Sedat /3 3 · 2012. 11. 19. · S. Checker Sedat Gokugh /3 3 9. Lead Design Enginear J. Wesft Davis...

CRWMS/M&0 I *. Calculation Cover-Sheet

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MOL .19990503 .0209

1.1 OA: LI Pe: I . of: 50

2. Calculatin Tide.N-PRea~w Spent Nucdea Fadl Criicalfty Calculatiozm

T. 0dWnator Paul 1. Sentierl

S. Checker Sedat Gokugh /3 3

9. Lead Design Enginear J. Wesft Davis

¶0. Remarks

Revision Filstory

11. Revision No. IL. Daft Approved 13. Desm%)*an of Revision

REV 00 Initial Issule

I vIM "WA> a7 ln W"a VWV- I of&WM 14

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Waste Package Operations Engineering CalculationTlEe: N-Reactor Spent Nudear Fuel Cditcality CalculationsDocument Identfier. BBAOOOOOO-01717-021000005 REV 00 of 50

Table of Contents

Item - PageI. PURPOSE 4

-4

3 ASUMONS 4

4 USEOFCObMUTERO0FWAlRE 6

4.1 SOFWAE PROV FOR QuAsIY SJRCE (QA) W (1R.C_____ 6A 1 UI J IZAM"Al __A _SA AC

42 SOF42.1

MWARE ROMS- - -7.Excel -nlf nfn.n n n . n ... ... 7

L CALCIgATION S

5.1.1 DEcription ofN-Rcactor SNF __5.1.1.1 N-Reactor Type Mak IA SNP .5.1.12WNaor Type Mark W SNF n__. 12

5.12 Description of N-Reactor SNF Storage ---. 12121 N-ReactorMa IASNF Sw .-adwM_. .. ..... 12

5.12.2N-Reaacor Mf N SNF Storage B s _. _145.13 Descrip of N-Reactor Multi-Caniser Overpacks (MOO'S) .i 1sS.IA Dscription of 4-Multi-CnisterOverpac Waste Package (4 MCO WP'5) ___ _ -18 IS

5.2 DESCRIPON OF CAS EVALA ............ _._ ...... 21521 Compariso Cases Between Type Mark IA and Mark IV N-Reactor SNF . __ _ 21

52.1.1 lnibd Mark lA and Mark i N-Reactor SNF _. 21521.2 Combinafion of ta and Degradd Scrap Mat 1A and Murc IV Nacor SNF . _ 215213 Degraded Scrap Mak IA and Mark IYN-Ractr SNF 22521A Evaluion ofCompason Cases Between Type Mark A and IV NRectr SN F 22

522 Cases Involving Various Degrees of Modeation n.. 2452±3 Carbon Steel as Moderator Exclusion andlor Neutronic Polson .. ......... 24S2A Basket Center Pipe Variations____ . n.n .nn _. .... 30S2.5 No Basket Center Pipe - Carbon Steel as Modetor Exclusion and/or Neutronic Poison _____ 3052.6 U-H%0 MKtr- Homogeneously Dispersed throughout the Basket - Mark lA Fuel._____ 3052.7 Zcaloy-2 Cladding Dispesed In Basket as Water Displacer-Mark IA F ud __ 30S2= Placement of Stanless Steel Ctuciform between MCO's in the 4 MOO WP 3 1529 Neutronlcally Posoned Stainless Steel Cruciform between MCO's hI te 4 MCO WP______. 315210 Incorporation of Stalnless Steel Process Tube __. 335±.11 Evaluation of Neutronc Poisoned Center Rod In 4 MOO WP ._____.._ __. 3352.12 Evaluation of Boron Carbide (BP4C Rods Placed in the Fuel Baskets _._ ___ . 35

Waste Package Operations Engineering CalculationTiMle: N-React Spent NuclearFuel &iticaWy CalculonsDocumentIdenifier.BBAOOOOO-01717-210-0005 REV00 of 50

6. ~~~~~~~~~~~~~~386.1 xESULTS FROM COMPARS CA OF TYPE MARIA AND MARK IV N-REACIoRSNF._.... 38

6.1.1 Results fom It act Mark IA and Mark TV N-Reactor SNF Cases ___ 386.12 Results from Intact and Degraded Scrap Mark IA and IV N-Reacr SNF Casms _ 386.13 Results fim Degrded Fel and Scrap Mark IA and Mark IV N-Reactor SNF Cases ... 39

62 REsxTs, FROM VAMoUS DEGREES OF MODMAM WmnN DIE MOO'S AD 4 MOO WP_._.._. 396.3 RESULTS FROM CABON STEEL AS MaDEATOR EWoHANAORNEUr CPOOSatL..... 406A RESULTS OF BSKE CEE Fi VARATON CASES B-MARx IAN-REACTOR._ 42

6A.1 Results fiOm Carb Steel within the Baskets -No Center Pipe hI Baskets - Mark IA 426A2 Othr Basket Centr Pipe Variation Cases -Degraded Mark IA Fuel . ____. 426A.3 Results fr= U-H Mixre Dbspersed In the Basket - Degraded Mark IA FueL___. 43

65 RESULTS FROM ZIRCAY2 CADDIASMODERATORECWSION W N rE BasE _ _ 436.6 RESULTS FROMNON-POLSONEDC CtUORMBTE MMOs NS4 MOO WP - __446.7 RESULTSFROM GADLIM POsNDCRUFORM BE UNMCOSIN4MCOOWP. _ 446A RESUTS FROM CASS coRPOA 304L STA S STECENTER PROCESS TLE._.-_._ _. 4S6.9 RESULTS FROM PIACEMTOFEU~NICAllY POSIONED RD DIN CERnOF4 MOO WP__.__... 466.10 RESULTS FROMPLACEMENTOFB 4C RODS NFUELBA _ . .____ . ... __-_47

7. REFERENCES 48

L* S. A _AUMNb .SO

Waste Package Operations Engineering CalculationTtle: N-Reactor Spent NuclearFuel Criticality CalculationsDocent Identifier. BBAOOOOOO-01717.021005 REV 00 . of 50

1. Purpose

Ihe purpose ofthis calculation is to characterize the criticality safety aspects of N-Reactor fuelstored in a Department of Energy spent nuclear fuel (DOE-SNF) canister that contains fourMuld-Canister Overpacks (MCO's). These calculations will be done to support the analysis thatwill be done to demotstrate concept viability related to pre-emplament storage and use in theMonitored Geologic Repository (MGR) envment for the pre-closure time frame. Thepurpose of these calculations is to evaluate the criticality issues related to these intact fueIls forthe pre-closure time firme, including providing criticality evaluations which can be used toevaluate various storage coand input into the preliminary DOE-SNF canister design

2. Method

The calculational method being employed in this calculation is MCNP4B2 computer code (Ref.7.2) to calculate effective multiplicaion factors (4 for various geometrical configurations of N-.Reactor Spent Nuclear Fuel.

3. Assumptions

The following assumptions were used in the calculations.

3.1 Beginnmg of Life pr-IraLdiation uraniumn values were ue b d up past criticaitysafety analysis (Ref. 73). Ibis assumption is used throughout Section 5.

3.2 Scrap fuel was evaluated at a pre-Iradiation enrichment of 1.25% thioughout all MarkIA SNF calculations in which scap fuel is inrporated into the case. Ihis is based uponthe fact that no control exists upon the placement of either the Mark IA inner assembly orouter assembly into the MCO scap baskets. Use of the 1.25% enriched material is basedupon this fact. Therefore all of the material in the Mark IA scrap basket is evaluated asthe higher enriched 12!i% material. This assumption is used throughout Section 5.

33 Past inspections by Hanfoid Fuel Operations indicate questionable structural integrt ofthe N-Reactor fuel. The fuel matrix comprising the elements was evaluated asheterogeneous spheres of fuel to account for possible structural questions. The mass ofthe fuel matrix was preserved and the spheres were spaced at the most reactive (optimum)pitch within the confines of the basket. This assumption was used throughout in all casesin Section 5, which the fuel was not evaluated as intact elements. The basis of thisassumption is engineeringjudgment.

3.4 When the intact Mark IA fuel matrix was evaluated as heterogeneous spheres, a weightedaverage of the inner and outer assembly pre-irradiation enrichment was used. Thisweighted average was calculated and evaluated as 1.15% enriched material. This

Waste Package Operations Engineering CalculationTile: N-React Spent Nuclear Fuel Criticality CalculationsDocument Identifrer. BBAOOOOOO-01717-0210- 0000REV 00 Page of 50

assumption Is used throughout Section 5. The basis of this assumption is engineeringJudgement.

3.5 A 30.48 cm thick water reflector of full density water (1.0 gem') was used in all casesthat were evaluated with reflection. This is based upon a 30A8 cm (12 in.) thick waterrefleor being effectively equivalent to an infinite reflector. (Rf. 7.10 p. 106) Thisassumption is used throughout Section 5.

3.6 The MOO baskets used to store Intact Mark IV elements have a 12-gauge stainless steelouter wall that extends 27.94 cm (11 in.) axially from the bottom of the basket la alcases, in SectionS, raig intact Mark IV elements, this outer wall has beenignored for simplicity purposes. This is conservative based upon the fact that thepresence of the outer wall will have a minor effect In reducing the interaction betweenadjacent MOO's and reduce.the effect of the water reflection present around the basketITis assumption Is used throughout Section 5.

3.7 Ihe basket used to store intact fuel elements, for both Mark IA and Mark I, has analuminum fuel spacer plate located in the bottom of the basket. In all of the cases, inSection 5, which included intact fuel elements, this basket was ignored. The presence ofthe spacer will have an insignificant negative effect on kdr This assumption is basedupon engineing judgement. This assumption is used thwoughout Section S.

3.8 The MCO has a shield plug assembly that is used as part of the closure mechanism. Theshield plug was replaced with a void in the no-flooded cases, and water at varied density(O to 100% theoretical) for the flooded cases. Ihis assumption is used throughout SectionS and is based upon engineering judgement

3.9 The top lid of the MCO was simplified and evaluated as "squared off" in all cases inSection 5 in which the MCO was used. This assumption was based upon engineeringjudgment Tbis assumption is used throughout Section 5.

3.10 The scrp, baskets contain six stainless steel plates that divide the basket into sixcompartments. These divider plates were ignored troughout all cases in Section S, inorder to simplify the case. This is a conservative assumption based upon the fact that theeffect of the divider plates will be a reduction in the ka of the system. This assumption isused throughout Section S.

3.11 Ihe basket bottom plates have small drain holes present throughout the entire plate.These holes were Ignored in all of the cases in Section S, to simplify the case. Basedupon engineering judgement, the presence of the holes will have an insignificant effect onkC. This assumption is used throughout Section 5.

Waste Package Operations * Engineering CalculationTite: N-React Spent Nuclear Fuel Criticality CalculationsDocment Identlfier. BBAOOOQOOO01717-021000005 REV 00 of 50

3.12 The trapezoidal shaped structural posts, which support the outer stainless steel shroud ofthe baskets, have been Ignored in all calculations. Tbis assumption is based upon the factthat the exclusion is conservative, since it will have a positive effect on kff Thisassumption Is used tughout Section S.

3.13 The inside cavity lengthof the MCO was evaluated as 406.7389 cm. The actual insidelength ppears to be 406.6286 cm (160.09 in.) (Ref. 7.8, Dwg. U H.2-828041). Due to thecomplicated design of the end cap and the design of the MOO cover, the inner length had-to be derived from the dimensions on multiple drawings. The small difference between*the actual length and the evaluated length wll have an insignficant effect on the kThs assumption is based upon engineeringjudgement and is used throughout Section S.

3.14. A uranium density of 18.882 gfc&' w assumed. Calculations that presrved fuel matrix.mass and volume yielded an average density-of 18.512 glcrA for the fuel matrix. Thisdensity was increased by 2%, to conseatvely bounds any uncertainties related to themasses and volumes. This 2% increase is based upon engineering judgement stemmingfrom review ofthe various data source documents (Ref. 7.5). This assumption is usedtIbroughout Section 5.

3.15 The entire length of the MCO was simpliied and evaluated wit a 60.96 cm outsidediameter over its entire length in all cases in Section S in which the MCO was used.. Thisassumption was based upon engineering judgment.

4. . Use of ComputerSoftware

4.1 Software Approved for Quali Assurance (QA) Work

41.1 MCNP4B and MCNP4B2

MCNP4B and MCNP4B2 computer code is used to calculate effective multiplicationfactors for the waste packages evaluated (Refs. 7.8 and 73).

* Program Name: MCNP* VersionitevisionNumber.:Version4B* Computer Software Configuration Item (CSCl) Number:. 30033 V4BLV* Computer ype: Hewlett Packard 9000 Woxkstations

The input files used are echoed in the output files. The outut files are listed inatachments VI. The output files are stored on compact disc electronic media (Ref. 7.12).

Waste Package Operations Engineering CalculationTite: N-Reactor Spent Nuclear uel Criticality CalculationsDocument Idenffier. BBAOOOOOO01717-0210 OM REV 00 of 50

a) The MCNP4B computer code (ReL 7.2) is an appropriae tool to be utilized todetermine the criticality potential, kd of fresh and spent lattices of intact anddegraded N-Reactor fiel assemblies.

b) This softwre has been validated over te rangeft was used.c) It was previously obtained from the Civilian Radioactive Waste Management

System (CRWMS) Management & Operating (M&O) Software ControlManagemenlt (SCM) in accordance with appropriate procedures.

* Program Name: MCNP* VersionRevisionNumber.Version4B2* Computer Software Configmation Item (CSCI) Number 30033 V4B2LV* ComputerType: Hewlett Packard 9000 Workstations

The input files used are echoed in the outt files. The output files are listed in attachments VILThe output files are stored on compact disc electronic media (Ref 7.12).

a) The MCNP4B2 computer code ef. 7.2) is an appropriate tool to be utilized todetermine the citicality potential, km of firesh and spent lattices of intact anddegraded N-React fued assemblies.

b) This software has been validated over the rmnge it was used.d) It was previously obtained from the Civilian Radioactive Waste M gement

System (CRWMS) Mangement & Operating (M&O) Software ControlManagement (SCM) in accordance with appropriate procedures.

4.2 Software Routines

42.1 Excel

* Title: Excel* Versionrevision Number. Microsoft Excel 97

The Excel spreadsheet program was used for simple numeric calculations as documentedin Section S ofthis calculation file. Ihe user-defined formulas, inputs, and results weredocumented in sufficient detail in the spreadsheets to allow an independent repetition offfie various computations. Ihe Excel files are stored on electronic media (Ref. 7.12) andas hardcopy in Attachment S.

