ML20198F729

Insp Rept 70-7001/97-04 on 970603-0714.Violations Noted. Major Areas Inspected:Plant Operations,Maint & Surveillance, Engineering & Plant Support

ML20198F729
Person / Time
Site:	07007001
Issue date:	08/05/1997
From:	NRC OFFICE OF INSPECTION & ENFORCEMENT (IE REGION III)
To:
Shared Package
ML20198F694	List: ML20198F688 ML20198F708 ML20198F729
References
70-7001-97-04, 70-7001-97-4, NUDOCS 9708130129
Download: ML20198F729 (32)

Text

0 U.S. NUCLEAR REGULATORY COMMISSION REGION lli Docket No: 70 7001 Inspection Report No: 70 7001/97004(DNMS)

Facility Operator: United States Enrichment Corporation Facility Name: Paducah Gasoous Diffusion Plant Location: 5600 Hobbs Road P. O. Box 1410 Paducah, KY 42001 Dates: June 3, through July 14,1997 Inspectors: K. G. O' Brion, Senior Resident inspector J. M. Jacobson, Resident inspector T. D. Reidinger, Senior Fuel Cyclo inspector, Region ll1 Approved By: P. L. Hiland, Chief Fuel Cyclo Branch 9708130129 970005 PDR ADOCK 07007001 C PDR

EXECUTlyf

SUMMARY

United States Enrichment Corporation Paducah Gaseous Diffusion Plant

. NRC Inspection Report No. 70 7001/97004(DNMS)

Plant Operations e The inspec. ors identified that Building C 310 operators had not been trained on the design, cporation, and safety implications of a newly installed sodium fluoride production oven. As a result, the operators were neither aware of the system alignment nor the actions required to respond to the off normal conditions which occurred on June 9. This is a Technical Safety Requirement violation.

(Section 01.2)

Maintenance and Surveillanco e The inspectors identified a Quality Assurance Program violation, in that, insufficient controls existod to preclude preconditioning of the autoclaves prior to the performance of a Technical Safety Requirement surveillance test._ A similar example of the violation was discussed in NRC Observation Report 70 7001/95004.

(Section M1.1)

• A weakness in the indopendent verification process permitted an instrument mechanic to lif t and incorrectly re land electricalleads during a Technical Safety Requiremont mandated surveillance test. As a result, the uranium hexafluoride release detection system activation logic for one of the Building C 310 withdrawal pumps was reconfigured in an unapproved, though conservative, manner.

(Section M1.2)

Enaineerina

• The inspectors identified that an engineering project to modify the autoclaves, a safety component, had been approved and initiated without the proper development, approval, and control of installation and test procedures. This is a TSR violation, in addition, some of the test procedure acceptance criteria were inconsistent with the associated engineering specifications. (Section E1.2) e The inspectors identified two Safety Analysis Report changes which were imptomonted without adequate evaluations, this is a 10 CFR 76.68(a) violation.

The evaluations of these potentially significant changes did not provide a basis for determining that the changes did not: 1) decrease the effectiveness of the safe;;.::"ds program for protecting special nuclear material, or; 2) increase the probability of safety equipment malfunction, an unreviewed safety question.

(Section E1.3) 2

• A violation of the double contingency principle for the storage of legacy process equipment containing fissile and potentially fissile material, as required under Technical Safety Rcquirement 3.11.5, was identified. The violation was related to two prior violations for legacy process equipment issuud in NRC Inspection Reports 70 7001/97002 and 70 7001/97003. (Section E1.4) flent Succort Emeroency Preoaredness

• Compliance Plan issues related to the Emergency Plan, emergency response organization training, public warning system and fire protection pre fire plans were adequately addressed. (Sections P1.1, P1.4)

• The certificatee's Emergency Plan and implementing procedures provided suf ficient guidance for responding to plant emergencies, and the emergency response organization training was good for responding to emergencies. Emergency response personnel were adequately trained and were knowledgeable of emergency response procedures and equipment. (Section P1.2)

• The certificatee maintained an effective tracking system for deficiencies identified during drills or exercises; proposed actions to correct a minor backlog of corrective actions for identified drill or exercise deficiencies were adequate. (Section P1.3)

• Emergency exercises and drills were very effective in exercising the certificatoe's emergency response organizatlon. (Section P1.4)

• The overall organization and management structure of the Emergency Plan function was consistent with the Emergency Plan and implementing procedures.

(Section P1.5)

SEL!!iW

• Tee inspectors identified that management had not developed and implemented administrative procedures to limit the work hours of security guards. As a result, dtring May 1997, several security guards worked hours in excess of the limits specified in the Technical Safety Requirements. (Section S1.1) 3 l

DEIALLR

l. Onorations

01. Conduct of Operations'

-01.1 Status of Plant (88100)

During the period, operations stali seconfigured the cascade to reduce the top product assay of uranium hexafluoride (UF.) from 1.9 to 1.5 weight percent uranium 235 (U 235). In addition, the operators implemented a reduction in the cascada power level (directly correlated to amount of gaseous UF In the cascade) from approximately 1900 to 1000 megawatts (MW). The inspectors noted no

. abnormal occurrences as a result of the changes.

01.2 Hydroaen Fluoride Release

a. insnection Scone (IP 88100)

As a followup to a June 9 hydrogen fluoride (HF) release, the inspectors reviewed the measures implemented to control operation of the Building C 310 sodium fluoride production oven, a recently installed system.

b. Observations and Findinas Eygnt Summarv On June 9, personnel in Building 310 identified an unusual haze in the air near the sodium fluoride (NaF) filter room and product _ withdrawal area. _The personnel were: 1) aware that hydrolyzation of UF would result in the production of uranyl fluorides and HF, and; 2) concerned that the observed haze indicated that a UF, release had occurred. The presence of the haze was communicated to building management and actions were initiated to assess the situation.

As an immediate response to the release, management evacuated the area and directed building personnel to relocate to the area control room (ACR).

In addition, an operations engineer responded to the building and opened the power supply breaker to the NaF production oven. Over the next few hours, plant personnel determined, based upon real time sampling, that the observed haze was the result of HF in the air. However, the source of the .

' HF was not readily identifiable. Af ter some time,- the NaF production oven,

- Topical headings such as 01 MS. etc., are used in accordance with the NRC standardized inspection report outline contained in NRC Manual Chapter 0610. Individual reports are not expected to address all outline

- topics, and the topical headings are therefore not always sequential.

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located in the fittor room, was identified as the source of the HF. Although an operations engineer had opened the power supply breaker to the oven shortly after the hazo was observed, the HF release continued. The continued HF roloase was due in part to heat stcied in the process and a positive system pressure created by communication of the oven with the plant air system.

During the implomontation of response actions to the release, operations personnel suspended the performance of most other routino activities. One activity that was not suspended was the performance of twico por shift testing of the UF, detector heads. The testing of the UF, detector heads was performed by assessment vice building operations personnel.

Managemont's recall of building personnel to the ACR following the release procluded their involvement in the testing. The use of assessment personnel to conduct testing ensured continued compliance with a Technical Safety Requirement defined survoillance interval.

Qoorational Controls Following the event, the inspectors interviewed the oporations personnel on-duty at the time of the release. The inspectors datormined that the building operators woro not awaro of or controlling ovon operations at the time of the roloaso. As a result, neither building management nor the operators were cognizant of the status of NaF oven activitios. The building staff were not aware of a proceduto directing normal and of f normal NaF oven operations.

