ML20202E882

Insp Rept 70-7001/97-208 on 970915-19 & 1103-07.Violations Noted.Major Areas Inspected:Mgt & Administrative Practices for Nuclear Criticality Safety,Plant Operations,Operating Procedures & Criticality Alarm Monitoring Sys

ML20202E882
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Site:	07007001
Issue date:	11/28/1997
From:	NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
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70-7001-97-208, NUDOCS 9712080171
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U.S. NUCLEAR REGULATORY COMMISSION OFFICE OF NUCLEAR MATERIAL SAFETY AND SAFEGUARDS Docket No: 70-7001 Certificate No: GDP 1

- Report No: 70 7001/97-208 Certificate Holder: United States Enrichment Corporation Location: Paducah Gaseous Diffusion Plant Paducah, Kentucky Dates: - September 15 -19,1997 November 3 - 7,1997.

Inspectors: Dennis Morey, Lead Inspector, NRC Headquarters Christopher Tripp, Inspector, NRC Headquarters -

Sandra Larson, NRC Contractor .

Andrew Prichard, NRC Contractor Approved By: Philip Ting, Chief, Operations Branch, Division of Fuel Cycle Safety and Safeguards, NMSS Enclosure 3 ,

9712080171 PDR 971128 ADOCK 07007001

_ .C - PDR

e UNITED STATES ENRICllMENT CORPORATION PADUCAll GASEOUS DIFFUSION PLANT NRC INSPECTION REPORT 70-7001/97-208 EXECUTIVE

SUMMARY

Areas Insnected An NRC lleadquarters team conducted an announced nuclear criticality safety inspection of the Paducah Gaseous DifTusion Plant (PGDP) in Paducah, Kentucky, on September 15 - 19 and November 3 - 7,1997. The inspection was conducted using NRC Headquarters staff and two NRC contractors. The focus of this inspection was to determine the level of nuclear criticality safety (NCS) through review of NCS program implementation as described in Part 76, the certi6 cation application, and Se compliance program.

The inspection was conducted using Inspection Procedure IP 88015 in the following areas:

• Management and Administrative Practices for NCS

• NCS Function l

i e Plant Operations

• Operating Procedures

• Maintenance for NCS

• NCS Training

• NCS Inspections, Audits, and Investigations

• Criticality Alarm Monitoring Systems

• NCS Emergency Re,ponse As a result of this inspection, three Level IV violations of NRC requirements, one deviation from a commitment, two unresolved items, and six inspector follow-up items were identiGed.

Results

• Flowdown of controls into procederes was adequate and operators were aware of the signincance of NCS controls, how controls were implemented, and how to take mitigative action in response to process upset conditions. (Section 1.0)

• A deviation was identined involving a change to a commitment made to the NRC one day before the due date. PGDP had committed to correct Violation 97-201-02 by August 20,1997, and by letter dated August 19,1997, the date was changed to November 12,1997. (Section 1.0) 2

• One Level IV violation was identified conceming inadequate validation of the SCALE criticality modeling code at 5.5wt% enrichment. (Section 2.0)

• A program weakness was identified in the area of rigorousness of NCS analysis.

(Section 2.0)

• The procedures for laboratory analysis were adequate to ensure independence of dual sampling required for double contingency, and the laboratory personnel were knowledgeable of the analysis procedures. (Section 3.0)

• The document control piogram was adequate to ensure version control of NCS-related documents, including NCS As and operating procedures. (Section 6.0)

• The elements of the Maintenance Program were found to conform to the SAR requirements. (Section 7.0)

• PGDP training programs are in compl ance with specific SAR requirements for specific NCS training and training curricula supported full implementation of SAR requirements.

(Section 8.0) l

• Communication between NCS and plant operations was inadequate. One Level IV l violation was identified involving failure to notify NCS of potential criticality problems.

l (Section 9.0)

• Characterization of surveillance findings and implementation of corrective actions is inadequate. One Level IV violation was identified involving inadequate corrective actions resulting from an NCS surveillance. The proposed corrective actions were not completed before restarting the refeed operations. (Section 9.0)

• CAAS detector placement was adequate to provide complete coverage of fissile material areas. (Section 10.0)

• Current emergency planning is adequate to protect the health and safety of the public and workers in the event of a nuclear criticality accident. (Section i 1.0) 3

REPORT DETAILS 1.0 Management and Administrative Practices for NCS Scope The inspectors reviewed selected operations based on potential risk significance to determine whether plant operating conditions were suflicient to reliably maintain criticality safety controls as identified in the applicable NCSEs Inspectors also reviewed two management issues with PGDP staff.

Qbservations and Findings A. Reliability of Controls The inspectors reviewed the C-400 Spray Booth operation, the Uranium Recovery operation, and the Process Coolant System. Review of the C-400 Spray Booth operation included the following documents:

• KY/S-244, Rev. 6," Nuclear Criticality Safety Evaluations of the C-400 Spray Booth Operations at the Paducah Gaseous Diffusion Plant"

• NCSA 3973-06, Parts A and B,"C-400 Spray Booth"

• Procedure CP4-CU-CH2108, Rev. 4," Operation of the C-400 Spray Booth" In addition to conducting a review of these documents, the inspectors walked down the C-400 operation. The flowdown from NCSA/NCSE requirements to field operations was well-developed. The inspectors found that the operations staff understood what parameters were being controlled from a criticality safety perspective and haw the control was achieved. The operators did not view the operational tasks as difficult or burdensome, and the controls were observed to be reliable.

The inspectors reviewed the following documents as part of their review of Uranium Recovery:

• NCSA 3973-09, Part A and B," Uranium Recovery System"

• Procedure CP4-CU-CH2111, Rev.1," Operation of the Contaminated Solutions Receiving and Storage Facility" In addition to conducting a review of these documents, the inspectors walked down the Cor.taminated Solutions Receiving and Storage Facility. The operations staff clearly understood what was being controlled from a criticality perspective. The tasks did not appear difficult and 4

the operators did not consider them unduly burdensome. Thus, the controls were observed to be reliable.

l l The inspectors reviewed the following documents as part of their review of the Process Coolant l System:

• KY/S-243, Res. 3," Nuclear Criticality Safety Evaluations for the Cascade Equipment and Facilities at the Paducah Gaseous Diffusion Plant"

• NCSA 3971-01, Parts A and B," Operation and Maintenance of the UF6 Enrichment Cascade" e Procedure CP4-CO-CN6069, Rev.1,"TSR Surveillance - Shiftly and Daily Checks"

• Procedure CP4-CO-CN2030. Rev.1, " Inspection, Removal, installation, and Handling of Uranium Contaminated Cascade Equipment"

