NUREG/CR-6748 

ORNL/TM-2001/33

STARBUCS: A Prototypic 

SCALE Control Module for 

Automated Criticality Safety 

Analyses Using Burnup 

Credit 

Oak Ridge National Laboratory 

U.S. Nuclear Regulatory Commission 

Office of Nuclear Regulatory Research 

Washington, DC 20555-0001
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NUREG/CR-6748 

ORNL/TM-2001/33 

STARBUCS: A Prototypic 

SCALE Control Module for 

Automated Criticality Safety 

Analyses Using Burnup 

Credit 

Manuscript Completed: August 2001 

Date Published: October 2001 

Prepared by 

I. C. Gauld, S. M. Bowman 

Oak Ridge National Laboratory 

Managed by UT-Battelle, LLC 

Oak Ridge, TN 37831-6370 

R_ Y. Lee, NRC Project Manager 

Prepared for 

Division of Systems Analysis and Regulatory Effectiveness 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

NRC Job Code W6479
NUREG/CR-6748 has been reproduced 

from the best available copy.
ABSTRACT 

STARBUCS is a new prototypic analysis sequence for performing automated criticality safety analyses of spent 

fuel systems employing burnup credit. A depletion analysis calculation for each of the burnup-dependent regions 

of a spent fuel assembly, or other system containing spent fuel, is performed using the ORIGEN-ARP sequence of 

SCALE. The spent fuel compositions are then used to generate resonance self-shielded cross sections for each 

region of the problem, which are applied in a three-dimensional criticality safety calculation using the KENO V.a 

code. This prototypic burnup credit analysis sequence allows the user to simulate the axial and horizontal burnup 

gradients in a spent fuel assembly, select the specific actinides and/or fission products that are to be included in 

the criticality analysis, and apply isotopic correction factors to the predicted spent fuel nuclide inventory to 

account for calculational bias and uncertainties. Although STARBUCS was developed to address the burnup 

credit analysis needs for spent fuel transport and storage applications, it provides sufficient flexibility to allow 

criticality safety assessments involving many different potential configurations of spent nuclear fuel to be 

simulated.

iii
CONTENTS 

Page 

ABSTRACT .................................................................................................................... 1i 

LIST OF FIGURES ................................................................................................................................................. vii 

LIST OF TABLES .................................................................................................................................................... ix 

1 I INTRODUCTION .............................................................................................................................................. 1 

2 TYPE OF PROBLEMS STARBUCS CAN ANALYZE .............................................................................. 3 

3 M ETH OD OLOGY ............................................................................................................................................. 5 

4 PROGRAM FLOW ............................................................................................................................................ 7 

4.1 Overview ................................................................................................................................................ 7 

4.2 Subroutine Descriptions ......................................................................................................................... 9 

5 CAPAB ILITIES AND LIM ITATION S ................................................................................................... 11 

6 SPECIAL SETUP REQU IREM EN TS ............................................................................................................. 15 

7 LOGICAL UNIT A SSIGN M EN TS ................................................................................................................. 17 

8 INPUT DESCRIPTION ................................................................................................................................... 19 

8.1 Overview of Input Structure ................................................................................................................. 20 

8.2 Sequence Specification Card ................................................................................................................ 21 

8.3 M aterial Inform ation Processor ....................................................................................................... 21 

8.4 Control Param eter D ata ........................................................................................................................ 24 

8.5 Power H istory D ata .............................................................................................................................. 29 

8.6 KEN O Input Data ................................................................................................................................. 30 

9 SAM PLE PROBLEM S .................................................................................................................................... 33 

9.1 Sample Problem 1 ................................................................................................................................ 33 

9.2 Sample Problem 2 ................................................................................................................................ 33 

9.3 Sample Problem 3 ................................................................................................................................ 36 

9.4 Sample Problem 4 ................................................................................................................................ 40 

9.5 Sam ple Problem 5 ................................................................................................................................ 45 

10 FUTURE DEVELOPME NT S .......................................................................................................................... 53 

11 REFERE N CES ................................................................................................................................................. 55

v
LIST OF FIGURES 

Fizure Page 

1 Modules executed by STARBUCS ........................................................................................................ 6 

2 STA RBU CS subroutine call tree ..................................................................................................................... 8 

3 Fuel and material mixture numbering convention used in STARBUCS ................................................. 12 

4 Plot of the CE 14 x 14 assembly array geometry in sample problem 3 .................................................... 37 

5 Plot of the generic 32-assembly burnup credit cask geometry .................................................................. 40 

6 Plot of the 2 x 2 array of CE 14 x 14 assemblies with bumup gradient ................................................... 51

vii
LIST OF TABLES 

Table Page 

1 File parameters for logical units .................................................................................................................... 17 

2 Valid data block keywords ............................................................................................................................. 19 

3 Outline of input data for the STARBUCS sequence .................................................................................. 20 

4 Standard composition specification ............................................................................................................... 22 

5 Table of control parameter data ..................................................................................................................... 26 

6 Built-in burnup-dependent axial profiles, NAX= -18 ............................................................................ 27 

7 Built-in burnup-dependent axial profiles, NAX= -19 ............................................................................ 28 

8 Table of power history data ........................................................................................................................... 30 

9 STARBUCS input listing for sample problem 1 ...................................................................................... 34 

10 STARBUCS input listing for sample problem 2 ....................................................................................... 35 

11 STARBUCS input listing for sample problem 3 ...................................................................................... 38 

12 STARBUCS input listing for sample problem 4 ....................................................................................... 41 

13 STARBUCS input listing for sample problem 5 ........................................................................................ 47

ix
1 INTRODUCTION 

The Nuclear Regulatory Commission (NRC) issued Revision 1 of the Interim Staff Guidance 81 (ISG8) on burnup 

credit in July, 1999. ISG8 provides guidance on the application of limited bumup credit in criticality safety 

analyses for pressurized water reactor (PWR) spent fuel in transportation and storage casks. Burnup credit is the 

concept of taking credit for the reduction in reactivity in spent fuel due to burnup. The reduction in reactivity that 

occurs with fuel burnup is due to the change in concentration (net reduction) of fissile nuclides and the production 

of actinide and fission-product neutron absorbers. In contrast to criticality safety analyses that employ a fresh

fuel assumption (i.e., conservatively assume unirradiated fuel compositions), credit for burnup requires the 

prediction of both fissile material and absorber nuclide inventories in spent nuclear fuel (SNF) and consideration 

of the many bumup-related phenomena, in addition to the criticality issues. 

Consideration of the depletion aspects in the criticality assessment of SNF places an increasing reliance on 

computational tools and methods, and significantly increases the overall complexity of the criticality safety 

analysis. The credit for fuel burnup in criticality safety also necessitates consideration of many additional 

uncertainties associated with fuel depletion. ISG8 highlights, for example, the need for applicants employing 

burnup credit in criticality safety assessments to address the axial and horizontal variation of the burnup within a 

spent fuel assembly, uncertainties and bias in the isotopic predictions, and the additional reactivity margin 

available from fission products and actinides not credited in the licensing basis. 

To assist in performing and reviewing criticality safety assessments of transport and storage casks that apply 

burnup credit, a new SCALE control sequence prototype, STARBUCS (_andardized Analysis ofReactivity for 

Bunup Credit using SCALE) has been created. STARBUCS automates the generation of spatially-varying 

isotopic compositions in a spent fuel assembly, and applies the assembly compositions in a three-dimensional 

(3-D) Monte Carlo analysis of the system. STARBUCS automatically prepares input files for each of the code 

modules in the sequence, executes the modules through the SCALE driver, and performs all flow control, module 

interface, and data management functions. The prototypic STARBUCS sequence uses the well-established code 

modules currently available in the SCALE-4 code system.2 

The STARBUCS sequence automates the depletion calculations using the ORIGEN-ARP methodology3

'

4 to 

perform a series of cross-section processing and depletion calculations to generate a comprehensive set of spent 

fuel isotopic inventories for each spatially-varying bumup region of an assembly. The spent fuel isotopics are 

subsequently input to the SCALE Criticality Safety Analysis Sequence (CSAS) 5 to generate the resonance self

shielded macroscopic cross sections for each burnup-dependent fuel region. Finally, a calculation using the 

KENO V.a code 6 is performed using the spatially-varying cross sections to determine the neutron multiplication 

factor (keff) for the system. Only minimal input is required by the user to perform a burnup credit analysis. The 

user may specify the assembly-average irradiation history of the assembly, the axial moderator density, the axial 

and horizontal burnup profile, and the nuclides that are to be credited in the criticality safety analysis. Isotopic 

correction factors may also be applied to account for known bias and/or uncertainty in the predicted SNF 

compositions.

