ML20203H691

Summary of 971208 Meeting W/Licensee & NRC in Rockville,Md Re Scale Computer Code Validation.List of Attendees & Presentation Matl,Encl

ML20203H691
Person / Time
Site:	Portsmouth Gaseous Diffusion Plant, 07007001
Issue date:	12/16/1997
From:	Yawar Faraz NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
To:	Pierson R NRC OFFICE OF NUCLEAR MATERIAL SAFETY & SAFEGUARDS (NMSS)
References
70-7002-97-203, NUDOCS 9712190025
Download: ML20203H691 (9)

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• ....,/ December 16, 1997 MEMORANDUM TO: Robert Pierson, Chief Special Projects Branch Division of Fuel Cycle Safety and Safeguards Office of Nuclear Material Safety and Safpuards THROUGH: Dan Martin, Section Chief ItfA f

Enrichment Section Special Projects Branch Division of Fuel Cycle Safety and Safeguards Office of Nuclear Material Safety and Safeguards FROM: Yawar Faraz, Project Manager M Enrit:hment Section -

Special Projects Branch Division of Fuel Cycle Safety and Safeguards Office of Nuclear Material Safety and Safeguards

SUBJECT:

MEETING WITH USEC ON SCALE COMPUTER CODE VALIDATION Members of SPB, FCOB and FCLB met with USEC representatives from USEC Headquarters ar.d Portsmouth Gaseous Diffusion Plant (PORTS) at NRC headquarters on December 8, 1997. Two members of the public were also present. Attachment 1 is a list of attendeas.

NRC had suggested this technical meeting in a letter dated October 28,1997, after it found USEC's response denying NRC's violation documented in Notice of Violation 70 7002/97-203 03, to be inadequate.

USEC distributed Attachment 2 at the beginning of the meeting. USEC acknowledged the violation by stating that benchmark data was not used to validate SCALE code calculatior.s in the intermediate enrichment range of 5 to 20 percent. However, USEC stated that they had obtained data from several Russian experiments in the 7 to 30 percent range. As part of their corrective actions, USEC committed to preparing, and providing to the NRC by January 14.1997, a detailed action plan for preparing a validation report which would account for the additional experimental data in the intermediate enrichment range. NRC staff concluded that the issue did not constitute an immodiate safety concern and that no new compensatory measures were needed to be in piace prior to the submittal of the validation report action plan, pc5 a 9712190025 971216-PDR C

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Responding to NRC's request made during the nieeting, USEC provided a copy of a memorandum dated November 21,1997, from W. C. Jordon of ORNL to D. M. D'Aquilla of '

PORTS that deals with cross section validation for intermediate enrichments (Attachment 3),

Questions, comments or suggestions with regard to this procedure should be forwarded directly to Yawar Faraz (415 B113/e mail YHF).

Attachments: As stated cc w/ attachment: S, Tveile, USEC R. DeVault, DOE DISTRIBUTION:

NRC File Center PUBLIC Docket No. 70 7001 & 70-7002 -

FSPB r/f FCSS r/f PHiland, Rlli RPierson ETen Eyck CTripp, FCOB DHaitland, Rlli 2WBrach WSchwink, FCLB PTing, FCOB WTroskoski, FCOB JDavis, FCOB DDamon,FCLB KWinsberg, OGC TCombs SGagner a:scalemtg.mmo OFC SPB E hPB FCOB SPB l SPB NAME YFaraz i h adley DMoreWeb Martin DATE 12./ 9 /97 V/k/ ~ ld- /97 l'4./) 1 /97 ik / I /97 / /97 C = COVER E = COVER & ENCLOSURE N = NO COPY OFFICIAL RECORD COPY

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USEC A Global Energy Company k .

UNITED STATES ENRICHMENT CORPORATION .

PORTSMOUTH GASEOUS DIFFUSION PLANT SCALE CODE PRESENTATION TO THE NRC December 8,1997 Attachmenc 2-

BACKGROUND NOV 97-203-03 cited a violation regarding the failure to properly validate the NCS computer code calculations in the intermediate enrichments range.

The code has been validated with benchmark data up to 5% and from 92.5% to 100%. Benchmark data included 235U and 238U throughout a wide range of moderation levels.

- USEC acknowledges that benchmark data was not used to validate SCALE code calculations in the intermediate enrichments range.

- At the time of the validation report, USEC was not aware of any benchmark data available in the intermediat enrichments range to validate the code.

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USEC A Global Energy Company

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CORRECTIVE ACTIONS ,

j USEC has assembled benchmark data applicable to the intermediate enrichments range (7-30% assay).

The code will be run using selected benchmark data to determire the impact '

on the bias of the code. The results of the impact of the bias will be utilized to determine any impacts on the current analyses.

- The new information from the benchmark data code run will be utilized to ,

l update the validation report.

. Completion dates for the above corrective actions will be provided to NRC by January 14,1998, as part of our supplemental response to NOV 97-203-

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USEC A Geotml Energy Company 2

SAFETY ASSESSMENT .

. USEC has reviewed the preliminary results of the SCALE code validation utilizing intermediate enrichment data.

. Preliminary results demonstrate that the code would perform as expected in the intermediate enrichment range.

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. Therefore, USEC believes that the results of current NCS calculations are still valid and that no compensatory measures are required.

. Thus, assurance exists regarding the safety of operation / controls which' utilize calculations that are based on intermediate range enrichments.