• Procedure CP4-CO-ON3006, Rev. O," Major Air or Nitrogen inleakage to Cascade"

• Procedure CP4-CO-ON3007, Rev. O, " Major R-114 Inteakage to Cascade" In addition to conducting a review of these documents, the inspectors walked down the Enrichment Cascade. The inspectors interviewed the operations staff, focusing on the off-normal event of a major R-114 coolant leakage into the process gas ana how this might compromise the barriers that keep the RCW water from the process gas. The operators clearly articulated the barriers keeping the water from the process gas, and understood the indicators of this off-normal event, the required response time, and the various process-related automatic operations that respond to a major R-114 inleakage. The inspectors concluded the operators have diverse indicators of a major R-114 inleakage and have a significant a nount of time to respond to a major R-114 leak. Insrsectors also concluded that the procedures adequately detail the necessary steps to preclude wy r inleakage into the process gas in the event of an R-114 leak, and that the steps required to respond to an R-114 leak can be easily achieved.

The inspectors concluded that the controls available to preclude water inleakage into the process gas from an R-114 leak are reliable.

B. Management Issues PGDP committed to correct Violation 97-201-02 on August 20,1997. By letter dated August 19, 1997, the date was changed to November 12,1997. Change to a commitment normally requires concurrence of the NRC. Inspectors discussed this issue with PGDP management and requested that concurrence be obtained prior to changing these commitments. Changing this ccmmitment without NRC concurrence is Deviation 70-7001/97-208-01.

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In respcnse to Unresolved item (URI) 97-201-05, PGDP is in the process of changing the self-assessment procedure to incorporate specific requiremems for CM program assessments.

Inspectors reviewed corrective actions with CM staff and determined that enough infomiation was available to resolve this issue. Corrective actions were not complete and revision of the Self-Assessment procedure will be tracked as IFl 70-7001/97-208-02.

Conclusions The inspectors determined that the field operating conditions were reliable and adequately maintain the controls identified in the NCSEs. The flowdown of controls into the procedures.

was adequate and the operators were aware of the significance of NCS controls, how they were implemented, and how to take mitigative action in response to process upset conditions.

2.0 NCS Function Scope The inspectors reviewed the validation of computer codes used in criticality analysis against the requirements of the PGDP Safety Analysis Report (SAR), to ensure that the codes were validated in the region in parameter space in which they were used. The inspectors also reviewed selected NCSAs for rigor of analysis and reviewed the NCS staff actions in response to a 91-01 issue identified at Portsmouth Gaseous Diffusion Plant. The inspectors reviewed recent NCSA violations of GEN-15, since this had the largest number of violations since October 1996.

Observations and Findings A. Validation The inspectors reviewed KY/S-221, Rev.1," Validation of the Paducah Gaseous Diffusion Plant Nuclear Criticality Safety Code System for the ENDF/B-IV 27 Group Cross Sections." This validation report was used for uranium enriched from 1.0 to 5.5wt%. The benchmarle experiments used ranged from 0.71 Iwt% to 4.98wt% enriched uranium. During the review of this document, some problems were identified.

The method used to establish that the k,, Upper Safety Limit (USL) is sufficient to ensure with 95% confidence that 99.9% of all future calculations with a k,y < USL will be subcritical depends on the normal distribution of calculated benchmark results. The analysis offered a qualitative assessment that the benchmarks are normally distributed, but does not provide a rigorous analysis to justify this assessment.

The validation document described the methodology used to determine that the k,y USL meets the SAR requirements and has the SAR-defined minimum margin of subcriticality. This method was based on a point estimator of both the average and standard deviation of the calculated 6

benchmark k,a's. The inspectors found this method to be rigorous when the stated area of applicability is inside the range sprumed by the benchmark data, but is not appropriate outside the range of data. Outside the range of experimental data, the expectation is that the uncerte;aty in

- the bias will increase and the USL will be required to decrease. Several calculations were done at an enrichment of 5.5wt% without adequate justification that the statistical requirements of the SAR can be met. The validation report did not rigorously demonstrate that the k,, USL was adequate to ensure with 95% confidence that 99.9% of all future calculations with a k,y < USL will be suberitical, as required in the SAR.

l SAR Section 5.2.2.1 requires compliance with ANSI /ANS-8.1-1983, which requires that l computer codes be validated and that the area of applicability of the validation may be extended

! by analyzing trends in the data, and by including allowances for uncertainties in the bias and

! uncertainties due to extensions of the area of applicability in the margin of suberiticality. SAR l Section 5.2.3.2 requires that a statistical analysis be used to establish calculational bias, and that the margin of suberiticality be sufficient to ensure with 95% confidence that 99.9% of all future calculations will be suberitical when the calculated k,y < USL. The failure to provide this

assurance is Violation (VIO) 70-7001/97-208-03.

B. Rigor of Analysis The inspectors also reviewed NCSE-207,"Five-Gallon Drum Waste Oil Storage at 5.5wt% 2"U at the Paducah Gaseous Diffusion Plant." This NCSE is the basis for much of the waste storage in arrays on-site. The inspectors found that this NCSE allowed the storage of an infinite array of five-gallon drums with a mass limit of 120 g2"U, at a maximum enrichment of 5.5wt%. The worst case calculational model lumped all of the uranium in a small portion of the drum and considered the remainder filled with oil. This causes the uranium a plane to be isolated from the planes above and below. Based on the available waste streams, the regulatee claimed this model was the most conservative credible configuration of the fissile material. The regulatee agreed to provide additional information to demonstrate the reasonableness of the lumped model. Because the model has not been demonstrated to be sufficiently conservative, this is being tracked as Unresolved Item (URI) 70-7001/97-208-04.

The NCSE-207 calculations also take credit for a portion of the steel wall of the five-gallon drum to reduce the system reactivity. The five-gallon drum is an AQ-NCS item in the BDM but a minimum steel-wall thickness is not given. The regulatee committed to updating the BDM to ensure this parameter is controlled. These discrepancies do not represent an immediate safety concern because the observed drums contained a maximum of 10 g2 "U.

Inspectors reviewed the PGDP response to a 91-01 issue that was identified at Portsmouth Gaseous Diffusion Plant. The reported event was a failure to valve out RCW water when the RCW condensers were empty. PGDP determined that the NCSA requirement to valve out RCW water prior to investigating a freon leak in the RCW condenser should include the RCW condenser in the Purge and Evacuation (P&E) system freon condenser. PGDP staffis modifying 7

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the NCSE and the NCSA to include the P&E system. Modi 0 cation of the RCW condenser NCSE will be tracked as inspector Follow up item (IFI) 70-7001/97 208-05.