1
2 TYPE OF PROBLEMS STARBUCS CAN ANALYZE 

STARBUCS is designed to facilitate burnup credit criticality safety analyses employing burnup credit by 

automating and linking the depletion and criticality calculations. The STARBUCS burnup credit sequence has 

been designed to readily allow analysts and reviewers to assess the subcritical margins associated with many of 

the important phenomena that need to be evaluated in the context of the ISG8 guidance. However, STARBUCS 

is sufficiently general to allow virtually any configuration involving spent nuclear fuel to be analyzed. Some of 

the key capabilities of the STARBUCS prototype are described below. 

1. The types of spent fuel configurations that can be analyzed are very general. STARBUCS can be used to 

perform criticality safety assessments of individual fuel assemblies, a spent fuel cask, a spent fuel storage 

pool, or any nuclear system containing SNF. 

2. A spatially-varying burnup distribution is accounted for using a finite number of axial and, optionally, 

horizontal burnup zones. An arbitrary axial burnup profile may be used. Several built-in burnup-dependent 

profiles are provided as options. 

3. Burnup calculations can incorporate any desired operating history. The user may specify the specific fission 

power, cycle lengths, cycle down time, post-irradiation cooling time, etc. The axial water moderator density 

variation may also be specified for the depletion analysis. 

4. The effects of assembly design, burnable poison exposure, reactor operating conditions, etc., are accounted 

for in the ARP cross-section libraries used in the depletion calculations. These libraries can be readily created 

for any reactor and fuel assembly design that can be represented in the SAS2H depletion analysis sequence of 

the SCALE code system. 

5. The user can select the specific actinide and/or fission product nuclides to be included in the criticality safety 

analysis. The user also has the option to perform a criticality calculation employing all nuclides for which 

cross-section data exist. 

6. Isotopic correction factors may be input to adjust the calculated nuclide inventories to account for known bias 

and/or uncertainties associated with the depletion calculations. 

7. STARBUCS may be extended to include an iterative search capability to automate the generation of loading 

curves, which define the acceptable enrichment and burnup combinations for a given configuration of spent 

fuel (e.g., SNF transport or storage casks, storage racks in a spent fuel pool, etc.).

3
3 METHODOLOGY 

The STARBUCS program is a burnup credit sequence designed to perform a 3-D criticality safety calculation that 

includes the effects of spatially-varying burnup in a configurations of spent nuclear fuel. The sequence performs 

an automated fuel depletion analysis for each of the spatially-varying burnup regions of an assembly, creates 

resonance-corrected cross sections based on the fuel assembly design and spent fuel compositions, and perform a 

3-D Monte Carlo calculation for k/using the KENO V.a code.6 STARBUCS prepares input files for each of the 

codes used in the sequence, handles the data management and interface functions between the various codes, and 

executes each of the codes in the required order. 

The burnup credit sequence uses the well-established code modules currently available in the SCALE-4 code 

system. These code modules include ARP3 and ORIGEN-S7 to perform the depletion analysis phase of the 

calculations. ORIGEN-ARP 4 is a sequence within the SCALE system that serves as a faster alternative to the 

SAS2H sequence8 of SCALE to perform point-irradiation calculations with the ORIGEN-S code using problem

dependent cross sections. ARP (Automatic Rapid Processing) uses an algorithm that enables the generation of 

cross-section libraries for the ORIGEN-S code by interpolation over pregenerated SAS2H cross-section libraries. 

The ORIGEN-S code performs the isotopic generation and depletion calculations to obtain the spent fuel 

compositions. Cross sections for the criticality analysis are processed through the resonance self-shielding codes 

BONAMI9 and NITAWL' 0 using the region-dependent compositions from the depletion analyses. Finally, the 

cross sections are applied in KENO V.a,6 which calculate the keff value for the 3-D system. 

The ORIGEN-ARP depletion analysis methodology represents a significant increase in computational speed as 

compared to equivalent calculations performed using the SCALE depletion analysis sequence SAS2H, without 

sacrificing accuracy. ARP uses an algorithm that enables the generation of cross-section libraries for the 

ORIGEN-S code by interpolation over pregenerated cross-section libraries using initial fuel enrichment, burnup 

and, optionally, moderator density as the interpolation parameters. STARBUCS creates input files for ARP and 

ORIGEN-S for each burnup-dependent region of an assembly and calculates the spent fuel compositions for the 

region using a user-specified assembly irradiation history, cooling time, and bumup profiles. The ARP cross

section libraries must be available in advance of a STARBUCS bumup credit calculation, and may be created 

using the SAS2H depletion analysis sequence of SCALE. These libraries include the effects of assembly design 

and operating conditions on the cross sections. Several ARP libraries are distributed in the SCALE code system 

(for boiling water reactors an 8 x 8 library, and for pressurized water reactors a 14 x 14, 15 x 15, and 17 x 17 

library) and can be applied in a burnup credit analysis. Alternatively, a user may create a specific ARP library for 

other assembly types or operating conditions to capture reactor operating or design effects not available in the 

default libraries. The generation of ARP libraries is discussed in Ref. 4. 

The depletion phase of the sequence uses ARP and ORIGEN-S to calculate the compositions of each fuel region 

(axial or horizontal). After a single ORIGEN-ARP depletion calculation is completed, control is passed back to 

the STARBUCS module which reads the spent fuel inventories generated by ORIGEN-S, saves them, prepares 

the ARP and ORIGEN-S input files for the next bumup region, and executes the codes in sequence. This cycle 

continues until the fuel compositions for all axial and horizontal regions have been calculated and saved, 

completing the depletion phase of the analysis. The depletion calculations for each axial and radial zone are 

performed using an initial fuel basis of 1 MTU (106 g U). 

After all depletion calculations are completed, the STARBUCS control module reads the spent fuel isotopic 

inventories for all regions and prepares input for the resonance cross-section self-shielding calculations, 

converting the saved nuclide concentrations from gram-atom units to atoms/b-cm. Resonance-corrected cross 

sections for the criticality calculation are generated using the capabilities in the CSAS control module. 

Specifically, STARBUCS prepares input to the CSASI sequence of the CSAS module of SCALE, which performs 

Bondarenko self-shielding of the unresolved resonance range using the BONAMI code, and applies Nordheim 

self-shielding corrections in the resolved resonance range for nuclides with resonance parameters. The CSASI 

sequence applies the ICE module1

" to produce mixed, or macroscopic, cross sections for each mixture (burnup5
Methodology

dependent region). This greatly reduces the data management complexity of STARBUCS since it eliminates the 

need to store and manage the concentrations and microscopic shielded cross sections for the individual isotopes in 

the spent fuel by reducing the mixture data to a single set of effective macroscopic fuel cross sections. Following 

each CSASI case, the WAX utility12 code takes the macroscopic cross sections generated by CSASI and 

renumbers the fuel cross-section identifier (to avoid generating duplicate numbers) to a unique identifier for the 

specific axial or horizontal bumup region, and appends the cross section to create a single cross-section library 

containing the cross sections for all fuel regions. The CSASI and WAX cases are stacked sequentially for each 

fuel region and are executed by the SCALE driver. Following completion of these cases the KENO V.a criticality 

calculation is performed using the input data provided by the user and the burnup dependent cross sections 

prepared from the depletion analysis calculations. The overall program structure and flow is illustrated in 

Figure 1. 

