IR 07100008/2011024

Insp Rept 70-7001/97-11 on 971008-1124.Violations Noted. Major Areas Inspected:Plant Operations,Maint & Surveillance, Engineering & Plant Support

ML20203E516
Person / Time
Site:	07100008, 07007001
Issue date:	12/09/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20203E451	List: IR 07100008/2011024 ML20203E447 ML20203E490
References
70-7001-97-11, NUDOCS 9712170081
Download: ML20203E516 (19)

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U.S. NUCLEAR REGULATORY COMMISSION REGIONlli Docket No: 70 7001 Report No: 70-7001/97011(DNMS)

Facility Operator: United States Enrichment Corporation Facility Name: Paducah Gaseous Diffusion Plant Location: 5600 Hobbs Road P.O. Box 1410 Paducah,KY 42001 Dates: October 8 throu9h November 24,1997 Inspectors: K. G. O'Brien, Senior Resident inspector J. M. Jacobson, Resident inspector R. G. Krsek, Fuel Cycle Inspector, Region til

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Approved By: Patrick L Hiland, Chief Fuel Cycle Branch 9712170001 971209 PDR 600CK 07007001 C PDR _ . _

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EXECUTIVE SUMMARY

United States Enrichment Corporation Paducah Gaseous Diffusion Plant NRC inspection Report No. 70-7001/97011(DNMS)

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Plant Operations

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+- Plant sta# identified a failure to make timely event reports to the NRC for two events l involving loss of power to the alarm annunciation function for the high-voltage process j-I gas leak detection system in Building C-310 for uranium hexafluoride condensers,

, l accumulators, and associated piping. The cause of the loss of power had not been '

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determined as of the end of the inspection per;od. (Section 01.1)

5 -+ The inspectors identified that a Technical Safety Requirement violation occurred, in that, a required four-hour verbal report was not made to the NRC upon identification of a -

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deficient criticality safety ar#ysis. The inspectors determined t'aat the report was not made as the resun of the citicality safety staff's non-rigorous evaluation of an operational anomaly and imprecise communications between criticality safety and operations staf (Section 01.3)

Maintenance and Surveillance

+ Plant ste'f conducted appropriate functional tests of the safety systems for the east

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. Normetex product withdrawal pump following a plant change to remove non-safety-

- related loads from the pump's overloaded control circuit. (Section M1.1)

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• Plant staff performed the monthly surveillance of the diesel high-pressure fire water pump

in accordance with Tachnical Safety Requirement and procedural requirements, but discovered that staff had not used a calibrated flow meter for the last annual surveillance of water flow for two of four high-pressure fire water pumps. (Section M1.2)

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• - Plant staff began a self-assessment of the process foridentifying and documenting

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Augmented Quality-Nuclear Criticality Safety systems and components in the Boundary Definition Manuals which the inspectors will continue to follow. (Section E1.1)

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• The inspectors determined that an engineering notice and evaluation, completed in

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response to NRC inspectors' ouestions, inappropriately authorized the continued use of fissile waste storage drums with wall thicknessen less than the dimensions assumed in the current nuclear criticality safety evaluations, a Technical Safety Requirement violatio The current nuclear criticality safety evaluations were also noted to included non-p conservative as3umptions and methods. (Section E1.2)

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• The inspectors identified that the assessment of and response to an anomalous Building C 335 holding-drum-room temperature reading did not identify the absence of a currently implemented criticality safety evaluation,' approval, or procedures for the holding drums, a d

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i Technical Safety Requirement violation. Some deficiencies in previous controls for volving orders were also identified. (Section E1.3)

Plant Support

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+ Plant emergency response staff demonstrated an effective knowledge of the

!. requirements of the associated emergency plan implementing procedure in a criticality accident alarm drill for Building C-400. (Section P1,1)

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. The inspectors identified a failure by the plant staff to make a required one-hour verbal

report of the discovery of an infraction of classified information controls, a Technical Safety Requirement violation. The violation occurred, in part, due to a non-rigorous evaluation of the event by operations management. (Section S1.1)

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• The inspectors observed several routine material receipt inspections, and noted the inspections were thorough and performed in accordance with approved procedures. A-non-cited violation was identified regarding the use of management memoranda in lieu of an approved, written procedure. (Section Q1.1) .

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DETAILS

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l. Operations

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0 Conduct of Operations-01.1 S.gibling C-310 Process Gas Leak Detector Inocerability

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a- Inspection Scope (88100)

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The inspectors reviewed the circumstances surrounding the tripping of a safety-related 4 breaker which supplied power for safety systems and alarms in Building C-310 (Product r . Withdrawal). , Observations and Findinas i . On October 17, at approximately 11:15 p.m., the east Normetex *thdrawal pump tripped while operators were starting the pump. Building C-310 operators sutsequently tested the alarms in the area control room (ACR) and identified that power to the alarm annunciator panel in the ACR had been lost. The panel included the alarms for the high-

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voltage process gas leak detection (PGLD) system, required by Technical Safety Requirement (TSR) 2.3AA, to be operable because operations in the uranium hexafluoride (UF.) condensors, accumulators, and associated piping involved UF above atmospheric pressure (liquid UF.).

After discovering the inoperable alarms, the operators re-directed the direct current (DC)

power for the alarms from the normal sitemating current (AC) rectifer to the backup

battery room. However, this action did not restore power to the alarm annunciator Continued investigation identified, at approximately 11:40 p.m., that the breaker supplying

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power to the annunciator cabinet had tripped. After resetting the breaker, the operators

> tested the alarms and noted that power was restore On October 18, at approximately 2:00 p.m.s the east Normetex pumped tripped again while it was on stream and compressing UF., Building C-310 operators found the annunciator breaker was tripped again and reset the breaker. Alarm power for the high voltage PGLD was lost for approximately two minutes.