This lack of system knowledge precluded the operators from realizing the need to open the NaF ovon power supply breaker, at the initial stages of the event, and may have contributed to the longthy timo required by the assossment personnel to determine the source of and driving force for the release.

The inspectors discussed those findings with management. The building manager indicated that the NaF oven was in the experimontal stage and had not boon released to operations staff, instead, operations management had assigned control and operation of the oven to the operations engineering group. One individual, within the operations engineering group, had knowledge of and responsibility for the oven. This was the same individual that took action shortly after the identification of the haze in Building 310 to open the power supply breaker to the NaF oven. The inspectors noted that operation of the NaF oven in the cascade building, where the system could affect ongoing operations and safety, appeared inconsistent with designation of the system as " experimental."

The inspectors reviewod the proceduto used to defino NaF oven operations, CP4-CO CA2033, " Conversion of NAHF, to NAF," Revision 0, dated December 6,199G. The inspectors noted that the procedure discussed both normal and off normal operations. The procedure included valving, electrical breaker, and combustion loading limitations. - The inspectors also reviewod 5

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1 operations procedure CP3 CO C01003, " Organization and Administration,"

Revision 2, change A, dated May 13,1997. Procedure step 5.7 assigned responsibility to the operators to initiate actions required by normal and off.

normal operating procedures and limited the operation of equipment or systems within buildings or areas to only qualified operators or trainees.

Operations management indicated that the operations engineer, assigned responsibility for the NaF oven, was not a qualified operator.

During a subsequent walkdown of the system, the inspectors observed a

" working copy" of the procedure in the oven area. The working copy included revisions to both the valve listings and the operational parameters.

The inspectors discussed these observations with the operations engineer and was informed that several changes to the operations protocol were being :nvestigated as a part of oven operations. The changes would be incorporated into the procedure when the process was finallred.

Technical Safety Requirement (TSR) 3.4 requires, in part, that individuals relied upon to operate the plant in a safe manner are properly trained. Plant procedure CP3 CO C01003, " Organization and Administration," required that only qualified operators or trainees operate equipment or systems in Building C 310. The failure to train the Building C 310 operators on the normal and off normal operations of the NaF oven is a Violation (VIO 70 7001/97004 01).

c. ConcluslDDa The inspectors identified that Building C 310 operators had not been trained on the design, operation, and safety implications of the newly installed sodium fluoride production oven. As a result, the operators were neither aware of the system alignment nor the actions required to respond to off normal conditions as experienced during the June 9 release. This is a TSR violation.

08. Miscellaneous Matterg 08.1 Certificatee Event Ronpris (90712)

The certificatoe made the following operations related event reports during the inspection period. The inspectors reviewed any immediate safety concerns indicated at the time of the initial vorbal notification. The inspectors will evaluate the associated written reports for each of these items following submittal.

Number Status Title 32571 Open Loss of Alarm Audibility of the C 310 Building Criticality Accident Alarm System.

(CER 70 7001/97004-02) 6

4 32610 Open Safety System Actuation of the C 360 Building West Crane Holst Brakes.

(CER 70 7001/97004 03 )

32622 . Open Loss of Alarrn Audibility for the C 310 Building.

(CER 70 7001/97004-04 )

32627 Open Actuation of the C 360 Number 1 Autoclave Steam Pressure Control System.

(CER 70 7001/97004 05 )

ll. Maintenance and Surveillance M1. Conduct of Maintenance and Surveillance M1,1 Autoclave Pinch Test

a. insoection Scone (88102)

The inepectors reviewed preventive maintenance and testing of the autoclave head to shellinterface,

b. Observations and Findinal On June 2, the inspectors reviewed work scheduled for completion during the preceding weekend. The inspectors noted that the schedule included both " pinch" and pressure decay testing of the number 4 north and south autoclaves. The schedule directed that the " pinch" test was to be performed prior to the pressure decay test The " pinch" test was a preventive maintenance check and adjustment, as necessary, of the autoclave head to shell alignment. The pressure decay test was a

- TSR mandated quarterly test used to assure continued operability of the autoclave containment function.

Based upon the scheduled testing sequence, the inspectors discussed the issue with the system engineer, The inspectors were informed that. due to other priorities, the testing had not been performed. However, the teats had been rescheduled for later the same week The system engineer indicated that the testing sequence was unchanged from that scheduled over the weekend, -

.The inspectors further reviewed the procedure used to perform the pinch test and determined that the procedure directed adjustments to the alignment which could invalidate the TSR mandated quarterly pressure decay test results by preconditioning the equipment Specifically, the procedure directed adjustments to the autoclave locking ring which would decrease the

' leak rate determined during a pressure decay test The inspectors discussed the findings with the system engineer and the test sequencing was revised.

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i Subsequently, a review was performed of pinch tests conductad on the other autoclavos since March 3. The review indicated thu on March 13, pinch testing and adjustments were performed on thu . south autoclave followed by a pressure decay tost. Immodlately prior to the testing, tho 2 south autoclave was in service based upon successful completion of a January 1997 pressure decay test. The next TSR mandated quarterly pressure decay test was schedulod for April 1997. The performance of preventivo maintenance adjustments to the autoclave head to shell alignment resulted in equipment proconditioning and a loss of "as found" data required for the April 1997 TSR mandated surveillance.

As a in:lowup to the inspectora questions, system engineering staf f initiated a review of other routine proventivo maintenance activities associated with safety related equipment. The review was to ensure that other examples of safety related equipment proconditioning did not exist.

The inspectors noted that the issue of preconditioning safety related equipment was previously discussed in NRC Observation hoport 70 7001/95004, issued October 23,1995. As an internal response to the Observation Report finding, plant management conducted a leview of activities to preclude preconditioning safety related equipment. The inspectors reviewed those corrective actions and noted that measures wore not implemented to prevent recurrence.

10 CFR 76.93, " Quality Assurance," requiros, in part, that the cortificatoo shall establish and execute a quality assuranco program. Section 2.11 of the Quality Assurance Program, " Test Control," requires in part, that the test control system provide measures to ensure that test proceduros include provisions establishing prerequisitos, such as appropriato equipment test j' conditions.

The scheduling and conduct of preventativo maintenance and adjustment to the autoclave head to shell alignment prior to the performance of the TSR mandated surveillance, did not ensure that appropriate test conditions were established for the TSR surveillance test. The failuro to ensure that the TSR mandated survoillance tests were performed under suitably controlled conditions is a Violation (Violation 70 7001/97004-06).

c. Conclusions The inspectors identified that insufficient controls existed to preclude preconditioning of the autoclaves prior to the performance of TSR mendated surveillance testing. This is a Quality Assurance Program violation.

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M1.2 East Normelex Withdrawal Pu.mn Surveillanta

a. [Qugstion Scono (88102)

The Jnspectors reviewed the circumstances surrounding NRC Event Report 32489, submittod on Juno 15. The report indicated that an instrument mechanic mado the Building C 310 west Normetox withdrawal pump UF, detection system inoperable during testing for the east Normotex withdrawal pump UF, release detection system,

b. Observations and Findinos On June 13, instrument mechanics performed a TSR required surveillance of the UF, detection system for the east Normotex withdrawal pemp. The UF, rolosse detection system was designed to trip the withdrawal pump and to shut the pump discharge valvo upon detection of a UF, release by any two adjacent smoko detector heads. TSR Surveillance Requiremont 2.3.4.31 required annual testing to assure that " smoking" each potential combination of two detector heads would cause the pump to trip and the dischargo valvo to closo. Each pump had independent detection systems. The detection systems were composed of two separate trains, each including a signal conditioner and a programmable controller. Either train could independently trip the pump.