STARBUCS

I, 

ARP 

is 

ORIG EN-S

STARBUCS SEQUENCE 

SPENT FUEL DEPLETION 

AND DECAY ANALYSES

REPEAT FOR ALL REGIONS

f RESONANCE CROSS SECTION 

CSASI .-a-BONAMI --i--NITAWL---ICE PROCESSING - REPEAT FOR ALL 

i I FUEL REGIONS 

WAX LABEL AND COMBINE CROSS SECTIONS 

FOR ALL SPATIAL FUEL REGIONS 

KENO CRITICALITY CALCULATION

Figure 1 Modules and flow of STARBUCS sequence

6

Section 3
4 PROGRAM FLOW 

4.1 Overview 

The STARBUCS program is designed to run as a control module within the SCALE system. The depletion phase 

of the analysis, used to generate the spatially-varying isotopic compositions in the spent fuel, is performed using 

repeated ARP and ORIGEN-S calculations. The input files for these codes are generated automatically by 

STARBUCS. A decision was made early in the development of this sequence to use formatted, standard text 

input files for these codes (ORIGEN-S has an option to read unformatted binary input which is used in the other 

control modules in the SCALE-4 code system). The formatted input allows the user to readily verify the input 

used by the depletion analysis modules. This decision however, required the input routines of both ARP and 

ORIGEN-S be slightly modified in order to function as modules within the STARBUCS sequence. The changes 

allow formatted input to be read from a unit other than unit 5. 

The cross-section processing phase of a STARBUCS analysis is performed using the Criticality Safety Analysis 

Sequence, CSAS. This is an unconventional approach since the SCALE system driver is not designed to allow a 

control module (i.e., STARBUCS) to call another control module (i.e., CSAS). However, this approach 

significantly simplifies the sequence and expedited development of STARBUCS. It is not necessary to generate 

input files for the functional modules BONAMI, NITAWL, and ICE, because this capability was already 

available in the CSAS sequence. In order to implement this capability within SCALE, the formatted input 

generated for CSAS is written to the input file used by the SCALE driver (SYSIN). During a bumup credit 

calculation, the STARBUCS program reads the input file transferred by the driver. After the depletion phase of 

the calculation has been completed, STARBUCS writes/appends the input files for all CSAS cases, plus the input 

for the KENO V.a criticality calculation, to the SYSIN file. STARBUCS instructs the driver that it has finished 

its processing, and the driver then continues to read the SYSIN file and executes each of the CSAS cases and the 

KENO V.a calculation. Consequently, any additional cases stacked after the STARBUCS case (after the END) 

will be overwritten during the sequence. Therefore, no additional cases may be placed after a STARBUCS case 

(e.g., multiple stacked STARBUCS cases or other modules following a STARBUCS case are not allowed). 

Ultimately, it is planned to integrate the STARBUCS control module more fully within SCALE to allow multiple 

cases to be run consecutively. This will be done by integrating the relevant subroutines from CSAS directly into 

STARBUCS and calling each of the functional modules in the sequence directly (see Section 10, Future 

Developments). Instead of using CSAS to generate the input for each of the cross-section shelf-shielding function 

modules BONAMI, NITAWL, and ICE (see Figure 1), all the functional modules would be controlled by 

STARBUCS alone. 

The overall program call structure is illustrated in Figure 2. Calls to the SCALE subroutine library, and intrinsic 

procedures, are shown as the routine name with no box. The main program module, STARBUCS, controls the 

program flow. This module is divided into two major sections; the first section reads the user input, processes the 

initial material compositions, and performs the depletion analysis; the second section processes the spent fuel 

isotopic compositions required for the cross-section processing, and subsequently performs the criticality safety 

analysis. The program subroutines are described below in the order of the program flow.

7
0 

ARP WAX 

OMCDEN 

00 F- FINDID 

SBINRCX SINR~ 

IOARRY NA 

NUARRYNORM 

Figure 2 STARBUCS subroutine call tree 

0
Program Flow

4.2 Subroutine Descriptions 

Routine Name: STARBUCS 

Routines Called: OPNFIL, IONUMS, KNIGHT, MIPLST, WI, RDINP, FINDID, SETMTU, RESETP, 

ARPCHK, ARP, ORIGEN, APPEND, RESETM, NOAH, INVENT, CZERO CSASI, WAX, SYSTEM, CLEAR 

Function: STARBUCS is the main program called by the SCALE driver. The main function is the overall 

control of the program flow. During the initial execution of STARBUCS in the sequence, the program reads the 

standard compositions (initial material specifications) via subroutine KNIGHT that is a standard routine in the 

Material Information Processor of the SCALE system. KNIGHT provides the array data containing the nuclide 

IDs, the mixture numbers, zone temperatures, and the atomic number densities for all materials in the problem. 

The OPNFIL subroutine is used to open files using standard SCALE filename conventions. IONUMS is used to 

initialize the file input and output units used by the SCALE read routines. MIPLST prints the results of the 

material compositions from KNIGHT. The WI subroutine is also used to print array data to the output if the 

debug option (BUG-=yes) is requested. The rest of the input file is read by RDINP. RESETM is another SCALE 

subroutine that is used to set the control parameters for communication between the control module and the 

driver. 

Routine Names: RDINP, FINDID, SETMTU, RESETP 

Function: RDINP reads all user input after the standard composition data. All reading is performed using the 

SCALE reading functions AREAD, CREAD, IREAD, LREAD and FREAD (for character*4 character*8, integer, 

logical, and floating-point input data, respectively). The character nuclide identifiers used to specify the burnup 

credit nuclides and isotopic correction factors are read and parsed by the NUARRY subroutine. A user-input 

axial or horizontal burnup profile is normalized, optionally, using a call to subroutine NORM. FINDID converts 

the SCALE nuclide IDs to ORIGEN-S IDs and determines the library type (e.g., actinide, light element, or fission 

product) of the nuclide. SETMTU calculates the volume of fuel that will result in 1 MTU. The volume is used to 

convert the atomic number densities calculated by KNIGHT into mass units used in the depletion calculation with 

a basis of 1 MTU. After all of the input has been read and processed, the input and other data parameters are 

saved to a formatted file by subroutine RESETP using calls to subroutine IOARRY. 

Routine Names: ARPCHK, ARP, ORIGEN, APPEND 

Function: ARPCHK verifies that the initial fuel enrichment and discharge burnup of each burnup-dependent 

region is within the range of the ARP library specified, and halts the sequence if the ranges are exceeded. The 

ARP and ORIGEN subroutines create the formatted input files for the depletion modules. The ORIGEN 

subroutine uses calls to BINRCX and BINRCF to write FIDO input fields of floating-point and integer type, 

respectively, for the ORIGEN-S code. These routines call WTEND and WTRECT to write data end and block 

terminators. Input files are created for each node, one node at a time. That is, after an ARP-ORIGEN calculation 

is performed for the first node, control is returned to STARBUCS where the nuclide inventories are saved to a 

binary file by APPEND, and the input data for the next node are created and the depletion modules are executed. 

This cycle is repeated until the spent fuel compositions for all nodes have been computed and saved.

9

Section 4
Program Flow Section 4

Routine Names: INVENT, CZERO, NOAH, CSASI, WAX 

Function: After the depletion analysis phase is complete, the spent fuel isotopic compositions are read and 

processed, and the criticality analysis modules are prepared. INVENT reads the binary file created by APPEND 

during the depletion calculations that contains the complete nuclide inventory for each node. Subroutine CZERO 

is then called to perform two functions. It first checks for the nuclides that are available on the cross-section 

library that will be used in the criticality calculation, and secondly checks for the nuclides that have been specified 

as burnup credit nuclides that are to be used in the criticality analysis. If neither criteria is met, the concentration 

of the nuclide is set to zero which is a flag that it will not be used in the criticality analysis. NOAH is called to 

convert the ORIGEN-S nuclide IDs for the nuclides in the criticality analysis into alphanumeric names recognized by the SCALE Material Information Processor (e.g., 922350 is the identifier for 235U). Subroutine CSASI is 

called to write the input for the CSAS for each of the fuel regions in the problem. After the input for each region is created, a WAX case is appended. The WAX case is required to copy the cross-section set generated by CSAS 

for the given fuel region to another file containing the cross-section data for all fuel regions. This is required since the cross-section file created by CSAS is overwritten during each run. 

