ML20196J041
| ML20196J041 | |
| Person / Time | |
|---|---|
| Site: | Portsmouth Gaseous Diffusion Plant |
| Issue date: | 07/28/1997 |
| From: | Allen D UNITED STATES ENRICHMENT CORP. (USEC) |
| To: | NRC OFFICE OF INFORMATION RESOURCES MANAGEMENT (IRM) |
| References | |
| 70-7002-97-203, GDP-97-2015, NUDOCS 9708010278 | |
| Download: ML20196J041 (16) | |
Text
{{#Wiki_filter:. _ _ _ _ _ _ _ _ _ _ _ _, E United States Enrichment Corporation 2 Democracy Center 6903 Rockledge Drive Bethesda, MD 20817 Tel. (304 564-3200 l'nited hes Fad ns64-3201 Enrichenent Corimnition July 28,1997 GDP-97-2015 United States Nuclear Regulatory Commission Attention: Document Control Desk Washington, D.C. 20:55 - 4'ar49mceth Gawous D: Guitar Phmt (PEdiTS) - e
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Docket No. 70-7002 Response to Inspection Report (IR) 70-7002/97-203 Notice of Violation (NOV) Nuclear Regulatory Commission (NRC) letter dated June 27,1997, transmitted the subject Inspection Report (lR) that contained three violations involving: 1) failure to comply with Nuclear Criticality Safety Approval (NCSA) administrative controls,2) failure to properly validate NCS computer code calculations, and 3) failure to correctly identify NCS-required system boundaries. USEC's response to these violations is provided in Enclosures I through 3, respectively. lists the commitments made in this report. Unless specifically noted, the corrective actions specified in each enclosure apply solely to PORTS. As indicated in our response, USEC does not agree with NOV 97-203-03 concerning the validation of certain NCS calculations. The basis for our denial is provided in Enclosure 2 In the cover letter transmitting the subject IR, NRC requested that USEC provide our " evaluation of the findings of two external audits, the planned corrective actions, interim compensatory measures, and the schedule for implementation." As was documented in our letter to NRC dated June 16,1997, which responded to PGDP Inspection Report 70-7001/97-201, USEC briefed the NRC on the General Manager Independent Assessment at a meeting on May 2,1997, at NRC Region III. USEC will also be providing further information on the results of the independent assessment prior to restart of the Paducah C-400 cylinder wash and spray booth operations as required by the February 28,1997, Confirmatory Action Letter issued by NRC. If necessary, USEC would be willing to brief the NRC further on the results of the independent assessments conducted M at both sies. pT- \\\\ 9708010278 970728 PDR t ADOCK 07007002 PDR ( Off ces in Paducah, Kentucky Por tsmouth. Ohio Washington. UC i
~ l F , United States Nuclear Regulatory Commission July 28,1997 Page Two Ifyou have Ony questions regarding this submittal, please contact Ron Gaston at (614) 897-2710. . Sincerely, Dale Allen i General Manager l Portsmouth Gaseous Diffusion Plant 4..... :....Enclomres. -.......... -.....'.-. --.*...---.>.--1 cc: Regional Administrator, Region III NRC Resident Inspector-PORTS j l i l i I i
7 't i: l L , United States Nuclear Regulatory Commission l July 28,1997 ~ Page Three l Distribution-Robert L. Woolley - 001 bec: J. Adams, HQ J. Adkins, HQ j J. Dietrich, LMUS { L. Fink, PORTS l R. Gaston, PORTS i J. Labarraque, PGDP i' A. Rebuck-Main, HQ j...... ..S* Routh, Hg,,,,,,,,,. .q D. Scott, fiQ D. Silverman, ML&B J. Slider, HQ B. Sykes, PGDP ) l R. Wells, HQ i f:\\ rob \\ ports \\nov97203. pts L Concurrence line CRS / /97 i I ) l I I f
~ 't UNITED STATES ENRICIIMENT CORPORATION (USEC) REPLY TO NOTICE OF VIOLATION (NOV) 70-7002/97-203-05 Restatement of Violation TSR Section 3.11.2 requires, in part, that "All operations involving uranium enriched to 1.0 wt.% or higher U-235 and 15 g or more of U-235 shall be... performed in accordance with a nuclear cr.:ticality safety approval [NCSA]." NCSA PLANT-045, item 7, requires that "Any limi*.ed safe volume containers that are used to store (either permanently or temporarily) uranium-bearing materials shall be labeled as to their contents and enrichment." ... NCSA-pfgT-018.A0}, ite,m,5, re,o,ui, rey thayy)AW [ Dry Activg,Wayte] cpnta,ipers,ang gags,,,,,,. j shall be spaced at least two feet edge-to-edge from uranium-bearing equipment or other types of uranium-bearing material." NCSA-PLANT-018.A01, item 17, requires that "B-25 boxes containing DAW material shall be spaced at least two feet edge-to-edge from uranium-bearing equipment or other uranium bearing material." Contrary to the above A. On April 30,1997, a limited safe volume container that was used to store uranium-bearing materials was not labeled as to the contents and enrichment. Specifically, a one-gallon bucket containing autoclave valve internals contaminated with uranium enriched to 1.0 wt.