Inspectors reviewed corrective actions for an NCS violation at Building C-400. Problem report PR.EN 971540 states that the Air Capture System at Building C-400 was installed and operated without NCS approval. inspectors walked down the Air Capture System and reviewed corrective actions. Corrective Actions include developing and publishing an NCSA for the system. The immediate compensatory measure was to shut down the system. The NCSA will be issued in November 1997. Corrective actions for the C 400 Air Capture system will be tracked as IFl 70 7001/97 208-06.

The inspectors selected several violations of the NCSA GEN 15 to detennine the reason for the large number of violations and whether corrective action was necessary. The review consisted of tha. fobowing documents:

• Memorandum C 102 T-06, MS 102TG, PGDP (6608), dated July 21,1991.

• Memorandum C 102 T-06, MS 102TG, PGDP (6608), dated November 6,1997.

• Criticality Safety Audit / Inspection Reports NCS INC 97 014, 023 029, 031,-047,

-049, -052, and -057.

• NCSA GEN 15, Parts A and B,"On-Site Generation,llandling, Accumulation, Staging, Transportation, and Storage of Fissile or PotentiaFy Fissile Waste up to a Maxin'um of 5.5 Weight Percent Enriched."

• NCSA WM-01, Parts A and B," Transportation of Temporary Fiscile Storage Areas (TFSAs), Associated Drum Monitors, and Consolidation Areas."

The inspectors also conducted a plant tour focusing on storage arrays governed by GEN 15.

During a review of GEN 15 violations, the inspectors identified an inconsistency in the NCSA.

GEN 15 required all containers in fissile control arvas (FCAs) to contain less that 5.5 gallons, unless specifically covered by another NCSA. The inspectors identined two drums with

>30 gallons in a GEN 45 NCSA; these drums were allowed under NCSA WM-01. Three out of the eight violations involving GEN-15 were related to containers larger than 5.5 gallons inside a GEN 15 FCA. The potential exists for operations staff to misconstrue the 5.5 gallon lirnit because another analysis pennits larger drums. The conditions under which larger d.umu are allowed inside an FCA appear to be unclear. The regulatee has initiated activity to analyze repeated violations of GEN-15 a f revise the NCSA as appropriate. Revision of NCSA GEN-15 will be tracked as IFl 70-7001/97-208 07.

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Conclusions The validation report did not demonstrate rigorously that calculations perforrned for S.Swt%

enriched uranium systems meets the above stated S AR requirements. Although the extension from 4.98 to 5.Swi% is small, the use of criticality safety codes outside the range of applicability (in enrichment or another parameter) is a safety concern when calculated results are close to the upper safety limit. The actual limits applicable for different confiden:e levels depend on the fomi of the distribution of calculated results; thu-). esults that are not normally distributed may yield a non-conservative USl This is not an immediate sa' ty issue, because th process enrichment is limited to 2.75wt% maximum.

3.0 Mant Operations Scope The inspectors reviewed operation of the Radiochemistry Laboratory and the TIMS operations, to ensure independence of dual analysis required to support double contingency in certain plant operations.

Observations and Findir gs Many NCSE comrols rely on uranium assay and/or concentration for criticality safety, and require dual analysis oflaboratory samples to support dauble contingency. The inspectors found that dual samples are received by the laboratory and assigned the same sample number followed by A or B. Procedures provide two options of analysis for laboratory techniciam: a) two difTerent technicians may analyze the two samples on difTerent machines on the same day, or b)

' the same tecimician may analyze both samples on any instrument on two difTerent days. The laboratory staffindicated that this second method would be more reliable. At a minimum, the machines are calibrated using standards at the beginning and end of each day, as well as after every tenth sample. The technician enters the resn'ts into the comuter database accessible to operations and the results are confinned by a supervisar.

Conclusions The procedures for laboratory analysis were sufficient to ensure independence of dual sampling required for double contingency, and the laboratory personnel were knowledgeable of the analysis procedures.

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6.0 Ooerating Procedurcs SCDDC The inspectors examined the document control program and sampled NCSAs and opeiating  !

procedures in the field to ensure that the latest revision of NCS related documents were in use, and to determine whether procedures existed for the approval, issuance, and revision of these documents.

Observations and Findings The inspectors reviewed procedure IJE2-TO RM1031," Document Control Program," to ensure that it properly implements the requirements of SAR Section 6.10. The inspectors determined that there was a centrally controlled database for NCSAs, NCSEs, and procedures and that the procedure met the requirements for the revision of controlled documents and transmittal of revisions to predetermined copyholders. The inspectors chose a sampling of NCSAs and audited several copies of the NCSAs and their associated operating procedures to ensure that the latest revision was in use by operations and by the NCS Section. This review did not reveal any discrepancies in revision, though there were several cases in which the procedure number referenced in the NCSA did not match the actual procedure number. The regulatee indicated that th!s was due to an ongoing change of the procedure numbering system.

Conclusions The document control program was adequate to ensure ve.rsion control of NCS related documents, including NCSAs and operating procedures. An audit of these documents did not reveal any inconsistencies and the program appeared to be effective.

7.0 hiain'.mance for NCS SCDDR The inspectors reviewed the work control process, including maintenance planning, scheduling, and the he,, ackage Control Center, against the requirements of the SAR in order to ensure continuing safe operation of the plant.

Observations 2nd Findings The Boundary Definition Manual (BDM) is used to classify plant equipment as quality (Q),

augmented quality (AQ), augmented quality for NCS (AQ NCS), and nonsafety (NS) systems, for purposes of work control, and prioritization of preventive maintenance (PM) and corrective 10

maintemnce (CM), as well as other maintenance activities. The wk control process is similar for Q and AQ NCS items, but less stringent for AQ and NS systems. Adequate operation of the maintenance program depends strongly on the accuracy of the IIDM.

The work control procedures, and the Computerized Maintenance Management System (CMMS) which tracks the maintenance activities, have been implemented for only a brief period of time, I and thus there is not a large database from which long tenn trends can be established. The computer program Primavera is used for maintenance scheduling and infomiation is transferred between the Primavera and CMMS systems on a daily basis. Priority is given to Q and AQ NCS items with respect to scheduling preventive and corrective maintenance. Work packages that can Le saved in the system include a system description, work instmetions, post maintenance testing requirements, and applicable procedures. NCS must approve all work packages that do not contain an approved procedure. Once the maintenance activity has been completed and verified l by the PSS, the activity is logged to the CMMS as a pleted. Quality Assurance and the maintenance plarmer must verify that the work package is compue before it is stored in the Work Package Control Center.