After the input for the cross-section processing modules has been created for each of the spatial zones, 

STARBUCS appends the input, in addition to the KENO V.a file, to the SCALE input file (SYSIN) using the 

intrinsic procedure SYSTEM. Control is then passed back to the SCALE driver, which executes each of the 

CSAS and WAX cases, and finally the KENO V.a criticality calculation.

10
5 CAPABILITIES AND LIMITATIONS 

STARBUCS is designed to perform criticality safety analyses that employ burnup credit. Minimal user input is 

required to perform many types of analyses. Default values are supplied for many of the input parameter 

keywords. The user may select from built-in bumup-dependent 18-axial-zone profiles,13" 4 or the user may input 

an arbitrary user-defined bumup distribution with up to 100 axial zones and up to 10 horizontal zones. The 

depletion analysis calculations for each zone are always performed for all nuclides (the ORIGEN-S data libraries 

contain cross-section and decay data for more than 1000 unique nuclides). The specific nuclides to be considered 

in the keff analysis may be input by the user. If no nuclide set is explicitly selected, then all nuclides that have 

cross-section data in the library will be automatically applied in the criticality analysis, resulting in a "full" 

burnup credit criticality assessment. A capability to adjust the calculated isotopic inventories using correction 

factors that can account for biases and/or uncertainties in the calculated isotopic concentrations is also provided. 

An appropriate ARP cross-section library must be available for the depletion analysis using STARBUCS. The 

user may use the libraries distributed with SCALE (e.g., 14 x 14, 15 x 15, 17 x 17) or the user may generate their 

own problem-specific libraries using SAS2H.8 Methods for generating the ARP libraries are described in 

Section D1.A.2 of the SCALE manual.4 

The user is required to provide a complete KENO V.a model of the spent fuel configuration. Although 

STARBUCS was designed with spent fuel transportation and storage casks in mind, any valid KENO V.a 

criticality model can be used for the analysis. For example, STARBUCS could be used to analyze a spent fuel 

storage pool, transportation cask accident configurations, fuel processing/reprocessing operations, or fissile 

solution storage tanks. 

The initial material composition information is defined in a standard composition data block. The fuel material is 

automatically depleted in the sequence for each of the axial and/or horizontal burnup-dependent regions of the 

problem, and macroscopic self-shielded cross sections are generated for each region. The macroscopic cross 

sections applied in the KENO V.a analysis contain all composition information, and therefore no compositions are 

explicitly required in the KENO V.a input model. The cross sections for each of the fuel regions are identified by 

unique mixture numbers assigned by STARBUCS based on the axial and horizontal region in the problem (see 

Figure 3). The user is required to specify the geometry/extent of the axial and horizontal zones in the KENO V.a 

model and apply the appropriate mixtures for the desired configuration. STARBUCS performs no checking of the 

criticality model to verify that all mixtures in the problem have been used or if the order of the mixture numbers 

in the KENO V.a model corresponds to the order of the input profile. This provides the user a great deal of 

flexibility in setting up problems. However, it also requires that the user accurately prepare the input files to 

ensure that the spent fuel region cross sections are assigned to the correct KENO V.a geometry region. For 

instance, the user could (intentionally) reverse the order of the axial material identifiers in the KENO V.a model 

to simulate inverted fuel, or simulate a problem using only a subset of the available fuel regions. 

There are several conventions that must be followed when using STARBUCS. In general, these relate to the 

specification of materials and mixture numbering of the cross-section mixing table. 

1) All structural materials in the problem (MX > 3) must be numbered contiguously.

11
Standard composition 

mixture (MX) 

1 I (fuel)

Modified KENO V.a mixture numbers (MX)

(1) 

101 

102 

103 

104 

105 

II00+NAX•

Horizontal region 

(2) 

201 ... 

202 

203 

204 

205 

1200+NAXI

(NHZ) 

NHZ*100+1 

NHZ*100+2 

NHZ*100+3 

NHZ*100+4 

NHZ*100+5 

NHZ*100+NAX

2 (clad) 

3 (mod) 

ININ

Figure 3 Fuel and material mixture numbering convention used in STARBUCS 

tJI

C,,) 

0

Mixture 

numbers 

for 

each 

axial 

burnup 

region

2 

3 

N
Capabilities and Limitations

2) Not all SCALE standard composition alphanumeric names"5 (see Section M8 of SCALE manual) are 

currently recognized by STARBUCS. The use of special materials (e.g., C-GRAPHITE, NIINCONEL, 

H-POLY) that have nuclide identifiers that are not readily translated to ZA numbers should presently be 

avoided. 

3) SCALE cases may not be stacked after the STARBUCS input file. Any cases placed after the STARBUCS 

case will not be executed by the SCALE driver. 

4) Only a single initial fuel material may be defined in the problem for depletion. Therefore, only one initial 

enrichment may be used.

13

Section 5
6 SPECIAL SETUP REQUIREMENTS 

The STARBUCS control module has been designed to create formatted, card-image text input files for all codes 

(functional modules and control modules) in the sequence. This allows the user to easily read the input files 

created by STARBUCS and verify the input. To accomplish this, minor modifications were required for the 

ARP and ORIGEN-S codes to allow the codes to read card-image input from a file other than logical unit 5 

(a restriction in SCALE-4). In addition, an updated version of the SCALE driver that allows the STARBUCS 

control module to call the CSAS control module is needed. The Standard Composition Library must be Rev. 17 

or later. The user must establish links, or place the modified versions of these codes and the Standard 

Composition Library in the SCALE working directory prior to executing a STARBUCS case. The links may be 

established prior to executing a STARBUCS case by overriding the links established by the SCALE script, or new 

links may be created in the STARBUCS input file using the =shell command followed by the new link 

directives, and terminated with an end. A link to the STARBUCS program executable must also be present in 

the SCALE working directory. 

Note that the special requirements needed to run STARBUCS under SCALE-4 will not be required with 

SCALE-5.

15
7 LOGICAL UNIT ASSIGNMENTS 

Table 1 lists the logical unit Nos. created and accessed by STARBUCS and the functional and control modules it 

executes. All input file interfaces are in text (BCD) format. 

Table 1 File parameters for logical units

Unit No. Variable Type of file 

5 N5 BCD 

6 N6 BCD 

71 N71 Binary 

72 NSAV Binary 

92 NICE BCD 

91 N91 BCD 

(33)a NDSET Binary 

93 N93 BCD 

94 N94 BCD 

99 N99 BCD 

a Units in parentheses may be reset by user input dat

Description 

Free-format input file 

Printed output 

ORIGEN-S nuclide inventories 

ORIGEN-S nuclide inventories (all cycles) 

CSAS input file (scratch) 

ARP input file 

ARP generated cross-section library (default) 

KENO V.a input (scratch) 

ORIGEN-S input file 

Data storage

17
8 INPUT DESCRIPTION 

The input is divided into different data blocks containing related types of information. The standard composition 

data block used to define initial (fresh) fuel composition and all other materials in the criticality analysis problem, 

is read and processed by the Material Information Processor of SCALE and conforms to the standard input 

conventions (summarized in Section 8.3). In addition to the standard composition data, STARBUCS defines 

three more main input data blocks. The data blocks are entered in the form 

READ XXXX input data END XXXX 

where XXXX is the data block keyword for the type of data being entered. The types of data blocks that are 

entered include general control parameter information, irradiation history and decay data, and the KENO V.a 

input specifications. The valid block keywords are listed in Table 2. A minimum of four characters is required 

for a keyword. However, the keywords can be up to twelve characters long, the first four of which must be input 

exactly as listed in the table. Entering the words READ XXXX followed by one or more blanks activate data 

input. All input data pertinent to block XXXX are then entered. Entering END XXXX followed by two or more 

blanks terminates data block XXXX. 

All input within a data block is entered using keywords and is free format. Keyword entries may be of variable or 

array type. Variable keyword entries include the keyword plus the "=", followed by the value. Array keywords 

are usually followed by a series of entries, each separated by a blank or comma, and must always be terminated 

with an END that does not begin in column one. In some instances a single value may be input as an array entry; 

however, the word END is still always required. Within a given input data block the keyword entries may be in 

any order. 