% or higher of U-235 was observed by NRC inspectors on the floor int the X-344 facility near the autoclaves without an identification label on the bucket showing the enrichment. B. On May 1,1997, DAW containers and bags were spaced less than two feet edge-to-edge from other types of uranium-bearing material. Specifically, a plastic bag containing wet / oily / sludgy material and visible liquid was observed by NRC inspectors to be stored directly adjacent to a DAW drum on the second floor of building X-330 which is a process building where operations involve uranium enriched ) to 1.0 wt % or higher U-235. C. On May 13,1997, a B-25 box containing DAW material was spaced less than two feet edge-to-edge from other uranium-bearing material. Specifically, a drum containing other uranium-bearing waste material was observed by NRC inspectors, near building 6619, in temporary storage less than two feet from a B-25 Box containing DAW. El-1
I I. Reasons for Violation Many of the NCSAs have been recently implemented and have in many cases established either new NCS requirements or modified / expanded NCS requirements. In the cases cited in the inspection.eport, the labeling requirement in the X-344 process building was different than past practice (example A) and the requirements for the X-6619 facility were "new" (example C). In the case of ex ample B, while the NCSA governing the collection and storage of DAW was new, the spacing requirement for the facility is not new. The reason for example B was that corrective actions for previously self-identified deficiencies were less than adequate.
Background:
PORTS has been experiencing NCS field compliance problems in this area throughout the period of transition from the "old" NCS program requirements and documentation to the a b"'NCS pr5i; rain Utilizing the NCS'A's Ed oth'er prEgradu$1atic' cEadges/r'e'q'uireniehts aI ' " " ^ ^ ~ --oc-- n described in Section 5.2 of the SAR. Many of the NCSAs were activated within the last six months and plant personnel are gaining experience in implementing the NCSA specific requirements. The plant self-assessments and surveillances have been identifying similar problems with field compliance and corrective actions are being implemented; these typically include both the immediate correction of the NCSA noncompliance and identification of programmatic corrective actions. Based on the assessments and inspections that have been performed since transition to NRC regulations, the plant has initiated a Quality of Operations Plan (QOOP) initiative that is directed toward programmatic improvements to correct what are perceived as " root cause" issues. As new challenges are identified via the problem l reporting system, deficiencies are evaluated against the QOOP initiatives to determine whether additional programmatic actions are needed. II. Corrective Actions Taken and Results Achieved l l 1.) On April 30,1997, the improperly labeled one-gallon can containing valve internals found in building X-344 building was properly labeled as less than 5 % U-235. (Example A) l 2.) On May 1,1997, the improperly spaced DAW container and bags in X-330 building i were moved to provide proper spacing. (Example B) 3.) On May 13,1997, the improperly spaced B-25 box near building X-6619 was properly spaced. (Example C) El-2
4.) NCS has increased the frequency of monitoring operations via the NCSA monthly walkdown program. The actions to assess the results of the walkdown program are discussed in section 111. 5.) On July 24,1997, an article was published in the Open Line to communicate to the general plant population the proper actions to take if a violation of an NCSA is identified. III. Corrective Steps to be Taken 1.) PORTS is using the information collected from the NSCA monthly walkdown program to trend NCS non-conformances. As data becomes available, the information will be evaluated for negative trends to determine what programmatic changes are needed. This may include additional training in the most commonly violated requirements (i.e., spacing, labeling); engineering controls to replace administrative .... l "E..ls g'sical spacers, segpg,ated lo,ck,ab,leytoragareapi re,visi,on of thq N,CS,,,,, a admimstrative requirements to establish accountability of waste contamers. By September 12,1997, the results of the NCS evaluation and recommendations will be presented to the Management Analysis and Assessment Team (MAAT) for approval. 