During the testing, the east Normotex pump was inoperable and withdrawal operations woro performed using the west Normotox pump.

Two instrument mechanics woro required to perform the survoillance 5

proceduro one stationed on the cell floor, at the Normotox pumps, and one on the operating floor, at the UFe rolease detection system signal conditioner panols. The surveillance proceduto directed one of the instrument mechanics to soquentially lif t two loads on each of the two output cards of the associated programmable controller to disablo a train of the system. This method was used to assure that each train of the system would trip the pump when the other mechanic smoked adjacent detector heads. While lifting the leads for the east Normetex pump, the mechanic incorrectly interpreted the procedural instructions and lif ted two additional leads. The additional loads were on the signal processor output card for the west Normatex pump. The additionalleads wore re landed following completion of each step.

On June 15, Building C 310 operators performed a twice per shift check of the UFedotection system for the standby withdrawal pump. At the time of the test, the east pump was operating and the west pump was in stand by status. The test involved smoking a single detector head and obstving the activation of an off normal condition light in the ACR. During this test, the operators observed indications that the standby pump tripped. This condition was suppose to occur only when two detector heads were 9

activated. The system engineer performed a follow-up investigation of the occurrence. The systam engineer identified that two leads for one train of the west Normetex pump programrnable controller output card had been reversed. Further investigation identified that the leads were reversed during l the annual surveillance for the east pmp.

Based upon the system engineer findings, the plant shift superintendent (PSS) notified the NRC that the west Normetex pump UF, release deter. tion system had been made inoperable during surveillance testing of the east Normatex pump UF, release detection system on June 13.

The inspectors reviewed the as-found configuration and system drawings.

Based upon a reconstruction;of the events and system design, the inspectors noted that the release detection system for the west pump may not have been inoperable during the porlod in question. Although the leads were reversed, the as-found configuration was more conservative in response to a detector actuation. Specifically, the as-found configuration would trip the pump upon activation of any single detector vice the two adjacent detectors required by the system design. The other train had not been involved in the incorrect lifting of leads. The review also indicated that the labeling of the leads and associated programmable controller input and output cards for the safety system trains was inconsistent and difficult to follow without the system drawings.

The inspectors noted that the surveillance procedure included independent verification steps to ensure that leads, lif ted to disable the specified train, were properly re-landed. However, the procedure did not include steps to independently verify that only those leads specified by the procedure were actually lif ted. As a result, the mistaken lif ting and reverso landing of leads for the west pump was not identified. The inspectors discussed this finding management and were informed that the independent verification process did not require verification of lif ted leads.

After discovery, the leads were correctly reversed and a successful surveillance test was conducted,

c. Conclusions A weakness in the independent verification process permitted an instrument mechanic to lift and incorrectly re-land electricalleads during a TSR mandated surveillance test. As a result, the UFe release detection system activation logic for one of the Building C-310 withdrawal pumps was reconfigured in an unapproved, though conservative, manner.

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M8 Miscellaneous Maintenance Matters M8,1 Certificatee Event Reoorts (90712)

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

Number Status Title 32489 Open C 310 West Normetex Pump Release Detection System Inadvertently Made Inoperable During Performance of Annual Surveillance.

(CER 70 7001/97004-07) H 32523 Open lon Pumps Contaminated with Potentially Fissile Material not Properly Spaced in Instrument Maintenance Shop. (CER 70 7001/97004-08) ll1. Enalneerina E1. Conduct of Engineering E1.1 Installation of Sodium Fluoride Production Oven in Buildina C 310

a. insnection Scope (88100) -

The inspectors reviewed the process used to approve installation of a sodium fluoride (NaF) production oven in Building C-310.

b. Observations and Findinas The plant staff installed an electric oven in Building C-310 during late 1996 and early 1937 to facilitate the production of NaF. The NaF was used as a filter medium to treat process gases prior to atmospheric release.

As a part of the installation project, engineering performed a preliminary safety review (PSR) of the project, PSR 96-101, Revision 1, dated September 17,1996. Plant procedure, CP4 CO-CA2033, " Conversion of NaHF2 to NaF," dated December 9,1996, was also developed and approved to control operation of the oven. The inspectors reviewed the PSR and noted that the potential for or impacts of HF releases from the oven to the building atmosphere had not been considered. As a result, the PSR did not identify that oven releases could create a building atmosphere which would preclude operator actions required by the TSRs or as discussed in the SAR.

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During the June 9 release discussed in Section 01.2 of this report, the HF atmosphere created b/ NaF oven releases and management's precautionary recall of building personnel to the ACR prevented the operators from performing a routine TSR surveillance. The twice-per shift TSR surveillance was required to ensure the continued sensitivity and operability of UF, detectors located throughout the building, The inspectors also noted that the PSR was used to satisfy a procedure development process requirement for an unreviewed safety question screening. At the time the procedure was approved, the plant had implemented a change control and unreviewed safety question determination process consistent with NRC requirements, specifically 10 CFR 76.68. In addition, the initial NRC certification had t een received and approved. The inspectors reviewed the PSR process and noted that the process: 1) was not consistent with the requirements specified in 10 CFR 76.68, and: 2) may not have been appropriate for use during epproval of the operations procedure.

The inspectors discussed the findings with engineering and regulatory affairs staff. The inspectors were informed that operation of the oven had been placed on hold pending resolution of allissues associated with the release, including the apparent incorrect use of the PSR process,

c. Conclusions .

The inspectors identified weaknesses in the plant change process used to authorire operation of the NaF ovens prior to NRC certification of the plant.

E1.2 Autoclave instrument Unarade Protect

a. inspection Scone (88105)

The inspectors reviewed portions of the engineering modification materials developed to support the autoclave instrument upgrade project,

b. Observations and Findinns During the period, plant engineering staff completed preparations for and had scheduled field work to implement the autoclave instrumentation upgrade project. The purpose of the modification was to install new pressure transmitters and ancillary instrumentation for process pressure measurements in the three autoclave facilities. The engineering work was described in engineering service order (ESO) Z90830.

The inspectors reviewed some of the installation and testing plans associated with the modification. The modification package, as approved by the Plant Operations Review Committee (PORC), included a master work order with generic installation instructions and test plans that could be 12

annotated to specify the involved autoclave. The inspect (,rs noted that neither the installation instructions nor the test plans were developed and controlled as procedures. Instead, the responsible engineer indicated that the materials were developed as work instructions. The engineer indicated that. plant procedures allowed the use of work instructions instead of procedures. ' Work instructions were not reviewed or controlled with the same amount of rigor as procedures.

During a further review of the lastallation instructions and test plans, the inspectors noted that the current installation instructions were 27 pages in -

length and had been revised at least three times, One of the test plans, TP 19083-03, " Soap and Vacuum Test Plan," was six pages in length and had been revised twice. The inspectors determined that the initial installation instructions were provided to the PORC, as a part of the original modification approval. However, neither the test plans or installation instructions revisions were provided for PORC review, in addition, the revisions were not evaluated, in accordance with the plant change process, to ensure that the revision did not constitute an intent change to the document.