The input records are limited to a maximum of 80 columns. A single data entry may be entered anywhere in a 

record but can not be divided between two records; however, array data entries may be divided over many 

records. The code identifies keywords using only the first four (maximum) characters in the keyword name. 

Beyond the first four characters, the user may enter any alphanumeric or special character acceptable to 

FORTRAN, including single blanks, before the "=" character. Floating-point data may be entered in various 

forms; for example, the value 12340.0 may be entered as: 12340, 12340.0, 1.234+4, 1.234E+4, 1.234E4, or 

1.234E 04. Also, the value 0.012 may be entered as 12E-3, 12-3, 1.2-2, etc. Numeric data must be followed 

immediately by one or more blanks. 

Table 2 Valid data block keywords 

Data block type Keyword* 

Control parameters CONTROL 

Irradiation history HISTORY 

KENO V.a input KENOVA or KENO-VA 

*Only the first four characters of keyword are required

19
Input Description Section 8

8.1 Overview of Input Structure 

An overview of the input to the STARBUCS sequence is given in Table 3. This table provides an outline of the 

input data block structure. The input in data positions 1 to 6 (see Table 3) are read and processed by the Material 

Information Processor16 of SCALE. These are the first data read by the code and must be in the order indicated. 

Data positions 7, 8 and 9 are read directly by STARBUCS and may be in any order.

Table 3 Outline of input data for the STARBUCS sequence 

Data 

position Type of data Data entry Comments 

1 Sequence specification =STARBUCS Begin in column 1. 

2 TITLE Enter a title 80 characters 

3 Cross-section library 27GROUPNDF4 The cross-section library to be used in the 

name 44GROUPNDF5 criticality analysis (available in SCALE 

238GROUPNDF5 system). 

etc. 

4 Type of calculation INFHOMMEDIUM These are the available options. See the 

LATTICECELL explanation in Section M7.4.3. of the 

MULTIREGION SCALE manual. 

5 Standard composition Enter the data Terminate this data block with END COMP. 

Data 

6 Cell geometry Enter the appropriate Omit for INFHOMMEDIUM 

specification data (Omit for See Table M7.4.7 for LATTICECELL. 

INFHOMMEDIUM) See Table M7.4.8 for MULTIREGION. 

7 Control parameter data Enter the desired data Start with READ CONT, and terminate with 

block END CONT. See Table 5. 

8 Power history data Enter the desired data Start with READ HIST, and terminate with 

block END HIST. See Table 6. 

9 KENO V.a data block Enter KENO V.a model Start with READ KENO, and terminate with 

END KENO. 

10 Terminate analytic END Must begin in column 1. 

sequence

20
Input Description

8.2 Sequence Specification Card 

The STARBUCS analytical sequence is initiated with "=STARBUCS" (or "#" can be used in place of"=") 

beginning in column 1 of the input. This instructs the SCALE driver module to execute the STARBUCS 

sequence. The input data is then entered in free-format. The input is terminated with the word "END" card 

starting in column 1. An "END" is a special data item, which may be used to delimit an input data block, end an 

array of input items, and terminate the input for the case. In the context of input data blocks, the "END" has a 

name or label associated with it. An "END" used to terminate an array of entries must not begin in column 1 as 

this instructs the SCALE driver to terminate input to the sequence. 

8.3 Material Information Processor 

The Material Information Processor is used to read and process the standard composition specification data that 

define the initial compositions of the fuel, and all structural materials in the problem, into mixing tables and unit 

cell geometry information that are used by STARBUCS. All composition data required for the problem are 

entered as standard composition entries. The formats and options available for the standard composition entries 

are only summarized here. A detailed description of this portion of the input can be found in Section M7 of the 

SCALE manual. 

The types of data required for the Material Information Processor are given in Table 4. The individual entries are 

briefly explained below. 

1. TITLE. An 80-character title is required. The title is the first 80 characters of the Material Information 

Processor data. 

2. CROSS-SECTION LIBRARY NAME. This item specifies the cross-section library to be used in the 

criticality analysis. Any library available in the SCALE system may be used. However, the user must be 

aware that some libraries have a limited set of nuclides. For a bumup credit calculation it is important that the 

library selected has the actinides and/or fission products required for the analysis. The libraries frequently 

used are the 44GROUPNDF5 and 238GROUPNDF5 libraries. 

3. TYPE OF CALCULATION. The options are INFHOMMEDIUM, LATTICECELL, and MULTIREGION. 

The type of calculation is used only to define the geometry for the cross-section self-shielding analysis. 

4. STANDARD COMPOSITIONS. These data are used to define the mixtures that will be used in the problem. 

These data are required for every problem. Only one fuel type is permitted. The composition defined for the 

fuel should correspond to the initial, fresh fuel compositions. Additional mandatory entries are required for 

the cladding and any moderator mixtures as required by the Material Information Processor. Other materials 

may be defined to specify compositions for any other material required for the criticality analysis problem. 

The words "END COMP" end the standard composition entries. 

5. CELL GEOMETRY SPECIFICATION. These data are used to define geometry specifications for the cross

section self-shielding analysis. The entries are dependent on the type of calculations selected 

(INFHOMMEDIUM, LATTICECELL, or MULTIREGION). The geometry is only used in the self

shielding analysis of the cross sections for the criticality analysis.

21

Section 8
Input Description

Table 4 Standard composition specification'

Entry Variable

n1

2 MX 

SI FD

Mixture ID number 

Fuel density grams 

of U

Entry 

requirement 

Always

Comments 

Enter once for each standard composition. Enter the 

alphanumeric description from Table M8.2.1 of the 

SCALE manual. Additional allowed names include 

those beginning with ARBM for arbitrary materials, and 

SOLN for solutions

umber name Type of data 

1 SC Standard 

composition 

component name 

Al ROTH Theoretical density 

of material (g/cc) 

A2 NEL Number of 

elements in the 

material 

A3 MS No longer used but 

must still be 

entered 

A4 ICP Compound 

indicator 

A5 IRS No longer used but 

must still be 

entered 

A6 NCZA ID number (from 

far right column 

of Table M8.2.1 

of the SCALE 

manual) 

A7 ATPM Number of atoms 

of this element 

per molecule of 

arbitrary material 

or 

Weight percent of 

this element in 

this arbitrary 

material

ARBM 

and 

ICP=I 

or 

ARBM 

and 

ICP=O

Repeat the sequence A6 and A7 for each element in the 

arbitrary material before entering entry 2. Do not enter a 

value unless ICP=l 

or 

Repeat the sequence A6 and A7 for each element in the 

arbitrary material before entering entry number 2. 

Do not enter a value unless ICP=O

Always Enter once for each standard composition component 

SOLN Enter once for a solution or Pu per liter of solution

22

ARBM Enter once for each standard composition component that is 

an arbitrary material 

ARBM Enter once for each standard composition component that is 

an arbitrary material 

ARBM Enter once for each standard composition component that is 

an arbitrary material. Enter 0 or 1. 

ARBM Enter once for each standard composition component that is 

an arbitrary material. 

Enter I for a compound and 0 for alloys, mixtures, etc. 

ARBM Enter once for each standard composition component that is 

an arbitrary material. Enter 0 or 1. 

ARBM Repeat the sequences A6 and A7 for each element in the 

arbitrary material before entering entry number 2. Enter 

the number from the far right column of Table M8.2. 1. 

(Premixed standard compositions cannot be used in an 

arbitrary material definition.)

Section 8
Input Description

Table 4 (continued) 

Entry Variable Entry 

number name Type of data requirement Comments 

S2 AML Acid molarity of SOLN Enter once for a solution. AML=0 if there is no acid in the 

the solution solution

01 SPGR Specific gravity of 

the solution

or 

ROTH Density of the 

basic standard 

composition

Optional If the specific gravity (SPGR) of the solution is known, it 

should be entered as SPG=SPGR 

or 

If the density of a basic standard composition (ROTH) is to 

be entered, use DEN=ROTH

3 VF Density multiplier See comment 

column

Enter the density multiplier (density fraction, volume 

fraction, or a combination). Default value is 1. This item 

can be omitted if entries 4, 5, 6a, and 6b are also omitted. 

