C13, D1, DS, D9, and D13 of Building 337. Discovery of branch lines not connected to high pressure fire water header for sprinkler systems A1, A5, A9, A13, B1, B5, B9, and C13 in Building 333. See discussion under Sections F1.2 and F8.1. The event report is open pending a review of the implementation of the corrective actions described in the written report discussed in FS.1 (CER 70-7001/97002-29).

F8.11 (Ocen) Certificatee Event Reoort 32170: Discovery of branch lines not connected l

to the high pressure fire water header for sprinkler systems 3 and 31 in Building l

331 and sprinkler systems A1, A5, A9, and A13 in Building 337. See discussion under Sections F12 and F8.1. The event report is open pending a review of the l

implementation of the corrective actions described in the written report discussed in F8.1 (CER 70-7001/97002-30).

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-l S1, Conduct of Security and Safeauard Activities S 1.1 Access to Classified Matter a.

Insoection Scone (90712) l l

The inspectors reviewed the circumstances surrounding the discovery of classified matter in an area available for use by individuals not possessing a "O" or "L" clearance.

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b.

.Qbservations and Findinas l

On April 15 and 17, the PSS provided one-hour notifications to the NRC l

Region lli office of potential compromises of classified matter. The

)

information was located in areas of the plant frequented by individuals not possessing a "O" or "L" clearance. The matter had not been controlled as classified matter (Confidential Restricted Data).

The issue was identified by security staff performing work for the Department of Energy. However, the matter in question was within the space leased by USEC and in areas which were in routine use. The classified matter remained within the plant staff's protected area, although it was accessible to individuals within that area who did not possess a "O" or "L" clearance. Following discovery, the classified matter was removed from j

the locations and promptly secured. The security management undertook a review of other documents and locations where the matter could have been l

located, but did not identify any additional cases of uncontrolled access l

being available.

)

I Although the plant staff identified the violation, the inspectors were aware that similar previous problems with the control of this particular classified matter. The repeat nature of this issue indicated that past corrective actions were inadequate. The problem appeared to be that the matter was at one time not classified, and efforts to find and control all the materials onsite after it was classified were not fully successful. Thus, the critoria of Section Vll.B.1 of the NRC Enforcement Policy (NUREG 1600) were not met.

10 CFR 95.35(a) requires, in part, that no person subject to the regulations in this part may receive or may permit any individual to have access to l

matter revealing Confidential Restricted Data unless the individual has a "O" or "L" access authorization. The failure of the plant management to control access to the classified matter in plant drawings is a violation (VIO 70-7001/97002-31).

c.

Conclusions Plant staff identified two examples of inadequate control of c%ssified matter, a violation of 10 CFR 95. The matter involved was immediate. y secured.

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5 S1.2 Security Radio Network Comoensatorv Measurgs a.

Insoection Scope (88100)

The inspectors reviewed actions taken in response to problems with the plant security radio network.

b.

Observations and Findinas On March 6, the inspectors noted a problem report indicated recurring problems with the site security radio network. Specifically, plant staff identified that the site security radio network did not always relay communications broadcast using the plant radio system."all-call" feature.

This feature was used to inter-connect the plant radio networks during events and other situations, allowing a single radio broadcast to reach all site staff carrying hand-held radios.

At the time of these problems, the security radio network was utilized as a i

compensatory measure for a Compliance Plan issue. Specifically, the Compliance Plan Issue 50 justification for continued operations (JCO) required personnel, entering known areas of criticality accident alarm system i

inaudibility, to have and test the functionality of a hand-held radio. The JCO l

further indicated that personnel were expected to c; otinuously monitor the radio for "a page delivered over all radio frequencies," the all-call feature.

The inspectors discussed with PSS staff those compensatory actions taken as an inimediate respor.se to the identified equipment problems. The inspectors were infcimed that both fire and security personnel were made aware of the problems and that other PSS staff were made knowledgeable of the issue through the PSS log. Fire and security personnel were also directed to listen to the public address system and building howlers for information during a site wide problem. The PSS staff also indicated that the Cascade Coordinators, the originators of most site wide "all-call" announcements, were made aware of the problem. The Cascade Coordinators were also verbally instructed to make a separate announcement on the site security radio network of any messages made using the "all-call" feature. However, these instructions were not documented in either the Cascade Coordinator logs or in a LTO.

During a review of the inspector's observations, the PSS manager discussed the issue with a sampling of Cascade Coordinator staff. Each of the staff indicated a clear understanding of the expected actions to be take whertsver the radio system "all-call" feature was used. The individuals were aware of the need to make a separate call over the security radio network. They also appeared knowledgeable of the bases for these instructions.

Notwithstanding these findings, the PSS manager issued an LTO to formally document these instructions regarding the "all-call" process and to ensure continued adherence to the JCO for Compliance Plan issue 50.

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The inspectors identified a failure to document the immediate corrective actions taken to a plant-identified problem with the site security radio network in the emergency response procedure. As a result, appropriate actions to notify personnel over the security network may not have been implemented in a real emergency over an extended period of time.

T1.

Conduct of Transoortation Activities T1.1 Handlina and Transoortation of UF. Filled Cvlinderg l

a.