The inspectors reviewed the changes made to TP-19083-03, since its original development, and discussed the changes with the responsible engineer. Most of the changes were made to improve the directions provided to the craft performing the work. Craf t personnel had specifically requested several of the changes. During the revision of the test plans, the engineering staff added a requirement for a vacuum test of the system following the modification. The system engineering specification require the test but it was omitted in initial versions of the test plan.

Based upon the operating parameters of the involved systems and some of-the instructions in test plan TP-19083-03, the inspectors evaluated-adherence of the test plan to the engineering specification. The inspectors determined that the test plan did not ensure that the engineering specification was completely met. Specifically, the plan approved testing the system at a lesser vacuum for a shorter period of time than required in the engineering specification, A technical justification, with management approval, for the change had not been developed. The inspectors also noted that the engineering specification referenced in the materials was for ,

systems with a maximum pressure of 100 pounds force per square inch gauge (psig). Section 3.2 of the SAR stated that the autoclaves, including piping and valves, out the second containment valve, have a maximum working pressure of 200 psig. None of the engineering modification materials available to the inspectors provided justification for use of this engineering specification during development of the testing procedures.

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l F-Technical Safety Requirement 3.9. requires, in part, that written procedures shall be developed, approved, implemented, and maintained for activities described in Appendix A to SAR Section 6.11. - Appendix A to SAR Section 6.11 describes changes to facilities and equipment and modification design control as activities requiring written procedures. The failure to develop, approve, implement, and maintain procedures for the installation

.and testing activities associated with the autoclave instrument upgrade project is a Violation (70-7001/97004-09).

Following the inspectors' identification of the procedural inadequacies, the engineering group placed a hold on current activities. As of the end of the inspection period, none of the upgrade projects had been completed and returned to service using the unapproved test plans or the unsupported testing acceptance criteria.

c. Conclusions The inspectom identified that an engineering project to modify the autoclaves, a safety component, had beer' approved and initiated without the proper development, approval, and control of installation and test procedures. This is a TSR violation in addition, some of the acceptance criteria, referenced in the test procedures, were inconsistent with the system engineering specifications.

E1.3 Plant Chanaes

a. Insoection Scooe (881051 The inspectors reviewed two changes to operations described in the Safety Analysis Report made using the provisions of 10 CFR 76.68. The changes involved special nuclear material possession limits, contained in Table 1-3 of the Safety Analysis Report, and surveillance testing for the cascade cell trip function,

b. Observations and Findinas Chance to Possession Limits The inspectors reviewed three documents developed to support a revision to the SAR special nuclear material procession limits. The documents included a request for authorization change (RAC) No. 97-C-015, Revision 0, dated May 9,1997, a plant change request (PCR) No. PCR-C-97-0094, Revision 0, dated April 2,1997, and a safety evaluation No.97-008, Revision 0, dated April 25,1997. The change, once approved, would allow the plant to possess " swipe samples" with an assay greater than 5.5 weight percent enrichment. The SAR revision involved changing the " physical form" and

" description" columns for certain entries in SAR Table 1-3, " Possession limits for NRC-regulated materials and substances."

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1 The inspectors noted that one of the changes made to Table 13 allowed the plant to possess 500 kilograms of uranium (kgu) " enriched in isotope 235 up to 10 percent by weight, for use as laboratory calibration standards, analysis '

f of samples, laboratory chemicals, or contained in equipment from other facilities." Previously, the table had allowed the plant to possess 500 kgU

" enriched in isotope 235 up to 9.95 pewent by weight." The NRC approved the originallimit of 9.95 weight porcent based on the definition in 10 CFR 76.4 for special nuclear material of low strategic significance. The NRC also approved the plant's physical security program assuming the possession of special material of low strategic significance, also known as Category 111 material.

The definition of special nuclear material of moderate strategic significance (Category 11) includes "10,000 grams (10 kg) or more of uranium 235 (contained in uranium enriched to 10 percent or more but less than 20 percent in the U 235 isotope)." If plant staff had acquired 500 kgU of 10 weight percent materials (i.e.,50 kg of U 235), the plant would have been in possession of a Category ll quantity of material. The 10 CFR 73 physical security requirements for possessing a Category 11 quantity of special nuclear material are more stringent than those included in the current physical security plan approved by the NRC.

The inspectors reviewed of the RAC, PCR, and safety evaluction and noted that the documents did not assess the impact that the proposed change would have en the authorized Category for possession of special nuclear material, in particular, the documents did not include a discussion about the physical security aspects of changing the possession limit or how the physical security requirements for safeguarding Category ll materialin 10 CFR 73.67 were met. Changing the possession category of special nuclear material without ensuring appropriate safeguards could lead to a decrease in the effectiveness of the physical security program for protecting special nuclear materials onsite, Chanae to Cascade Cell Trio Surveillance Freauency in a letter dated June 23,1997, the certificatee responded to an NRC request for additional information that was needed to process a certificate amendment request (CAR) to change the current TSR testing requirements for the cascade cell trip system. Technical Safety Requirement 2.4.4.12,

" CASCADE CELL TRIP FUNCTION," required tha cell trip system to be tested on each planned cell shutdown. The proposed change would allow testing of the system at each planned shutdown or prior to startup. The SAR t accident analysis took credit for the operator's ability to use the cell trip system to de-energizo cascade cell stage motors, thus bringing a cell below atmosphere pressure. -The decrease of a cell's pressure to below atmospheric levels would reduce the amount of material released during an accident. The SAR also stated that cell trip system reliability was ensured through the manual shutdown of each cell within a five year period. The 15

certificatee letter also informed the NRC that a recent SAR change had been approved to delete the phrase, "through manual shutdown of each cell within a five year period,"_ from the sentence discussing how the periodicity of functional testing of the cell shutdown system ensured system reliability.

The inspectors reviewed the I bCR (RAC No. 97C014, Revision 2, approved June 20,1997) and the safety evaluation (No.97-018, Revision 1, approved June 20,1997) used to approve removal of the SAR five-year surveillance frequency. The PCR referenced the safety evaluation as the basis for determining that the proposed change did not constitute an unreviewed safety question. However, the inspectors noted that the safety evaluation did not address the long-term reliability concerc or safety impacts associated with removing a SAR requirement at testing the trip circuit every five years. Instead, the change was simply identified as a " clarification" of the surveillance requirement. In addition,-the inspectors noted that the PCR identified the change as an editorial change only.

The inspectors also noted that the PCR included a review of other proposed TSR, TSR basis statement, and SAR changes required as a part of the cell trip TSR frequency change. Because of the PCR included multiple changes, the inspectors could not determine which " change" was being addressed in the answers to individual questions.

The inspectors reviewed the associated vendor manuals for the circuit breakers and motor breakers associated with the cell trip function. _The vendor __ manuals recommended an annual preventive maintenance and testing of the breaker trip function to ensure reliability. The inspectors also noted that standard industry guidance recommended breakers be tested at least every five years. However, the PCR and safety evaluation did not address

- why vendor recommendations or standard industry guidance were not appropriate for ensuring reliability of the breakers causing cell trip.