VF=0 is not allowed for SOLN or ARBM

4 ADEN Number density 

(atoms/b-cm) 

for the nuclide 

5 TEMP Temperature, in 

degrees K

VF=0 Enter only if VF=0.0

See comment 

column

Default value is 293 K. The temperature is applied in the 

criticality model. This entry can be omitted if entries 6a 

and 6b are also omitted

6a IZA 

6b WTP 

7b END 

END 

COMP

Isotope's ZA 

number 

Weight percent of 

the isotope 

Terminate a 

standard 

composition 

Terminate the data 

block

VF•O Enter for each isotope in the standard composition 

component. Omit if VF=0. Entries 6a and 6b are entered 

in pairs until each isotope in the component is defined 

VF*0 Enter for each isotope in the standard composition 

component. Omit if VF=0. Entries 6a and 6b are entered 

in pairs until each isotope in the component is defined 

Always Enter once for each standard composition component. 

This terminates the data for a standard composition 

component. Enter END to terminate the component. 

Repeat entries 1 through 7 until all the mixtures have been 

defined. At least two blanks must separate entry 7 from 

the next entry 

Terminus Enter once for a problem. Enter the words END COMP 

when all the standard composition components have been 

described. At least two blanks must follow the keyword 

END COMP

23

"aSee Section M7.4.4 of the SCALE manual for detailed description of each input variable. 

"bNOTE: Entry 7 should not begin in column 1 unless a name is associated with it. At least two blanks should separate the 

last entry 7 from the keyword END COMP.

Section 8
Input Description Section 8 

8.4 Control Parameter Data 

The control parameter data block allows the user to specify control parameters and array data for the problem. All input is by keyword entries. All keywords are three-character identifiers that must be followed immediately by an equals sign ("="). The keywords may be in any order within a data block. Input to the parameter data block is initiated with the data block keywords READ CONTROL (only first four characters of block name are required). The data block is terminated by the keywords END CONTROL. 

The types of control parameter data that may be input is summarized in Table 5. The individual keyword entries 

are briefly explained below. 

1. NCY= NUMBER OF IRRADIATION CYCLES defined in the burnup history data block. This is an 

optional input entry, and is automatically determined from the user input in the power history data 

block if not input explicitly. If NCY is input, the value is only used to check the value determined from the power history data block. If multiple libraries are requested for a cycle using the NLIB= keyword in the power history data block, the code subdivides the input cycle into NLIB cycles. A separate depletion cross-section library is generated for each cycle. 

2. ARP= NAME OF THE ARP LIBRARY TO BE USED. A character string with the name of the ARP 

library to be used in the depletion calculation. This is a required entry. The library must be 

defined in the SCALE text file ARPDATA.TXT that contains the cross-section library names and 

interpolation data used by ARP. 

3. NAX= NUMBER OF AXIAL ZONES. This is the number of axial burnup subdivisions. For a user

input profile the value of NAX is determined automatically by the code, and the NAX keyword is 

optional, provided the AXP= array has been entered. The maximum value of NAX is 100. By default, the profile is automatically normalized by the code unless NPR-no. Built-in 18-axial 

zone profiles may be selected with entries of-18 or -19. These built-in profiles, and the burnup 

range over which they are applied, are listed in Table 6 and Table 7. These profiles have been 

proposed elsewhere (Refs. 13 and 14) as bounding axial profiles, and are included as options for 

convenience only. The default value is -18 (built-in profiles). 

4. NHZ= NUMBER OF HORIZONTAL ZONES. This is the number of uniform horizontal bumup subdivisions in the assembly. Optional entry if no horizontal profile is requested. Maximum 

value of 10. 

5. NUC= THE BURNUP CREDIT NUCLIDES USED IN THE CRITICALITY CALCULATION. A list 

of actinides and/or fission products that are to be included in the criticality safety calculation. 

Any nuclides in the fuel matrix that are neither actinides nor fission products are always automatically included in the analysis. This is an array entry keyword and is delimited by an 

END. The nuclides are entered using their standard composition alphanumeric names, as listed 

in Section M8 of the SCALE manual. Isotopic correction factors (multiplied times the nuclide 

concentration) may be entered, optionally, immediately following the nuclide name. The 

concentration of any nuclide that does not have a correction factor is not adjusted. To select all 

available nuclides (with cross-section data) for the criticality calculation, the user may select 

NUC= ALL, without an END terminator. This is the only situation where an array entry does not 

require an END.

24
Input Description

6. AXP= AXIAL BURNUP PROFILE. The axial burnup profile of the assembly to be used in the analysis. 

This is an array entry and must be delimited by an END. If NAX is input, the AXP array must 

contain NAX entries. Otherwise, the value of NAX is determined automatically by the code. By 

default (NPR=yes), the profile is automatically normalized by the code unless NPR=no. If the 

burnup profile is normalized, it is implicitly assumed that the height/volume of each axial region 

is uniform when determining the average fuel burnup (i.e., the burnup of each axial region is 

equally weighted). The user is cautioned that if nonuniform axial subdivisions are used, 

normalization should not be used and the user must ensure a correct correspondence 

between the axial profile input and the axial regions specified in the criticality calculation. 

7. HZP= HORIZONTAL BURNUP PROFILE. The horizontal burnup profile of the assembly. This is an 

optional array entry and must be delimited by an END. If NAZ is input, the HZP array must 

contain NAZ entries. Otherwise, the value of NAZ is determined automatically by the code. The 

profile will be normalized if NPR=yes (default). 

8. FIX= FIXED ASSEMBLY POWER OPTION. Option to select a constant specific power level for the 

depletion analysis for all axial and horizontal zones of the assembly. For FIX=yes, the depletion 

analysis for all zones is performed using the specific power input in the power history data block 

for the POWER= keyword. The irradiation time is adjusted to achieve the desired burnup. The 

default of FIX=no applies a variable power for all zones and a constant irradiation time as defined 

by the BURN= keyword. 

9. NPR= NORMALIZE PROFILE. Option to control whether the user input axial and horizontal burnup 

profiles will be normalized. The input profiles are automatically normalized using NPR=yes 

(default). 

10. PRT= PRINT LEVEL. Option to set the type and level of printed output. Values range from 0 to 10. 

This option is currently not functional. 

11. BUG= DEBUG PRINT OPTION. BUG=yes will print program debugging variables and arrays in 

STARBUCS. The default is BUG=no.

25

Section 8
Input Description

Table 5 Table of control parameter data 

Keyword Data Default 

Name Type Value Comments 

READ CONTROL Initiate reading the control parameter block of data

NCY= 

ARP= 

NAX= 

NHZ= 

NUC=

Character

END CONTROL

None 

None 

-18 

1 

None

Optional keyword. Number of irradiation cycles. If not used, 

NCY is set automatically from the power history data 

input. 

Name of the ARP library to be used. Required. Library must 

be defined in SCALE text file ARPDATA.TXT. 

Number of axial burnup subdivisions in fuel assembly. The 

value of NAX is determined automatically if an axial 

profile is input using AXP= entries. The maximum value 

of NAX is 100. Default value (-18) applies a built-in 

18-axial region bumup profile. 

Number of uniform horizontal burnup subdivisions. 

Maximum value of 10. No entry is required if horizontal 

profile is not used. 

List of bumup credit nuclides, and optionally the 

corresponding isotopic correction factors, to be included in 

the criticality calculation. Array entry generally delimited 

by END, unless ALL is selected. Nuclides are input using 

their standard composition alphanumeric identifiers.

Integer variable 

Character variable 

Integer variable 

Integer variable 

Character and real 

mixed array' 

Real arrayl 

Real array' 

Real array' 

Character 

Character 

Integer variable

None 

I .

Horizontal bumup profile array. Required if NHZ > 1. Array 

containing NHZ entries that define the horizontal, or 

radial, bumup profile for the analysis. Array is 

automatically normalized by the code. Delimited by END. 