Waste Package Operations -. Engineering CalculationTlUe: N-Reactor Spent Nuclear Fuel Criticality CalculationsDocument Identfier. BBAOOOOO001717.021000005 REV 00 of 50

5. Calculation

5.1 Calculation Inputs

'he number of digits for cited values does not necessarily indicate accuracy. It may reflect tevalue as reported in a reference or be an artifact of conversion.

5.1. Description ofN-Reactor SNF

Criticality calculations of the intact form of N-Reactor Spent Nuclear Fuel (SNF) wereperformed using Beginning of Life (BOL) fuel loading quantities. This was shonm to beconservative for these applications (Ref. 7.5, p. 4-i)

Ihe details of the N-Reactor fuels were obtained from the N-Reactor data package provided byHanford N-Reactor Operational Support personnel. The N-Reactor fiel considered consisted oftwo types, Mark IA and Mark IV, which are very similar in design Both ofthese fuels aremetallic uranium fuels clad in Zircaloy-2. Each of the elements consists of a nested assemblyconfiguration, in which an inner annular assembly is nested mnside an outer annular assembly.Figure 5-1 shows an example of the Mark IA and Mark IV type fuels.

5.1.1.1 N-Reactor Type Mark IA SNF

The Mark IA fuel consists of an inner assenmbly with a pre-iudiation enrichment of 0.947%0 and...an outer assembly with a pre-irmrdiation enrichment of 125%. The length ofthe Mark 1A fuelranges from 37.846 cm to 53.035 cm. The Mark 1A case used the 53.035 cm long elements.The pertinent dimensions and weights comprising the MarkIA elements are given in Table S-1(Ref. 7.5, p. 2-3). The Mark IA fuel element inner and outer assemblies have Zircaloy-2 endcaps having an axial length of 0.483 em on each end.

The original configuration consikts of forty-eight Mark IA elements per basket (Fige 5-2). Sixbaskets containing Mark IA SNF are then placed into a MCO. Previous criticality calculations'Cf. 7.5) allow two of the six baskets within the MCO to contain scrap or degraded Mari IAfuel. These calculations indicate that these baskets need to be placed into the MQO as the bottomand top baskets. Four MCO's wil then be placed into a 4-Multi-Canister Overpack WastePackage (4 MCO WP). The data nPuts for the baskets, MCO's, and 4 MCO WP's will bedescrbed later in this report

Waste Package Operations . Engineering CalculationTWe: N-eacr Spent Nuclear Fue Qiicaity CaIcailonsDocument Identifier: BBAOOOOOO-0171702 10-00005 REV 00 of 50

MARKWO

-EndCXps:

;up

MARKIA

calm

Figure 5-1. Schematic of N-Reactor MarklA an! Mark n Fue Elements

Waste Package Operations .; Engineering CalculationTWi: N-Reactor Spent Nulear Fuel Ckiticality CalculatiosDocument Identfier. BBAOOOOOO.01717-0210.O0 S REV 00 of 50

Table S-1..NReactorFuelAssemblies Dimensions andWeights

Mark [V Mark IAOuter Tube Diameters: cm cmZconium clad outer diameter 6.160 6.106Uranium outer diameter 6.032. 5.979Uranium iner diameter 4.422 4.592Zircoium clad inner diameter 4.321 4.481

Outer Tube Enrichment: wVM wV%mu0 0.94700 12500V . 0.0392 0.0392

VMU 99.0138 98.7108

Inner Tube Diameters: cm cmZiroium clad outer diameter 3249 3.165Uranium outer diameter 3.096 2.962Uranium inner diameter 1321 1.245Zircon}u clad inner diameter 1.219 1.118

Inner Tube Emichment: wt% wt%U . 0.94700 0.94700

WV 0.03920 0.03920_U 99.0138 99.0138

Fuel Assembly Dimensions: cm cmMaximum length 66294 53.035End-cap thickness 0.483 OA83

Fuel Assembly Wcight: kg kgMaximm weight 23A. 16.6

Waste Package Operations ---Ena-n-r- a -- …nifin

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Title: N-Reactor Spen Nuclear Fuel CrIicality C nculaflsDocument Identfier. BBAOOOOOO-01717-0210R00005 EVa 00 Pane 11i nfqn

- . as . . _. TV

Pna�ii A�fl

LoadingAlrrangementfor~farkiVFuelIn UCO Cornfainer

Loading Armangement forMarkIA Fuelln MCO Container

Figure 5-2. Example of Loading Arrangements In MCO'i

Waste Package Operations Engineering CalculationTMte: N-Reactor Spent Nuclear Ful C(icality CalculationsDocument Identifier. BBA000000-01717-0210 OOOOSREV 00 of 50

51.12 N-Reador pe Mark IV SNF

The Mark IV fiuel consists of an inner assembly with a pre-irradiation enrichment of 0.947% andan outer assembly with a pre-irradiation enrichment of 0.947%. The length of the Mark IV fuelranges from 44.196 cm to 66294 cm. The Mark IV case used the 66294 cm long elements. Thepertint dimensions and weights comprising the Mark IV elements wre given in Table 5-1 (Ref.7.5, p. 2-3). Ihe Mark IV fuel element inner and outer assemblies have Zircaloy-2 end capshaving an axial length of 0.483 cm on each end.

The origi configat on consists of fifty-four Mark IV elements per basket (Fe 5-2). Fivebaskts containing Mark IV SNF are then placed into a MCO. Previous criticality saft.calculations (Rsf: 7.5) allow two ofthe five baskets within the MCO to contain scrap or degradedMark IV fuel. These calculations indicate that the scrap baskets need to be placed into the MCOas the bottom and top baskets. FourMCO's will be placed into a 4 MCO WIP. The datainputs for the Mark IV baskets are described later in this report.

5.1.2 Description of N-Reactor SNF Storage Baskets

The Mark lA and Mark IV N-Reactor SNF will be placed into storage baskets. Either six (MarIA) or five (rk V) baskets will be placed into an MCO.Q he baskets are designed such thatwhen stacked, the lower end of the center pipe of the basket fits over the top end of the centerpipe of the basket below. Thus, IesngW the baskets together in a vertical stacL

5.12.1 N-Reactor Mark IA SNF Storage Baskets .-

Ihe basket is an annular type basket constructed of 304L stainless steel (Figr 5-3). The centerpipe is comprised of nominal 6.0 in. XS stainless steel material (actual dimension 6.625 in. or16.8275cm)), and the outer wa compised of 18-gauge stainless ste dsheet metal. The outer radiusofthe centerpipe is 8A1357cm withthe radius ofthe outer shell equl to 28.70708cm. Thewall tbickness ofthe centerpipe is 2.19456cm. Ihe forty-eightelements are housed inthe annulrsection. The inside height of the basket Is 55.83428 cm, with an overall outer height of 58.88228cm (Ref 7.7, Sketches 5 and 6). The basket bottom plate is constructed from 3.048 cm thickstainless steel with drain holes drilled through. Each of the baskets contains an aluminum elementspacer guide at the bottom of the basket. Tis spacer is approximately 5.0 cm thick and aranges theelements in the triangulr pitch configuration as shown in Figure 5-2, (Wef. 7.8, Dwg. # H-2-828060).

Waste Package Operations Engineering CalculationTle: N-Reactor Spew Nuclear Fuel Qiticality CalculationsDocumet Identfier. BBAOOOOOO-01717-0210 OOOREV 00 of 50

.1

Figure 5-3. Mark LA SNF Intact Element Storage Basket

Waste Package Operations Engineering CalculationTitle: N-Reactor Spent Nuclear Fuel Criticality CalculationsDocument Identifler: BBAOOOOOO-01717-0210.00005 REV 00 of SO

In addition, the same basket used to store intact Mark IA elements is also used to store Mark IAscrap material. The scrap baskets do not have the aluminum spacer element guide located at thebottom (Figure 5-4). Scrap Mark IA SNF consists of various sized pieces and sectionscomprised of Mark IA elements that have structurally failed.

lhe scrap baskets ae divided into six copartments separated by six stainless steel plates (Ref7.7, Sketch 6). These compartnt dividers are conservatively ignored in this evaluation. Inaddition, the structural posts, which support the outer I tgauge stainless steel shell, are alsoignored in the ca

5.122 N-Reactor Mark IV SNF Storage Baskets

The Mark IV basket is also an annular type basket constructed of 304L stainless steel (Figure .5-5).The fifty-four elements are housed In the annular section. The center pipe Is comprised of a 6.985cm outside diameter stainless steel pipe, with a wall thickess of 1.27 cm (Ref. 7.8, Dw& # H-2-828070). Unlike the Mark IA fiuel, in which the reguar and scrap baskets are the same, the Mark[V regular baskets and scrap baskets have slight differences.

The basket used to house intact Mark IV feIl assemblies has an outer shell and six compadividers that extend a tely half the height of the bucket (Ref 7.7, St 8). The outsideshell is constructed of 12-gauge 304L stainless steel as well as the basket dividers (Ref 7.8, Dwg.# H-2428070). The outer radius of the outer shell is 22.8290 cm (22.70 in. diameter). Each of thebaskets, for intact assemblies, contains an alumimum element spacer guide at the bottom of thebasket This spacer is oximately 7.62 cm thick axially and arranges the Mark IV elements inthe triangular pitch configuration at a typical center to center pitch of 6.985 cm (Ref. 7.8, Dwg. #H-2-928070), shown in Figure 5-2. The inside height of the basket is 67.818 cm, with an overalouter height of 70.66788 cm Ref. 7.7, Sketch 8).

The Mark IV scrap basket Is also an annular type basket constructed of 3 O4L stainless steel (Figure5-4 Ihe Mark IV scrap material is housed in the annular siction. The center pipe is comprised ofa 6.985cm outside diameteratainess steel pipe, with wall thickness of 127cm with an outershellcomprised of 12-gauge stainless steel sheet metal (Ref. 7.8, Dwg. # H-24828070). The outer radiusofthe outer shell is equal to 28.8290cm (22.70 in. diameter). The basket bottom plate is constructedfiom 0.635 cm thick stainless steel with drain holes drilled through (Ref. 7.7, Sketch 8). The scrapbaskets do not contain the aluminum element spar guide at the bottom of the basket.

The baskets are divided into six compartments separated by six stainless steel plates (Ref 7.8,Dwg. # H-2-42807S). These compartment dividers are conservatively ignored in this evaluation.

Waste Package Operations Engineering CalculationTile: N-Reactor Sp=n NucearFuel Qiicaiy 1cmans

. Docuent Identifier. BBAOOOOO-01717-021O00005 REV 00 Page IS of S0

-Centerrpost6 In. XXS Ss pipe0.84 In a219S on) walittickmess

. I

Figure 5-4. Mark IA SNF Scrap Material Storage Basket

2R -- & n-a -tala aata.

wwvabu rwwutmlt upulauwam Enuineering LCaKcuiIof

Ttle: N-Reactor Spen Nuclear Fuel Criticaty CalculatiosDocument Identifier. B oBA0ooo-om01717-0210-OOO5 REV o0 of So

as.-

Figure S-S. Mark IV SNF Intact Eement Storage Basket

- w -

Waste Package Operations . Engineering CalculationThe: N-ReaorSpe Nuclear Fuel CdtcaIty CalculationsDocument Idengfier. BBAOOOOOO-0171702 100005 REV 00 of 50

D2.655 In. (6.744 an)2.645 In. (6.718 cn)

-Y

0.6tn (27cn Ut ~wat.

j

30A07 In.(77234 ar#30387 In.

(77.183 an)

27.822 hI.(70.688 CM)27.812 In.

(70.642 ac)

[I _ ,,In

Is h50I. (6M68 an)qv v &- - - --- -- -

*-D2.720 hI. (6809 cmn2.710 In. (6.683 an)

E MAX D22.700 In. b(57.658 cr)

Eigure 6. Mark IV SNF Scrp Material Storage Basket

Waste Package Operations Engineering CalculationTtl e: N-Reactor Spew Nuclear Fuel Criticality CalculationsDocumentldentfier:BAOOOOOO-01717-IT00 OOOO5REVOO Pape 18 of 50

5.L3 Description of N-Reactor Multi4CnIster Overpacks (MCO's)

The MCO's are constructed out of304L stainless steel having an outside diameter 60.96 cm anda wall thickness of 2.54 cm (Figure 5-7). The top portion of the MCO has a slightly larger radiusthan the overall tube body in order to accommodae the top mechanical closure device. TheMCO was evaluated with a 60.96 cm outside diameter over its entire length The overall lengthofthe MCO is 420.09 cm with an Inner cavity height of 406.6286 cm (160.09 in.) (Ref. 7.8,Dwg. # H-2-828041).

Ihe bottom plate has a thickness of 4A704 cm with the top lid having a thickness of 8.9916 cm(Ref. 7.8, Dwg. # H.2-828041). he design ofthe top lid and the mechanical closure device Is avezy complex design and was simplified in these cases. The lid was squared offwith themechanical closure device being ignored.

In addition a central process tube constructed out of 304L stainless steel is present in the MCO's.This central process tube is located down the center of the MCO through the stacked basketscenter pipes. The central process tube has an outer diameter of 3.3401 cm with a thickness of0.90932cm (Ref. 7.8, Dwg. # H-2428051).

S.IA Description of 4-Multi-Canister Overpack Waste Package (4 MCO WP's)

The final waste package will consist offour MCO's placed into a 4 MCO WP. The 4 MCO WP(Figure 548) has an inner diameter of 160.1 cm with an inner cavity height of 423.5 cm. The 4MCO WP is constructed of an inner corrosion resistant shell and an outer corrosion allowanceshell. The inner corrosion resistant shell is constructed of Hastelloy Alloy C-22 with 2.0 cmthick walls and 25 cm thick end shell lids. The outer corrosion allowance shell is constructed ofA516 Grade 70 Carbon Steelvwith 10.0 cmthickwals and 11.0cmend shell lids. The outershell lids are recessed on both ends. The overall outer height ofthe 4 MCO WP is 498.5 cm.

Waste Package Operations En-alneerina CalculationTitle: N-Reactor Spen Nucla Fuel Cificaity CalculationsDocument Identfler. BBAOOOOOO-01717.021040005 REV 00 oT SO

642Va7 CM

_ I Icowiecnp-~..

COWa-=.

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d.Vlrr/lvlm4- +

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45 In.tO.145

419.722an*0*8a

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148.D0 h*0.095

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140.375 IM*0210

*0.533

I I IIB - �I*- -. - a - I - - - -

Shel boam- IK- 2 16hOIm I

c-24.00 k. -S.(80.98 MO~ mm

Figure 5-7. Multi-Canister Overpack - MCO

Waste Package Operations -- --- - - .- -. Engineering CalculationTitle: N-Reactor Spent Nuclear Fu Criticaitr CalcuatfosDocument Idetier. BBAOOOOO-01717-021000005 REV 00 wPale 20of SO

1.