10 CFR 76.4 defines a unreviewed safety question (USO), in part, to be a change which may increase the probability of occurrence of a malfunction of equipment important to safety previously evaluated in the SAR. The safety evaluation, used to approve deletion of the SAR requirement to test the cell trip system for each cell within a five year period, did not address how the change would affect the system reliability. A decrease in system reliability would increase the probability of occurrence of'a malfunction of the cell trip system, equipment important to safety previously evaluated in the SAR.

Therefore, the safety evaluation did not address whether or not the change constituted a USO.

10 CFR 76.68(a) provides, in part, that the certificatee may make changes to the plant's operations as described in the SAR without, prior Commission-approval, provided that the changes will not decrease effectiveness of the plant's safety, safeguards, or security programs or involve a USQ. The failure to evaluate whether: 1) a change to the SAR possession limit for 16

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uranium enriched to 10 weight percent impacted the effectivaness of the safeguards program, and; 2) the deletion of a SAR requireme.t to test the l cell trip system for each cell within a five year period constituted a USO is a Violation (VIO 70-7001/97004 10),

^

c. Conclusions The inspectors identified two examples of inadequate evaluations used to support changes to the SAR, a 10 CFR 76.68(a) violation. The evaluations of potentially significant changes did not provide a basis fo.- determining that the changes did not decrease the effectiveness of the safeguards program for protecting special nuclear material or increase the probability of safety equipment malfunction (a USO).

E1.4 Failure to Imolement Double Continoency Princiole for Leaacv Eauioment

a. Insoection Scone (88020)

The inspectors performed follow-up inspection for Certificatee Event Report (CER) 32516, dated June 19,1997,in which the plant staff reported a failure to fully implement the double contingency principle for legacy process equipment containing fissile and potentially fissile material,

b. Observations and Findinas On June 19, plant criticality safety staff discovered that two nuclear criticality safety approvals (NCSAs), written to establish double contingency for legacy process equipment, had not been fully implemented. As a result, plant staff could not ensure that double-contingency controls had been implemented for certain pieces of legacy process equipment containing fissionable materials (fissile and potentially fissile materials). The NCSAs, GEN 27, for legacy equipment originating from Paducah, and GEN-20, for legacy equipment originating from Oak Ridge or Portsmouth, utilized assay (enrichment) or mass anc' spacing as the two controls relied upon to meet the double contingency principle. The NCSA allowed the assay or mass assessments to be either independent visual observations, non-destructivo analyses, or independent confirmation of the assay based upon the removal location. The NCSAs also permitted legacy equipment, designated as requiring only uncomplicated handling, to be stored with two-foot spacing.

Uncomplicated handling meant the equipment contained less than a safe mass of fissile materials. Equipment containing greater than a safe mass of fissile material required spacing of a least six feet to preclude neutron interaction.

Prior to June 19, plant staff had reported violations of the two-foot spacing requirement for legacy equipment controlled under NCSA GEN-27 to the NRC (see CERs 32049 and 32407). At the time of the last report, the inspectors asked the criticality safety staff if the other control for legacy 17

equipment, specifically assessment of mass or assay, remained in place.

The inspectors were informed that, to the best of the staff's knowledge, the mass or assay controls had not been violated.

While preparing the written response to a GEN 27 related posting violation, issued in NRC Inspection Report 70 7001/97002(DNMS), the criticality staff identified that the mass-or-assay control had not been established for certain pieces of equipment. The failure to implement this control occurred because of a misunderstanding on the part of various plant staff as to the proper implementation of NCSA GEN 20 and GEN 27. Specifically, some plant staff believed the NCSAs allowed legacy equipment to be spaced two feet apart from other items containing fissile or potentially fissile material while waiting (for up to months) for an assessment of mass or assay.

After the discovery of the failure to fully implement GEN 20 and GEN 27, the PSS made a four-hour report to the NRC pursuant to Bulletin 91-01. The report identified that the double-contingency _ controls for storing legacy process equipment had not been fully implemented. - The General Manager immediately stopped all movement of fissile and potentially fissile materials onsite, except for cylinders of uranium hexafluoride, in a letter to the NRC dated June 20,1997, the General Manager committed to perform a walkdown of all buildings containing potentially fissile material by June 23.

In addition, he committed to complete a full chsrecterization of alllegacy equipment, identified during the walkdown, within 45 days (by August 4).

The inspectors' review of the event and follow up by plant staff during the inspection period identified:

1) A walkdown of the entire leased p'>rtion of 'the site was completed by

- midnight on June 23. Plant staff developed detailed inventories of all legacy equipment identified. The project coordinator provided information indicating that the walkdown identified approximately

-450 pieces of legacy equipment or equipment which could not be verified to be new or rebuilt. The project coordinator stated that approximately 220-260 pieces of equipment had not been identified during previous walkdowns implementing GEN-20 and GEN-27.

Other plant staff stated that the more rigorous walkdown approach, i.e., requiring two independent verifications of new or rebuilt equipment, added a significant number of items. The number of addition items included cabinets and boxes which contained valves, pipes, instruments, and miscellaneous items.! Therefore, exact quantification of the number of items lacking double contingency controls could not be made until all the follow up assessments have been completed. Also, a number of the items could qualify for a safe geometry exemption included in GEN 27.'

18

L

2) The procedure implementing GEN 27 was modified to require ten-foot spacing of alllegacy equipment which had not been properly

, characterized as to assay or mass. A criticality safety evaluation documented that, at this spacing, no neutronic interaction would

, occur between fissile items. The inspectors identified some instances where the required 10-foot exclusion area was not achieved for all sides of the equipment. Operations staff immediately corrected these barricade problems.

3)- Plant staff identified eight items which contained potentially fissile materiallocated outside of the area of criticality accident alarm system (CAAS) coverage. After sampling and analysis, four of the items were not fissile, that is, had assays below 1.0 weight percent.

The other four had assays above 1.0 weight percent. Plant staff entered the CAAS Limiting Conditions for Operation (LCO) Action Statements and developed specific movement plans for each piece of equipment to bring the equipment back under CAAS coverage.

4) Plant staff began a campaign to perform assay assessments by smearing the items identified and analyzing the smears. Plant staff identified that approximately 300 smears had to be taken. Some of the smears had been scheduled for collection prior to June 19. Over 95 percent of the smears taken, by the end of the inspection period, indicated material assays below 1.0 weight percent. The associated equipment was then categorized as "non fissi's." For items with smears indicating material with assays grea'ar than 1.0 weight percent, plant staff began performing non-r.estructive assays to determine the mass in the equipment.

5) Of the equipment characterized by the e id of the inspection period that contained fissile material, the maxirnum assay was 16 weight percent. The maximum mass was approximately 600-720 grams of U 235 (based on two independent non-destructive analyses) for an assay machine contaminated with 3.2 weight percent material. None of the equipment characterized to date had a deposit near or greater than safe mass.