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The pump trip was caused by a blown fuse in the control circuit which further review by

engineering staff identified as electrically overloaded. The pump trip was initially suspected to be involved with the annunciator cabinet breaker trip, insofar as three alarms actuate in the ACR when the east Normetex pump trips. However, further review and testing demonstrated that the pump trip could not have caused the annunciator

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breaker trip. As of the end of the inspection period, the root cause for the breaker trip had not been determined.

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i I Topical headings such as 01, MS, etc., are used in accordance with the NRC standardized inspection report i

, l outline contained in NRC Manual Chapter 0610. Individual reports are not expected to address all outline sopic and the topical headings are therefore not always sequentia .

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l For each event, the appropriate TSR Limiting Condition for Operation (LCO) Action .-

Statement was entered and a smoke watch for the area was initiated. However, the ever ts were not recognized by the plant shift superintendent (PSS) to be reportabl Upon retum to work on October 20, staff from the Nuclear Regulatory Affairs Division j

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identifk.1 that the events were reportable in accordance with 10 CFR 76.120 and the ,

controlling procedure, and that the NRC had not been notified within the required 24-hour !

period from the time of the events, j i

Technical Safety Requirement 3.g.1 required that written procedures be prepared and ,

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implemented to cover the activities descnbod in Safety Analysis Report (SAR)

. Section 6.11.4.1 and listed in Aopendix A to SAR Section 6.11. Appendix A to SAR i

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Section 6.11 identified " investigations and reporting" as an activity requiring an administrative procedure, Step 6.1.2 of Procedure UE2 RA-RE1030, * NUCLEAR REGULATORY EVENT REPORTING," Revision 2, dated February 28,1997, required

that the plant shift superintendent (PSS) review problem reports and determine

reportability of the event cr condition using Appendix D. Stop 6.2.1E of the procedure i required that the PSS vert > ally notify the appropriste NRC office within the time requirements shown in Appendix D. Appendix D, "NRC REPORTING CRITERIA," of Procedure UE2-RA RE1030, specified the criteria and reporting time for events and conditions. Specifically, Criterion .l.1.a requirtd a 24-hour report for an event in which equipment is disabled or fails to function when it is required by TSR to mitigate the consequences of an accident; is required by TSR to be available and operable and either

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should have been operating or should have operated on demand; and no redundant

oquipment is available and operable to perform the required safety function (that is,

alerting operators to a release of UF.). The failure to verbally notify the NRC within the ,

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24-hour period required by Procedure UE2-RA-RE1030 is a Violation of Technical

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Safety Requirement 3.9.1 (VIO 70-7001/9701101a). Although identified by plant staff, i the violation is being cited because of the number of examples of missed reportability determinations (some identified by the NRC) during the inspection period, s

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The PSS subsequently made two 24-hour repurts (Certificatee Event Reports 33124 and

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33126) to the NRC on October 20 and 21, over two days from the last event on October 18. The first event report contained an incorrect event date and time and TSR

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reference. The written notification was later revised and resubmitted to the NRC with the correct date and time, in followup discussions, the inspectors noted that part of the l

confusion derived from the belief of some plant staff that the 24-hour reporting window 1 began at the conclusion of the assessment which identified the event as reportable, not l 1 at the time of the event itself. The inspectors noted that while this approach could be l reasonable for certain situations involving complex operability or engineering evaluations, ,

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this approach was not reasonable for self-revealing events where the inoperability of the i

safety system was readily apparent as evidenced by the immediate entry into the

< associated LCO Action Statements, j c. Conclusion 1

Plant staff identified a failure to make timely event reports to the NRC for two events involving loss of power to the alarm annunciation function for the high voltage PGLD l'

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system in Building C-310 for UF, condensors, accumulators, and associated piping. The

!- cause of the loss of power had not been determined as of the end of the inspection period.-

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01.2 - Buildino C-333A Release and H!nh Steam Pressure isolation Events ,

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a, lnspection Scope (88100)

1 The inspectors reviewed the circumstances surrounding two event reports (Certificatee Event Reports 33038 and 33039) made on October 7 for a minor release in Building -

}, C 333A (Feed Facility) and two subsequent actuations of the steam pressure control i systems for Autoclaves 3 North and 4 North.

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, Observations and Findinos

On October 6,1997, the PGLD head above the heated housing for Autoclave 3 South in I Building C-333A fitad. Operators responded to the alarm por the alarm response l _ procedure, donned protective equipment, and took samples of the area for hydrofluoric

acid (c byproduct of a UF, release). All samples were negative.' After an attempt to reset ,

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the PC D head was unsuccessful, operators established a smoke watch for the area.

Subsequent troubleshooting led to the identification of a " pinhole" leak in a three-eighths *

f- inch copper purge line associated with the autoclave. Upon discovery of the minor leak, operators placed all autoclaves operating in Building C-333A into containment and evacuated the affected piping. During the subsequent response by the emergency l

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squad, the leaking copper line was temporarily plugge After mitigating the release and obtaining negative results for samples taken in the area,

[ the emergency response was terminated by the incident commander. Building operators then retumed five autoclaves to service. Within appr0ximately 10 minutes of re-initiating

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steam heating, the sicam pressure control system for Autoclave 4 North actuated i- because of a high steam pressure. The steam pressure control system isolates the steam supply and vent lines in order to limit the temperature of the cylinder in the I autoclave (thus preventing a rupture) by limiting the steam pressure. Subsequent to the Autoclave 4 North actuation, the steam pressure control system for Autoclave 3 North g

L actuated. Both actuations appeared to result from actual high steam pressures of seven

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pounds per square inch gauge (psig). No firm causes for the actuations were identified at the time of the even Subsequent to the event, two operators and an instrument mechanic assigned to the area

,- provided urine samples which contained slight amounts of uranium (15 micrograms per-

liter or less) which were well b< low the plant administrative limit for consideration of work restriction. These personnel were present prior to the identification of the leak and could have been exposed to very minor concentrations of UF.. Followup sampling of these personnel yielded results that were below the plant's minimum detectable activity.