I D703 TYPat 12Tan each end

Figu rc 5-8. Four Multi-Canister Overpack Waste Pickage - (4 MCO WP)

Waste Package Operations. Engineering CalculationTitle: N-Reactr Spent Nuclear Fuel Criicality calculationsDocment IdenUrier. BBA0OOD01717.02l0-00005 REV00 of 50

5.2 Description of Cases Evaluated

52.1 Comparison Cases Between Type Mark 1A and Mark V N-Reactor SNF

The first set of calculations included comparison cases between the type Mark lA and Mark IVin order to determine if one of the fuels could be used as a bounding fuel to encompass bothtypes. Various cases were used which incorporated intact, degraded, and a combination of intactand degraded material in the same MCO. The two types offuels, Mark lA and Mark IV, are notevaluated as being contained in the same MCO. Both single MCO and four MOO's in the 4MCO WP cases are evaluated. In these cases the MOO(s) ae evaluated containing intact fuel inall of the baskets, intact in the center basts and degraded fuel in the top and bottom baskets, ordegraded baskets in all of the baskets. All of these initial comparison cases aet evaluated withthe MCO(s) and 4 MCO WP flooded with full density water. The systems are also fiuly reflectedwith a 30.48 cm thick water reflector at fhul density.

5±2.1.1 Intact Mao IA and Mark IV N-Reactor SNF

The first set of calculations included comparison cases between the type Mark IA and Mark IVintact fuels. The fiuls arm evaluated as intact and stored as outlined in preceding sections. Ineach of the cases the maximum number of baskets were placed into the MCO.

51.12 Combination of Intact and Degraded Scrap Mark 1A and Mark IV N-Reactor SNF

This set of calculations included comparison cases between the type Mark IA and Mark IV fuelswhich are evaluated as a combination of Intact and degraded scrap material. In this set of casesthe MCO(s) are loaded with a scrap basket on the top and bottom levels with the middle baskets

c aing intact fiel.

The Mark IV scrap baskets are evaluated containing 0.947% enriched material The scrap masswas that equivalent to fifty-four Mark IV fuel elements. The Mark lA baskets are evaluatedcontaining 1.25% enriched material, equivalent to the material mass of 48 intact elements. Thiswas done due to the fact that both degraded Mark IA inner (0.947% enriched) and outer (1.25%)assemblies can be placed into the scrap baskets, without any control over the proportionsbetween the two types of materials.

An evaluation (Attachment I) was completed to determine the most reactive size of sphericalparticle to use in the evaluation of the scrap material, along with the most reactive spacingbetween the spheres within the confines of the baskets. This study was completed using the typeMark lA fuel. In addition, the sphere diameter was limited to the thickest radial fuel matrixdimension of a Mark 1A fuel clement, 0.8585 cm.

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In the cases evauated, the maximum calculated kfwas shown to occur with the spheres at thismaxm= diameter. This assumption was extrapolated to the Mark IV type fuel wih the scrapevaluated at the thickest radial fuel matrix dimension of a Mark IV fuel element, 0.8875 cm(Figure 5-9).

52.13 Degraded Scrap Mark 1A and Mark IV N-Reactor SNF

Some question adsts relating to the structural integrity of the N-Reactor fuels. The question isconcerned about whether or not credit can be taken for the strucura integrity of those fuelsclassified as intact. In order to address this situation a set of cases was developed in which all ofthe material contained in the baskets was evaluated as degraded.

he spherical case mentioned in the previous section was used to evaluate these cases. The MarkIV fuel was evaluated as 0.947% enriched material in each of the baskets housed in the MCO(s)..

The Mark A sap fuel case has a slight vaation from the Mrk IV sp fuel case. The MarkIA fuel scp was evaluated as 125% enriched material and was placed into the top and bottombaskets. Once again the spherical configuration was used to evaluate this case. The middlebaskets however, are evaluated containing 1.15a enricWh fuel materials. This is a weighted-average ofthe lark IA inner and outer fiel tubes and is used to simuate the degraded inner andouter fuel tube scenario.

*5±1A Evaluation of Comparison Cases Between Type Mark IA and IV N-Reactor SNF

Compison cases show that the Mk IA and Mark lV fule have very simir calculated .'s forthe normal case in which the fuel is evaluated as small spherical particles and no fuel is presentin either the Mark IA or IV basket center posts. Due to the smaller center post and the slightlyhigher mass loading, the intact Mark TV cases yielded a higher kfghan the intact Mark IA fuel.However, the limiting cases are those in which the fuel Is evaluated as degraded and not as intactAn egineering judgment is used to determine that Mark IA fuel is enveloping for the type of

criticality calculations necessary to evaluate the N-Reactor fuel. In the normal cases the basketcenterpost forthe Mak IA fuel is much largerthan for the MarkV fuel. Iffuel is included inthese centl regions the calculated kffis greater for the Mar lA fuel than for the Mak IV fult,with the degraded spherical case. Therefore, the remnaining cases only consider the Mark IA typeN-Reactor fiel. It is possible for small fragments to get into the center post area of the basket.Since this is a low enriched fuel, a homogeneous m te of small particles and water Is notconservative. For conservative purposes, the fuel in the center post region was evaluated aslarger particles, even though they could not actually fit through the gap into the center region.

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f I'

;'

ExPanded ViewcdkAnangemertotSpheres Used to Modflihe

Degraded Mark IA Fuel

1.0cmC-C SpacliBetween Spheres-

Fire 5-9. MCNP Plot - X-Y View of Heterogeneous Sphere Case Used

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522 Cases Involving Various Degrees of Moderaution

A set ofases was evaluated in which the density of the water in the MCO and the 4 MCO WPwas varied to determine the most reactive configuration.

Th density ofthe insitl waterpresent In a MO conning Mak A fuel was varied form10°h to filll density (1.0 g/cm&). AlB of the voids with the MCO are evaluated at the variousdensities. The MCO itselfwas s fully reflected onthe outside by30.48 cm thick layeroffuildensity water (Figure S-10).

Once the most reactive water density wffiin the MCO was determined, a case was constructedthat consisted of four optimally moderated MCO's in a 4 MCO WP which was filled with waterwhich varied in density from 10% to 100%/s. The 4 MCO WP was fidly reflected on the outsideby a 30.48 cm thick laer of full density waler (Fiues 5-11 and 5-12).

S.2.3 Carbon Steel as Moderator Exclusion and/or Neutronic Poison

If the Mark IA fuel is evaluated as optimally spaced small spherical particles to simulatedegraded fuel, high k.;'s result. If no credit can be taken for the structure ofthe fuel, then stepsmust be taken to ensure that the resulting kd!s are within the acceptance criteria. A possiblesolution is to include a material in each of the baskets that would act to exclude water and/or to.neutricall poson the system, thus reducing . The matial chosen was Grade SS A S16.Carbon Steel.

A case was constructed which consisted of an MCO containing degraded Mark IA fueL Theintrsitial space between the small sphes was filled with a miture of water and carbon steel atvaried proportional densities (Fiure 5-13), ie., if the materi contained 90% full density carbonsteel, the mai 10% void was filled with water.

A set of cases evaluated the effects of placing the carbon steel - water mixture, at variousproportional densities as previously described, in the radial void spaces between intact elementswithin each basket (Figure 5-14). In these cases the central void and annular void regions of theelements were filled with full density water.

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--

Five Mark iA IsedsInaFullyWaterFloodsdand Refected MCO

Enlarged VlewofArea

FIgure 5-10. MCNP Plot - X-Z View of Single MCO

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Vold In 4MCO WP

Figure S-1I. MCNP Plot - X-Y View of 4 MCO WP

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i, _

Side View of Two of the FourMCO'a contained In the

4 MCO WP

Figure 5-12. MCNP Plot -X-Z Viewof 4 MCO WP

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Carbon Steel - Water Mxture Between Spheresand in Center Pipe - Carbon Steel combined

with Water at Various Volume Factions

Figure 5-13. MCNP Plot - X-Y View of Interspersed Carbon Steel-Water Cases

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:4

4zr!1

Carbon Steel and WaterMburePlaced Around ant In Betweenlth

IrdactMauk 1AIElements

Figure 5-14. MCNP Plot - Intact Mark 1A Elements with Carbon Steel - WaterMixture in Radial Spaces

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S2IA Basket Center Pipe Variations

In all of the following cases the'MCO's were fly flooded and reflected with full density water.Also each ofthe MCO's were evaluated contaning degraded Mark lA fuel as previouslydiscussed using the spherical case. In each of the following cases the central process tube was notincluded.

A case was constructed in which the basket center pipes ar removed in order to evaluate the k);effects.

Another case was evaluated which considered the reactivity effects of shifting the center pipe5.08 cm radialy In one direction. This case was evaluated to provide comparison to a caseincluded in the original criticality reporL (Ret 7.5, p. 5-5)

The final case in this section considered the reactivity effects of placing degraded fuel into thecenter pipe while each ofthe baskets was fully loaded with degraded fuel.

S2.5 No Basket Center Pipe - Carbon Steel as Moderator Exclusion and/or NeutronicPoison

A set of cases was used to evaluate the decrease in kf due to the introduction of carbon steel, at.various densities, into a basket which had the center pipe removed. The same raio of carbonsteel and water, as previously described, was used in this set of cases. -

52.6 U-HIO Mixture - Homogeneously Dispersed throughout the Basket - Mark lA Fuel

In order to determine reactivity effects from particulate, a set of cases was developed. In thesecases, the MCO was evaluated fully loaded with degraded Mark lA fuel. The MCO was fiulyflooded and reflected with full density water. Ihe water intmpersed within the MCO wasevaluated containing various gram quantities of the uranium fuel matrix. The gram quantitiesranged from 0.001 g/L to I gIL. In these cases the basket center pipes were evaluted intact.

52.7 Zircaloy.2 Cladding Dispersed In Basket as Water Displacer - Mark 1A Fuel

Previous cases took no credit for the presence of Zircaloy-2 material comprising the cladding ofthe elements. The uranium metal fuel matrix of each of the elements is clad in Zircaloy-2.Although the structural integrity of the cladding is in question, it has been verified as beingpresent Zircatoy-2 has a low neutron absorption cross section. But when credit is taken for thevolume ofthe Zircaloy-2, it will lower the available space (volume fraction) for water currentlycontained between the spaced heterogeneous spheres of fuel material.

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Ike presence of the Zircaloy-2 material from the cladding was evaluated. In these cases, theamount of material representing cladding was dispersed thrugh the water mixture within each ofthe baskets in the annular region only. Casm variations were evaluated which considered theeffects of placing Zrcaloy-2 material in the center pipe along with the degraded fuel. In thelimiting cases, no Zircaloy-2 was placed in the center pipe region along with the degraded fiel.

Cases were developed which considered fuel contained in the center pipe and fuel not containedin the center pipe. The 4 MCO WP was considered in this case. Ihe individual MCO's wereevaluated fully flooded with no water between them within the 4 MCO WP. Full density waterreflection was placed aound the outside ofthe 4 MCO WP.

52$ Placement of Stainless Steel Crciform between MCO's in the 4 MCO WP

The next set of cases developed was based on the case that was discussed in Section 5.2.6. Inthis set of cases however, each of the cae contained fuel in the center pipe. A set of cases wasdeveloped to determine the effect on the calculated k~ ifa stainless steel cruciform was placed -into the 4 MCO WP inbetween the individual MCO's.

The top collars of the MCO's have an outer diameter of 642874 cmn (2531 in.) while -the lowerportion of the MCO has an outer diameter of 60.96 cm. The tolerances on loading four MCO's-into a 4 MCO WP are quite tight but some room between the lower portion ofthe MCO's exists.Since this is where the fuel is housed the cruciform would not need to extend the full length ofthe4 MCOWP. -

Stainless steel plates, ranging in thickness from 1.0 to 4.0 cm, were evaluated. The type ofmaterial chosen for this ease was 3 04L stainless steel. An example of the physical layout isshown in Figure 5-1S.

52.9 Neutronically Poisoned Stainless Steel Cruciform between MCO's in the 4 MCO WP

Cases were evaluated that were similar to those discussed in Section 527. The case chosen asthe base case is that which includes the 2.0 cm thick stainless steel cruciform. Ihis was donesince preliminary evaluations indicate that thickness greater than 2.0 cm would not allow enoughroom to plac the individual MCO's into the waste package.

An evaluation was made to determine the magnitude of the effect on the calculated kff if aneutron absorbing material were integrated into the cruciform structr The material of choicewas gadolinium. Various gadolinium weight percents ranging from 0.01 to 2.0 wt0h wereconsidered.

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4 MOO's in a4 MQOOWPMOO's Full Flooded

No Watcr In4 MOO ViPBetween MOO'.

Figure 5-1. MCNP Plot -4 MOO WP with Stainless Steel Cruciform Present

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£2.10 Incorporatlon of Stailess Steel Process Tube

The last case evaluated consisted of the incorporaton of the cetra process tube. This tube isconstructed of3D4L stainless steel. It Is an annular tube having an outer diameter of 1.3 1S in.(3.3401 cm) and a thickness of 0.358 in (0.90932 cm). This tube was ignored in all of theprevious cases.

The process tube rums the entire length of the MCO down the center of each of the baskets.Various cases were considered to evaluate the effects ofthis center process tube. Cases forsngle MCO's, the 4 MCO iYP with and witout the stainless steel cruciform, and the 4 MCOWP with the Udfpoisoned stainless steel cruciform present, were evaluated.

In this next set of cases, the fuel was placed between the central process tube and the ceer tube.No Zincaloy-2 material was Interspersed between the heterogeneous spheres of fuel contained in

this region. The fuel spheres contained in this region were evaluated at two kadii. The firstradius anabyzed was the same radius used in the evaluation of the fuel spheres contained in thebasket annular region, 0.42925 cm. An example of this configuration is shown in Figure 5-16.

The next case analyzed and evaluated fuel spheres having a radius of 0.25 cm to determine theeffects ofthe smaller particles in the center tube. The change in the calculated kf due to thedifferent sized particles occuping the basket center pipe was evaluated.

.52.11 Evaluation of Neutronle Poisoned Center Rod in 4 MCO WP

An evaluation was made to determine if other materials were a better choice for the neutronic* poisoning of the system. In these cases the loading in the scp basket were reduced to 575 kg of

1Uanium, which is the actu loading limitaions on the Mark IA scrap baskets. Previous casesconsidered the most reactive configurations based upon a loading that was slightly higher.The previous cases considered the scrap as being comprised of the uranium from appoximately36 lements. This 575 kg mas imit is equivalent to approximately 34 elements..