2.) By September 12,1997 Portsmouth will conduct shift briefings with Lockheed Martin Utility Services (LMUS) personnel responsible for the spacing, labeling, and movement of uranium bearing material covered by NCSAs. The shift briefings will be conducted within the work areas to discuss the specifics of this violation, building specific trending information, corrective actions, and the importance of maintaining a questioning attitude. 3.) USEC will develop and distribute an NCS bulletin to communicate to the general plant population important NCS issues (i.e., spacing, labeling, etc...). This action will be completed by August 29,1997 IV. Date of Full Compliance Full compliance was achieved on May 13,1997, when the improperly spaced B-25 box near building X-6619 was properly spaced (The date selected for establishing compliance is the later of the three examples). Programmatic corrective actions designed to prevent recurrence will be evaluated and approved by the MAAT by September 12,1997. El-3
-l UNITED STATES ENRICliMENT CORPORATION (USEC) REPLY TO NOTICE OF VIOLATION (NOV) 70-7002/97-203-03 Restatement of Violation TSR Section 3.11.1 requires, in part, that "A Criticality Safety Program shall be established, implemented, and maintained as described in the Safety Analysis Report... " SAR Section 5.2.3.2. requires, in part, that "When.NCS is based on computer code calculations of K,y, controls and lirnits are established to ensure that the maximum K y e complies with the applicable code validation for that type of system being evaluated." Contrary to the above, as of May 30,1997, three Nuclear Criticality Safety calculations, NCS-CALC-97-009, NCS-CALC-97-010, and NCS-CALC-97-012, were observed to be based upon
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report. USEC's Response PORTS believes the SCALE Code is adequately validated for the referenced NCS calculations. The range of enrichments and materials examined in the SCALE Code validation is limited by the number of critical experiments which have beea performed, and the availability of data. The benchmark critical experiments used to valida.e the SCALE Code at PORTS can be generally summarized as covering low enrichments ranging between 1.4% to 5.0%, and high enrichments ranging between 92.5% to 97.7%. Included in the validation are cases with 2"U-metal, UO F, UO (NO )2, U 0s, and UF, which are moderated by materials 22 2 3 3 4 l such as water, sterotex, and paraftin. In the cited violation, NRC quotes PORTS SAR section 5.2.3.2 as the basis for this violation. It is noteworthy, that SAR section 5.2.3.2 also states: " Computer codes are validated using experimental data from benchmark experiments which, ideally, have geometries and material compositions similar to the systems being modeled." " Validation of the computer code detennines its calculational bias or uncertainty as well as the effective margin of suberiticality. The PORTS i validation involves the modeling of benchmark critical experiments over a range of applicability. Because the K,g value of a critical experiment is essentially 1.0, the bias of the code is taken to be the j deviation of the calculated values of K,g from unity. Statistical d E2-1 4 ~
analysis is employed to estimate the calculational bias, which includes the uncertainty in the bias and uncertainties due to extensions of the area of applicability, as well as the effective margin of suberiticality. Uncertainty in the bias is a measure of both the precision of the calculations and the accuracy of the experimental data. The validation of the computer code specifically defines the maximum acceptable K,g used to determine subcriticality."' USEC believes that the NCS calculations referred to in the cited violation are validated, considering the above quoted sections of the SAR, based on the following justification: The SCALE computer code (as used at PORTS) uses a 27 energy group cross section library, which is based on ENDF/B-IV nuclear data. In other words, for each cross section type (fission, absorption, etc.), and each nuclide type (2"U,238U, etc.), there are 27 energy groups representing the microscopic cross section as a function of neutron energy. The validation of the computer code determines its calculational bias or uncertainty as well as the effective " mar'jjin ' ~f sIib'critiLility* TM ialidation also involves the'modEling oflieilchhiark critical " " ~ * * *
- o experiments over a range of applicability. The validation of the computer code specifically defines the maximum acceptable K nused to determine suberiticality.