Conclusion After a small release in Building C-333A which resulted in minor intakes of uranium by-three personnel, two autoclave steam pressure control systems actuated upon retuming

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_ the associated autoclaves to service, The inspectors will review the written reports

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i associated with the events for root causes and corrective action .

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01.3 BLMna C-335 Holdina Drum Room Low Tomoerature j

' Inspection Scope (88100)

The inspectors reviewed the reportability of an event documented in the plant problem -

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reporting system, Observations and Findinas  !

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On November 10,1997, plant staff reported discovering that the Building C-335 hol:"'g e drum room temperature was 100 degrees Fahrenheit. Normally, plant surge and hc.dmg drum rooms were maintained above 105 degrees Fahrenheit as one of two nuclear a criticality safety controls. The discovery was documented in Problem Report 6630 and

) was evaluated by both criticality safety and operations staff.

! The criticality safety staffs review of the issue was documented in incident Report 97-061. The inspectors reviewed the report and noted that it contained conflicting l

i information and direction. Specifically, the report indicated: 1) that the current nuclear criticality safety approval for the separation system did not include requirements for the

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holding drums; 2) that a nuclear criticality safety approval did not exist for the drums;

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3) that controls for nuclear criticality safety were maintained; and,4) the loss of a nuclear

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criticality safety approval requirement occurred that did not violate the double contingency

- principle. Since the nuclear criticality safety approval for the Froon and UF, reparation

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system did not address the holding drums, and since no controls were specified in the approval, the inspectors could not determine how any controls were maintained or how the operation was in accordance with current requirements. The incident report did state

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that a revised criticality safety evaluation and approval had been developed; howes 3r, no procedures had been developed to implement these controls. During subsequent

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discussions between the inspectors and the criticality safety staff, the staff could not v explain how each of the above statements could be concurrently correct. Another l pioblem report was issued to_ document, that at the time of discovery, the holding drums

were not controlled using a currently approved criticality safety evaluation or approval,

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l- The operations staffs review of the finding was documented on the problem reporting i form. The problem report indicated that the event was not reportable; however, the basis for this determination was not stated. The inspectors discussed this conclusion with several operations and engineering staff and noted that the decision not to report the finding was based upon the assessment in the incident report that double contingency l

was maintained.- The inspectors questioned the basis for this conclusion given that the 5cident report also stated that the safety approval did not include controls for the holding drums. Additionally, the inspectors noted that a low drum room temperature would

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constitute a loss of one of two criticality controls implemented as a compensatory ,

i- measure and now relied upon. The other control would be moderation contro l On November 19, after review of the additionalinformation discussed above and the supplementary problem report filed by criticality safety engineers, operations j management determined the original issue was reportable and made the required verbal I report.- l

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-l Technical Safety Requirement 3.9.1 required that wdtten m,oodures be prepared and j implemented to cover the activities described in SAR Section 6.11.4.1 and listed in Appendix A to SAR Section 6.11.~ Appendix A to SAR Section 6.11 identified '

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" investigations and reporting" as an activity requiring an administrative procedur Step 6.1.2 of Procedure UE2-RA-RE1030, * NUCLEAR REGULATORY EVENT

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REPORTING," Revision 2, dated February 28,1997, required that the PSS review - *

problem reporis and determine reportability of the event or condition using Appendix ,

Step 6.2.1E of the procedure required that the PSS verbally notify the appropnate NRC .

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office within the time requirements shown in Appendix D. Appendix D, "NRC-REPORTING CRITERIA," of Procedure UE2-RA RE1030, specified the criteria and l

j reporting time for events and conditions. Specifically, Criterion A.2.c required a four hour report for operations that comply with the double contingency principle in which -

moderation is used as the pnmary wiMij comrol and where it is detoimined that <

l criticality safety analysis was deficient and the necessary controlled parameters were not *

< established or maintained. The failure to make a required vert >al report on the deficient

. criticality safety analysis between November 10 and 19,1997 is an example of a Violation of Technical Safety Requirement 3.9.1 (VIO 70-7001/9701101b).

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j-c. ~ Conclusion

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i The inspectors identified that a Technical Safety Requirement violation occurred, in that,

! a required four hour verbal report was not made to the NRC upon identification of a

' deficient criticality safety analysis. The inspectors determined that the report was not made as the result of the criticality safety staff's non-rigorous evaluation of an operational F anomaly and imprecise communications between criticality safety and operations staf ,

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0 Certificatee Event Reports (90712)

5- The certificatee made the following operations-related event reports during the inspection

- period. The inspectors reviewed any immediate safety concems indicated at the time of the initial verbal notification. The inspectors will evaluate the associated written reports

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Number Status Iilla

, 33124 Open Loss of Power to High-Voltage Process Gas Leak Detection i System Alarms in Building C-310 (CER 70-7001/9701102)

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33125 Open Loss of Power to High-Voltage Process Gas Leak Detection System Alarms in Building C-310 (CER 70-7001/97011-03)

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- 33130 Open Safety System Actuation of Building C-333A Autoclave 3

• South Steam Pressure Control System (CER 70-u

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7001/97011 04)

! 08.2 ' Bulletin 91-01 Recoris (97012)

' The certificatee made the following reports pursuant to Bulletin 91-01 during the  !

inspection period. The inspectors reviewed any immediate nuclear criticality safety concems associated with the report at the time of the initial verbal notification. Any l

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- significant issues emerging from these reviews are discussed in separate sections of the repor .

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33126 - 11/4/g7 Failure to include Criticality Safety Requirements for Certain Process Gas Coolers in Nuclear Criticality Safety Approval 3g71-01 -

11. Maintenance and Surveillance l M Conduct of Maintenance and Surveillance M1.1 -. - gast Normetex Withdrawal Pumo Tests

- Inspection Scope (88102. 88103) -

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The inspedors observed the functional tests performed to retum the east Normetex product withdrawal pump in Building C-310 to service following a change to the loads on

- the pump control circuit in response to continued pump trips involving blown fuses.