Technical Safety Requirement 3.11.5 requires, in part, that the double contingency principle, as stated in the Safety Analysis Report, be used as the basis for the design and operation of piocesses using fissionable materials. Section 5.2.2.3 of the SAR, " Process Evaluation and Approval,"

defines the double contingency principle as foliews: " Process designs should, in general, incorporate sufficient factcis of safety to require at least two unlikely, independent, and concurrent changes in process conditions before a criticality accident is possible. The PGDP (Paducah Gaseous Diffusion Plant] NCS Inuclear criticality safety) program applies this principle by implementing controls either on two different parameters or by implementing two controls on one parameter." The failure to establish 19

double contingency (i.e., two controlled parameters - mass or assay and spacing) for the process of storing legacy equipment containing fissile and potentially fissile materials (fissionable materials)is a Violation of TSR 3.11.5 (VIO 70-7001/97004-11). This certificatee-identified violation is being cited because of the significant number of items for which double contingency was not established,

c. Conclusions A violation of the double contingency principle for the storage of legacy process equipment containing fissile and potentially fissile (fissionable) materials occurred. The violation appeared to result from an incomplete understanding of the nuclear criticality safety requirements by operations and maintenance staff. The violation was related to two prior violations for legacy process equipment issued in NRC Inspection Reports 70 7001/97002 and 70 7001/97003.

E8 Miscellaneous Enaineerina Matters E8.1 iClosed) Violation 70 7001/97002-04: Criticality accident alarm system (CAAS) clusters discovered to be inoperable on two separate occasions because of background readings below design criteria.

The certificatee responded to the violation in a letter dated June 30,1997. The letter committed to implement long-term corrective actions to: 1) change the trouble alarm or fault set point for each CAAS cluster module (detector and associated electronics) to 2.5 milliroentgen per hour (mR/hr) from 5 mR/hr;

2) change the fixed background reading for each module to 4.0 mR/hr from 10.0 mR/hr, and: 3) change the alarm set point to 10.0 mR/hr from 20.0 mR/hr.

Implementation of these set-point changes would ensure that management were aware of decreased CAAS sensitivity prior to the system failing or the SAR assumed radius of coverage for a CAAS module was decreased.

Plant staff developed a temporary procedure for changing the set points and background reading for the CAAS modules. The inspectors reviewed the procedure and confirmed that the changes outlined in the response letter were included. A random check of cluster background readings verified that the CAAS clusters had been modified. Based on the review, the inspectors concluded that the certificatee's corrective actions had bcen appropriate.

E8.2 (Closed) Certificatee Event Reoort 31892: Discovery of an inoperable CAAS cluster in Build C-337.

The corrective actions documented in the written report were essentially the same as those for the violation discussed in Section E8.1. Based on the review discussed above, this certificatee event report is closed. (CER 70-7001/97002-05) 20

9 E8.3 [ Closed) Certificatee Event Reoort 31968: Discovery of an moperable CAAS cluster

-in Building C 337A.

The corrective actions documented in tha written report were essentially the same as those for the violation discussed in Section E8.1 Based on the review discussed above, this certificatee event report is closed. (CER 70 7001/97002-08)

E8.4 ' Certificatee Event Reoorts (907121 The certificatee made the following engineering-related event reports during the inspection period. The inspectors reviewed any immediate safety concerns indicated at the time of the initial verbal notification. The inspectors will evaluate the associated written reports or follow-up for each of these items following submittal.

Numbet Statui 11112 32516 Open Failure to implement double contingency principle for legacy process equipment.

(CER 70 7001/97004-12)

- IV. Plant Suonort P1.1 Emeraency Plan and imolementina Procedures

a. Insoection Scoco (88050)

The inspectors reviewed and discussed the plant's emergency plan (EP) and implementing procedures, and organization and staffing with plant staff to determine if the emergency program was current with site conditions and being maintained in a state of operational readiness. The inspectors also .

reviewed training records and interviewed selected emergency response supervisory staff and_ technicians to evaluate their awareness of emergency procedures. Compliance Plan (CP) issues related to emergency preparedness and fire protection were also reviewed.

b. Observations and Findinni The emergency response organization (ERO) and emergency response personnel responsibilities were consistent with that described in the EP. The EP and implementing procedures provided good guidance on classification and mitigation of the consequences of emergencies, assessment for any potential releases of radioactive materials and hazardous chemicals, personnel accountability, site evacuation, and internal and off-site notification of emergencies. _ _

21

The inspectors noted that selected training records were up-to-date and l training lesson plans for the Plant Emergency Operations Director, Crisis l Manager, incident Commander Shif t Engineer, and the Public Information i

Manager were of high quality and adequately covered subject matter relative to their respective assignments. As part of emergency response capabilities, a state-certified emergency medical technician rescue staff (usually firefighters), who are trained to handle emergencies, was maintained on site.

The staff's training included emergency response in the areas of industrial hygiene, radiological response procedures, hazardous materials decontamination and spill control procedores, and first aid, including mass casualties, procedures.

The staff also maintained an on-site Fire Department which was equipped with a ladder truck and pumper truck to respond to fices and emergencies.

The inspectors noted that the fire protection pre-fire plans had been updated to reflect current facility configurations and conditions as required in the CP.

In addition, a combustible loading analysis was completed as scheduled in the CP. The fire safety staff had developed a draft procedure, .

CP2 SS-FS1038, " Combustible Storage in Process Building." The procedure was in the review and approval process.

3 The certificatee maintained a current roster cf qualified Plant Emergency Operations Center Duty Roster in the Emergency Operations Center (EOC).

Interviews were conducted with several ERO members in different capacities. The inspectors outlined an accident scenario with the shift engineer responsible for plume modeling for the offsite chemical / dose assessment program. Personnelinterviewed were knowledgeable of their responsibilities and procedures for their respective assignment in the emergency organization.

The inspectors determined upon review of the EP, the ERO training records, and staff interviews that the certificatee had satisf actorily addressed the emergency management issues addressed in the CP.

c. Conclusion The EP and implementing emergency procedures provided sufficient guidance for responding to plant emergencies. The emergency organization was adequate for responding to emergencies. Emergency response personnel were adequately trained and were knowledgeable of emergency response procedures and equipment. Compliance Plan issues were adequately addressed.

22

__a

t P1.2 Offsite Suonort Aaencies

a. insoection Scoce (88050)

The. inspectors evaluated the plant staff's involvement with offsite support agencies as described in the EP.

b. Observations and Findinas The emergency plan contained current agreement letters with offsite agencies for response or assistance during emergency events. The certificatee had for mally notified local, county, state, and federal support agencies regarding the biennial emergency management tabletop exercise that occurrod on May 8,1997. Attendance was good from offsite support agencies during this tabletop exercise. Plant staff contacted offsite support ageacies on a quarterly basis to verify telephone numbers and points of contact.

c. Conclusions Plant staff maintained adequate support from offsite agencies for responding to or assisting during an emergency event.

P1.3 QIjlis. Exercises, and Audits

a. Insoection Scoce (88050)

Reco ds of drills, exercises, and audits were reviewed and discussed with cognizant personnel,

b. Observations and Findinas in accordance with the EP, periodic drills or table-top exercises were conducted by the plant's ERO each year. Every two years a major field exercise, consisting of an accident scenario, activation of the ERO, and activation of emergency response f acilities, had been conducted. The biennial field exercise with the NRC and other support agencies participating was scheduled for 1998.

Periodic drills included any of the following: a plant accountability drill, a chlorine drill, an uranium hexafluoride release drill, a fire drill, or a confined space drill. The inspectors noted that the informal expectations for completing plant wide accountability during drills and exercises was between 30 to 45 minutes. The plant staff agreed and stated that they will formalize that expectation in the implementing procedures, as appropriate. Criticality 23

(- .

evacuation drills were performed quarterly. Since March 3, plant staff conducted several fire evacuation drills from all site buildings. In addition, the staff conducted an emergency management tabletop exercise on May 8, ,

with the NRC and offsite agencies.