;

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4 IkO'sina 4MCO WPMCO'sFully~ooded

NoWaterh4MC0OWP-Betwmc MMO'

Figure 5-16. MCNP Plot - 4 MCO WP - Center Process Tube Included In MCO's

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he first thee cases considered incorporated a 14.4 cm rod located within the center of the 4MCO WP within the ar-y of the 4 MCO's. In the first case the center rod was included asgaphite containing 2 wtl% Gd. The rmaining 4.0cm 1.0 wda Gd poisoned stainless steelcruciform was included in the case. No fuel was present in the center pipe region.

he second case is similar to the first case descbed in this section with thie exception tbat thegraphite center rod was enclosed with a 1.0 cm thick stainless steel shell.

Two variations on the previously outlined cases were also evaluated. The first incorporated a14A cm boron carbide (B 4C) rod in the middle of the waste package instead ofthe previouslydescribed graphite rod. In this case the 4.0 cm thick 1.0 wtK Gd poisoned stainless steelcruciform was evauated. The last case considered only the presence of the BC rod and didn'tinclude the stainless steel cruciform.

An example of these configurations can be found In Figure 5-17.

S2.12 Evaluation of Boron Carllde (B4C) Rods Placed in the Fuel Baskets

Cases were evaluated to determine the effects of placing BjC rods into the MCO in various-configuatimons. The B4C rods were evaluated at the same outer dimensions of the N-Reactor

Mark lA fuel elements. hree rods were placed into a basket at 120-degree intervals., In the firstcase, three rods were placed into each of the six fuel baskets. The next case considered only rods*in the top and bottom scrap baskets. The last case considered B 4C rods in all six baskets with thefuel baskets evaluated with center ppes equivalent to the diameter of the Mark IV baskets.. -

Once agai, the scrap baskets contained 575 kg of uranium. The spacing between the fuelspheres, in the scrap baskets, optimized at a slightly larger center to center pitch due to theconfiguration that included the B4C poison rods.

The cruciform was not included in these cases.

An example of these coiguions is shown in Figue 5-18.

Wn*fA MqrikanA rInAmfinnc FmInpArInn (' ple-mWinnam-po a ma-1-wr -r-1-43011 % . _ll|s|gswss~

Tlte: N-Reactor Spet Nuclear Fuel Citifcality CalculationsDocument denffier. BBAOOOOOD-01717210 -0005 REV 00 of 50

, _ _~~~~~~~~~~~~~

.

7 1 1

Neufardca Poisond Center fled14Aan Elub - I.DanSS Oadh~U-dsC*3O

Figure 5-17. MCNP Prot - Neutronically Poisoned Rod Placed into Center of4 MCO WP

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II

CIO Dkm3 td

=Mr j I

A

4:9UMM!-,-mm 'AA014Mwommomwe�

Figure -S18. MCNP Plot-Neutronically Poisoned B 4C Rod Placed Into MCO Baskets

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. Document IdenUlfer. BBAOOOOOQ-01717-021G 00005REV O0 Pae 38 of 50

6. Results

Exting data were used in the development of the results presented in this section. Therefore,the use of any data fEom this calculation for Iput Into documents supporting p emt,fabrication, or construction Is required to be Identified and tradked as To Be Verified (FBV) inaccordance with appropriate procedures.

6.1 Results from Comparison Cases of Type Mark 1A and Mark IV N-Reactor SNF

Cases were evaluated using both Mark IA and Mark IV N-Reactor SNF as outlined InSection S2. Results from the various comparison cases are presented below.

6.1.1 Results from Intact Mark iA and Mark IV N-Reactor SNF Cases

MCO's and 4 MCO WP's, containing intact Mark IA or Mark IV N-Reactor SNF wereevaluated. In these cases all of the MCO baskets were evaluated as described in Section 5.2.1.1.The results from the various intact cases are given in Table 6-1.

Table 61. Resuts from Intact Mark 1A and Mark W N-Reactor SNF Cases

Input File Type of Fuel Configuration kc±1Ur 1 : +2a |

mkla intactl Mak 1A . Single MCO 0.7922 ± 0.0006 0.7934mk4 intacti Mark IV Sngle MCO 0.8289 + 0.0005 i 0.8299

Idpnmkla intactl Mark IA 4 MCO's in 4 MOO WP 0.8151 + 0.0006 0.8163 a

Ildp mk4intact1 MrkIV 4MC0'sim4MMOWP 0.8411±0.0004 0.8419

6.1.2 Results from Intact and Degrded Scrp Mark 1A and IV N-Reactor SNF Cases

MCO's and 4 MCO WP's, containing both intact and degraded Mark lA or Mark IV N-ReactorSNF, were evaluated. These cases, as described In Section 5.212, were conservatively evaluatedwith the scrap material as optimally spaced spheres with the enrichments corsponding to thefuel type stored. The results from the various mbinations of intact and degraded fuel cases aeglveninTable6-2.

Table 6-2. Results from Intact and Degraded Scrap Mark 1A and Mark IV N-Reactor SNF

Input File Type ofFuel Conflguration k,1flc Ir ______|

mkla combla| Mark IA | Single MCO 0.8783 ±0.0005 | 0.8793mk4 combl Mark IV Single MCO 0.8407 ± 0.0005 0.8417

ldp nkla combla| Mark IA 4 MCO's in 4 MCO WP 0.9026 + 0.0005 0.9036Idpnk4=oombl1 Mark IV 4 MCO's in 4 MCO WP 0.8562 + 0.0004 0.8570

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6.13 Results from Degraded Fuel and Scrap Mark 1A and Mark IV N-Reactor SNF Cases

MCO's and 4 MCO WP's containing degraded fuel and scrap material from Mark lA or MarkIV N-Reactor SNF were evaluated. These cases were evaluated as described In Section 5213.The degraded fuel and the scrap material were both evaluated as optimally spaced spheres withthe enrichments cosponding to the fuel store. Th results trom the warious combinatons ofthe degraded and scrp fuel cases are given in Table 6-3.

Table 643. Results from Degraded and Scrap Mark 1A and Mark IV N-Reactor SNF

Input Fle Type of Fuel Configuration k,1 + la k + 2amklaL spheresla Mark IA SingleMCO 0.8816 ±!0.0005 0.8826mk 4spheresi MarkI V Single MCO 0.8989 + 0.0005 0.8999

ldp ;mklaespheresla Mark IA 4 MCO's in 4 MCO WP 0.9054 ± 0.0005 0.9064I Idp mk4spheresla I Mark IV 4 MCO's in 4 MCO WP 0.9118 + 0.0004 0.9126

6.2 Results of Various Degrees of Moderation within the MCO's and 4 MCO WP

MCO's containing Mak IA fuel, degraded and scrap, were evaluated with various densities ofwaterplaced into the i tial paces between the heterogeneous fuel spheres witin the'MCO'baskets. In these cases, the MCO's were reflected by full density water. The results of thesecases are found in Table 6-4.

I ... Table 64. Results from Cases Incorporating Water Moderator at Various Densities toDetermine Optimum Moderation - Shle MCO

InputFile % Densityof ' k.,lc±les k+2aWater

_within MCOrmkla b2o010%a 10 % 0.3841 ± 0.0003 03847mkla b2o 20°/%a 20 % OA788 ± 0.0004 0.4796mkla h2ao30)1oa 30 % 0.5660±0.0005 0.5670mkla. h2o 40°/%a 40 % 0.6404 ± 0.0005 0.6414mkla h2o_50%a 50% 0.7003+ 0.0005 0.7013mkla b2o_60/ca 60% 0.7514±0.0005 0.7524mkla b2o 70°/%a 70% 0.7939 ± 0.0005 0.7949mkla bh2o,80oa 80 % 0.8286 ± 0.0005 0.8296mkla bh2o 900oa 90 % 0.8561 ± 0.0005 0.8571rklak.spheresla 100 % 0.8816 + 0.0005 0.8826

Once the optimum degree of moderation within the MCO was determined, four MCO's were

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placed into a 4 MCO WP. The water in between the MCO'sv within the 4 MCO WP wasevaluated at various densities (O to 100% theoretical density). The results of these cases aregiven in Table 6S. :

Table 6"5. Results from Cases Incorporating Water Moderator at Varioui Densities toDetermine Optimum Moderation - Four MCO's within a 4 MCO WP

% Densty ofInput File Water within Ic&±1a kdr +2c

4 MCO WP. dp pnkla h2o 0°0oa 00%0 0.9440±0.0004 0.9448

1Idp mnkla h2o, 10Yea 10 % 0.9325 ± 0.00D4 0.9333.ldpemklaa.2o_20%a 20% 0.9248 + 0.0004 0.9256 _ldp klaj2O 30%a 30 % 0.9200 ± 0.0004 0.9208Idp_nlak }2oq..40%a 40% 0.9170±0.0004 0.9178ldpmkla. b2oS'0%a 50% 0.9143 ± 0.0004 0.9151Idp pmkla b2d6_0%a 60% 0.9120 + 0.0004 0.9128_dp mkla h2o_70%a 70% 0.9099 ± 0.000 0.9107ldp nkla bh2o..80%a 80% 0.9094 ± 0.0004 0.9102.:ldppmklab.h2oL90%a - 90% 0.9067 ± 0.0004 0.9075Idpjmkla.pheresla 100% 0.9054 0.0005 0.9064

6.3 Results from Carbon Steel as Moderator Exclusion andlor Neutronic Poison

MCO's containing Mark IA fuel, degraded and scrap, were evaluated with various densities ofcarbon steel placed into the interstitial spaces between fuel within the MCO baskets for etherintact Mark IA elements or the heterogeneous fiiel sphere case of Mark IA elements. In thesecases the cabon steel was combined with water at various proportional densities to occupy thevoid spaces within the basket.

ln the case ofthe intact elements, the carbon steel - water mixture only occpied the radialspaces between the intact fuel elements. The cental void and annular regions within the elementitself were filled with full density water. Th results from these cases are found in Table 6-6.

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Table 6-6. Results from Cases Incorporating Carbon Steel as Moderator Exclusion and/orNeatronic Poison within the Baskets as Carbon Steel-Water Mixture -

Degraded Mark 1A Fuel Basket Center Pipes IntactInput File % Density of % Density of kIC, la Ie+ 2a

Carbon Steel Watermkla;shr.cs9O%a 90% 10% 02456±0.0002 0.2460mklaspr cs80%a 80% 20 % 03053 ± 0.0O02 03057mklaiphr cs7O0M1a 70% 30% 03647±0.0003 03653mkla sphLr. cs60%a 60Y% 40% 0.4237±0.0003 0.4243mklajcpbr. cs50%a 50 % 50 % OA842±0.0004 0.4850mklaL sph cs4OPoa 40 % 60 % 0.5458 ± 0.0004 0.5466*mk1 aphr cs3OMM 30% 70% 0.6146±0.0004 0.6154mkla..pbri..cs2O**a 20% 80 % 0.6901 0.0004 0.6909mkla.sphrb.cslO0oa 10 % .90 % 0.7783 ± 0.0005 0.7793

Intact elements were also evaluated containing various densities of crbon steel placed into theinterstl radial spaces between the intact fuel elements within the MCO baskets. In these casesthe crbon steel was combined with water at various proportional densities to occuy the radial.void spaces between the intact elements within the basket he center and annular regions oftheelement itself were filled with full density water. The results from these cases are given inTable 6-7.

Table 6-7. Results from Cases Incorporating Carbon Steel as Moderator Exclusion and/orNeutronic Poison within the Radial Space Between Intact Mark IA Elements within the

Baskets-Carbon Steel Combined with Water at Various Densities-Basket Center PipesIntact

InputFile % Density of % Density of kd lIcrc 1 2aCarbon Steel Water

mkla intact cs900/oa 90% 10% 0.5129 0.0005 0.5139mkla Intact cs80Ka, 80°% 20% 0.5349±0.0005 0.5359

a iantact=70/oa 70% 30 % 0.5597±0.0005 0.5607mkla intact cs6O°0/a 60% 40 % 0.5839_±0.0005 0.5849mkla intact csSOFea 50 % SO % 0.6114 ± O.O05 0.6124mkla_ intactcs4ooa 40% 60 % 0.6408 ± 0.0006 0.6420mkla intact cs3Ooa 30 % 70 % 0.6722 ± 0.0006 0.6734mklaintact cs2oo1/a 20 % 80 % 0.7070_ ±.0007 0.7084mkla intact.cslO0/a 10 % 90%K 0.7468± 0.0007 0.7482

Waste Package Operations Engineering CalculationThie: N-Reactor Spent Nuclear Fuel Criticality CalculationsDocument Idenflerw BBAO00000-01717-0210-00005 REV 00 of 50

6A Results ofBasket Center Pipe Variation Cases - Mark IA N-Reactor SNF

6A.1 Results from Carbon Steel within Baskets -No Center Pipe In Baskets - Mark IA

MOO's containing Mark IA fuel, degraded and scrap, were evaluated with various densities ofcarbon steel placed into the interstitial spaces between the heterogeneous fuel spheres within theMCO baskets. In these cases, the carbon steel was combined with water at various proportionaldensities to occupy the void spaces within the basket. The baskets were evaluated without thecenter pipes present The cases were evaluated as described in Section 52.5 and the results canbe found In Table 6-8.

Table 6-8. Results from Cases Incorporating Carbon Steel as Moderator Exclusion mnd/orNeutronic Poison within the Baskets Containing Degraded Fuel - Carbon Steel-Water

fMixture at Various Densities -Mark IA fuel-No CenterPipe In Basket

Input File %Densityof %Dnsityof kli ; kl+ 2aCarbon Steel Water __

-mkIclancp.~.cs90O/a 90 % 10% 0.2861 0.0002 02865mlkla ncp s80°/oa 80 % 20% 03562± 0.0003 03568mlkla ncpcs70%a 70 % 30% 0.4222 + O.O0Q3 0.4228mlkla ncpcs6Oa 60%Y 40% 0.4854±0.0003 0.4860nmkla ncppcsS5Oa 50% 50%s 0.5496±0.0003 0.5502mklancpcs40O/a 40 % 60% 0.6154 ± 0.0004 0.6162mkla ncp .cs3Oa 30% 70% 0.6886+0.0004 .0.6894m1cla ncp cs20%a 20% 80% 0.7717±0.0004 0.7725mklajncp .csl0%a 10% 90% 0.8656 ±0.0005 0.8666

mnka ncp j2oa 0% 100% 0.9802 ± O.0005 0.9812

6A.2 OtherBasket CenterPipe Variation Cases -Degraded Mark lAFuel

MCO's containing Mark IA fuel, degraded and scap, were evaluated with the center pipes of theMCO baskts removed, ofiset, or filled with degraded fuel, as descrbed in Section 5.24. Theresults from these cases can be found in Table 6-9.