, Observations and Findinas After dis,;overy of continued blown fuses in the control circuit for the east Normetex

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product withdrawal pump, engineering staff undertook an electrical load study of the circuit to identify whether or not the circuit was overloaded when the pump was operating within normal parameters. The study indicated that the circuit was different from the other control circuits for Normetex pumps at Paducah and that because of additicaal non-

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safety-related relays connected to the circuit, the fuse protection for the circuit was inadequate. Engineering staff subsequently initiated a plant change pursuant to

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10 CFR 76.68 to remove the leads from the non-safety-related relays associated with an inoperable side withdravcal buffer system.

l Prior to declaring the east Normetex pump operable and retuming it to service, plant operators and instrument mechanics performed functional tests of the high discharge -

pressure and UF, release detection safety systems. While observing the tests, the inspectors noted ti,at the tests were performed in accordance with an appropriate work package and procedures. The high discharge pressure safety system tripped the pump  ;

- before the limiting control setting of 42 pounds per square inch absolute (psia) was j

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- reached. In addition, the PGLD safety syrtem also tripped the pump when two heads were smoke tested as required by Surveillance Requirement 2.3.4.31. After the successful completion of the functional tests for the east Normetex safety systems, the ,

- pump was declared operable and retumed to servic j l

I Conclusion Plant staff conducted appropriate functional tests of the safety systems for the east Normetex product withdrawal pump following a plant change to remove non-safety- l related loads from the pump's overicaded control circui l

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- M1.2 L Qimaal Fire Water Pumo Surveillance inspection Scope (86102)

The inspectors observed the montNy surveillance for the diesel high-pressure fire water

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(HPFW) pum Observations aind Findinas The inspectors observed the monthly manual start, mandated by Surveillance Requisement 2.4.4.81, and mechanical checks for the diesel HPFW Pump Number The pump was successfully started from both the pump house'and the Building C-300 Central Control Facility. The sarveillance was performed in accordance with the appropriate procedure and the pump was property retumod to service for automatic start upon loss of fire water pressure following the surveillanc On November 10, some days after the surveillance, plant staff ider.tified that the ultrasonic flow meter used to perform the annual flow test (Surveillance Requirement

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2.4.4.8-4) in December 1996 was not calibrated. As a result, the two HPFW pumps

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checked with this meter (Numbers 5 and 6) were declared inoperable pending a re test with a property calibrated instrument. This left the plant with the TSR limit of two operable pumps (Numbers 2 and 3). The inspectors discussed the LCO actions for loss ,

of another pump or the fire water table with various operations and fire safety staff and found all staff to be knowledgeable of the requirements, Conclusion

Plant staff performed the monthly surveillance of the diesel HPFW pump in accordance with TSR and procedural requirements, but discovered that staff had not used a L calibrated flow meter for the last annual surveillance of water flow for two of four HPFW

pump Ill. Enaineerina E Conduct of Engineering

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E Boundary Definition Manual Self-Assessment - Inspection Scope (88101) '

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The inspectors discussed a self assessment that plant staff had initiated for the review of

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nuclear criticality safety items in the plant Boundary Definition Manuals with the Nuclear Safety Manage Observations and Findinas

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The Boundary Definition Manuals (BDM) provided plant staff with the definition of quality

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(Q) and augmented quality nuclear criticality safety (AQ-NCS)_ systems relied upon for safety._ The AQ-NCS designation was used to denote systems and components relied upon to ensure double contingency or evaluation assumptions identified in applicable i

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nuclear criticality safety evaluations (NCSE) and nuclear criticality safety approvals j

- (NCSA). Prior to Match 3,1997, during the transition penod from Department of Energy l

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to NRC regulatory authority, plant staff reviewed, revised, and updated all of the plant ,

NCSEs and NCSAs as well as developed the BDMs as part of the quality assurance  :

program. Since March 3,1997, plant staff had identified some components or items  !

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which should have been in the BDMs as AQ-NCS, but had been ovotiooked during the .

transition process. A review of the items identified did not indicate any safety-significant l

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items were involved or missing from the BDMs.

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As a result of these findings, the Nuclear safety Manager began a systematic self-

assessment of the BDMs and the process for incorporating NCSE and NCSA-required

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systems and components into the AQ NCS definitions in the BDMs. The inspectors will monitor the progress of and corrective actions resulting from this BDM self-assessment

+ as an inspector Followup item (IFl 70-7001/9701105).

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- Plant staff began a self assessment of the process for identifying and documenting-

' Augmented Quality Nuclear Criticality Safety systems and components in the Boundary F Definition Manuals which the inspectors will continue to follow.

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j' E1.2 - FisM!e Waste Storace Drum Nue!eer Criticality Safety Evaluations

, Insoection Scope (88100) -

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- The inspectors reviewed an engineering evaluation and notice associated with a current

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nuclear criticality safety evaluation and approval.