, CMMS is also used to track performance indicators. such as overdue PM and time to complete CM. The current backlog for these activities is close to the target. The Compliance Plan item for trend analysis of equipment failure was due on September 30,1997. Procedures are in place to perform this analysis, but insufficient data has been compiled to permit a thorough inspection.

Conclusions The elements of the Maintenance Program were found to conform to the SAR requirements.

8.0 NCS Training SE@c SAR Section 6.6 requires specific criticality safety training during General Employee Training, Operations and Maintenance Technical Training, Nuclear Criticality Safety Engineer / Specialist Training, and Manager / Supervisor Training. Inspectors reviewed the PGDP training program to verify that required NCS programmatic training objectives were implemented in conformance with the SAR. Inspectors also reviewed training records of selected employees with specific NCS responsibilities to verify that training and qualification requirements were completed.

Observations and Findings initial training for employees at PGDP consists of General Employee Training which includes an NCS orientation film and an overview of the physics and practices of criticality safety.

Supervisors and Managers receive additional training in NCS policy.

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Personnel who handle fissile material or work in Fissile Control Areas (FCA) or supervise these operations receive training in the principles and practices of NCS. The fissile material handler training includes instruction in the physic- . criticality, controls, historical accidenu, and plant policy. Upon completion of the training, staff are required to pass a written examination.

In addition to the above training, employees receive task specific training when tasks involve NCS. This training may consist of crew briefings or required reading and is intended to inform workers of NCS requirements that are unique to a particular job or system such as specific NCSA controls. An NCS Engineer has specific training and qualification requirements defined in SAR Section 5.2.2.3 which are: 1) a bachelors degree in engineering, mathematics or related science,

2) one year at the facility,3) a KENO V.a course,4) complete four evaluations under a senior NCS engineer,5) perfonn walk through inspections under a qualified NCS engineer,6) NCS surveillance team training,7) a nationally recognized criticality safety course, and 8) one year of physics if the employee does not have a nuclear engineer or physics degree A senior NCS engineer is also required to complete 1) four technical reviews of NCS evaluations under a qualified senior NCS engineer or NCS Section Manager, and 2) complete one year as a qualified NCS engineer and two years at the facility.

Inspectors reviewed the training and qualification records of a qualiikd NCS engineer and a qualified senior NCS engineer. PGDP did not maintain a file for either employee to document compliance with this specific SAR requirement. The employees eventually produced documentation that they maintained at their own initiative to demonstrate that they met the requirements. A new qualification card process was partially completed an can be expected to fully document SAR requirements when implementation is complete.

Conclusions PGDP training programs are in compliance with specific SAR requirements for specific NCS training. Training curricula supported full implementation of SAR requirements.

Documentation of NCS engineer training and qualification requirements was informal; however, selected employees were able to demonstrate compliance with personal records.

9.0 NCS Insocctions. Audits. and Investigations Scope The inspectors teviewed the nuclear criticality safety audit and inspection program to verify that operations are inspected by the NCS staff on a periodic basis adequate to ensure safe operation of the plant, that prompt and effective corrective actions are taken in response to surveillance findings and problem reports, and that violations of criticality controls are trended and tracked to resolution. The inspectors also reviewed examples ofintemal surveillances and walk throughs.

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Ohsrvations and Findings A. Problem Reports Procedure CP4 EG-NSI104," Nuclear Criticality Safety Engineer Response to Emergency, Off Normal, and Procass Upset Conditions," covers the review of problem reports, including violations of nuclear uiticality controls, by the Nuclear Criticality Safety function. The problem reports are maintained in the Business Prioritization System (BPS) and assigned to a relevant organization by the Problem Report Scre:ning Committee (PRSC)in Commitment Management.

The inspectors reviewed the problem reports for the period March 3 September 16,1997, and identified 35 problem repons relating to violations of nuclear criticality controls. These were examined to determine whether they were resolved promptly and effectively, whether investigations were conducted to determine root cause, and whether the events were correctly classified as reportable or non reportable under NRC Bulletin 91-01.

Upon discovery of a potential loss of criticality controls, Procedure UE2 IIR-C11030, " Problem Reporting," requires that a problem report is issued and the Plant Shift Superintendent (PSS) is notified. Procedure CP4 EG NSI104 requires notification of the NCS ftmetion, generally by the PSS or facility manager as well as issuance of an NCS Audit / Incident Repon within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> of discovery. This report identifies the nature of the violation, the apparent cause, and the recommended mitigative corrective actions. This is an informed evaluation of the situation and a recommendation of those actions necessary to restore the operation to a state of double contingency, rather than a comprehensive root cause investigation. The inspectors res ewed the 35 NCS Audit / Incident Reports and concluded that three had not been completed within the required 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br />. These were Audit / Incident Reports NCS INC-97-013, NCS INC-97-015, and NCS-INC-97-046. The time between discovery of the NCS violation and filing, of the Audit /

Incident Report varied from one to three months for these three cases. In addition, in the case of 97-015, immediate corrective action was taken although NCS did not issue the required report for six weeks. The corrective action was to remove incorrectly placed plastic bags from an FCA in the C-720 building, in violation of NCSA GEN-15. Upon interviewing the NCS and PSS staff, it was detennined that in these cases NCS had not been notified as required by the PSS on duty when the problem report came in. NCS staff had later identified these as instances when they should have been r otified, and then filed the appropriate Audit / Incident Report. The failure of NCS staff to evaluate these potential violations of nuclear criticality safety controls and to issue the equired Audit / Incident Reports within 24 hours2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> is VIO 70-7002/97-208-08.

The failure to review violations of critiesiity safety controls promptly is a safety concem because it may allow operations that are no'. doubly contingent to persist for long periods of time, and because inappropriate or ineffectual corrective action may be taken without proper guidance.

An additional concem exists because in one case the configuration of an operation found in an abnormal wndition was modified without NCS concurrence.

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in two of these cases, the fact that the required notification of the NCS function was not made was noted (one in an infonnal memorandum and one in the Audit / incident Report), but was not ,

flowed into a problem report and corrective action was not taken. (Audit / Incident Report NCS INC 97-046 was issued on September 9,1997, and an inquiry was ongoing at the time of the inspection to determine whether a problem report should be issued.)

B. Incident Classification The inspectors also found that in 7 of the 35 Audit / Incident Reports, the cause of the violation was listed as " unknown" and in one case was not filled in on the fonn. In two cases, the cause was listed as insufficient training, but the proposed corrective action did not address the cause.