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MUSECA Global Energy Company

. *. A f CAN RIDGE NATIONAL. LABORATORY PMcam:tanisnans inminess er Lee =Ime naamru essmov nessnacH componAtion Pall *(48Bf SPedf18 PeRTIS u.8.sWWITMWff er WIWhof *wissueT:w eemLew YYmneouNaesuren Ta"sm"tIIE"ene Deta: - November 21,1997 To: D. M. D'Aq,dla From: W. C. Jordan W4

Subject:

Cross Section Validation ofIntermediate Enriched Systems Per our discuuion, the purpose of this memo is to review some of the wa 41ons that haw been used when evaluating intermediate enriched uranium systems. Also presented are some considerations for extending the existing validations for highly enriched and low enrichment systems to include some intermediate enriched systems in general, the " safe" parameters presented in K 1019. " Criticality Data and Nuclear Safety Guide Applicable to the Oak Ridge Gaseous Diffbsion Plant" wem used for the evaluation of uranium systens encountered in the operation at the ORODP Safe geometry (spherical, cylindrical, and slab), safe volume, and safe mass are tabulated over the full range of enrichment. GAT 225 is the equivalent document that is used at the Portsmouth gaseous diffusion plant. He safe data presented in K-1019 (and GAT-225) was generated from critical experimental data with guided extrapola and interpolation using neutan transport theory. An important series of documents which describe some of the critica experiments that were used to derive the safb parameters in K-1019 are the " Critical Mass Studies", Parts I throug The safe parameters from K 1019 were generally accepted and applied to the operation of the diffusion plants with n additionaljustification.

There are several items of significance in the past use of K 1019 safe para 4i. All of the safb parameters are based on optimally water moderated systems for the enrichment and parameter considered, and all of the systems are fbily water reflected. At the diffusion plants advantage was generally taken of the larger safb geometries for enrichments below about 10%, and safe geometry for fhlly enriched uranium (~90%) was used for greater than 10% enrichment.

ne actual enrichment was generally considered when safe mass was used for evaluation, where the safe mass is 43.5% of the minimum critical mass for an optimally moderated, fully reflected sphere. De safb parameters presented in K 1019 are limited to systems that do not exceed the uranium density obtained in water moderate i UOf: systems, ie contain uranium compounds whose maximum theoretical uranium density in the pure state does not exceed 3.2 gU/ce. This wu generally applied to UOh, UF., and UF. systems.

In the past, fbw calculations were actually required or performed in evaluating the diffusion plant agah==it. When calculations were required they generally fbil into two enrichment categories; less than 5% enrichment and -93%

enrichment. The codes and crosc sections have been validated over a broad range of moderation for wMr=*

between 2 % and 5%, and for ~93% (for exampic see ORN!JCSDfrM-238). De validation is considered applicable

'o systems which are "similar" to those that have been validated. Similar is taken to be similar in geometry, fuel type, moderator, moderation range, etc. The average energy group of the neutron causing fission (or more recently, the average energy of the neutron causing fiss on) has been nsed also, to compare spectral similaritics of systems to those that have been validated.

Attachment 3

W Some meansion~of the existing validations fbr applicabdity to intermediate enriched systems can bejusti6ed. In .

, validation oflow eriched umnium systems'over a broad range of hydrogen moderation, the "U cmse sections haw  !

l_ - been validseed over a broad energy range. If there were sigmflamat eners in the~ cross section they would appear as i significant trends and/or bias as the hydrogen moderation and enrichment are varied. While there may be some trends 4 and bias, their magnitudo is small sad they can be easily bounded by applying a safbty margia to the at-lasiam.

Similar arguments can be applied for claiming that the mU cmse sections haw also basa validssed over a broad range

- of energy. Based on these arguments the validation can be extended to hydrogen moderated systems with intermediate ,

enrichments as long as the spectrum is similar to the spectrums of the validation esperbnents. _De averess saargy of -  !

the neutron causing fluion has been used to determine the spectral similarity. De validation opuid be opplied to -

intermediate enrichments where optimum tr&.Goa fbr minimum mass or solution geometry is considered.  !

There are fbw if any validation experiments with intennediate energy spectrums. Examples of systems with intermediate spectrums are water rnoderated hishly enriched uranium systems with an H/X in the range between about -

2 to 20.- Systems with intermediate mass moderators can also have intermediate spectrums. Caution should be used in 1 applying the existing validation to systems with intermediate spectrums, especially those with moderstars other than i hydrogen. Systems with intermediate enrichments can be physically similar to systems that have been validstad but may be significantly diffbrent neutronically. For example, a highly ennched UF. system at an H/U-2 (H/X=2) would be predominantly fast with some intermodiate contribution, the same system at 20% enrichment would haw a much greater intermediato contribution because the H/X=10 when H/U-2.

There are some experiments with intermediate enriched uranium that could be added to the existing ulidation. Sowrml -  !

intermediate enrichment experiments have been evaluated in NENNSC/ DOC (95)03, "Intemational Handbook of .

Evaluated Criticality Safety Benchmark Experiments". Another possible source is ORNL 2968, " Critical Mass L Studies, Part X - Uranium ofIntermediate Enrichment". . I believe that additional critical experiments are needed for ,

validation if calculations are going to be pfer.cd for systems at intermediate enrichments and especially if the spectrums are intermediate. h the absence of validation experiments in the intermediate enrichinent range, a larger margin of suberiticality should be used ( compared to the margin of suberiticality applied to < 5% and >90%

enrichments). The magnitude of an acceptable margin could be highly system dependant.

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