e A large range of moderation and reficction conditions were included in the benchmark cases used in the validation. By varying the amount of moderation and reflection, the energy spectrum of the neutron flux in a critical system can be shifted more towards the thermal or fast energy groups. A convenient indicator of whether or not a particular benchmark case is dominated by thermal or fast energy group cross sections is the Average Energy Group (AEG) causing fission. Figure C-1 [ attached] from the SCALE validation report (POEF-LMUS-13)2 shows the K,y results from each benchmark case plotted versus AEG. Figures C-2 and C-3 [ attached] show the same information, broken down into low and high enriched cases, respectively. The figures referenced above clearly demonstrate that the PORTS validation test cases have thoroughly tested the SCALE cross section library over a wide range of neutron energy spectrums, for both low and high enriched cases. Based on the good agreement between the calculated results (see Figure C-1) for the critical experiments (measured K,g= 1.0), and the wide range of neutron energy spectrums examined, the uranium microscopic cross section data contained in the 27 energy group library is valid through the entire enrichment range. Therefore, we conclude, based on the low enriched and high enriched results, that the U microscopic cross sections are valid over a wide energy 23 range. 'See page 5.2-15 of the PGRTS SAR, " Computer Calculations," second and third paragraphs. l l 2The upper safety limit for PORTS calculations is less than or equal to 0.9605 established by the most l conservative of the Validation Reports used at PORTS (POEF-T-3636, Rev i and POEF-LMUS-13) E2-2 l
in order for the relative accuracy / bias of a particular SCALE calculation to be affected by the enrichment of the fissile material, the 23 U or 235U microscopic cross sections in the energy groups where the most neutrons exist would have to be different. Such a problem would have to manifest itselfin the results of either the hi@ or low enriched validation test cases, resulting from the large range of AEG investigated during the validation. Since the validatica test cases demonstrate that no such deviations in the 238U or 235U microscopic cross sections exist, the use of SCALE 4.3 to perform a calculation for enriched material between 5% anJ 92.5% does not constitute performing a calculation outside the range of applicability of the validation. The quality check occurs when the AEG of the results are verified to be within the range analyzed in the validation. Calculations identified to date have not identified any neutron energy spectrums which are outside the analyzed range of the validation test cases. PORTS NCS Engineers check to j ensure that the AEG is within the analyzed range of the validation for each case (XP4-EG-NSl 100;" Nuclear CriticalitySafety Calculations"). If the AEG of a case is outside this range, the results are considered suspect, rejected, and the condition investigated. Also, no conclusions are ever based on just one SCALE calculation. Usuall'y~, calculations are performed over a range of moderation and reflection conditions, similar to the validation test cases. This allows both the engineer and peer reviewer to identify any questionable or suspect results by examining trends in K-effective. Based on the above discussion, PORTS has concluded that the use the SCALE code for enriched material between 5% and 92.5% and the NCS calculations are validated in accordance with SAR Section 5.2.3.2. l l I i k E2-3
d Attachinent to Enclosure 2 UNITED STATES ENRICIIMENT CORPORATION (USEC) REPLY TO NOTICE OF VIOLATION (NOV) 70-7602/97-203-03 Figures C-I, C-2, and C-3: Validation of the CSAS25 Calculational Sequence ir. SCALE-4.3 and the 27 Energy Group ENDF/B-IV Cross Sections of the Portsenouth Gascous Diffusion Plant Nuclear Criticality Safety Section IBM RS/6000 Workstation 1 l l I l E2-4 'l
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-~^ 1 t, UNITED STATES ENRICHMENT CORPORATION (USEC) REPLY TO NOTICE OF VIOLATION (NOV) 70-7002/970-203-08 Restatement of Violation Title 10 Code of Federal Regulations (CFR) 76.93, " Quality Assurance," requires that the Corporation shall establish, maintain, and execute a quality assurance program satisfying each of the applicable requirements of ASME NQA-1-1989, " Quality Assurance Program Requirements for Nuclear Facilities. " ASME NQA-1-1989, " Quality Assurance Program Requirements for Nuclear Facilities," Basic Requirement 8, " Identification and Control of Items," states that controls shall be established to assure that only correct and accepted items are used or installed. -Pottsinduth Q&atfly dsn*rtince Frc? grani, App *elidiPA','S'ectiofl.8, iX$NuEleFC'riticality' ~ *
- Safety Identification and Control of Items," states that ASME NQA-1-1989 Basic Requirement 8 shall be applied to those SSCs identified in NCSAs/NCSEs required to meet the double contingency principle.
l Contrary to the above, ASME NQA-1-1989 Basic Requirement 8 was not applied to SSCs l identified in NCSA-PLANT-028 and NCSA-330-013, as required to meet the double l contingency principle. Specifically, as of May 30, 1997, the metal covers plates and associated systems, which are designed to prevent the introduction of moderator into PEH equipment, were not identified as AQ-NCS. L Reasons for Violation USEC agrees that ASME NQA-1-1989 Basic Requirement 8 was not appropriately applied to SSCs identified in NCSA-PLANT-028 and NCSA-330-013, as required to meet the double contingency principle. The reason for the violation was a lack of knowledge and experience l in applying the criteria for determining AQ-NCS SSCs.