' ~

Critiques of fire and criticality evacuation drills and the tabletop exercise were detailed and ccmarehensive to identify and correct deficiencies. The inspector noted that although the staff had established an effective system to track corrective actions for identified deficiencies from exercises and drills, some of the expected due dates for correcting the deficiencies were not met. However, the inspectors noted, during a review of several corrective action deficiencies, that no items were identified as having significant safety significance or serious deficiencies in the EP content or implementation. The staff indicated that emergency management resources were committed to resolving the corrective action backlog by the end of third quarter of 1997, The inspectors reviewed an annual emergency preparedness intemal audit that was issued January 22,1997. The audit was performed by members of the Independent Assessments Group and a contractor. None of the auditors had direct responsibility for implementing the emergency response program. The audit evaluated performance relative to the requirements of the EP and procedures, with special emphasis on training, emergency equipment, and memorandurns of understanding from offsite agencies.

There were no audit findings that required a written response and no significant weaknesses were identified. Comments from the audit were being addressed by the emergency response staff,

c. Conclusions Emergency exercises and drills were consistent with the commitments in the EP and adequately exercised the ERO.- Although actions for recornmendations and deficiencies identified in annual exercises were effectively tracked, a minor backlog for corrective actions was identified.

P1.4 Emeraency Eauioment and Facilities

a. insoection Scone (88050)

The inspectors inspected the EOC, other facilities and equipment, and related surveillances to determine whether the emergency response equipment, instrumentation, and supplies located in emergency repositories were maintained in a state of operational readiness. The decontamination trailer and field team monitoring kits were also inspected. The inspector also reviewed the CP issue related to the installation and testing of the public warning sirens and controls. The following reports were reviewed.

24

I e Paducah Gaseous Diffusion Plant, Offsite Public Alerting System Testing and Findings (OPASTF), Revision 1, dated January 1997.

• Siren Sound Propagation Study (SSPS), Paducah Gaseous Diffusion

. Plant, dated January 1.997.

b. Observations and Findinos The inspectors noted that emergency equipment repositories contained the quantities and equipment identified in the EP and implementing procedures.

Cabinets containing emergency equipment and field kits were clearly identifiable, contents were orderly, and well maintained. Survey meters examined were calibrated and operational, and self-contained breathing apparatus air tanks were full. In addition, the inspector verified via docurnentation (in support of maintenance, periodic tests or surveillances) that inventory and operability checks were timely, and that equipment and instrumentation stored at selected locations (Fire Service's emergency response vehicles, Plant Shift Superintendent's emergency response vehicles) were operational and properly maintained.

The inspectors determined that the completed upgrade and augmentation of the public warning siren and controls per the CP were satisfactory.

Surveillance records for the public address system were consistent with the requirements in the implementing procedures in addition, the review of the SSPS and the OPASFT indicated that the reports were consistent with the requirements in the American National Standards Institute (ANSI) S12.14, dated 1992, for fixed outdoor public alerting devices and Federal Emergency Management Agency (FEMA) Radiological Emergency Response (REP) plan, REP-10, for public alert and notification systems.

The EP stated, in part, that decontamination equipment and supplies on emergency response vehicles were adequate for decontamination activities in the field. However, a self-contained personnel decontamination trailer was available, if needed. The inspector noted from review of surveillance records that the decontamination trailer has not been available to conduct radiological decontaminations of contaminated personnel since 1995 due to pending Nuclear Criticality Safety (NCS) evaluations of the various shower drain holding tanks. At the exit, the Enrichment Plant Manager committed to resolving the NCS concerns by November 1997. As an interim compensatory measure, the plant staff was using a portable decontamination system (series of wading pools with various water hoses).

Although no emergency event occurred that required the decontamination trailer, the inspector noted that ample opportunity existed for the plant staff to resolve the NCS concerns related to the decontamination trailer through a review of the quarterly emergency preparedness surveillance program records.

25

c. Conclusions The emergency preparedness staff maintained a good inventory of well-maintained emergency response equipment and supplies that were in a i

state of operational readiness._ The CP issue related to the public warning system was completed adequately.

P1.5 Emergency Pseparedness Organization and Administration

a. insoection Scooe (88050)

The inspectors conducted discussions with the EP staff regarding the current plant organization and reviewed the current organizational chart,

b. Observations and Findinas The overall organization and management structure of the EP function was consistent with the EP and implementing procedures. The Emergency Management Drill and Exercise Coordinator (DEC) and two emergency management specialists (EMS) reported directly to the Emergency Management Manager (EMM), who reported to the Plant Site and Facility Support Manager, who reported to the Plant General Manager. Recently, an emergency management analyst was transferred to another organization and an emergency management consultant left due to budget constraints.

Discussion with the 6MM indicated that his primary emphasis has been the plant emergency management program although one emergency management trainer had been assigned to other than EP duties. As a result, the staff involved in the emergency preparedness program utilized the administrative office manager to help conduct required surveillances in addition to maintaining various EP databases. The inspectors reviewed selected surveillances conducted by the administrative office manager and determined that the surveillances were consistent with the requirements in the EP and reflected a high level of competence. Interviews of the administrative office manager indicated a high degree of knowledge regarding the EP, implementing procedures, and associated EP surveillances requirements. The administrative office manager's training records indicated that ERO and supplemental hazardous training requirements have been completed as required to conduct required EP surveillances and EP training,

c. Conclusions The structure of the emergency preparedness function was consistent with the requirements in the EP.

26

S1 Conduct of Securit < and Safeauards Activities S1.1 Securitvf,tuards Hours of Work

a. IDinection Scone (88100)

The inspectors reviewed the controls on and work hours of security guards.

b. Observations and Findinas During the period, the inspectors discussed with site security personnel their normal work hours and the use of overtime since the NRC assumed .

regulatory authority on March 3. Several of the individuals interviewed indicated that the use overtime hours was not significant during the March and April time period; however, overtime had increased and become routine in May and June. The guards attributed this change to management's use of the security force to ensure staff compliance with TSR exclusion areas required fer inoperable criticality accident alarm systems.

Based upon discussions with the security force, the inspectors reviewed the work hours of all guards since March 3. The inspectors noted an increase in the frequency and use of overtime beginning in the May. The overtime include guards workin] in excess of 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in a seven day period.

Plant procedure CP2-HR-LR1030, Limitations on Hours of Work," change B, dated May 7,1997, limited the hours of individuals performing safety work to less than 16 hours1.851852e-4 days <br />0.00444 hours <br />2.645503e-5 weeks <br />6.088e-6 months <br /> in 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />, 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> in 48 hours5.555556e-4 days <br />0.0133 hours <br />7.936508e-5 weeks <br />1.8264e-5 months <br />, and 72 hours8.333333e-4 days <br />0.02 hours <br />1.190476e-4 weeks <br />2.7396e-5 months <br /> in seven day period. An internal document, HR-001, " Personnel Performing Safety Functions," defined those individuals covered by the procedure. The inspectors reviewed HR-001 and noted that the security guards were not included as individuals performing safety work.

The inspectors discussed the findings with the security manager. The manager concurred with the inspectors findings. The manager also agreed that use of the guards to ensure compliance with the TSR exclusion area requirements was the primary cause for the increased use of overtime.