Axial moderator density, applied in the fuel depletion 

analysis. Note that MOD= is only used if the ARP library 

contains variable moderator density cross sections. NAX 

entries ordered as AXP= array. Delimited by END.

No Option to select a constant specific power level for all axial 

and horizontal zones of the assembly using FIX=yes.

Yes Option to normalize user-input axial and horizontal bumup 

profiles. Default is to automatically normalize profiles. 

Optional print level control. Currently not implemented.

No Optional debug printout with BUG=yes. 

End of the control parameter block of data

" Terminate with END. Do not place this END in column 1.

26

See NAX Axial burnup profile array. Required if NAX > 0. NAX 

entries that define the axial burnup shape. The profile is 

automatically normalized ifNPR=I. Delimited by END.

AXP= 

HZP=

MOD=

FIX= 

NPR= 

PRT= 

BUG=

Section 8

I
Input Description

Table 6 Built-in burnup-dependent axial profiles, NAX= -18 (from Ref. 13) 

Fraction of >30 GWd/t 18-30 GWd/t < 18 GWd/t 

Axial node core height 1 2 3 

1 0.0278 0.649 0.668 0.652 

2 0.0833 1.044 1.034 0.967 

3 0.1389 1.208 1.150 1.074 

4 0.1944 1.215 1.094 1.103 

5 0.2500 1.214 1.053 1.108 

6 0.3056 1.208 1.048 1.106 

7 0.3611 1.197 1.064 1.102 

8 0.4169 1.189 1.095 1.097 

9 0.4722 1.188 1.121 1.094 

10 0.5278 1.192 1.135 1.094 

11 0.5833 1.195 1.140 1.095 

12 0.6389 1.190 1.138 1.096 

13 0.6944 1.156 1.130 1.095 

14 0.7500 1.022 1.106 1.086 

15 0.8056 0.756 1.049 1.059 

16 0.8611 0.614 0.933 0.971 

17 0.9167 0.481 0.669 0.738 

18 0.9722 0.284 0.373 0.462

27

Section 8
Input Description

Table 7 Built-in burnup-dependent axial profiles, NAX= -19 (from Ref. 14) 

Fraction >46 42-46 38-42 34-38 30-34 26-30 22-26 18-22 14-18 10-14 6-10 <6 

Axial Aiafctor of core GWd/t GWd/t GWd/t GWd/t GWd/t GWd/t GWd/t GWd/t GWd/t GWd/t GWd/t GWd/t 

node height 1 2 3 4 5 6 7 8 9 10 11 12 

1 0.0278 0.573 0.674 0.660 0.585 0.652 0.619 0.630 0.668 0.649 0.633 0.662 0.574 

2 0.0833 0.917 0.949 0.936 0.957 0.967 0.924 0.936 1.034 1.044 0.989 0.930 0.947 

3 0.1389 1.066 1.053 1.044 1.091 1.074 1.056 1.066 1.150 1.208 1.019 1.049 1.091 

4 0.1944 1.106 1.085 1.080 1.121 1.103 1.097 1.103 1.094 1.215 0.857 1.059 1.105 

5 0.2500 1.114 1.095 1.091 1.126 1.108 1.103 1.108 1.053 1.214 0.776 1.108 1.094 

6 0.3056 1.111 1.095 1.093 1.111 1.106 1.101 1.109 1.048 1.208 0.754 1.144 1.087 

7 0.3611 1.106 1.093 1.092 1.094 1.102 1.103 1.112 1.064 1.197 0.785 1.168 1.086 

8 0.4169 1.101 1.091 1.090 1.093 1.097 1.112 1.119 1.095 1.189 1.013 1.183 1.087 

9 0.4722 1.097 1.089 1.089 1.092 1.094 1.125 1.126 1.121 1.188 1.185 1.189 1.091 

10 0.5278 1.093 1.088 1.088 1.091 1.094 1.136 1.132 1.135 1.192 1.253 1.190 1.096 

11 0.5833 1.089 1.086 1.088 1.092 1.095 1.143 1.135 1.140 1.195 1.278 1.183 1.102 

12 0.6389 1.086 1.084 1.086 1.099 1.096 1.143 1.135 1.138 1.190 1.283 1.167 1.105 

13 0.6944 1.081 1.081 1.084 1.096 1.095 1.136 1.129 1.130 1.156 1.276 1.135 1.105 

14 0.7500 1.073 1.073 1.077 1.087 1.086 1.115 1.109 1.106 1.022 1.251 1.079 1.096 

15 0.8056 1.051 1.053 1.057 1.073 1.059 1.047 1.041 1.049 0.756 1.193 0.976 1.066 

16 0.8611 0.993 0.987 0.996 1.003 0.971 0.882 0.871 0.933 0.614 1.075 0.806 0.986 

17 0.9167 0.832 0.800 0.823 0.796 0.738 0.701 0.689 0.669 0.481 0.863 0.596 0.806 

18 0.9722 0.512 0.524 0.525 0.393 0.462 0.456 0.448 0.373 0.284 0.515 0.375 0.474

28

Section 8
Input Description

8.5 Power History Data 

The power history data block defines the irradiation history for the assembly. These data are entered by keyword. 

The keywords are summarized in Table 8. Only the first four characters of the keywords are used, and any 

characters after the first four characters may be input. A minimum of two entries are required for each cycle, 

(1) the average assembly power (POWER=) and (2) the irradiation time (BURN=). The decay time (DOWN=), 

if any, at the end of the cycle, and the number of cross-section libraries (NLIB=) are optional. The word END is 

required to delimit the entries for each cycle. The entries within a given cycle may be in any order. Logarithmic 

time intervals may be requested for the final cooling time by entering a negative value for the down time of the 

last cycle. 

If the optional NCY= keyword (in control parameter data block) is omitted, the number of cycles is determined 

automatically from the power history data. If the keyword is input, it is used only to check the number of cycles 

input in the power history data block. 

POWER= THE AVERAGE SPECIFIC POWER OF THE ASSEMBLY FOR THIS CYCLE. The units of 

the specific power are in W/g of initial uranium. The axial and horizontal profiles are multiplied 

by the specific power to achieve the desired spatially-dependent burnup profiles for the assembly 

when FIX=no (default). If FIX=yes, the specific power input using this keyword is assumed to 

be uniform over all fuel regions (axial and horizontal) and the code will adjust the irradiation time 

to obtain the desired bumup for each region. 

BURN= THE IRRADIATION TIME FOR THIS CYCLE. The cycle irradiation time in days. 

DOWN= CYCLE DOWN TIME. An optional entry to specify the down time, in days, at the end of an 

irradiation cycle. The down time is simulated as an irradiation time step of effectively zero 

power after the irradiation cycle. The down time for the last cycle is simulated as a separate 

ORIGEN-S decay case with nine equally spaced time steps. If a negative down time is input, the 

time steps are spaced logarithmically. 

NLIB= LIBRARIES PER CYCLE. An optional entry to request multiple cross-section libraries during a 

depletion cycle. If requested, the code automatically subdivides the cycle in NLIB segments of 

uniform duration and generates a separate library for the depletion analysis for each segment 

using ARP. Generating multiple libraries provides a more accurate representation of the time

dependent cross-section variation during the burnup analysis. Each segment of the cycle is 

assumed to have the same specific power, and no down time is assumed between each segment of 

the cycle. 

END The word END is required to terminate the input for each cycle. 

Repeat the above entries for each cycle to define the complete assembly power history. The word END is 

required to terminate the input for each cycle.

29

Section 8
Input Description

Table 8 Table of power history data 

Keyword Data Default 

Name Type Value Comments 

POWER= Real variable None Average assembly power for this cycle (Wig) 

BURN= Real variable None Cycle irradiation time (days) 

DOWN= Real variable 0 End-of-cycle decay time (days). Optional. A negative down 

time may be used to select logarithmic decay time intervals 

for the last decay case. 

NLIB/CYCLE= Integer variable I Number of ARP libraries to be applied in this cycle. 

Optional. If multiple libraries are requested for this cycle, the cycle is subdivided into equal time segments, and an 

updated library is generated for each segment. No down 

time is simulated between segments. 