Observations and Findinas

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During a recent inspection of the criticality safety program, NRC inspectors raised questions regarding the plant staffs failure to track or control the wall thickness of waste '

drums used to store liquid and solid fissile' wastes. The inspectors noted that criticality safety calculations, performed to demonstrate that the waste operations conformed to

TSR limits, assumed a minimum wall thickness and treated the drum wall material as a neutron poison; Therefore, control of the wall thickness appeared necessary in order to )

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ensure that assumption 9 made in the criticality safety calculations were maintained, in response to the NRC inspectors' questions, the plant staif measured thc wall thickness of a number of the drums used to store fissile wastes and performed additional criticality t

safety calculations - The additional calculations were initially performed in demonstrate that the wall thickness did not need to be tracked or controlled; The staffs .

measurements indicated that some drum walls were thinner than previously assume The inspectors noted that the thinner drum walls could, in some cases, decrease the

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[ margin of ssfety demonstrated by the previous criticality safety analysis. The revised

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criticality safety calculations were performed using a wall thickness equal to

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approximately forty percent of the minimum well thickness previously assumed for the

" -- drums and well below the measured thicknesses. The calculational results indicated a decrease in the margin of safety; however, the final results were still within the TSR  :

limits. Both the measurements and calculational results were communicated to the plant i

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. staff through an Engineering Notice, EN C432-g? 050, %oceptable Wall Thickness for Maximum 5.5 Gallon Drums," Revision 0, dated November 7,1997, and Engineering Evaluation, EV-C432-97-017, %cooptable Wall Thickness for Maximum 5.5 Gallon c Drums," Revision 0, dated November 7,1997.

• The inspectors reviewed the engineering notice and evaluation and noted that the

. documents appeared to authorize a change to the nuclear criticality safety evaluations approvod by the Phrd Operations Review Committee (PORC). Specifically, the enginoedng notice: 1) documented that some drums' measurements were outside the limits assumed in the current nuclear enticality safety evaluations for the drums; and,

' 2) provided engineering's approval for the plant staff to continue to use the drums witt

- thinner walls, based upon the revised nuclear criticality safety calculations documented in

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the engineering evaluation. During a comparison vsview of the original and revised nuclear criticality safety evaluations, the inspectors also determined that the original 1 l evaluations included non-conservative assumptions. Specifically, the original evaluations

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l i.' did not use the most conservative combination of physical parameters associated with the two drums authorized for use in the storage of fissile wastes, Additionally, the ,

i e evaluations did not use the smallest dimension allowed for the drum wall thickness,

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The inspectors discussed the findings with the criticality safety staff and engineering l

management. The staff acknowledged that the calculations were performed to ensure

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that safety was always maintained and that the field measurements indicated that some drums had dimensions outside the limits assumed in the original criticality safety

evaluations. The staff also concurred that the original evaluations did not use the most conservative combination of parameters or the smalleat wall thickness. However, the

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staff did not believe that the original nuclear criticality safety evaksations required changing since the engineering evaluation demonstrated that the TSR limits were not exceeded. During discussions with engineering management, the inspectors highlighted i i

that the staff's approach substituted the engineering notice and evaluation for the PORC-

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approved nuclear criticality safety evaluations. Subsequent to the inspectors'

l discussions, the PORC reviewed the engineenng notice and evaluation and approved the i documents as an amendment to the current nuclear criticality safety evaluations.

Technical Safety Requirement 3.10.5 required, in part, that the PORC be used to conduct reviews of all nuclear criticality safety evaluations and approvals, including the use of

,

storage drums for fissile wastes. The. failure of the PORC to conduct a review of the decrease in the approved margin of safety documented in the evaluations for use of j fissile waste drums, due to wall thicknesses less than the minimum value previously

,

assumed, was a Violation of Technical Safety Requirement 3.10.5 (VIO 70-

'-

7001/97011-06).

, Conclusions The inspectors determined that an engineering notice and evaluation, completed in

- . response to NRC inspectors' questions, inappropriately authorized the continued use of fissile waste storage drums with wall thicknesses less than the dimensions assumed in the current nuclear criticality safety evaluations, a TSR violation. The current nuclear

'

critical:ty safety evaluations were also noted to include non-conservative assumptions

and methods, b

- 12 P

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

j,, -

q

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E1.3_ Buildina C-335 Freon and Uranium Hexafluoride Separation System '

m 4 Inspection SooneD8100)

. .

The inspedors reviewed the nuclear criticality safety evaluation and approval - '

,

' implemented for the Building C 335 Froon and uranium hexafluoride separation system.

._

!

l Observations and Findmps .

,

During routine tours of Building C 335, maintenance staff observed that the temperature

of the holding drum room was below the normallevel and was than the minimum 105

- degrees Fahrenholt maintained for al! other drum storage rooma. The observations were l

discussed with plant criticality safety staff and actions direreted to retum the room temperature to greater than 105 degrees Fshrenheit. The problem report, which documented the finding and directed corrective actions, ind6cated that room temperature

was not maintained as a criticality control.'

As a followup to the initial low temperature finding, the inspectors reviewed the criticality l

safety evaluation and approval designated for control of the Freon and uranium  ;

separation system, performed a walkdown of the system, and reviewed operations a

implementation of the criticality safety directed controls, During review of the criticality safety evaluation and approval, the inspectors determined that the documents:

- 1) assumed that the general system was shut down and void of any significant amounts

of fissile materials; and,2) did not consider the holding drums as a part of the separation system. As a result, neither the evaluation nor the approval specified any controls for the j - holding drums. At the time of the discovery, the separation system was shut down and

.

the holding drums contained an undefined amoun';f Freon and uranium hexafluorloe

! (estimated at approximately two pounds) enriched to greater then ene weight percent in

uranium-23 During their initial review of the low-temperature concom, the criticality safety engineers

advised plant management as to the seriousness of the issue and necessary corrective measures based upon a newly approved criticality safety evaluation and approval for the system. The newly approved documents, though not yet implemented, were developed

.

to allow system operation. While the criticality safety engineers' assessment and actions

assured immedikte safety; the abssnce of further reviews of the differences in criticality

' controls directed by the two evaluations precluded the self-identification of a currently p

inadequate criticality safety evaluation and approval.

l The inspectors performed a walkdown of the system using plant diagrams and the

current, implemented criticality safety evaluation and approval. During the walkdown, the-inspectors determined that all system valves discussed in either the evaluation or

4 approval were in the proper position and were appropriately controlled using caution tags.

l The inspectors also identified that valves associated with the holding drum room were controlled under a separate valving order. The inspec' ors reviewed the valving order and 4 the current plant configuration and noted discrepancies. Specifically, the valving order

-' directed the closure and corArol of two valves in each piping run extending from the

holding drums pending the development of a formal criticality safety evaluation for the

'

system. However, the inspectors identified that one set of valves, though closed, was not -

controlled.' Additionally, the inspectors observed that opening the two valves would -

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_ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ _ . _ _ __ _ _ ._ _.___.___ __._._ _

i

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create a pathway around all of the other closed valves. During discussions with the

,

operations staff, the inspectors were informed that the one set of valves was used during operation of the separation system in 1996. Operation of the system using the valves

. appeared to conflict with the controlling valve order in place at that tim ,

.,

.