Table 6-9. Results from Basket Center Pipe Variation Cases - Mark IA Fuel

Center Position of DegradedInput File Pipe Center Pipe Fuel in kIc la k±I + 2a

Present Center Pipemklajipji2oa No Not Present NJA 0.9802±0.0005 0.9812

rIkladlegpfi11a Yes Centered Yes 0.9200 ± 0.0005 0.9210nklajlea.cpoffa Yes Offset 5.08 cm No 0.8978 ± 0.0005 0.8988

Waste Package Operations Engineering CalculationTie: N-Reactor Spent Nuclear Fel Criticality CalculationsDocument Identifier. BBAOOOOOO-01717-0210-00005 REV 00 of 50

6A.3 Results from UJH-20 M Dispersed In the Basket-Degraded Ma A FueI

In order to determine reactivity effects from particulate, a set of cases was developed. In thesecases the MCO was evaluated fifly loaded with degraded Mark 1A fuel. The MCO was fullyflooded and fMly reflected with full density water. The water interspesed within the MCo wasevaluated containing various gram quantities of the ranium fuel matri The results ofthesecases can be found in Table 6-10.

0

Table 6-10. Results from Uranium in Solution - Homogeneously Dispersed throughout theBasket - Degraded Mark 1A Fuel Case

Concentration of Uranium VuelInputFile Matrixin SolutionX dL) l_± la lc; +2cr

mkla deg_0.OOOlgcca 0.1 0.8766 + 0.QOO5. 0.8776mikladtegOQ.OOOSgcca 05 0.8693 + 0.0005 0.8703mkla deg_0.O0lgcca 1.0 0.8707 + 0.0005 0.8717mkla deg_0.O05gcca 5.0 0.8636+ 0.0005 0.8646mkla.deg_0.01gcca 10.0 0.8671 + 0.0005 0.8681

6.5 Results from Zircaloy-2 Cladding as Modentor Exclusion within the Baskets

Cases were evaluated as descnrbed in Section 5.2.7. These cases were developed whichconsidered both fuel contained In the center pipe and fuel not contained in the cent pipe. The 4MCO WP was considered In this case. The individual MCO's were evaluated fully flooded withno water between them within the 4 MCO WP. Full density water reflection was placed aroundthe outside ofthe 4 MCO WP. Two cases were evaluated. One with fuel in the center pipe, andanother cases had no fe in the center pipe. In each of tese cases, the Zhicoy-2 matial wasonly int e in the water in the annular region of the basket.

The results from these cases can be found in Table 6-11.

. .1

.- 0

Table 6-11. Results from Cases Incorporating Zrcaloy-2 from Element Cladding asModerator Exclusion within the Baskets

'Zfrcaloy 2 -Water-Inpt File Fuel Mixture Present k& ±1c k r 2a

_________________ In Center Pipeldppmkla zr h2Poc No 0.9374±0.0004 0.9382

Idpj.nkla cpfll~zr. 2oOY/b Yes 0.9501±0.0004 0.9509

a

Waste Package Operations -. % .Engineering CalculationTil: N-Reactor Spent Nuclear Fuel Cricality Calculations.Docment Identifier. BBAOOOO0 1717.02 1000005 REV 00 of 50

6.6 Results from Non-Polsoned Cruciform between MCO's In 4 MCO WP

A set of cases was developed as described in Section 52.8 to determine kr if a stainless steelcruciform was placed into the 4 MO0 WP in between the individual MCO's. Thickness'fiom 1.0 to 4.0 cm were evaluated, In all of these cases, the fuel was present inside the centerbasket tube. The Zircaloy-2 material was intpersed only in the water between the fuel spheresin the annular region. The results from these cases are given in Table 6-12.

Table 6-12. Results from Cases Incorporating Non-Poisoned Stainless Steel Cruciform

ThIckness ofInput File Stainless Steel kdrf lcr k,+r2a

Cruciform Between._ _ _ _ _ _ _ ___ M CO's .__ _ _ _ _ .- _ _ _

ldp mkla cpflzl_ss1.Oa 1.0 0.9424±0.0004 0.9432ldp_;nkla cpfillzr ss2.0a 2.0 0.9402±0.0004 . 0.9412*:Idp.mnkla OcpfilLz ss3.0a 3.0. 0.9377 ±0.0004 0.9385Idp jkla.cpfihllzr.ss4.0a 4.0 -0.9363 ± 0.0004 0 0.9371

6.7 Results from Gadolinium Poisoned Cruciform between MCO's In 4 MCO WP

A set of cases was developed as descrbed in Section 52.9 to determine kc1, ifa neutronicallypoisoned stainless steel cruciform was placed into the 4 MCO WP in between the individualMCO's. A thickness 2.0 cm was chosen as the base case to be evaluated. Various wt% ofgadolinium was incorporated into the stainless steel cruciform. In all of these caseS the fuel waspresent inside the center basket tube. The Zrcaloy-2 material was interspersed only in the waterbetween the fuel spheres in the annular region.

The results from these cases are given In Table 6-13.

Table 6-13. Results from Cases Incorporating Poisoned Stainless Steel Cruceform

wt%ofGdInput File Incorporated In IC&+a1cT dr 2

Stainless SteelCruciforn Between

_ ______________ MCO'sIdp nmklcl lpfillzr Ogd.la 0.01 0.9484 +0.0004 0.9492ldp_;nkla.cpfilLzr.ss2.OgdO.la 0.1 0.9470 0.0004 0.9478ldpmklsppfillzr. ss2.Ogdl.Oa 1.0 0.9446+0.0003 0.9452

Waste-Package Operations- Engineering Calculation-- -

Waste Package Operations Engineering CalculationTlte: N-Reactor Spent Nuclear Fuel Oiticality Calculations

. Document Identifler. BBAOOOOO-01717-0210-00005 REV 00 of 50

A last set of cases was evaluated in which only water, and no Zircaloy-2, was present betweenthe fu~el sphees located i the cetter pipe. The results from these cass are given in Table 6-14.This case consisted of 4 MCO's in the 4 MCO WP as previously outlined. The neutronicallypoisoned 2.0 cm thick stainless steel cruciform was included.

Table 614. Results from Zfrcaloy-2 between Fuel Particles In Annular Region Only

w._t% of GdInputFile Incorporatedin kIre1cr k+2o

Stainless SteelCruciform Between

._ MCO's .Udp mk1a Ocpiil..ss2.0gdO.01 0.01 0.9408 ± 0.0003 0.9414

qlkla cpfllrzrss2.OgdO.1 0.1 0.9392 ± 0.0004 0.9400Idp mlclascpfillfzr.ss2.0gdl.0 1.0 0.9367 ± 0.0004 0.9375

6.8 Results from Cases Incorporating 304L Stainless Steel Center Process Tube

The last set of cases considered consisted ofthose descnbed in Section S-10. In these cases,the change in kT, due to the presence of the stainless steel central process tube, was evaluated.The first subset of cases consisted of the region between the center process tube and the inside ofthe basket center pipe being filled with fiel spheres at a radius of OA2925 cm. This was themuaum posfiduted radius prdicle size, and shown to be most reactive in the normal baseconfiguration cases.

No Znraloy-2 material was present in the wat that was interspersed between the fiel spheres inthis centra region. Zircaloy material was evaluated in the water that was inespersed betweenthe fuel spheres in the basket annular region. Once again four flooded MOO's were evaluatedcontaied in a 4 MCO WP. No water existed between the MOO's within the 4 MCO WP. Theneutonically poisoned 2.0 cm thick staless steel plate was also present

Table 6-15. Results from Inclusion of Center Process Tube - 0A292S cm Radius Fuel

wst% of GdInput File Incorporated In kIC1cr kf +2cr


.__ _ _ _ _ _ _ _ _ _ M CO '9

Idp pikla cpfll~z ss2.OgdO.Olb 0.01 0.9442 + 0.0004 0.9450Idp npklakcpfiuzr.ss2.0gdO.lb 0.1 0.9438±0.0004 0.9446Idp_#lacpfilOzr.ss2.Ogdl.Ob 1.0 0.9412 ± 0.0004 0.9420

This set of results outline the effect on I due to replacing the Lager fuel spheres (0.42925 cm

Waste Package Operations Engineering CalculationTe: N-Reactor Spent Nuclear Fuel Criticality Calculations,Docment Identfer. BBAOOOOOO-01717-02100OOO REV 00 o0SO

radius), located in the central region, with the smaller 025 cm radius fuel spheres. Due to thepresence of the center process tube an argment can be made that this is the largest sized particlethatcan migrate into the central region outside of the center process tube and inside of the basketenter pipe. The fourth case listed varies from the first three cases in that no zircalloy wasevaluated in between the heterogeneous spheres in the scrap inserts. The results fiom this set ofcases are given in Table 6-16.

Table 6-16. Results from Inclusion of CenterProcess Tabe -0.25cm Radius Fune

wt% of GdInput File Incorporated in kli la kr+ 2a


_ _ _ _ _ _ _ _ _ _ _ ___ M OC O 's _ _ _ _ _ _

ldp mkla cpfllLzrss2.0gdD.Olc 0.01 0.9414 ± 0.0003 0.9420Idp mkla cpfillzrss2.OgdO.lc 0.1 0.9406±0.0004 0.9414lpjnklad pfcilzrl ss2.Ogdl.Oc 1.0 0.9383 ± 0.0004; 0.9391ldp ;mkla.cpfif.zr._s2.Ogdl.Od 1.0 0.9492 ± 0.0004 0.9500

6.9 Results from PlacementofNeutronicagy Poisoned Rod iCenterof4 MCO WP

These ae the results fiom those cases as outlined in Section 5.211. As described in Section52.11lanevahlaionwas maderelatngtotheeffecofplacinganeutonicay poisonedrod inthe center of the 4 MCO WP. In these cas no fuel was placed in the center pipe of the baskets.Full water reflection exists outside ofthe 4 MCO WVP with no water within the 4 MCO WE. The

results we given in Table 6-17.

Table6417.Results from Inclusion ofNeutronicany PobonedCenter Rod

Input File Cruciform Material of Rod k.;f la r+ 2crPresent

inp42b Yes Graphite 0.9076 0.00O4 0.90842 wt% Gd

aiteimp42d Yes 2 vPO/ Gd 0.9073 ± 0.0004 0.9081

Enclosed In SSinp42c Yes B4C 0.8976 ± 0.0004 0.8984

No B4C 0.9112±0.0004 0.9120

Waste Package Operations Engineering CalculatIonTie: N-Reactor Spen Nulear Fuel Criticality CalculationsDocument Identfier: BBAOOOOO-01717-0210-00005 REV 00 ; age47 of SO

6.10 Results from Placement of B 4C Rods In Fuel Baskets

These are the results from those cases as outlined in Section 52.12. As descnbed in Section52.12 an evaluation was made relating to the effect of placing a neutronically poisoned rod inthe fuel baskets of the MCO's. In these cases no fuel was placed in the center pipe of thebaskets. Full water reflection exists outside of the 4 MCO WP with no water within the 4 MCOWP. The results are given in Table 6-18.

Table 6-I. Results from Iclusion of Neutrnecally Poisoned Center Rod

InputFile Location of B4C- Center Pipe of I1ff + Ur kc + 2crRods Fuel Basket

b4cmooa, Top & Bottom MakIM A J 0.8583 ± O.Q004 0.8591b4cmco Evey Basket Mark IA 0.9172 ± M.OON4 0.91890

bc4mcob Every Basket Mark IV j0.9171 ± O.O04 0.9179

I .

*,'

Waste Package Operations Engineering CalculationTtle: N-Reactor Spent Nuclear Fuel Ctiticaly CalculationsDocument Identifier. BBAOOOOO-01717-0210-00005 REV 00 of SO

7. References

7.1 [Reserved].

7.2 Los Alamos National Ldboaoy (LANL) 1997. MCNP-A General Monte Carlo Codefor Neutron andPhoton Transport Code, Verslon 4B - UC 705 and UC 700. LA-12625-M, Version 4B. Los Alamos, New Mexico: LANL. ACC: MOL.19980624.0328.

73 Cvlian Radioactive Waste Management System (CRWMS) Management and Operating(M&O) Contractor 1998. Software uaflication Reportfor the MCNP Version 4B, A

* General Monte Carlo N-Partcle hmport Code. Computer Software ConfigurtionIdentifier (CSC1): 30033 V4BLV, Document Identifier Number. 30033-2003 REV 00.Las VegasNevada: M&O. ACC: MOL.19980217.0561.

7;4 CRWMS M&O 1996. Material Compositions andNumber Desies For NewronicsCalculations. BBAOOOOO-01717-0200-00002 REV.00. LasVegas Nevada- M&O.ACC: MOL.19960624.0023.

7.5 Fluor Daniel Northwest 1997. Criticallty Safeny Evaluation Reportfor SpentNuclearFuel ProcessingandS&orage FaciitleL WHC-S-SNF-CSER-005, Rev.3. Hanford,Washingtx. FluorDanielNorthwest ACC: MOL.19980729.0238.

7.6 [Reserved]

7.7 Parsons Infrastruc and Technology Group, Inc. 1997. Mult-Canister IverpackDrawlng4 HNF-SDF-DR-003, Rev. 0 Appendices 1 and 19, Draf Hanford,Washington' Parsons Infrastuctur and Technology Group, Inc. TIC: 239396.

7.8 CRWMS M&O 1998. Software Qua ation Report for the MCNP Version 4B2, AGeneral Monte Carlo N-Partcle 7)ansport Code. Computer Software ConfigurationIdentifier (CSCI): 30033 V4B2LV, Document IdentifierNumber. 30033-2003 REV 00.as Vega Nevada. M&O. ACC: MOL 19980622.0637.

7.9 [Rsered

7.10 LANL 1987. Crical Dimensions ofSystems Containing 235U, 239Pu, and 233U, 1986Revision. IA-10860-MS. Los Alanos, New Mexico: LANL. MC: 209447.

Waste Package Operations .. Engineering CalculationTile: N-Reaco Spewt Nuclear Fuel Qiticality CalctmsDocument IdentIfier. BBAOOOOOO-01717M210-00005 REV 0O of 50.