In two cases, the cause was listed as lack of NCSA clarity, but the proposed corrective action did not address the cause. Although the cause is not meant to be a fonnalized root cause determination, it represents the expert judgement of the NCS engineer and should be considered in detennining corrective actions. Problem reports are sent along with the Audit / Incident Reports to Commitment Management for review and disposition by the PRSC. The PRSC screens problem reports as either Significant Conditions Adverse to Quality (SCAQ) or Conditions Adverse to Quality (CAQ)in accordance with the criteria in Appendix A of UE2 ilR C11031, Rev. 0," Corrective Action Process.'~ SCAQ items are required by this procedure to undergo a fonnally documented root cause investigation. The PRSC then issues a Problem Report Response Sheet (PRRS). This sheet conte' .s the conunittee's recommendations and designation of a " problem owner" and the owner's Corrective Action Plan (CAP) to resolve the problem. Although the PRSC is provided with i \udit/ Incident Report as part of the initial package, the problem report alone is incorporated in <latabase and provides the basis for the committee's recommendations and ultimate conrectivt .ms. Tbc inspectors tracked the four problems in which the immediate corrective action did not appear to address the underlying cause to detennine whether the ultimate corrective action addressed the cause, in two cases, the ultimate corrective actions did not appear to address the cause as identified on the Audit /

Incident Fonn or take this into consideration in determining the corrective actions. The cause was captured in those instances when NCS became the " problem owner."

Level IV victations involve compromising a criticality control without losing double contingency. Four of the 35 violations were classified as Level Ill violations,in which double contingency is lost. The inspectors reviewed the disposition of these Level lli violations to detennine whether they were considered reportable under NRC Bulletin 9101 and whether they were investigated to determine root cause and corrective action. The inspectors verified that all Level lli violations are considered reportable under the criteria in SAR Table 6.9-1 and have been reported under Bulletin 91-01 as required. However, the inspectors noted that in only one case the Ims of criticality controls had been classified as SCAQ instead of CAQ. Although the SCAQ criteria from Appendix A of UE2 IIR-C11031 encompasses events involving loss of double contingency, these criteria are not requirements but merely PRSC guidelines. The only-incidents requiring a root cause investigation are events that require a 30-day written report to NRC (in the case of criticality safety incidents, an actual criticality accident). Tims, there is not 14

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violation of regulatory requirements, although the expectation of the NCS staffis that this type of event should normally receive a formal root cause investigation. That a loss of double contingency can occur without such an investigation means that underlying causes and potential  !

precursors could remain uncorrected and could recur.

in response to these findings, new guidance in the form of a Crew Briefing to the PRSC has resulted in classification of all Level lll violations as SCAQ items. Procedure UE2 HR-C11031, Rev.1, is under development and will formalize these changes in the SCAQ criteria. Because it is in draft fonn, this is being tracked as IFl 70 7001/97-208-09. ,

I C. Surveillances and Investigations The inspectors reviewed the investigation for Audit / Incident Report NCS INC-97 017, which received SCAQ status. This incident involved a fissile material operation that was started without an approved NCSA. The investigation determined the root cause as failure to follow procedure and two additional contributing causes. The investigation resulted in six recommended corrective actions, which adequately addressed the root and contributing causes, and were flowed into the corrective action plan. The conduct of this investigation and scope of tne corrective actions was adequate.

SAR Section 5.2.2.9 requires trending of NCS violations by building, organization / group, and the type of criticality control violated. Trending of NCS violations is implemented in accordance with CP2-EG NS1107,"NCS Trending." The inspectors examined trending records of these NCS violations and verified compliance with these requirements. The trending of these items is maintained as a quality record and was found to be adequate.

SAR Section 5.2.2.9 requires several types of surveillances and audits to ensure that NCS controls and limits are adhered to. These include internal surveillances of operations conducted by the NCS Section, annual operating orga iization surveillances, internal audits of the NCS

. program, and biennial walk throughs of plant operations, in addition, NCS conducts special field verifications of new or modified operations prior to start up. The inspectors reviewed several intemal surveillances and biennial walk-throughs, as well as the resultant corrective actions, and reviewed the implementing procedures for all types of surveillances.

The following intemal surveillances were reviewed during the inspection:

e 97 31 05 t/22/97 Enrichment Cascade Facilities Surveillance

-e 97 31-01 4/30/97 Chemical Operations Facilities Surveillance e- 97-31-12 7/31/97. NDA Program Surveillance e 97 31-06 7/31/97 C-360 Facility Surveillance  ;

e 97 31 10 7/31/97 Feed and Vaporization Facilities Surveillance ',

15 w, _ . . . . . - - - - -.. - - - - _ . _ . - . -. -

These surveillances generated five " satisfactory with negative observation" findings, eight

" negative observation" findings, and one " unsatisfactory" finding. A review of the findings revealed inconsistencies between the number enumerated in the cover memo and the enclosed

" Internal Surveillance Repart." Procedure CP2.QA-QS1031," Conduct ofInternal Surveillances," Rev. 4, defines three levels of findings (satisfactory, satisfactory with negative obsen ations, and unsatisfactory). Further review indicated that the term " negative observation" is used inconsistently in the memoranda, meaning " satisfactory with negative observations" in one surveillance and " unsatisfactory"in another.

The inspectors then examined the problem reports generated by the surveillances to determine whether corrective actions were effective and timely. Each " satisfactory with negative observation" or " negative observation" resulted in a problem report that became part of the ilusiness Prioritization System (BPS) tracking system. In most cases, the corrective action appeared to adequately address the underlying cause of the problem as identified in the surveillance and was taken in a timely manner.

The inspectors reviewed the internal procedures CP4 EG-NS1101," Evaluation of Requests for Criticality Safety Approvals," Rev.1, and CP2-QA-QS1031," Conduct ofInternal Surveillances," Rev. 4. The inspectors verified that these procedures implemented the requirements of SAR Section 5.2.2.9 for field verification of reissued NCSAs and intemal surveillances, respectively. The schedules and documentation for internal surveillances, walk-throughs, and field verification were in accordance wi6 these procedures and SAR requirements.

Ilowever, the informal schedule for performing biennial walk-throughs of fissile material operations did not explicitly schedule all satellite buildings of main process building, most of which do not contain significant quantities of fissile materials.