Finally, the inspectors assessed the piant staff's implementation of the interim controls .  !

' recommended by the criticality safety engineers following the initial low-temperature finding. Based upon discussions with operations staff and reviews of applicable logs, the

_ inspectors determined that no formal methods had been used to ensure that the interim E

'

'

criticality safety controls were implemented. Instead, ope ions manage:reent had relied on verbal direction given at the time of the initial finding. Foi discussions of the

, finding with operations management, a long term order was implemented to ensure

'

continued implementation of the controls pending approval of formal procedure Technical Safety Requirement 3.11.2, tsquires that all operations involving oranium .

enriched to 1.0 weight percent or higher and 15 grams or more of uranium 235 be based 7 upon a documented nuclear criticality safety evaluation and performed in accordance with

- a documented nuclear criticality safety approval. Operation of the Building C 335 Freon and uranium hexaflouride system holding drums in a standby mode, while containing -

Froon and uranium hexafluoride enriched to greater than one weight percent in uranium-235, without a documented cr.ticality safety evaluation or approval from March 3, igg 7,

,

until November ig, is a Violation of Technical Safety Requirement 3.11.2 (70- ,

7001!97011 4 7).

' Conclusions ,

The inspectors identified that the assessment of and response to an anomalous Building C-335 holding-drum-room temperature reading did not identify the absence of a currently implemonted criticality safety evaluation, approval, or procedures for the holding drums, a

> Technical Safety Requirement violation. Some deficiencies in previous controls for valving orders were also identifie E1A Modification of the Criticality Accident Alarm System Operatina Ranaq t

. inspection Scope i The inspectors reviewed ongoing plant efforts to assess and authorize operation of the

,

criticality accident alarm system at temperatures below the manufacture's approved limits,

. Observations and Findinas J

During the inspection period, the inspectors became aware and perfoimed an initial review of plant activities to extend the operating range of the criticality accident alarm system. The activities were conducted in order to resolve a recurring problem with the exposure of some criticality accident alarm systems outdoors or otherwise exposed to

winter temperatures. The manufacturer- recommended operating lower tempmu:re limit was 14 degrees above zero Fahrenheit while historical winter temperatures at the plant

,

.wore recorded as low as 20 degrees below zero Fahrenheit Currently, plant staff have placed space heaters in the immediate area of the criticality accident alarm systems in

,

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_ . _ . _ _ . _ .. _ _ _._. _ _ _ _ _ . _ . _ __._ _.______~. .

-  ;

l order to maintain the local temperature within the manufacturer recommended operating i range. -

<

-

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The inspectors nt ted that all of the testing was performed in a laboratory setting and on - i out of-service eq sipment. Engineering developed and approved procedures were used to control the testing; however, none of the procedures or the engineering basis for the testing had boea reviewed and approved by the Plant Operations Review Committe ,

i

~ Aoditionally, the testing approach, methods, and critical equipment parameters had not '

'

been reviewed with or concurred upon by the equipment manufacture i Initial testing results included numerous failures of the equipment to maintain operating '

s parameters within acceptable limits; however, some equipment appeared to perform

acceptably. Engineering staff indicated that the basis for the failures o the acceptable

performance had not been detonnined. Through discussions with engineering ,

! management, the inspectors were informed that ersgineering management planned to

. Install and authorize operation of those units that passed the lower temperature testin Based upon the passed tests, operations would be authorized at temperatures down to t l

q 20 degrees below zero Fahrenheit.- The inspectors questioned the acceptability of this -

approach and the need for reviews and approvals under the plant change control '

. process, in that, raanagement planned to use the testing results to change the plant *

design basis.

.

' As of the end of the inspection period, none of the tested systems were installed in the h plant with the expanded operating range. Additional inspection effort will be required to assess the adequacy of the testing methods, the choice of critical parameters, and the appropriateness of making changes to the operating parameters without performiN

,

safety evaluations required for plant design changes, Engineering management #g

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l plar.ned to review the testing methods, and results with the equipment manufacturer prior I

'

to using the equipment with the expanded operating range. The actions described above will be tracked as an Unresolved item (URI 70-7001/9701108).

J j Conclusions ,

l

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An unresolved item was identified regarding the assessment of modifications made to the criticality accident alarm system. Additionalinspection effort was required to assess the

.

i

' adequacy of testing methods for the modified system, and to determine the need for '

reviews and approvals under the plant change contros process for the modif'mations to the cnticality accident alarm syste lV, Plant Succort P1.0 Conduct of Emeroency Preparedness Activities P Cnticality Alarm Drill Inspection Scoce (88100)

The inspectors observed selected aspects of a Building C-400 criticality accident alarm

- dril l i

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l- <

l w _

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

l

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l b.- Observations and Findmos l

The inspectors observed the evacuation of building personnel, segregation of 1

- ocntaminated personnel, and exposure evaluation techniques for the emergency  :

response to the simulated criticality alarm. Staff observed were knowledgeable of the requirements in the associated emergency plan implementing procedure (EPIP). In j

-

addition, a timely accountability of the staff in the affected area was accomplished and - '

provided to the incident commander The inspectors noted that the simulated response .

appeared to effectively address the actions required by the controlling EPIP for that . l portion of the drill observed.-  :

.