Insoection Scooe (88100)

The inspectors reviewed issues associated with the plant staff's handling and transportation of UF, filled cylinders.

b.

9bservations and Findinas UF. Cvlindar Valve Tinnina On April 9, plant engineering staff identified that the current plant practice for the tinning of new and replacement UF, cylirider valves was not consistent with American National Standards Ir.stitute (ANSI) Standard 1

N14.1, " Uranium Hexafluoride Packaging for Transport." Specifically, ANSI N14.1-1990 required the use of ASTM B32, SOA solder for cylinder valve tinning. However, plant practice was to use a mixture of this and another solder for the process. The staff also noted that the NRC approved transportation quoEy assurance program, the NRC Certificate of Compliance (CofC) for Radioactive Materials Packages, No. 6553, Revision 11, and 10 CFR 71.5 required conformance with ANSI N14.1. As a result, the plant staff halted all UF, filled cylinder shipments.

Subsequent to this initial finding and immediate action, engineering and regulatory staff held numerous discussions with other cylinder users. The purpose of these discussions was to ensure that the extent of the issae was fully understood and could be immediately addressed. In addition, an engineering equivalency review was initiated in an effort to essess the safety significance of using an incorrect solder mixture. The inspectors reviewed these evaluations and noted that none of the solder's critical parameters appeared to have been significantly changed by mixing the solder types.

Based upon engineering evaluations, site and corporate regulatory staff requested, from both the NRC and the Department of Transportation (DOT),

modifications to the NRC CofC and DOT regulations to allow resumption of shipping. Temporary changes and/or exemptiens were authorized by both the NRC and DOT, pending further reviews.

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l During review of the issue, the inspectors were informed that the current practice, to use a solder mixture vice the ANSI N14.1-1990 specified solder, had been in place for in excess of ten years. Additionally, the inspectors determined that plant staff had previously performed a review of all NRC l

CofC and Quality Assurance Program requirements; however, this l

discrepancy was not identified.

l The plant resumed shipments of both enriched and normal cylinders on April 16,1997.

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Cvlinder Tare Weichts l

As a followup to previous inspection efforts, the inspectors reviewed the plant staff's marking of UF, cylinders with revised tare weights, as determined during their periodic hydrostatic testing program. Through walkdowns of cylinders, located in the feed facility yards and awaiting heating, the inspectors determined that most toll normal cylinders were not marked with revised tare weights. The inspectors also noticed that the cylinder's bore other marking indicating that one to several hydrostatic retest l

had been performed. Some cylinders were observed that had the tare l

I weights revised. These were from one of the plant's toll normal material suppliers.

l The inspectors discussed these observations with engineering staff. The l

l engineering staff stated that the periodic hydrostatic testing process included a rebaselining of the cylinder tare weights. However, the plant i

staff did not remark the cylinders with these revised tare weighs. Instead,

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the data was maintained in a separate site-specific log book. The inspectors noted that this practice appeared to limit the availability of accurate tare l

weight data to only plant staff. Additionally, this approach appeared to provide incorrect data (i.e., the outdated tare weight marked on the cylinder) l to non-site users of the cylinders. During these discussions, the inspectors identified at least one scenario by which the plant's current practices could result in the heating of an overfilled cylinder.

Following these discussions, the inspectors reviewed ANSI N14.1 and noted that the Standard required that cylinder nameplates incorporate specific rrarkings, including the cylinder tare weight. The presence of accurate tare weight data on the cylinders appeared significant, in that, the maximum al; owed UF fill weight for a cylinder was determined, in part, based upon its i

tare weight. In addition, the standard acknowledged that cylinders were often used and filled by entities other than the owners.

L Gubsequently, the inspectors reevaluated a previous plant study of cylinder tare weights.- The study showed that cylinders, exposed to environmental and process conditions, experienced exterior surface metal wastage. The study also indicated that cylinder tare weights changed, as a result of the,e conditions, from a few up to sixty plus pounds between the periodic 34

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to 6..

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hydrostatic testings. The inspector estirr.ated, based upon generic data, that each 100 pound change in tare weigh resulted in an approximate one half percent decrease in minimum ullage available for,:ylinders filled to the maximum weight. The ANSI Standard requires a minimum five percent l

ullage for cylinders filled to the maximum weight. Therefore, the inspectors l

determined that the ANSI ullage limit could be exceeded if the cy!inders did j

not have accurate stamped tare weights.

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10 CFR 71, requires, in part, that shipments of licensed material be made m accordance with applicable NRC and DOT regulations. The NRC CofC for the shipment of licensed materials, using the Paducah Overpack, requires, in part, that cylinders valves and plugs are tinned in conformanca with ANSI l

N14.1-1990. 49 CFR 173, requires, in part, that packages (cylinders) used j

for shipments of uranium hexafluoride must be marked in accordance with ANSI N14.1. The failure to tin cylinder valves and plugs with the proper solder and the failure to mark the correct tare weigh on cylinders offered for shipment is a Violation of 10 CFR 71 (VIO 70-7001/97002-31).

c.