D, Corrective Actions The inspectors identified two problems with corrective actions in connection with surveillance findings. Surveillance 97 31-06 identified inconsistencies in procedures that cover the 2S-cylinder refeed operation in the C-360 building. The C-360 building, the Toll and Sampling Building, contains steam autoclaves which are used for heating 10-ton UF. cylinders from which samples are drawn into 2S cylinders. These cylinders are then analyzed in C-720 and any residual heels can be refed to the C-360 surge drums, which hold the material before it is transferred back into the cascade in C-337. The assay in these surge drums, and in the enishment cascade, is limited to 2.0wt% 2"U by procedure CP4-CO-CN2007,

" Operation of Surge Drums." However, the refeed operation is not limited in assay by Procedure CP4 CO-CN2051i," Returning 2S Sample Containers and C-710 Mass Spectrometry Metal Cold Traps to the C-360 Surge Drums." As a result of this surveillance, Problem Report PR-EN-97-4094 was issued which recommended revising the refeed procedure (CP4-CO-CN2051i) to limit assay to a maximum of 2.75wt% 2"U.

16

This particular example is of minimal safety concern; the surge drums were analyzed in NCSA 3972 03 to be safe to 2.75wt%. hioreover, the typical amount of material in the sample cylinders was much less than a minimum critical mass, and there was no evidence that material of greater than 2.0wt% assay had been fed to the surge drums. Ilowever, this remains of regulatory concem for several reasons. The proposed corrective action was inadequate in that it still permitted transfer of material to the surge drums in excess of the procedural limits. The proposed corrective action addressed the inconsistency between the procedures but not the NCSAs from which the assay limits were flowed down, NCSA 3972 08-01," Refeeding 2S Cylinders and hietal Cold Traps at the C-360 Sampling hianifold," and 3972-03," Operation and hiaintenance of the C-360 Surge and Relief Drums." hiost significantly, the procedure continued in use even afler the problem report was filed, and the problem report was closed out on the basis of a procedure change request, PDF C-97-02127. This is of concern because an operating procedure ,vas used even though it was known to be deficient, because corrective actions were l

i not effective, because there were deficiencies in the flowdown of criticality safety limits into l procedures, and because the problem report was closed on the basis of an action item that was not completed. The failure to take proper corrective action is VIO 70-7001/97-208-10.

In response to these inspection findings, the regulatee has shut down the 2S cylinder refeed operation and has placed a hold on procedure CP4 CO CN2051i, pending resolution of this

! issue. The regulatee has also committed to reviewing the procedures for closing out problem repons on the basis of uncompleted actions, and is reviewing old problem reports that may have been closed on this basis.

The second case concerned the oil level in a Stokes-Pennwalt seal exhaust pump in the C 337 building. Surveillance 97-31-05 dated June 11,1997, contained an unsatisfactory finding that resulted in Problem Report PR-EN-97-3047. The oil level in a particular pump was observed to exceed 4.75", in violation of a criticality safety control established at 5.5wt% assay, and was classified as a Level-IV criticality control violation. The oil reservoir was drained until the level retumed to normal on the sight glass, and the pump was restaned. No pump abnormalities were observed during the recovery, but the problem report indicated suspected mechanical failure. A CAP was instituted to determine the cause of the event and determine whether design improvements could be made to lessen the possibility of overfilling the pump; the closing date is January 15,1998.

This represents a safety concem because the pump was retumed to service without verifying that the passive overflow intended to control the level of oil functioned as intended, or taking corrective action to ensure that the problem did not recur. The regulatee indicated that the level control was a secondary criticality control and not required for double contingency. However, the regulatee did not state what controls are required for double contingency or provide a demonstration that those controls are reliable. The NCSE did provide an analysis showing that the pump could be filled with oil at 2.0wt% *U, but it relied on not exceeding 40wt% uranium load in the uranium ail mixture. Several pumps have failed at different uranium loadings, but there is no statistically significant data to demonstrate that the pumps will fail before this limit is 17

reached, or that a sudden dumping of uranium into the pump cavity is incredible. There is no immediate safety issue because the level has been restored and the assay in this portion of the cascade is limited to 2.Gwt%. If the oil level was a primary control required to meet the double contingency principle, the corrective actions taken would have been inadequate and the event would have been reportable. Because the status of the oil level control is ambiguous, this issue is being characterized as URI 70-7001/97-20811.

Conclusions The system for tracking and trending probem s ,mrts and NCS violations was found to be adequate, with the following qualifications. Communication between NCS and plant operations was found to be inadequate, resulting in several cases where the required NCS review and timeliness in filing Audit / Incident reports was in violation of established procedure. In addition, the flowdown of NCS incident assessment and recommendations into the problem reports and corrective action plans, except when NCS becomes the responsible owner, was a program weakness. These weaknesses appear to stem from a lack of communication of NCS relevant information between the varic organizations involved in problem resolution and corrective action. The actual cases examined are of minimal safety concern because the immediate violation was corrected and compliance restored.110 wever, these deficiencies are of concern in that they may prevent adequate corrective action from being taken, or corrective action being taken without being informed and advised by NCS, Furthermore,it leaves open the possibility that root causes may not be tied to corrective actions or may not otherwise be adequately addressed.

There is an also apparent lack of rigor in the characterization of surveillance findings; the ambiguity in the term " negative observation" could lead to confusion. The lack of rigor extends to flowdown of nuclear criticality controls across related processes, as exemplified by the surge drum and GEN 15 issues, and to identification of those controls relied on for double contingency as in the seal exhaust pump issue.

10.0 Criticality Alarm Monitoring Systems Scope The inspectors reviewed analyses performed to verify adequate CAAS coverage for all plant areas with the potential to be affected by a nuclear criticality accident. The inspectors walked down areas without CAAS coverage to ensure that no fissile material operations were in progress.

Observations and Findings The analyses for CAAS coverage were completed using MCNP for the Cascade and C-360 buildings. The MCNP analysis determined the radius of CAAS coverage based on the point at 18

which the dose fell below 10 mlMr. with a maximum calculational error of 5%. A line of sight method which included shielding from walls and other large structures was used for all other facilities.

Each alarm cluster contains three detector modules and a coincidence logic analyzer. Two of the three modules must alarm before the cluster activates the alarm horns. The detectors are set to alarm when a dose rate of 10 mIMr is detected, and contain a light emitting diode that simulates a background level of 4 mlWr. The detector ; dts at a dose rate of 2.5 mIMr. upon which the detector module turns on an indicator light in C-300. Thus, the detector threshold and fault setpoint are adequate to ensure that all CAAS detector modules remain functional.Some buildings only have partial CAAS coverage, as indicated in the exception request of August 15, 1996. The inspectors found all cascade cells enriched to > 1.0wt% were cover:d, but the detectors did not cover the entire operating floor of some cascade buildings. The inspectors walked down the uncovered area in C 331 and verified that it contained no fissile materials in USEC jurisdiction. Four rooms in C-710, the Laboratory Analysis building, also did not exhibit CAAS coverage. These areas were also verified to contain no fissile material. The personnel working in these uncovered areas were aware of the fact that material could not be transferred between rooms because of the incomplete CAAS coverage. The inspectors also observed one portable CAAS alami cluster in C-710, which was electronically tied into the building CAAS system such that it would alarm if another detector in the buildir g alarmed.