Background:
The designation of SSCs as AQ-NCS was implemented as part of the certification application. l The SAR and the Configuration Program Manual describe the process and criteria for classifying SSCs; however, in this case, instead of classifying the installed SSCs as AQ-NCS, the SSCs were considered administrative control items needed to maintain a buffer as indicated by the NCSA. E3-1 1 1
l* l i 's' l Double contingency for moderation control for handling PEH equipment is achieved by I maintaining a positive buffer on an identified deposit. This positive buffer is achieved by installing a metal cover over equipment c.penings (TSR 2.2.3.16), applying a nitrogen or dry air bufTer and periodically monitoring the nitrogen pressure. Because the requirement to apply l the cover and monitor the pressure were believed to be administrative controls (i.e., TSR i i requirements) and there were no special quality or design requirements beyond " standard I commercial grade material", the Quality Boundary Evaluator / Engineer concluded that the l cover and associated equipment should not be classified as AQ-NCS. After the NRC Inspector raised the question about the classification of the SSCs, the above referenced NCSAs and the requirements of the Configuration Program Manual were l reevaluated to determine if the SSCs in question were misclassified. This evaluation l determined that the cover plates were AQ, similar to the cascade quality boundary classification,'vhile the relief valve and purge gas fittings should be classified as AQ-NCS since they provide controls for maintaining the buffer. ,e,,,, +.... -,,... ...,...,~.... .n.....,,,.. a H. Corrective <'.ctions Taken and Results Achieved 1.) Using procedure XP3-EG-EG1037;" Establishing & Controlling Quality Boundaries," and considering the uniqueness of the PEH handling process, on June 6,1997, work began to reclassify PEH handing and storage equipment. The SSCs used for PEH handling and storage were reclassified on July 23,1997 as AQ-NCS. This action l implemented ASME NQA-1-1989 Basic Requirement 8 to SSCs identified in NCSA-PLANT-028 and NCSA-330-013, as required to meet the double contingency l principle. 2.) On July 25,1997 NCS and Configuration Management engineers were briefed _ on the l reasons for this violation and on the newly implemented classification for.PEH handling and storage equipment for the process buildings. Hl. Corrective Steps to be Taken By August 29,1997 Portsmouth will conduct a lessons leamed session with Quality Boundary Evaluators / Engineers and NCS engineers to review procedure XP3-EG-EG1037;" Establishing & Controlling Quality Boundaries." This action will improve the knowledge base and experience level of personnel establishing quality boundaries and classifying equipment. IV. Date of Full Compliance Full compliance was achieved on July 23,1997, when associated equipment vrad to ensure moderation control of PEH equipment was classified as AQ-NCS. The corrective actions to prevent recurrence with this specific violation were completed on July 25,1997 when NCS engineers and CM engineers were briefed on this violation and the newly implemented classification for PEH handling and storage equipment. The remaining corrective action will be completed by August 29,1997. E3-2
4' 'o UNITED STATES ENRICHMENT CORPORATION (USEC) LIST OF COMMITMENTS 70-7002/97-203 NOV 97-203-05 s 1.) PORTS is using the information collected from the NSCA monthly ~walkdown program to trend NCS non-conformances. As data becomes available, the information will be evaluated for negative trends to determine what programmatic changes are needed. Some potential reconunendations may include additional training in the most commonly violated requirements (i.e., spacing, labeling); evaluating engineered i controls to replace administrative controls (physical spacers, segregated lockable storage areas); revision of the NCS administrative requirements to establish accountability of waste containers. By September 12,1997, the results of the NCS l evaluation and recommendatians will be p~ resented to the Management A.ialysjs and ~ ^^
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l 2.) By September 12,1997 Portsmouth will conduct a lessons learned shift briefing with Enrichment Plant personnel to discuss the specifics of this violation, building specific l trending information, corrective actions, and tl a importance of maintaining a questioning attitude. 3.) USEC will develop and distribute an NCS bulletin to communicate to the general plant population important NCS issues (i.e., spacing, labeling, etc...). This action will be completed by August 29,1997 I NOV 97-203-08 1.) By August 29,1997 Portsmouth will conduct a lessons learned session with Quality Boundary Evahtators/ Engineers and NCS engineers to review procedure XP3-EG-EG1037;" Establishing & Controlling Quality Boundaries." This action will improve the knowledge base and experience level of personnel establishing quality boundaries and classifying equipment. i l 1 i a i E4-1 ?I 4 i -. - -}}