1 Conclusions Plant emergency response staff demonstrated an effective knowledge of the EPIP

requirements in a criticality accident alarm drill for Building C-40 .0 Ggnduct of Security Activities

,

a S Classified Document Shredded in Buildina C-300 i

l a. Inspection Scope (88100)

'

~

The inspectors reviewed the plant stoff's identification and reporting of the inappropriate shredding of a classified documen b. Observations and Findir1gg on November 18, plant staff alertly identified matter in a Building C-300 waste basket that was classified as Confidential Restricted Data. The document had been shredded in a machine associated with the waste basket; however, the shredding machine was not the ,

proper type nor was it authorized for the disposal of classified matter. Following the l

' discovery, operations management took action to property control the document and to l assess the reportability of the findings,

'

Bated upon initial reviews of the findings, operations management concluded that the event was not reportable. This determination was communicated to the inspectors ,

'

approximately two hours later. At the time of notification, the inspectors questioned the basis for not reporting the finding, Specifically, the inspectors noted that NRC regulations required one-hour verbal reports for any infractians, losses, compromises, or possible compromises of classified information. Additionally, the inspectors noted that the

.

- infraction, that is, the improper shredding of the classified matter, was not directly related to an NRC-authorized reporting exemption. Following a more rigorous review of the event specifics and the reporting requirements, operations management made the required report approximately 6 hours6.944444e-5 days <br />0.00167 hours <br />9.920635e-6 weeks <br />2.283e-6 months <br /> after the finding was identifie Technical Safety Requirement 3.g.1 required that written procedures be prepared and

'

implemented to cover the activities described in SAR Section 6.11.4.1 and listed in Appendix A to SAR Section 6.11. Appendix A to SAR Section 6.11 identifwwi

" investigations and reporting" as an activity requiring an administrative procedure. Step -

6.1.2 of Procedure UE2 RA-RE1030," NUCLEAR REGULATORY EVENT REPORTING "

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

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o -,

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i- .  !

Revision 2, dated February 28,1997, required that the PSS review problem reports and--

! determine reportability of the event or condition using Appendoc D. Step 6.2.1E of the i procedure required that the PSS vert > ally notify the appropriate NRC office within the time

' +

requirements shown in Appendix D. Appendix D, *NRC REPORTING CRITERIA," of --

.

> Procedure UE2-RA RE1030, specified the criteria and reporting time for events and -

conditions. _ Specifically, Criterion L.4 required an immediate report (within one hour) for

-

any infractions, losses, compromises, or possible compromise of classified infom,ation or_

classified documents. The failt.ro to notify the NRC within one hour of a discovery of an -

infraction of classified information control at 5:15 p.m., on November 11,1997, is an exemple of a Violation of Technical safety Requirement 3.9.1 (VIO 70-7001/97011-

'

01c). Conclusion The inspectors identified a failure by the plant staff to make a required one-hour verbal report of the discovery of an infraction of classified information controls, a Technical -'

Safety Requirement violation. The violation occurred, in part, due to a non-rigorous

+ evaluation of the event by operations management.

S1.2 - Control of Classified Matter

.

The certificatee continued to notify the NRC pursuant to 10 CFR 95.57(b) of discoveries of classified matter which was not property marked or controlled. -The discoveries included classified mate.ials found both inside and outside the controlled access area (CAA). The discoveries were made as a result of the certificatee's corrective actions for the deficiencies in the control of classified matter identified as an apparent violation discussed in Section S1,1 of NRC Inspection Report 70-7001/97007(DNMS). The

,-

inspectou will review the certificatee's corrective actions as a part of the response to the i apparent violation associated with the pending escalated enforcement action (eel 70- 3

-

7001/97009).

!: Q1.0 Quality Control

!

' Recelot inspection Procram Q1.1 3 Inspection Scope (68105)

.

The inspectors reviewed selected aspects of the receipt inspection program within the Quality Control Organization. The inspection consisted of interviews with quality centrol inspectors and managers, and a review of quality control procedures, intemal surveillances and audit Observations and Findinas The inspectors observed several receipt inspections conducted for items defined by the Quality Assurance Program as quality or augmented quality (AQ) materials and equipment. The inspectors noted that the receipt inspections were conducttd in accordance with Procedure CP2-QA-Ql2O34, " Receipt inspection," and the corresponding use reference procedures. The re^eipt inspectors were knowledgeable of the procedural V

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L i _ requirements and performed thorough receipt inspections for materials and equipment l receive During a review of an intomal surveillance conduuted for the receipt inspection program

(Surveillance KP-SU 897083), the inspectors notod that a surveillance finding dealt with the use of memoranda in lieu of an approved, written procedure.' The surveillance finding was documented as Problem Report PR-QA-97-3569, and highlighted that two interoffice memoranda, both dated Aptli 11,1997, instructed receipt control inspectors to only

- specify catalog numbers and storage levels on AQ ltem serviceable tags if the catalog l

,

. number and storage level were listed on a purchase order. The storage level and catalog i

' number were entries on the AQ item serviceable tag required by Procedure CP2-MA-  ;

MT1031, " Material Traceability," The memoranda highlighted that the traceability of  ;

c

!

i augmented quality items would continue to be mairitained through the purchase order.

l The memoranda were written in response to Problem Report PR-SU-97-1901, which l

! documented on April 10,1997, that receipt inspectors did not have the storage level and l catalog number information on purchase orders for some AQ items. The two interoffice l

'

i~ memoranda were tvritten to provide guidance to the receipt inspectors until Procedure .

CP2 MA-MT1031 was revised and approved.