The inspectors observed that the NCSA GEN-08, which covers the transportation and storage of laboratory samples, indicates that CAAS is not installed in certahi areas of C-710. The request for exemption from CAAS coverage covers transportation but not storage of fissile materials.

The regulatec committed to writing a problem report and clarifying the CAAS coverage requirements in the revision to GEN-08 which is in progress.

The inspectors confirmed that each building under CAAS coverage contained its own detectors, and found no cases in which the detector in one building provided coverage for another building.

Conclusions The inspectors found the CAAS detectors adequate to provide complete coverage of all fissile material areas in those buildings where there was partial CAAS coverage, workers were aware of the uncovered areas and the work restrictions in those areas.

11.0 NCS Emergency Response Scope The inspectors reviewed the emergency response function to verify that the Emergency Plan and the associated implementing procedures are adequate to assure the public health and safety in the event of a nuclear criticality accident. The inspectors reviewed the emergency drill program, 19

1 1

54 1

I maintenance of the criticality alarm system (CAAS), and flowdown of appropriate nuclear criticality controls and limits into firefighting instructions and pre-fire plans.

Observations and Findings A. Emergency Planning The inspectors reviewed the current Emergency Plan and the associated implementing procedures to determine whether they were adequate to ensure an adequate response in the event of an accidental nuclear criticality. The inspectors reviewed the following implementing procedures from Appendix A of the Emergency Plan:

• CP2 EP EP5038, Rev. I " Criticality and Radiation Emergencies" e CP2 EP-EP5033, Rev. O " Evacuation and Take Cover (Plant and Buildings)"

• CP2 EP-EP5042, Rev. I " Termination and Recovery After Emergencies"

• CP2-EP-EP5052, Rev. I " Emergency Response Drills and Exercises"

• CP2 EP EP5044, Rev. I " Mutual Emergency Assistance"

• CP2-EP-EP5043, Rev. I " Medical Emergencies"

• CP2-EP-EP5058, Rev. 0 " Maintenance of Emergency Facilities and Equipment" The Emergency Plan Implementing Procedures (EPIPs) were found to be adequate in providing instructions for personnel evacuation, drills, assistance from off site firefighting and medical personnel, monitoring for radiation and establishing radiation boundaries, and recovery following a nuclear criticality accident, as required by ANSI /ANS-8.19. Ilowever, the criteria for re-entry following a nuclear criticality accident are not provided in CP2 EP EP5042. Listings of telephone and pager numbers for members of the Emergency Response Organization are maintained and available to the PSS, and these numbers are checked quarterly as specified in CP2 EP-EP5058. The inspectors also examined procedure CP4 EG-NSI104," Nuclear Criticality Safety Engineer Response to Emergency, Off-Normal, and Process Upset Conditions." This procedure governs the actions of NCS Engineers during various criticality-related scenarios including loss of criticality safety controls, planned expeditious handling (PEli) spacing violations, discovery oflegacy equipment, discovery of UO 2F 2deposits, UF. releases, and cell shutdown. The procedure includes lists of technical information that must be gathered and pertinent questions for each scenario. In general, the EPIPs and procedure goveming the actions of the NCS staff during abnormal and accident conditions were adequate.

The inspectors examined the recoids of evacuation drills and exercises, which are performed in accordance with a predetermined schedule. Plant-wide evacuation drills are conducted annually as part of the " table-top" drill involving the on-site Emergency Response Organization (ERO). Semiannual full-scale drills involving both on site and off site responders are conducted in which local police, firefighting, and medical personnel are involved. Upon completion of the drills, the responders are critiqued and regulatory or safety findings are fed into the problem 20 l

reporting system for tracking and resolution. No deficiencies were identified as a result of the review of the drill system.

The inspectors walked down the X 333 building to examine the evacuation routes, exit signs, an .

emergency lighting systems. There are illuminated exit signs along the exterior walls and emergency lights scattered throughout the operating and cell doors. The evacuation routes are marked with yellow lines on the finors. The emergency lighting is checked annually and there is a monthly power failure check to ensure activation of the lighting circuits and the emergency diesel generators. The inspectors examined KY/D-4655, Rev. 0,"PGDP/Paducah DOE Reservation Cascade Facilities Emergency Action Plan for C-331, C 333, C-335, C-337." This Emergency Action Plan (EAP) contains evacuation maps for both the operating and cell floors for the cascade buildings, with routes radially outward to the nearest exit on the operating floor and across to the exterior stairwells on the cell floor. Assembly points for all buildings containing fissile materials are also prc vided. Evacuation routes and assembly points were found to be reasonably direct and appropriately located, However, the evacuation routes as drawn ran along the exterior walls of the buildings rather than directly away from the exit. While the walls may offer some protection in the event of a fire or chemical release, this could have individuals running along the wall next to an ongoing criticality event and receiving unnecessary radiation doses. The EAP was found to be concise and appropriate for use in the event of a criticality emergency.

B. Pre Fire Plans The inspectors reviewed pre-fire plans for buildings containing fissile material to detennine whether nuclear criticality restrictions on firefighting activities were included. The regulatee had issued a pre-fire plan supplement dated April 4,1997, which covered the C-310, C-331, C-333, C-335, C 337, C-360, C-400, C-409, C-710, C-720, and C-746-Q1 buildings. This supplement addresses the use of water as a possible moderating agent and as a means of changing the spacing and/or configuration of storage array elements. The instructions limit the use of tight sprays in fissile control areas (FCAs), as well as the use of water around equipment containing deposits and breached cylinders. The inspectors reviewed EN C 832-96-007, which is part of the supplement and identifies the location of all FCAs. The inspectors reviewed the Criticality Safety Guidelines contained in the supplement as well as the plant coverage and found them to be adequate, with the exception that the pre fire plans did not cover the cylinder storage yards, flowever, Procedure CP2 EP-EP5031," Oil and llazardous Materials Spills and Releases,"

contains specific instructions implementing a restriction on the use of water on a breached UF.

cylinder because of nuclear criticality safety concerns. The procedure was found to adequately address the cylinder yards, although it was not clear how these instructions would be communicated to firefighters on the scene in the absence of a pre fire plan. The incorporation of criticality safety guidelines into the pre-fire plans closes VIO 97-201-03, but the communication of the criticality concerns in the cylinder yards to firefighters is being tracked as IFI 70-7001/97-20812. The regulatee indicated during the inspection that a pre-fire plan would be developed.