7.11 General Electric (GE) NuclearEnergy 1996. Nuclides and Isotopes, Fifteenth Edition.San Jose, California: GE Nuclear Energy. TIC: 233705.

7.12 CRWMS M&O 1999. Electronic Media for. N-Reactor Spent Nuclear Fuel CQ*cafltyCalculations: Intact Mode. BBAOOOOOO-01717-210-00005 REV 00. Compact Disk.Las Vegas, Nevada: M&O. ACC: MOL.19990225.0001.

Waste Package Operations Engineering CalculationMle: N-Reacor Spent Nudlear Puel Criticality CalculationsDocument Identfier. BBAOOOOOO-01717.02 1000005 REV 00 - of 50

8. Attachments

The hardcopy attachments are listed in Table 8-1 below. Electronic attachments are provided oncompact disc and are listed in Table V-I forExcel Spreadsheet, Table VI-1 for Rev OOA MCNPcases and Table VII-l for Rev OOB MCNP cases (Ref. 7.12).

Table 8-L Attachments of Supporting Documentation for N-Reactor SNF DisposalCriticality Analysis

achment Number Description PagesI Summary Tables of Cases Used to Determine Optimum Particle 4

size and SpacingH Surmaiy Table of Cases Used to Determine Most Reactive 1

Loading Per MCO BasketIII Summary Table of Cases Used to Evaluate Various Neutronic I

Absorber ConfigurationsIV Summary Table of Cases Used to Detemine Optimum Pitch I

and Particle Sizes for Cases in which the Scrap Baskets wereLimited to 575 kg of U

V Summary Table and Excel Spreadsheets used in Engineering 19.________ Calculations (Also Stored Electronically.on CD)

VI Summary Table of Rev OOA MCNP4B case inputs and outputs 9stored in electronic format

VII Summary Table of Rev OOB MCNP4B2 case inputs and 9outputs stored in electrnic format

VIII Drawing/Sketch of the Physical Dimensions of the MCO Waste IPackage

Waste Package Operations Engineering CalculationTMle: N-Reactor Spent Nuclear Fuel Criticality CalculationsDocument Identf er: BBAOOOOOO-01717-0210-00005 REV 00 Attachment IPage 14

Attachment L Parametric Study to Determine Optimum Pitch for HeterogeneousN-Reactor Fuel Sphere Calculational Cases

J1

A parametric study was completed to determine the most reactive (optimum) sized sphere andmost reactive (optimum) spacing between spers for the N-Reactor heterogeneous fuel spherecase. Due to equestionable sructul init of the N-Reactor fel a very conseraiveapproach was taken in this evaluation.

This approach entailed dispersing heterogeneous pieces of fuel thrmughout the entire basket. Theoptimum spherical size was determined along with the optimm spacing between spheres. lhispammetric study considered an infinite by infinite by infinite array of f spheres. The.madmum reactiviwas determined.

The sizes of the particles (spheres) were limited to the thickest radial thickness of the actual fiuelmatrix, as constrained by the Zircaloy cladding. The results fiom the various particle sizes andspacing configurations arc given below.

i

Table I-l. N-Reactor Spheres, Square Pitch Infinite LatticeOA2925 cm Radius for Spheres

Output Fiename Distance Between Distance BetweenSpheres Centerto Spheres Edge to kdr _la Yka+2aCenter (C-C (cm) Edge (E-)(cm)

0.9cm.o 0.9 0.0415 1.0475 0.0006 1.0487I.Ocm.o 1.0 0.1415 1.1193±0.0006 1.1205l.lcM.o 1.1 02415 1.1341 ±0.0005 1.13511.2cm.o 12 0.3415 1.1213 ±0.0004 1.12211.3cm.o 1.3 0A415 1.0874 0.0004 1.08821.4cm.o 1.4 0.5415 1.0393+0.0004 1.04011.Scm.o 1.5 0.6415 0.9814±0.0004 0.98221.6cn.o 1.6 0.7415 0.9197±0.0004 0.9205

Waste Package Operations Engineering Calculation

I.

Title: N-Reaor Spent Nuclear F Criticality CaulatfonsDocument Identlfier. BBAOOOOO-01717-210-00005 REV 00 .. , Affachment1, Page 1.2

Table 1-2. N-Reactor Spheres, Square Pitch Infinite LatticeOA2 cm Radius for Spheres

Output Filename Distane Between Distance Between.___________ Spheres C-C (cm) Sphems E-E (cm) kf + 1c l. k+ 2cr

0.BScm.o 0.85 0.01 1.0050±0.0007 1.00640.9cmao 0.9 0.06 1.0628 ± 0.0006. 1.0640L.Ocmo * 1.0 0.16 1.1204±0.0006 1.1216Mml.lco 1.1 0.26 1.1325±0.0005 1.13351.2cmo 12 0.36 1.1137±0.0004 1.114513cmco .13 0.46 1.0748+0.0004 1.07561.4c=io 1A. 0.56 1.0222 + 0.0004 1.02301.ScmLo 1.5 - 0.66 0.9619±0.0004 0.96271_ c lo 1.6 0.76 0.8979±0.0004 0.8987

Table 1-3. N-Reactor Spheres, Square Pitch Infin te LatticeOA1 cm Radius for Spheres

Output Filename Distance Between Distance BetweenSpheres C-C (cm) Spheres E-E (cm) k.C +lar k.fr+ 2a

0.85cmLo 0.85 0.03 1.0305 ± 0.0006 1.03170.9cm.o 0.9 0.08 1.0789±0.0006 1.08011.Ocmno 1.0 0.18 1.1265+0.0005, 1.12751.lcmeo 1 0.28 1.1303 +0.0005 1.1313

1.2cm.o 1.2 0.38 1.1032+0.0005 1.10421.3cm.o 1.3 0.48 1.0589 + 0.0004 1.0597lAcm.o 1.4 0.58 1.0031+0.0004 1.00391.Scm.o 1.5 0.68 0.9409±0.0004 0.9417

1.6cm.o .1.6 0.78 0.8738 + 0.0004 0.8746

Waste Package Operations Engineering CalculationMe: N-Reactr Spent Nulear Fuel cality CelculatiosDoument Idenlfier. BBAOOOOOO-01717.021000005 REV 00 . Attachment I, Page 1.3

Table I4. N-Reactor Spheres, Square Pitch MmiIte LatticeOA cm Radius for Spheres

Output Filem Distance Between Distance Between.___________ Spheres C-C (cm) Spheres E-E (cm) k.r ± la kff + 2c

0.8cm.o 0.8 0.0 0.9893 ±0.0006 0.99050.9cm.o 0.9 0.1 1.0938 ± 0.0006 1.0950l.Ocmo 1.0 0.2 1.1298±0.0005 1.1308l.lcmLo 1.1 0.3 1.1248±0.0004 1.125612cm.o 1.2 0.4 1.0928±0.0004 1.09361.3cm.o 1.3 0.5 1.0429+0.0004 1.0437lAcm.o 1A 0.6 0.9829 + 0.0004 0.9837.5cmno 1.5 0.7 0.916+±0.0004 0.9176

Table I-6. N-Reactor Sphercs, Square Pitch Infinite Lattice038 cm Radius for Spheres

output Filenamne D- ane Betweei Distance Between___________ Sphers C-C (cm) Spheres E-E (cm) k; ± 1a ki + 2cr

0.8cm.o 0.8 0.04 1.0414±0.0006 1.0426. 0.9cm.o 0.9 0.14 1.1146±0.0005 1.1156

* l.Ocm.o 1.0 0.24 1.1299 ± 0.0005 1.1307. .lcn.o. 1.1 0.34 1.1071±0.0004 1.1079l.2cm.o .1.2 0.44 1.0625±0.0004 1.06331.3cm.o 13 0.54 1.0029+0.0004 1.0037IAcKILo 1.4 0.64 0.9355+0.0004: 0.9363.cULo 1.5 0.74 0.8659+0.0004 0.8667

. .

I

Waste Package Operations Engineering CalculationTitle: N-Reactor Spent Nuclear Fuel Qiticality CaclationsDocument Identifier. BBAOOOOOO-01717.021040005 REV 00 Attachment L Page 14

Table I6. N-Reactor Spheres, Square Pitch Infinite Lattice036 cm Radius for Spheres

Output Filename Distance Between Distance BetweenSpheres C-C (cm) Spheres E-E (cm) kf ± la kff + 2a

0.8cmLo 0.8 0.08 1.0811±0.0006 1.08230O9cm~o 0.9 0.18 1.1252±0.0005 1.1262l.OCa.o 1.0 028 1.1197±0.0005 1.1207.lImUo 1.1 0.38 1.0825±0.0004 1.0833

1.2cmLo 12 0.48 1.0254 + 0.0004 1.02621.3cmLo 13 0.58 0.9579+0.0004 0.95871.4cno IA 0.68 0.8837±0.0004 0.8845

, .

Table 1-7. N-Reactor Spheres, Square Pitch Infinite Lattice0.34 cm Radius for Spheres

Output Filename Distance Between Distance BetweenSpheres C-C (cm) Spheres E-E (cm) kdf ± ic kf + 2a

' 0.7cmLo 0.7 0.02 1.0823 0.0006 1.08350.8cmLo 0.8 0.12 1.1123±0.0006 1.12350.9cm~o 0.9 0.22 1.1313±0.0005 1.1323l.OcULo 1.0 0.32 1.1066±0.0005- 1.1076Mc1.l=o 1.1 0.42 1.0543±0.0004.. 1.05511.2cm.o 1.2 0.52 0.9867 + 0.0004 0.98751.3cn.o 1.3 0.62 0.9104 + .0004 0.9104

Waste Packape Operations Englneering CalculationTitWe: N-Reaor Spent NulearFuel Citicality CalculationsDocument Identfier. BBAOOOOO-01717-0210-00005 REV 00 Aftachment 1I Page 11.1

AttachmentIL Parametric Study to Determine Optimum Loading for HeterogeneousN-Reactor Fuel Sphere Calculational Cases

Machmnent I considered infinite anays of fuel spheres to determine the most reactive (opimum)spacing versus size for the heterogeneous spheres of N-Reactor fuel. As shown in Attachment Ithe 0.42925 cm radius Mark IA fuel sphere was the most reactive- Another set of cases wasevaluated in which the spacing between the spheres was varied, while constained by thegeometry of the fuel baskets and MCO. In these cases the 0.42925 cm radius fuel sphere wase evaluated with the actual number of spheres present in the basket varying depending uponspacing, thus altering the fuel loading in each of the baskets. The results ofthese cases for the

- Mark A fal are given I Table l-l.

Table [-1. Sunmamy Table of Cases Used to EvaluateMost Reactive (Optimum) Spacing Between Particles

Equivalent to#Out Filenamne Spacing Between of Mark A kr± Ic - . k + 2a

Spheres (cm) Elements p_ _ _ _ _ _ Basket.

0.9cmO 0.9 65.5 0.8122 +0.0005. 0.81321.Ocm.o 1.0. .45.0 0.8675+0.0005 0.8685

_.05cm.o 1.05 412 0.8801±0.0005 0.8811l.lcnLo 1.1 36.0 0.8816±0.0005 0.8826

1.15cm.o 1.15 31.3 0.8783 ±0.0005: 0.879312cm.o 1.2 27.5 0.8680±0.0005 0.8690

: . I . ..

As shown by these results, the highest kff was not achieved with the maximum fuel loading perbasket (48 elements) but with a lower loading at a more optimized pitch. Ihe most reactive pitchfor these specific geometrical configurations compared well with those determine in the infiniteaay study completed in Attachment I. Ibis arangement was therefore used in the evaluation ofthe cases for this calculation.

Waste Package Operations Engineering Calculationlitle: NReatow Spent Nuclear F Citicality CalculationsDocument IdenUtfer. BBA00000-01717M10-OOQOS REV 00 A-achment Il Page 1111

Attachment i1L Evaluation to Determine Effects of Various Neutronic PolsoningConfigurations, N-Reactor Fuel Sphere Calculational Cases

An evaluation the effects of placing B4C neutronic poison in various locations within the MCOwas performed. The results fiom these cases are given in Table M-1.

Table rn-i. Summazy Table of Cases Used to Evaluate Various Neutronic AbsorberConfigurations.

Output Spher Unit Cell Absorber Form Absorber Jnner kff 2aFilename Radius Pitch and Outer

0.4 1.0782 3 B4C rods in 0-3.053 0.9777± 0.000middle four

basketsOA 1.0782 3 B4C rods in ii 0-3.053 0.9161 0.0007

__ basketsoutubea, 0.4 1.0782 cylindrical B 4C 3.6-6.39706 0.9294± 0.007

shell aroundcenterpotin all

.____ _ .basketsoumewta 0.4 1.0782 Cylindrical B 4C 3.6-6.6 0.9269±0.0007

shell aiidCentexrpost in all

_ _ _ _ _ _ _ _ _ _ _ _ _ _ _b ask ets:_ _ _ _ _ _ _ _ _ _ _ _

I

I

Waste Package Operations Engineering CalculationT~le: N-Reactor Spent Nuclear FIel Ctiicality CalculationsDocument Identfier. BBAOOOOOO-01717-I10-00005 REV 00 Aflachment IV, Page IV-1

Attachment IV. Parametric Evaluation to Determine Optimum Pitch and Particle Size forthe Cases Evaluated With Limited Mass of Uranium, N-Reactor Fuel Sphere Calculational

Cases

An evaluation the effects of arying the particle size and pitch between particles in the MCO wasperformed. The results fiom these cases are given in Table IV-1.