,

The surveillance finding, Problem Report PR-QA 97-3569, was written on June 30,1997,

'

and closed out on August 8,1997. The inspectors noted that the management response used to close out PR-QA 97 3569 did not address or reference any corrective actions for

,

the use of memoranda in lieu of procedures. The inspectors subsequently interviewed quality control and problem reporting personnel, regarding PR-QA 97-3569. The 4 interviews revealed that approximately five days after the memoranda were issued,

.

quality control management realized that the use of memoranda in lieu of the approved, l . viitten procedure to direct which information was required to complete a1 AQ serviceable ,

!- tag, was a violation of the procedures program. Quality control management subsequently instructed the receipt inspectors to immediately disregard the memoranda nd continue to follow the approved procedure. The AQ items received during the five

. day period when the memoranda were used, were checked to ensure that the AQ serviceable tags contained the catalog number and storage level of the AQ ite Interviews with problem reporting personnei, highlighted that although the issue as described in PR-QA 97-3569 was corrected immediately at the time of occurrence in April 1997, the management response used to close out PR-QA 97-3569, did not address the

,

issue of using memoranda in lieu of procedures. Problem Reporting personnelindicated that the problem report would be re-closed to ensuru allissues are adequately addressed in the management response. The use of memorar'da in lieu of the writtan procedure to direct which entries were required on the AQ item serviceable tags constitutes a violation of minor significance and is being treated as a Non-Cited Violation, conslatent with Section IV of the NRC Enforcement Policy (NCV 70-7001/9701109).

y

c. Conclusions

,

Routine receipt inspections observed by the inspectors were thorough and performed in

. eccordance with written procedures. A non-cited violation was identified, regarding the

. use of memoranda in lieu of a written procedure. Upon issuance of the memoranda, quality control management realized the error and retracted the memoranda. A problem

> report addressing the issue, generated several months later during a routine surveillance, i

i 18

_.____..______L_i_________L_____.,___, .,,, . _. . .v, , _ ,_

- . . . . - . - . . - . . --- - -... - . - . . . ._. . - - -..

l

. l l

I - was closed by management without specifying the corrective actions taken for the i inciden V, Mananoment Meetina

' Exit Meetina Summary The inspectors presented tha inspection results to members of the plant staff and

.

management at the concitsion of the inspection on November 24,1997. The plant staff acknowledged the findings presented. The inspectors asked the plant staff whether any materials examined during the inspection should be considered proprietary. No proprietary information was identifie PARTIAL LIST OF PERSONS CONTACTED United States Enrichment Corooration J. H. Miller, Vice President - Production

- 'J. A. Labarroque, Safety, Safeguards and Quality Manager

'

>

Lockheed Martin Utility Services (LMUS)

'

'S. A. Polston, General Manager

• H. Pulley, Enrichment Plant Manager i *W. E. Sykes, Nuclear Regulatory Affairs Manager S. R. Penrod, Operations Manager United States Department of Enerov (DOE)

G. A. Bazzell, Sito Safety Representative Nuclear Reaulatory Commission (NRC)

• K. G. O'Brien, Senior Resident inspector J. M. Jacobson, Resident inspector

• Deenotes those present at the November 24,1997 exit meet)n Other members of the plant staff were also contacted during the inspection period.

~

INSPECTION PROCEDURES USED IP 88100 Plant Operations

',

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IP 88102 Surveillance Observations IP 88103 Maintenance Observations IP 88105 Management Oversight and Controls 4 IP 90712 In-office Review of Events

'IP 92702 Follow up of Events 4 19

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' *

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. 3 l ITEMS OPENED. CLOSED. AND DISCUSSED n g

- 70 7001/97011 01 VIO failure to properly determine the reportability of events and make ti.wely notifications to the NRC

.- 70 7001/97011-02- CER loss of power to high-voltage process gas leak detection system alarms 70 7001/97011-03 CER loss of power to high-voltage process gas leak detection system alarms 70 7001/97011-04 CER safey system actuation (autoclave pressure control system) '

70 7001/97011-05 IFl self-assessment of Boundary Definition Manuals

- 70 7001/97011 06 VIO failure of the Plant Operations Review Committee to review's i change to the u,;;ceity safety analysis for waste drums

- 70-7001/97011-07 VIO operation of the Freon and uranium hexafluoride separation system

holding drums without a documented criticality safety evaluation or

. approval

_ 70 7001/97011-08 URI low-temperature wsting of criticality accident alarm modules

. 70 7001/97011-09- NCV use of memoranda in lieu of approved procedures to direct receipt inspection activities i Closed None l

piscussed E 70-7001/97007 09 eel failure to property control classified matter

- LIST OF ACRONYMS USED

.

AC- Altemating Current ACR Area Control Room AQ Augmented Quality

-

AQ-NCS - Augmented Quality - Nuclear - Criticality Safety BDM Boundary Definition Manual -

CAA Controlled Access Area CAAS Criticality Accident Alarm System

,

- CER Certificatee Event Report CFR Code of Federal Regulations DC- Direct Current i

DNMS Division of Nuclear Materials Safety

- DOE Department of Energy eel - Escalsted Enforcement item

,

EPl - Emergency Plan implementing Procedure

- HF- Hydrogen Fluoride j - HPFW High Pressure Fire Water HPT- Health Physics Technician IFi= Inspector Follow-up item

.

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i s

+

w-g- S - m +in- n - = e- r- - 1 - ypr -- w-y yi -.,- y- , w ,-w-, - - -u-- -, + , -w-w--- e 1 ---,-

. . _ _ . - -. - . . . . - .- . .-- ..

.

LCO Limiting Condition for Operation NCSA Nuclear Criticality Safety Approval NCSE Nuclear Criticality Safety Evaluation NOV Notice of Violation NRC Nuclear Regulatory Commission POR Public Document Room PGLD Process Gas Leak Detection PORC Plant Operations Review Committee PSIA Pounds Per Square Inch A:: solute PSIG Pounds Per Square Inch Osge PSS Plant Shift C Jpervisor Q Quality SAR Safety Analysis Report TSR Technical Safety Requirement UF6 Uranium Hexafluoride URI Unresolved item USEC United States Enrichment Co'poration

- VIO Violation

21

.- .. - .-