Conclusions Two examples of inadequate implementation of ANSI N14.1 were identified, a violation of 10 CFR 71. Each example occurred, in part, due to the continuation of long standing past practices without assurance of their conformance to the applicable requirements.

V. Manaaement Meetinas X.

Ex'it Meetino Summarv The inspectors presented the inspection results to members of the plant staff management at the conclusion of the inspection on April 21,1997. The plant staff acknowledged the findings presented.

The inspectors asked the plant staff whether any materials examined during the i

inspection should be considered proprietary. No proprietary information was identified.

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i PARTIAL LIST OF PERSONS CONTACTED l

United States Enrichment Corooration

'J. H. Miller, Vice President - Production

'J. M. Brown, Engineering Manager "J. A. Labarraque, Safety, Safeguards and Quality Manager Lockheed Martin Utility Services (LMUS) 1

• S. A. Polston, General Manager l

• H. Pulley, Enrichment Plant Manager

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

• S. R. Penrod, Operations Manager i

"C. Hicks, Site and Facility Support Manager United States Deoartment of Enerav (DOE)

• G. A. Bazzell, Site Safety Representative l

Nuclear Reaulatory Commission (NRC)

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

'J. M. Jacobson, Resident inspector

• Denotes those present at the April 21,1997 exit meetings.

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

INSPECTION PROCEDURES USED IP 88100 Plant Operations IP 88102 Surveillance Observations IP 88103 Maintenance Observations IP 88105 Management Oversight and Controls IP 88020 Regional Criticality Safety l

IP 90712 Inoffice Review of Events IP 92703 Confirmatory Action Letters l

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ITEMS OPENED. CLOSED, AND DISCUSSED

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Onened I

l 70-7001/97002-01 VIO failure to implement minimum staffing requirements j

70-7001/97002-02 VIO failure to control hours of work j

i 70-7001/97002-03 VIO failure to follow an alarm response procedure 70-7001/97002-04 VIO failure to maintain CAAS system operable 70-7001/97002-05 CER inoperable criticality alarm system 70-7001/97002-06 CER autoclave safety system actuation: wics

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70 7001/97002-07 CER 10 CFR 21 report on cylinder valve packing nuts l

70-7001/97002-08 CER inoperable criticality alarm system 70-7001/97002-10 CER failure to follow NCSA Gen-27 70-7001/97002-11 CER autoclave safety system actuation: wics j

70-7001/97002-13 VIO failure to follow NCSA Gen-27 70-7001/97002-14 IFl SAR and TSR inconsistency for cylinder UF, level 70-7001/9700215 IFl cal for inadequate implementation of ncs requirements 70-7001/97002-16 IFl need for crane brakes TSR inaccuracy 70-7001/97002-18 CER failure to implement an ncsa for fissile operations

{

70-7001/97002-19 VIO failure to make a required 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> report 70-7001/97002-20 CER inoperable fire protection sprinkler systems a

J 70-7001/97002-21 CER inoperable fire protection sprinkler systems 70-7001/97002-22 CER inoperable fire protection sprinkler systems 70-7001/97002-23 CER inoperable fire protection sprinkler systems 70-7001/97002-24 CER inoperable fire protection sprinkler systems i

70-7001/97002-25 CER inoperable fire protection sprinkler systems 70-7001/97002-2G CER inoperable fire protection sprinkler systems 1

70-7001/97002 27 CER inoperable fire protection sprinkler systems 70-7001/97002-28 CER inoperable fire protection sprinkler systems 70 7001/97002-29 CER inoperable fire protection sprinkler systems 70-7001/97002-30 CER inoperable fire protection sprinkler systems 70-7001/97002-31 VIO inadequate control of classified materials 70-7001/97002-32 VIO failure to follow transportation requirements (ANSI)

Closed 70-7001/97002 09 CER untimely implement TSR Action Statement requirements 70-7001/97002-12 CER inoperable CAAS system due to use of non-safety-related components

' 70-7001/97002-17 CER failure to develop an ncsa for fissile operations Discussed None 37

so c a LIST OF ACRONYMS USED l

i ANSI American National Standards institute j

ARP Alarm Response Procedure

]

ASME American Society of Mechanical Engineers i

ASTM American Society for the Testing of Materials CAAS Criticality Accident Alarm System CAL Confirmat6ry Action Letter CER Certificatee Event Report CFR Code of Federal Regulations i

CP Compliance Plan IM instrument Mechanic l

l LCO Limiting Condition for Operation L

LS Limited Spacing LTO Long Term Order NCS Nuclear Criticality Safety NCSA Nuclear Criticality Safety Approval t

NCSE Nuclear Criticality Safety Evaluation l

NCV Non-Cited Violation l

NOED Notice of Enforcement Discretion NOV Notice of Violation NRC Nuclear Regulatory Commission PEH Planned Expeditious Handling PGLD Process Gas Leak Detection PIDs Piping and Instrumentation Drawings PSS Plant Shift Supervisor SAR Safety Analysis Report TSR Technical Safety Requirement UF6 Uranium Hexafluoride UH Uncomplicated Handling USEC United States Enrichment Corporation VIO Violation i

3 38 l