Table IV-1. Summary Table of Cases Used to Evaluate Pitch and Particle Size

Ou~tPu Particle Unit Cell Pitch Sphere Unit Cell k1tE2aFilename Radius in Top In Top & Radius in Pitch in

& Bottom Bottom . Middle MiddleBskeas Baskets Baskets Baskets

. (CM' (Cm) (C) (C)owaal OA 052581 0.4 0.48276 0.9307*0.0007outaa2 0.4 052581 OA 0525 0.9323 * 0.0007outaa4 04 052581 0.42 05069 0.9312:*0.0007outaaS 0.4 052581 0.42 054 0.9326*0.0008outaa7 0.4 0.52581 0.42925 0.51807 0.9312* 0.0007outans 0.4 052581 0.42925 0.55 0.9332* 0.0008outbal 0.42 05521 OA 0.A4276 0.9309* 0.007outba2 0.42 05521 OA 0.525 0.9323 :0.0008outba4 0.42 0.5521 0A2 05069 0.9323*0.0007outba7 0.42 05521 0.42925 051807 0.9322 *0.0008outbag 0.42 05521 0.42925 055 0.9330 *0.0008outcal 0.42925 056426 0.4 0.48276 0.9314:* 0.0008outca2 .OA2925 056426 0.4 0525 0.9340 * 0.0007outca4 0.42925 056426 0.42 05069 0.9321*0.0008outca5 0.42925 056426 OA2 054 0.9339:t0.0007outca7 0.42925 056426 .OA2925 0.51807 0.9315: O.0008outea .OA2925 0.56426 0.42925 0.55 0.9341 *0.0007

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Waste Package Operations Engineering CalculationTitle: N-Reactor Spent Nuclear Fel CQticalt CalculatonsDocument Ideniffler. BBAOOOOOO-01717-0210t OOS REV 00 Attachment VI, Page VI-1

Attachment VITable VI-1. Rev. ODA MCNP Inputs and Outputs used in Engineering Calculations Stored

in Electronic Format on CD (Ret 7.12) ' -

Fie Sie Date Lest Access Fle Name Table Used(bytes)

11,867 11-02-98 1:21P ldp-x.klaJntaCtl 6-1354.749 11-02-98 1:21P 1 1dp.mklajntac1.d 6-110,086 11-02-98 1:21P ldp-jn4 Jntactl 6-1365,249 1102-98 1:21p Idp-mk4jntact1o 6-1

2329 11-02-98 1:21P inklaintactl 6-1348,211 11-02-98 1:21p mkllajnactl.o 6-17,097 11-02-98 1:21p mkl4ntal 6-1346,702 11-02-98 121P mk4nftcl.o 6-1

11,820 10-12-98 10.02a Idmarklas.combla 602369,698 10-1298 10:02a 1dpjmk1a-ccmb1ao -210,245 10-12-98 102a ldpjnml4_combl 6-23698,00 1G-12-98 10.02a Ipdink4 comb1.o 6-28,420 10-12-98 1002a mkla.combla 6-23S2.439 10-12-98 10:02a mkla.combla.o 6-27.148 10-12-98 10.02a mk4_combl 6-2348,8so 10-12-98 1O:a mk4_combl. 6-2

11,849 11-02-98 1:45p Idpmkla.spheresla 6-3365,4S5 11-02-98 1.4SP dpklak.sPhe=sla.o 6-310,300 11-02-98 14Sp dP-nk4-splheresl 6-3364,805 11-02-98 1:45p Wdpj n4_hsl.0 6-38,435 11.02-98 15p I mklaspheresla 6-3350,483 11-02-98 1.45p mkla.sphereslao 6-38,715 10-12-98 10:02a mklasphereslb 6-3348,033 10-12-98 10:02a mkla.sphereslb.o 6-37,183 11-02-98 1:45P p mk4sphersl 6-334S,798 11-02-98 1A45p ml.sphesl.o 6-3

8,600 10-12-98 10:03a mkla_b20j_0%a 6-4355,924 10-12-98 10:03a mlnla2o.10%o 6-48,600 10-12-98 10:03a mtklaji2.20%a 6-43S2,475 10-12-98 10:03a mkla.Ji220%ao 6-48,600 10-12-98 10:03a mkla h20_30%a 6-4352,673 10-12-98 10:03a Imklahi2O30%aLo 6-48,601 10-12-98 10:03a mkla.h20_40%a 6-4353,394 10-12-98 10.03a mklah_20 40%a.o 6-4

Waste Package Operations Engineering CalculationTife: N-Rmac Spen Nuclear Fud Ctimcality CalculationsDocument Identfier. BBAOOOOOD-01717-0210-O0 REV 00 Aitacbmnt 'A Page V2

File Size(bytes)

IDate Last Access Hie Name Table Used

8,S60 10-12-98 10:03a mk1naJi205O%a 6-4353,308 10-12-98 10:03a nmkla h2O0_50%ao 6-48,600 10-12-98 10;03a InklaJi20..60%a 6-4352,672 10-12-98 10:03a mklai2.W,60%ao 6-4,-600 10-12-98 10C03a I =kai2_70%a 6-4

352,672 10-12-98 10:03a mkla_b2O_70%ao 6-48,60 10-12-98 1003a mklajib2080%a 6-4350,566 10-12-98 I0.03a 3 mkla h20_80%ao -48,600 10-12-98 10.03a mklaJb20.90a 6-4350,530 10-12-98 10:03a mklaJi2O.90%a~o 6-48,435 10-12-98 10:03a mldasphersl 6-4

12,017 11-02-98 1:49p 1dpn*kla 1i2o_0%a 6-5368,652 11-02-98 1:49p Idpnklaji2o_0%a&o 6-512,030 11-02-98 -1:49p IdppmklaIh2_Ol%a 6-5368,548 11-02-98 .1:49p Idp.mka~L2o605o 6S12,030 11-02-98 1.49p Ildpklaji2o_20%a 6-5368,541 11-02-98 1:49p ldpm klal20_o20%ao 6-512,030 11.02-98 1:49p ldpjmklab2o0_30%a 6-5369,821 11-02-98 1:49p ldp-jnk1a&J2o030%ao 6-512,030 11-02-98 1:49p Idpmklaa 12o 40%a 6-5

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369,821 11-02-98 150p ldpjpkmlaji2qS0%Lo 6-5

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8,864 11-02-98 1:50p kIla.sphr-cs1O%a 6i6351,Q07 11-02-98 150p mkla.sphrcslOa.o 6-68,866 11-02-98 150P mkla.spbr-cs2O%a 6-6352,195 11-42-98 :SOp nkia.spphrqcs2O%ao 6-68,864 11-02-98 I50p Imklasphr-cs30%a 6-6352,003 11-02-98 1:Sfp mk1a.sphr.cs30%Lo 668,866 11-02-98 150p mklajsphr..cs4Oa 6-6352,297 11 -98 150p nkla-spbtrcs4Oao 6-68,864 11-02-98 150p mkla_sphr csSO%a 6-6

Waste Package Operations Engineering CalculationTitle: N-Racto Spen Nuclear FeL Criticali CalculatonsDocument Identifier. BBAOOOOOO-01717-0210-00005 REV 00

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A-tachment VI, Page VI-3p p

Ele Size006S)

Date Lost Access I le Name Table Used

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8,824 11-02-98 1 50p mkklaJntact-sOa 6-7348,504 11-02-98 1:50p mk}lajntcslO%ao 6-78,824 1-02-98 1 50p mklajnt__cs2O%a 6-7348,504 11-02-98 151p mklajntd cs20%ao 6-78,824 11-02-98 1:51p mklajntact-cs30%a 6-7372,305 11-02-98 1:51p mklajntacics3O%ao 6-78,824 11.02-98 151p mwklantact.cs4.0%a 6-7347,220 11-02-98 1 51P mklajtacdcs40%ao 6-7g8,24 11-02-98 1 51p mklaintactcs5Oa 6-7348,408 11-02-98 1:51p mkllajntactcs5O%ao 6-78,824 11-02-98 1:51p mklaJntactcs60%a 6-7346.164 11-0-98 1:51p mkla CiCS6O%Lo 6-78.824 11-02-98 1:51p mklajntacics7O%a 6-7348,504 11-02-98 1-51P mkla Intact-cs70%ao 6-78,824 11-02-98 1*51p mklajntaci.cs8O%a 6-7348504 11-02-98 151p mkanJntactcsSO%ao 6-7,824 11-02-98 151p mklantactLcs9Oa 6-7

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8,825 11-02-98 1:Slp mklancpcslO%a 6-8350,271 11-02-98 -:51p mklanpcs1O%a.o 6-88,824 11-02-98 1:51P mklajncp.cs2O%a 6-8351459 11.02-98 :51p mklaincpscs20%ao 6-88,824 11-02-98 151p nmklaincp.cs30%a 6-8350,589 11-02-98 151p mklalcpS.cs30%ao 6-88,824 -11-02-98 151p mklajncpscs40%a 6-8350Q582 11-02-98 1:51p mkla.ncp.cs4O%ao 688,824 11-02-98 1:51p mkla %a68352,731 11-02-98 1:51p mklaicpcsSO%ao 6-88,824 11-02-98 1:51P mnkla.ncpcs6O%a 6-8352,732 111-02-98 1:51p mkl4jncp.cs6O%ao 6_ _

8,827 11-02-98 1:51p mikla_ncpcs7O%a 6-8356,004 11-02-98 1:Slp mk~llacpcs7O%a.o 16-8

Waste Package Operations Engineering CalculationTMe: N-Reactor Spent Nuclear Fuel Criticality CalculationsDocument Identifier. BBAOODOD041717-0210 0005 REV 00 Attachment M, Page VI4

m, .

HIe Size(bytes)

Date Last Access Mie Name Table Used

mm 11-02-98 1:51p mkla-ncp-CS80%a 6-8357,492 11-02-98 1:51p mklajwcp_cs8oaO 688,2 11-02-98 1:51p mklajicp-cs90%a 6-8359,655 11-02-98 1:51p mklajncp-cs90%ao 6-8.8,394 11-02-98 1:51P mkla.ncpji20a 6-8350,598 11-02-98 1:51p mkla.ncpji2oao 6-8

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8,770 11-02-98 152p mkcla.deg.0.0001goca 6-10352,032 11-02-98 1:52p kla.deg.0.OOlgccao . 6-108,770 11-02-98 1:52p mkla&deg_0.0005gcca 6-10350,936 11-02-98 1:52p mkla-deg-0.0005gccamo 6108,769 I1-02-98 1:52p mkIa.deg_.0001gca 6-10352,130 11-02-98 1:52p ikla.deig_0.001gccLo 6&108,769 11-02-98 1:52p znkla.degO0.O5gcca 6-10350324 11-02-98 1:52p mkla...deg..O0S0goca o 6-108,767 11-02-98 1I52p pk1sadegO.01o~cai. 6-10350,194 11-02-98 152p rkladegO.Olgzccao 6-10

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14,435 11-02-98 153p ldp-xikla.-cpfil-zr-ss2.Og&Ola 6-13372,682 11-02-98 I:53p ldpmkla-cpfill~zrss2.OgdOlao 6-1314,432 11-02-98 1:53p 1dpmkla..cpfillzrSs2.0gdO.1a 6-13378,095 11-02-98 1:53p Idp_ lcpfiUzr ss2.01Q.lao 6-1314,435 11-2-98 1:53p Idpjnkla_cpffll zvss2.dl.Oa _ 6-13379,615 11-02-98 1:S3p Idp mkla-cpfill-zr-ss2.Ogdl.Oa1o 6-13

Waste Package Operations Engineering CalculationTitle: N-React Spent Nuclear Pud Citicaflty CadclaionsDocument ldentier. BBAOOOOOO-01717-0210-0005 REV 00 Attachment VI, Page WS

Mie Size(bytes)

Date Last Access FMe Name Table Used

14,435 11-02-98 1:54P Idp-mMla-cpf lUx-vss2.Og4.Ola 6-14372,682 11-02-98 1.54p Idp xmklA.cPfilHzr.s2.Ood.Olao 6-1414,432 11-02-98 154p 1dpjxakl acp zr waftd0.a 6-14378,095 11-02-98 1:54p IdpjraklaspailjdXs2.OgdO.lao 6-14.14,435 11-02-98 1:54P IdP-mklascPfl zr. a.aOgdl.Oa 6-14379,615 11-02-98 1 54p ldpjnkla.._cPfill zcs2.0gad.0a.o 6-14

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10,647 11-02-98 I S5p b4cwp 6-17362,900 11-02-98 165p b4cwpo 6-1711,093 11-02-98 I:SSP Inp42b 6-17364,472 11-02-98 I55P inp42bo 6-1711,132 11-02-98 1:55p inp42c 6-17364,619 11-02-98 1:55p lnp42co 6-1711,178 11-02-98 155p hp42d 6-17364,695 11-02-98 1:55p inp42do 6-17

10,803 11-02-98 1:SSp b4cmco 6-1811,043 11-02-98 1:55p b4cmcoa 6-18366,168 11-02-98 1:.5p .Mcoao 61811,042 11-02-98 1:S5p b4cmcob 6-18367,164 11-02-98 1I55p b4cmcobo 6-18364,351 11-02-98 1:55p b4cmcoo 618

1,392 11-02-98 11:48a 0.9cm Appeadixf1347,971 11-02-98 11:48a 0.9cLo Appendix I-11,392 11-02-98 1I:48a 1.0cm Aienix I-I

Waste Package Operatlons Engineering CalculationTl~e: N-Reactor Spent Nudear 1iel Citicality CallatonsDocument IdentIfier. BBAOOOOO -01717-0210-00005 REV 00 Attachment Vi, Page VI.6

File Size(bytes)

Date Las Access Fle Name Table Used

347,971 110298 I1:48a 1.0cmuo Appendix I-11,388 11-02-98 1148a 1.1cm Appendix I-1347,872 11-02-98 1:48a 1.lCmLo Appendix I-11,395 11-02-98 11:48a 1.2cm Appendix 1-1346,783 11-02-98 11:48a 12an.o Appendix I-11,390 11.02-98 11:48a 13cm Appendix I-1347,872 11-02-98 11:48a 13M.o AppendixI-I1,394 11-02-98 11:48a Acm AppendixI-I348,094 102-98 IlAU lAcmo Appendix I-I1,393 11-02-98 11:48a 1.5cm Appx I-348,9 11-02-98 11:48a 1.5cmLo Appendix I-11,394 11-02-98 11:48a 1.6cm ! Appendix -1348,607 11-02-98 11:48a 1.6cr.o Appendix 1-11,392 11-02-98 11:48a 1.7cm Appendix -1348,607 11-02-98 11:48a 1.7cm.o Appendix I-i

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Waste Package Operations Engineering CalculationTWe: N-Reactor Spent Nuclear Fuel Criticality CalclationsDocment Identier: BBAOOOOOO-01717-0210-O0000 REV 00 Atachment V, Page VI'7-Y P P p p -

Mile Sizebye)

Date Lost Access Fie Name Table Used

1,387 11-02-98 11:50a 12cm Appendix 1-3348,827 11-02-98 11:50a 12cmIo Appendix 1-31,389 11-02-98 l:S50a 1.3cm Appendix 1.3348,830 11-02-98 11:50a 1.3cmo Appendix 1-31.387 11-02-98 1150a Il4cm 'Appendix 1-3349,148 11-0-98 11:50a lAcnLo Appendix 1-31,389 11-02-98 11:50a 1M5m_ Appendix 1-3348,374 11-02-98 1150a 15Scm.o Appendix 1-31,389 11-02-98 11:SOa 1.6cm Apendix 1-3349,880 11.02-98 11:50a 1lo . Appendix 1-3