However, the manager noted that other factors could also cause the guards to exceed the procedure limits. In regard to the lack of application of the hours of work limitation to the security guards, the manager indicated that the certificates on June 26 sent a letter to the NRC requesting clarification of this issue.

27

On July 18, the NRC, in a letter from Mr. R. C. Pieraon, Chief of the Special Projects Rranch, informed the certificatee that the TSR 3.2.2.b, limiting the hours of work, applied to at least some of the security guards. Specifically, the TSR applied to the extent that the guards were used to meet the minimum staffing requirements of TSR Table 3.2.2.1 or to perform other safety functions.

Technical Safety Requirement 3.2.2.b requires, in part, that administrative procedures shall be developed and implemented to limit the working hours of facility staff who perform safety functions. Technical Safety Table 3.2.2.1,

" Minimum Staffing Requirements," specifies that four security guards shall be onsite at all times to perform safety functions. The failure to develop and implement administrative procedures to limit the security guards hours of work is a Violation (70 7001/9700413),

c. Conclusions The inspectors identified that management had not developed and implemented administrative procedures to limit the work hours of security guards. As a result, during May 1997, several security guards worked hours in excess of the limits specified in the Technical Safety Requirements. This is a violation.

S8 Miscellaneous Security Matters S 8.1 IQpen) Certificatee Event Reoort 32529: A one-hour report made by the certificatee pursuant to 10 CFR 95.57(b) for a possible compromise of classified information due to a subcontractor being allowed unescorted access to the C-100 Building af ter his clearance had expired.

A subsequent review by plant security staff identified that the individual's expired access was discovered directly af ter he entered the building. The individual was never alone and there was no classified information in the office he entered. Upon discovery, the individual was immediately escorted to the gate where he obtained an escort required badge in accordance with site procedures. Based on the review, the certificatee concluded that no compromise of classified iaformation had occurred and retracted the initial report. The inspectors concluded that the certificatee's review and retraction appeared reasonable. The inspectors will review the certificatee's corrective actions for ensuring individuals with expired clearances are not allowed unescorted access to the site. (CER 70-7001/97004-14)

F8 Miscellaneous Fire Protection Matters F8.1 (Closed) Certificatee Event Reoort 32320: Building C-337 sprinkler system D7 head installed too far from ceiling.

The plant engineering staff performed an engineering evaluation te demonstrate that one head located too far from the ceiling would have no impact on the ability of the 28

sprinkler system M perform its safety function. The evaluation concluded that the system would still aorm with a proper response time due to the actual ceiling jet and coverage from adjacent heads. Based on the engineering evaluation, the certificatee retracted the event report. The inspectors concluded that the evaluation and retraction appeared reasonable. (CER.70 7001/97003 24)

F8.2 (Closed) Certificatee Event Renort 32324: Foreign material discovered in N Building C 337 sprinkler system D 7 piping drop for ventilation duct.

The plant engineering staff performed an engineering evaluation to demonstrate that blockage of one head would not affect the ability of the sprinkler system to control-a fire in the ventilation ductwork because of a sufficient number of heads in close-proximity.to the affected head to control the fire. Based on the evaluation, the certificatee retracted the event report, The inspectors concluded that the evaluation and retraction appeared reasonable. (CER 70-7001/97003 25)

V. Manaaement Meetinas X. Exit Meetina Summary The inspectors presented the inspection results to members of the plant staff and management at the conclusion of the inspection on July 14,1997. 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 identified.

4 i

29

PARTIAL LIST OF PERSONS CONTACTEQ United States Enrichment Corooration

'J. H. Miller, Vice. President Production

'J. A. Labarraque, Safety, Safeguards and Quality Manager-Lockheed Martin Utility Services (LMUS)

'S. A. Polston, General Manager

'H. Pulley, Enrichment Plant Manager -

'W E. Sykes, Nuclear Regulatory Aff aire, Manager

'S. R. Penrod, Operations Manager United States Deoartment of Enerav (DOE)

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

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

'J. M. Jacobson, Resident inspector

'T. D. Reidinger, Senior Fuel Cycle Inspector

' Denotes those present or who participated by telophone in the either the July 11 or 14, 1997 exit meetings.

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

INSPECTION PROCEDURES USED IP 88050 Emergency Preparedness IP 88100 Plant Operations --

IP 88102 Surveillance Observations .

IP 88103 Maintenance Observations

'IP 88105 Management Oversight and Controls IP 88020 - Regional Criticality Safety.

IP 90712 Inoffice Review of Events 30

ITEMS OPENED, QJ,OSED, AND DISCUSSED Onened 70 7001/97004-0,1 VIO failure to train building operators par TSR 3.4 70 7001/97004-02 ~ CER loss of building'C 310 CAAS audibility 70 7001/97004-03 CER actuation of building C 360 crane hoist brakes 70 7001/97004-04 CER loss of building C-310 CAAS audibility 70 7001/97004 05 CER actuation of building C-360 autoclave steam pressure isolation 70 7001/97004-06 VIO preconditioning of autoclave head to shell alignment 70 7001/97004-07 CER building C 310 west normetex pump detection system-modified

' 70-7001/97004-08 CER lon pumps not stored in accordance with ncs controls 70 7001/97004-09 VIO failure to develop and install modification per TSR 3.9 70 7001/9700410 VIO inadequate 76.68 review for cell trip / material limits in sar 70 7001/9700411 VIO failure to ensure double contingency per TSR 3.11.5 70 7001/9700412 CER failure to ensure double contingency per TSR 3.11.5 70 7001/97004-13 VIO failure to limit security guards hours of work per TSR 3.2.2.b 70-7001/9700414 CER failure to prevent unauthorized sits access

.Cl.Q.Sfd 70-7001/97002-04 VIO- caas inoperable due to low background reading i 70-7001/97002-05 CER discovery of inoperable caas cluster in building C 337 70 7001/97002 08 CER discovery of inoperable caas cluster in building C-337A 70-7001/97003 24 CER building C 337, D 7 sprinkler system head too far from ceiling 70 7001/97003-25 CER building C 337, D 7 sprinkler system foreign materialin pipes Discussed None 31 c

l

9 e

LIST OF ACRONYMS USEQ ACR Area Control Room ANSI American National Standards Institute CAAS- Criticality Accident Alarm System CER: Certificatee Event Report '

CFR- Code of Federal Regulations CP - Compliance Plan

- DEC Drill and Exercise Coordinator EOC Emergency Operations Center EMM- Emergency Management Manager EMS Emergency Management Specialist j EP Emergency Plan ESO Engineering Service Order - l

- ERO Emergency Response Organization FEMA Federal Emergency Management Agency HF _ Hydrogen Fluoride LCO Limiting Condition for Operation MW_ Mogawatt NaF Sodium Fluoride NaHF, Sodium Bi Fluoride NCS. Nuclear Criticality Safety NCSA Nuclear Criticality Safety Approval NCSE Nuclear Criticality Safety Evaluation - '

NOV --- Notice of Violation NRC. Nuclear Regulatory Commission PORC- Plant Operations Review Committee PSS _ Plant Shif t Supervisor REP Radiological Emergency Plan SAR Safety Analysis Report TSR Technical Safety Requirement

- U-235 Uranium 235 =

UF6- Uranium Hexafluoride UH . Uncomplicated Handling USEC United States Enrichment Corporation -

VIO Violation 32

. . . . . . _ _ _ _ _