21

., .. l 3

'lhe inspectors examined the pre Grc plans in the Emergency Operations Center (EOC), the C 300 building, and venfled that the criticality safety information had been incorporated into those plans. The inspectors also verified that Fire Services personnel had been trained in the criticality safety guidelines as part of their annual llazMat Refresher training.

Conclusions The current Emergency Plan and EPIPs were adequate in protecting the health and safety of the public in the event of a nuclear criticality accident, and in accordance with ANSI /ANS 8.19. The criticality safety guidelines included in the pre-fire plans and in the hazardous material spill procedure were found to be adequate, although communication of these guidelines in the cylinder yards remains in question. The mass of the material in cylinders means that controlling the amount of moderator is a rignificant nuclear criticality concern. There is fire loading in the cylinder storage yards, such as straddle carrier fuel and grass, as well as the potential for rupturing a heated cylinder through thennal expansion. No other safety significant findings were identified.

12.0 Resolution of Previous Findings A list ofitems opened, closed, or discussed is included after the body of this report. Newly opened items are discussed in the applicable report section.

IFl 97 201-01 concerned inadequate controls to assure double contingency in the cylinder wash operation. Corrective actions are underway to review and revise the applicable NCSA/NCSE, and so this item remains open.

VIO 97 201-02 concerned failure to evaluate a NAM storage array for nuclear criticality safety.

Corrective actions identified in correspondence between PGDP and NRC have been determined to be adequate. The response to the Notice of Violation stato: "The reason for the violation is the failure to provide sufficient detailin KY/S 249 to support the assertion that the NAM / Fixed HEPA filter storage array was bounded by the analysis in KY/S-253." The NCSA GEN-09 has been modified to add postings in the used NAM filter accumulation area; the NCSE KY/S 249 has been augmented by adding several new accident scenarios and the analysis now encompasses used filter storage. This item remains open because the upgrade goes beyond an assertion of insufficient documentation, and the response was inadequate in addressing the true nature of the noncompliance.

IFl 97 201-04 concemed the inclusion of the NCS Section in the review process for all criticality related items. Procedure changes are underway to assure NCS participation in these "ters. Procedure CP3-EG-EG1070 is under development to mandate inclusion of NCS in the t- oge Control Board (CCB) and Procedure UE2-TO EG1031 is under revision to ensure that NG concems are addressed, and thus this item remains open.

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. _. _ _ - _ - _ - ~ _ - _ _ _ _ _ _ _ _ _ - - _ . . ____ --_---_ _____ __ __

IFl 97-201-06 concerned a baseline verification and review of the ilDhi by Configuration hianagement, but the regulatee disagreed that this baseline assessment was required. Thus, this item remains open.

URI 97 201-05 concerned the Configuration hianagement (Chi) self assessments procedure.

Ilecause this procedure is being revised to incorporate specific requirements for Chi program assessments, this Unresolved item is being closed and tracked as IFI 70-7001/97 208-02, 13.0 Exit hicsling An exit meeting was held between NRC Headquarters and Resident inspector staff and plant management on November 7,1997 and daily throughout the inspection. No classified or proprietary information was discussed.

a 23

, - - - - , - , , -c

=* . , , -

ITEMS OPENED. CLOSED. OR DISCUSSED ltems Opened DEV 97-208 01 Change to a commitment one day before the due date.

IFl 97 208 02 Incorporate specific CM self assessment requirements.

VIO 97 208-03 Inadequate Validation.

URI 97-208-04 Adequacy of NCSE 207.

IFl 97-208 05 P&E Air NCSE Revision.

IFl 97 208 06 Air Capture System NCSA.

IFI 97 208 07 Revision of NCSA GEN IS VIO 97-208-08 Failure to make 24 hour2.777778e-4 days <br />0.00667 hours <br />3.968254e-5 weeks <br />9.132e-6 months <br /> NCS Notification.

IFI 97 208 09 Classification of NCS violations.

VIO 97 208 10 Failure to take Corrective Action.

URI 97 208 11 Ambiguous oil level status.

IFl 97 208 12 Fire fighting instructions for cylinder yards.

1 Items Closed VIO 97 201 03 Failure to incorporate firefighting restrictions into procedures / training.

URI 97-201 05 CM program assessments. Reopened as IFI 70 7001/97 208-02.

IFI 97 201 07 PCR reviews.

IFl- 97-201 08 Potential changes to Compliance Plan.

Items Discussed VIO 97-201-02 Inadequate analysis of NAM storage array.

IFI 97-201-01 Revision of NCSA/E for cylinder washing.

IFl 97-201-04 inclusion of NCS in review process.

IFl 97-201-06 . Verification and review of BDM by CM.

24

PARTIAL LIST OF PERSONS CONTACTED

, lheed Manin Utility Servi. css l

Keith J. Ahern, Systems Engineering W. D. llattimore, Nuclear Criticality Safety JefTD. Fletcher, Production Support J. D. Sohl, Manager. Nuclear Safety W. E. Sykes, Nuclear Regulatory Affairs D. C. Stadler, Nuclear Regulatory AITairs J. C. Dean, Nuclear Criticality Safety USEC J. A. Labarraque floward Pulley NRC 1

John Jacobson, Resident inspector Kenneth G. O'Ilrien, Senior Resident inspector 25

ACRONYMS USED AQ-NCS Augmented Quality - Nuclear Criticality Safety llDM Boundary Definition Manual llPS Ilusiness Prioritization System CAAS Criticality Accident Alann System CAP Corrective Action Plan CAQ Condition Adverse to Quality CM Configuration Management DOE Depanment of Energy EAP Emergency Action Plan EOC Emergency Operation Center EPIP Emergency Plan implementing Procedure ERO Emergency Response Organization FCA Fissile Control Area llEPA '

liigh Efliciency Particulate Air IFl Inspector Follow up Item MCNP Monte Carlo Neutron and Photon (a computer code)

NAM Negative Air MacM :

NCS Nuclear Criticality Safety NCSA Nuclear Criticality Safety Approval NCSE Nuclear Criticality Safety Evaluation NRC Nuclear Regulatory Commission P&E Purge and Evacuation PEli Planned Expeditious liandling PGDP Paducah Gaseous Diffusion Plant PRRS Problem Repon Response Sheet PRSC Problem Report Screening Committee PSS Plant Shift Supervisor RCW Recirculating Cooling Water SAR Safety Analysis Report SCAQ Significant Condition Adverse to Quality USL Upper Safety Limit 26 l

.