1,391 11;02-98 1150a 0.8cm Appendix 1-4348,926 11-02-98 -1150a 0.8cm.o Appendix 1-41,388 11-02-98 1:50a 0.9cm AppdixI-4.348,509 11-02-98 11:50a 0.9an.o Appendix 1I41,390 11-02-98 I1:50a 1.0cm AppendixI-4346,252 11.02-98 11.50a lO.o Appendix 1-41,390 1l02-98 11:50a 1.1cm Appendix -4347,71 11-02-98 1150a 1.lcm.o Appendix1-41,391 11.2-98 11.:50a 112cm Appendix 1-4

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.1,391 11-02-98 11:51a 1.6cm App edix 1-4

349,880 114-2-98 11:51a 1.6cmco= Appendix 1-4

1,390 11-02-98 1151a 0.8cm Appendix I-S349,242 11-02-98 11:51a 0cm.o Appendix I-S1,388 11-0-98 1151a 0.9cm Appendix 1-5347,736 11-02-98 11:51a 0.9cm.o Appendix I-SI' 11-02-98 1151a l.Ocm Appendix I-S348,828 11-02-98 11:51a 1.Ocn~o Appendix I-S1,87 11-02-98 11.51a 1.1cm Appendix I-S349,146 11-02-98 1151a 1.lcm.o Appendix I-51,380. 11-02-98 11-51a 1cmAppendix 1-5349,047 11-02-98 1151a 1.2cm~o Appendix I-51,385 11-02-98 1151a 1.3cm Appendix I-S348,408 11-02-98 11:Sla 1.3cm.o AppdixI-S1,390 11-02-98 11:S1a IA4cm Appendix I-5

Waste Package Operations Engineering CalculationTltle: N-Reactor Spent Nuclear Fue Critcality CalclationsDocument Identfer. BBAOOOOOO01717-0210-OO05 REV 00 Attachment V, Page VAS

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Eile Size(bytes)

Date Last Access Fe Name Table Used

349,878 11-02-98 ll:51a lAcmo AppendixI-S1,389 1102-98 1151a 1.5cm Appendix I-S350,100 11.02-98 11:51a 1.Scmo Appendix l -S1,386 11-Q2-98 11:51a 1.6cm Appedix 1-S350,514 11-02-98 11:Sla lcM.o Appendix [-5

1,389 11-0-98 118.52a 08cm Appendix1-6346,584 11-02-98 11:52a 0.cmo . Appendix I-61,387 11-0-98 11:52a 0o9cm AppendixlI-6348,924 11-02-98 11:52a 0.9crLo AppendixI-61,390 11-02-98 11:52a 1.0cm Appendix 1-6348,924 11 -0-98 11:52a 1.0cm o Appendix 1-61,385 11-02-98 1152a 1.1cm Appendix 16347,958 11-02-98 11:52a 1.1an.o A di 61,391 11-02-98 11:52a 1.2cm Appendix 16349,461 11-02-98 11:52a 12mo Appendix 161,388 114-0298 11:52a 1.3cm Appendix 1-6348,726 11.02-98 11:52a 1.3cm o Apedix 1-61,388 11-02-98 11:52a 1mA . AppendixlI-6.350,196 11.02-98 1152a lA1 .o Appendix 1-61,389 _ 11-02-98 11:52a 1.5cm _ AppendixI-6345,675 12-08-98 9:l0a 1.Scm.O Appendix 1-61,389 11-02-98 11:52a 1.6cm -Appendi1-6350,832 11-02-98 11:52a 1.6cm o Appendix 1-6

1,387 11-02-98 11:52a 0.7cm Appendix I-7347,937 11-02-98 11:52a 0.7cmLo Appendix I-71,391 11-02-98 11:52a 0.8cm Appendix 1-7347,739 11-02-98 I1:52a 08Cm.O Appendix 1-71,387. 11-02-98 11:52a 0.9cm Appendix-7348,609 11I02-98 11:53a 0.9CULO Appendi I-71,386 11.02-98 11:53a 1.0cm x . Appendix 1-7348,828 11-02-98 1153a I.0cmO_ Appendix 1-71,389 11-02-98 ll53a 1.1cm Appendix 1-7347,994 11-02-98 11:53a 1.1cmno Appendix -71,377 _11-0-98 11:53a 12cM Appendix1-7349,563 11-02-98 1153a 12cm.o Appendix I-71,384 11-02-98 11:53a 1.3cm Appendix 1-7349563 11-02-98 11:53a 1.3cm.o Appendix 1-71,389 11-02-98 11:53a 14cm Appendix 1-7348,729 11-02-98 11.53a 1.4cm.o Appendix 1-71,388 1102-98 11:53a 1.5cm Appendix I-7350,199 11-2-98 1153a 1.5cmo Appendix 1-7

Waste Package Operations Engineering Calculation* The: N-Reactor Spent Nucdear Fuel Cicality Calculadons

Document Identifier. BBAOOOOOO-01717-0210-005 REV 00 Aftacmnt Vi WaeV1-9_

Fle Size(bytes)

Date IAe Access FIle Name Table Used

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365,334 11-2-98 11:54a outa- Appendix -1366,304 11402-981154a outaax Appndix rn-366,637 11402-98 1154a ou- .. Appendix m-1364,481 11042-98 11:54a out.bea Appredix U-

353,842 114298 11:54a outaal Appendix IV-354,994 1102-9811:55a outa.2 Appendix W-1355,001 114M-98 11:55a out=aa4 Appenix IV-1353,849 11-02-98 1155a out- - App:ndix W4354,994 11402-98 1155a outaa7 APPedixIv-1354,041 11402-98 1155a a8 Appen5ix -357.045 1142S98 11:SSa cutbal Appendix IV-1360,655 11402-98 11:SSa outba2 AppendixIV-1356,832 1142-98 11:55a outba4 Appendix IV-1355.872 1102-98 11:55a outba7 Appendix IV-1358,224 1142-98 1155a outbag Appendix IV-1355,680 110298 1155a outcal Appendix IV-1355,001 11-02-98 11:55a outca2 Appendix JV-1356,832 1142-981 1:5Sa outca4 Appendix IV-1355,001 1102-98 11:55a outca5 Appendix Ky-i356,832 11-02-98 11:Sa cutca7 Appen IV-1354,994 11-02-98 11:55a outcas Appendix

Waste Package Operations Engineering Calculation

Ttle: N-Reactor Spent Nuclear Fuel Criticaliy culationsDosunent dentlfier. BBA000000-01717-0210-00005 REVO10 . Atftachment V!.Page V1U-

Attachment VIITable VII-l. Rev. ODB MCNP Inputs and Outputs used In Engineering Calculations Stored

In Electronic Format on CD (Re£ 7.12)

Elle Size Date Iast Access EMle Name Table Used(bytes)

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Waste Package Operations Enilneering Calculation- En -

Tte: N-Reator Spen Nudear Fudl ticaity CalculationsDocment Identifier. BBAOOOOO0-01717-02M0M00 REV 00 Affachment VII, Page VU-2

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Waste Package Operations Engineering CalculationTitle: N-React Spean Nuclear Fuel CriicaW CalcationsDocument Identfier. BBAODOOOO-01717-0210-OO0S REV OD Attachment VI!, Page VI3-. I. I - - -

FRe size(bytes)

Date Last Access Fie Name Table Used

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Waste Package Operations Engineering CalculationTitle: N-Reacor Spent Nuclear Fud 0iticality CaculationsDocumert Identlrler. BBAOOOOO-01717-0-000 5REVOO Atachment VII. Pane VII-6

FIle Size(bytes)

Date Ast Access EMe Name Table Used

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1,391 11-02-98 1150a 0.85Cm Appendix I-3347,966 11-02-98 11:50a 0.85Cm.o Appendix 1-31,389 11-02-98 1150a 0.9cm Appendix 1-3348,608 11-02-98 11F5Qa 0.9cm.O Appendix 1-31,392 11-02-98 11:50a 1.0Cm Appedix 1-3

Waste Package Operations Engineering CalculationTue: N-Rector Spen Nuclear Fuel Qiticality CalclatDocmmnt Identifier. BBA0000OO-01717-0210-OOOO REV 00 Attachment VII, Page VII-7

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1,390 11.02-98 1151a 0.8cm Appendix I-S349,242 11402-98 11.51a 0.8cm.o Appendix 1-S1,388 11-02-98 11:51a 0.9cm Appe 1ix -SP347,736 11.02-98 11:51a 0.9cILo Appendi I-S1,384 11-02-98 Il:Sla 1.0cm Appendix 1-5348,828 11-02-98 11:51a l.Ocm.o Appendix 1-51,387 11-02-98 11 51a 1.1cm Appendit 1-5349,146 11-02-98 1151a I lcm.o ,Appeidix1-5

1,380 11-02-98 11:51a 12cm Appendix I-S349,047 11-02-98 11:51a 12cm.o Appendix I-5

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Waste Packaie Operations Engineering CalculationTiMle: N-Reactor Spent Nuclear CFe idcaity CalclaonsDocment Idetfier. BBA00006-1717-0210-00 REV 00 Atiachment VII, PtgelV%-

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Date Iast Access FEIe Name Table Used

1,385 11-02-98 1151a 13cm Appendix 1-5348,408 11-02-98 11:51a 13mo Appendix 1-51,390 1102-98 11:5a 14cm Appndi1-5349,878 11-02-98 11:51a lAcmo Appendix i-S1,389 11-02-98 11:51a 1.5cm Appendix I-S350,100 11-02-98 11 51a lScm.o Appendix 1-51,386 11-02-98 11:Sla 16cm Appendix 1-5350,514 11-02-98 11Sla 1.6an.o Appendix I-S

1,389 11-02-98 11:52a 0.8cm Appendix1-6346,584 11-02-98 11:52a 0.8cm.o Appendix-61i387 11-02-98 11.52a 0.9a Appendix-6348,924 11-02-98 11:52a 0.9cm=o Appendix 1-61,390 11-02-98 1152a 1.0cm Appendix i6348924 11-02-98 1152a I.Ocm.o AppdIx 1-1,385 11-02-98 11:52a 1.1cm Appendix 16347,958 11-02-98 11:52a .lmo Appendix1-61,391 11-02-98 11*52a 1.2cm Appendix 1-6349,461 11.02-98 1152a 12cm.o Appdix 1-61;388 11-02-98 11:52a 1.3cm AppendixI16348,726 11-02-98 11:52a 13co _ Appendix1-61,388 11-02-98 11F52a A4cm AppI-6350,196 11-02-98 11.52a lAcm.o Appendix 1-61,354 12-08-98 9:10a 1.5cm Ap.dixl-6345.675 12-08-98 9:10a 15cmCo Appendix 1-61,389 11-02-98 11:52a 1.6cm App=dix -6350,832 11-02-98 11.52a 1.6cm.o AppendixI1-6

1,387 11-02-98 11:52a 0.7cm Appendix 1-7347,937 11.02-98 11:52a 0.7cmuo Appendix I-71,391 11-02-98 11:52a 0.8cm Appendix 1-7347,739 11-02-98 1152a 0.8ncmo -Appendix1-71,387 11-02-98 11:52a 0.9cm Appendix I-7348,609 11-02-98 11:53a 0.9cULo Ap 1-71,386 - 11-02-98 11:53a 1.0cm Appendix 1-7348,828 11-02-98 11:53a 1.Ocm.o Appentix 1-71,389 11.02-98 11:53a 1.1cm Appn xI-7347,994 11-02-98 11:53a l.lcmo Appendix 1-71,377 11-02-98 11:53a 12cm Appendix 1-7349,563 11-02-98 11:53a 1.2cm.o Appendix 1-71,34 11-02-98 1153a 1.3cm Appendix 1-7349,563 11-02-98 1153a 13cmno 1ppendix1-71,389 11-02-98 1153a 1A4 Appendix 1-7

Waste Package Operations Engineering CalculationTMe: N-React Spent Nuclear Fuel Criticalit alculationsDocumentIdentgrier. BBAOOOOO 1717-010-00005 REV OD Attachment VYl. Pane VII9my I - 'p - --

FMle Size(bytes)

Date LUst Access Mile Name Table Used

348,729 11-02-98 11:53a lAcm.o Appehdix 1-71,388 11-02-98 11:53a 1.5cm .Appendixs-7350,199 11--98 11:53a 1.ScM.o ApPendix 1-7

1,392 11-02-98 1i53a 0.9cm Appendix I 1347,971 1102-98 1153a 10.9cmo Appendix 11-11395 11402-98 11:53a 105cmn AppendixI-i346,981 11402-98 11:53a .l5cm.o Appendix l-11,392 1102-98 11:53a 1.0cm Appendix 1-347,971 11-02-98 1153a lAk1Lo Appendix[-11,396 11-0M-98 1153a 1.15cma Appendix I-347,011 11-02-98 11:53a 1.lScmno Appendix 1-11,388 1102-98 1153a I.cma Appendix [I-1347872 1142-98 1153a 1.lcma.o Appendix .141,395 1142-9811:53a 1.2cm Appendix 1-1346,783 1102-98 11:53a 1.r.o .Appendix H-1

365,334 11042-98 11:54a outaa Appendix 1-1366304 11-02-98 11:54a outaax Appendix rn-i366,637 11042-98 11:54a outnewta Appendix M-i364.481 11402-98 1154a outubea Appendix 1El-1

353,842 11402-98 11 54a outa1 Appendix [V-1.354,994 114-2-98 11:SSa outna2 APIenidrx-1355,OO1 1142-98 1155a outaa4 Appendix 1.353,849 11-02-98 1155a outaa5 Appendix IV-i354,994 11402-98 11:55a outaa7 Appendix IV-354,041 11402-98 1155a outaa8 Appendix IV-1357,045 11402-98 1155a outbal Appedix[V-360,655 11402-98 11:55a outba2 Appendix V-I356,832 11402-98 11:55a qutba4 Appendix V-I355,872 1142-98 11:55a outba7 Appendix IV-358,224 1142-98 1155a outba8 Appendix [I-355,680 1142-98 1155a outcal Appendix [V-I355,001 114-kD98 1155a outca2 Appendix V-I356,832 1142-98 1155a outca4 Appendix IV-I355,001 11402-98 11'55a outCa5 Appendix[V-i356,832 11402-98 1155a outca:7 Appendix [V-i354,94 Q1142-98 11'55a outca8 Appendix IV-i

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Sedat /3 3 · 2012. 11. 19. · S. Checker Sedat Gokugh /3 3 9. Lead Design Enginear J. Wesft Davis...

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Transcript of Sedat /3 3 · 2012. 11. 19. · S. Checker Sedat Gokugh /3 3 9. Lead Design Enginear J. Wesft Davis...