END Required. Defines the end of the data for the current cycle. Repeat the above entries for each cycle in the irradiation 

history. An END, not to begin in column 1, must 

terminate each cycle definition. 

END MIST End of the power history block of data 

8.6 KENO Input Data 

The KENO V.a input for the problem is specified in the KENO data block. Input to the data block is initiated 

with the data block keywords READ KENO and is terminated by the keywords END KENO. STARBUCS 

performs no error checking of the KENO V.a input. The data within the data block delimiters is copied, without 

change, and executed. The user is advised to first verify the KENO V.a input using the SCALE Criticality Safety 

Analysis Sequence, CSAS, before using the input in STARBUCS. 

The input requirements for KENO V.a are not described in this section, but are described in detail in Section F 11 

of the SCALE manual. This section only describes the input requirements as related to execution of KENO V.a 

within STARBUCS and the conventions used for module compatibility. Any valid KENO V.a input may be used 

within STARBUCS, providing a large degree of flexibility in the types of spent fuel systems that may be 

effectively analyzed. 

The STARBUCS module automatically generates a macroscopic cross-section library for each of the spatially dependent spent fuel regions of a fuel assembly. Therefore, no isotopic composition information is required in the 

KENO V.a input since the macroscopic cross sections already include all concentration information. The cross 

sections are prepared as an AMPX working library stored on logical unit 40. The library contains a cross-section 

set for each region of the spent fuel assembly, plus cross sections for all of the structural and non-fuel materials 

required for the problem, as defined in the standard composition data (see Table 4). 

The cross-section identifiers (MIXTURE ID) for each of the non-fuel materials are defined as the mixture numbers 

(MX) specified in the standard composition input (see Table 4). The control module automatically defines the

30

Section 8
Input Description

MIXTURE ID for each of the fuel regions according to the axial and/or horizontal zone. The first axial zone 

mixture is assigned a MIXTURE ID of 101, and is incremented by one for each additional axial zone. Therefore, 

in a problem that defines 18 axial zones, spent fuel mixtures will be generated with identifiers that range from 

101 to 118. The correspondence of these mixtures to the assembly locations is determined by the ordering of the 

AXP= array input that define the axial burnup profile for the assembly. If the AXP= array is ordered from the 

bottom of the assembly to the top of the assembly, the resulting MLXTURE ID 101 would correspond to the 

bottom axial zone, and MIXTURE ID 118 would correspond to the top axial zone. If multiple horizontal zones are 

defined, then the numbering sequence of the second horizontal zone will start at 201 and, in the example given 

above, would range up to 218. This number convention allows a maximum of 99 unique axial zones, and 

9 unique horizontal zones, to be defined. The mixture-numbering scheme is illustrated in Figure 3. 

The AMPX cross-section library generated for KENO V.a is stored on logical unit 40. Therefore, the library unit 

number (LIB=) in the KENO V.a parameter data block must equal 40 in order to access the macroscopic cross 

sections for the problem, e.g., LIB=40. If the LIB= keyword is input by the user, it must equal 40. If the LIB= 

keyword is not entered in the parameter data block it is automatically added by STARBUCS, with the correct unit 

number assignment. An AMPX library requires that a cross-section mixing table data (MIXT data block) in 

KENO V.a of the form 

READ MIXT MIXING TABLE END MIXT 

The MIXING TABLE is used to specify each mixture and the nuclide IDs and number densities used in the 

mixtures. This data block is always automatically added to the KENO V.a input file by STARBUCS. Therefore, 

entering this data manually is neither required nor permitted. 

The following input illustrates a typical KENO V.a mixing table data block added by STARBUCS for six burnup

dependent axial zones: 

READ MIXT 

MIX=101 101 1 

MIX=102 102 1 

MIX=103 103 1 

MIX=104 104 1 

MIX=105 105 1 

MIX=106 106 1 

MIX= 2 2 1 

MIX= 3 3 1 

MIX= 4 4 1 

MIX= 5 5 1 

END MIXT 

In this example, the six axial burnup regions are defined with mixture numbers of 101, 102, 103, 104, 105, and 

106. Non-fuel mixtures 2, 3, 4, and 5 are also defined. These MIXTURE IDs defined by the first entry after the 

MIX= keyword are the cross-section identifiers for each region available in the library, and the second entry is the 

material identifier used in defining the KENO V.a geometry. The last entry for each mixture card is the atomic 

number density for this material. Note that the largest mixture number allowed for the non-fuel materials is 

MIX=100, since larger values will conflict with the predefined fuel mixture numbers. This example is provided 

only to illustrate the data required by KENO V.a. The mixing table is entered automatically by STARBUCS and 

no mixing table entries in the user input are needed or allowed.

31

Section 8
9 SAMPLE PROBLEMS 

A series of sample problems are presented to illustrate the application of STARBUCS to burnup credit criticality 

safety analyses. Sample problem 1 is a simple pin-cell problem with one axial burnup zone (i.e., uniform axial 

burnup). It is useful to illustrate the main features of the system and demonstrate functionality of the system 

modules within SCALE. Problem 2 illustrates the same problem with 18-axial burnup-dependent zones. 

Problem 3 extends the pin-cell model to an array of spent fuel assemblies residing in water. The burnup credit 

model applied 18-axial burnup-dependent zones. Problem 4 is a generic cask model, and this problem exercises 

more of the options available in STARBUCS. Problem 5 illustrates the use of the horizontal burnup option for a 

simple array of spent fuel assemblies. 

9.1 Sample Problem 1 

Sample problem 1, listed in Table 9, defines a simple infinite U0 2 pin-cell model with uniform axial burnup. 

The initial fuel enrichment is 2.0 wt %. The control parameter data block specifies that the standard 17 x 17 ARP 

library is to be used for the depletion analysis. The burnup credit criticality calculation uses a subset of the 

actinides in SNF. An illustrative power history is input that selects three cycles having different specific power 

levels, with a down time of 100 days after the second cycle, and a 1 year cooling time after irradiation. Multiple 

libraries per cycle are requested in the first and last cycle to provide a more accurate representation of the cross

section variation with burnup over the extended irradiation times of these cycles. 

9.2 Sample Problem 2 

Sample problem 2, listed in Table 10, illustrates a simple pin-cell model using 18-axial burnup-dependent zones. 

In this example, the built-in axial profiles for three burnup ranges are applied using the NAX= -18 option 

(see profiles in Table 6). STARBUCS determines the average assembly burnup from the power history data 

input, and automatically selects the appropriate profile based on the discharge burnup. The axial profile data are 

based on a predefined axial zoning structure (i.e., fraction of the core height). It is important that the KENO V.a 

model therefore also reflects this axial structure. That is, the height of each axial zone in the criticality model 

must conform to the axial zones for profiles applied in the analysis. In this example, the total pin height is 

365.7 cm (144 in.), which is subdivided into 18 equal-height segments of 20.32 cm each. 

The burnup-dependent cross sections generated for the criticality analysis have material identifiers ranging from 

101 (bottom) to 118 (top). There is no constraint on how the fuel materials can be applied in the KENO V.a 

model. For example, the order of the material numbers could easily be reversed, which would effectively invert 

the profile and could be used to simulate an assembly loaded upside down. It is also not necessary to use all of 

the materials in the problem. For instance, all fuel regions in the KENO V.a model could be assigned the same 

fuel mixture number to represent a flat axial profile having a burnup value equal to that of the particular mixture 

used. The average assembly burnup would also be equal to that of the particular mixture used, and not that 

defined by the power history data block.

33
Table 9 STARBUCS input listing for sample problem 1

=starbucs 

w17x17 pwr uniform axial 

44groupndfS latticecell 

"uo2 1 den=10.96 0.95 

zircalloy 

cr 2 0 7.5891E-05 

fe 2 0 1.4838E-04 

zr 2 0 4.2982E-02 

water 

h2o 3 1 end 

end comp I

--- fuel-pin-cell geometry: 

squarepitch 1.2598 0.8192 1 3

burnup pin cell model 

293.0 92235 2.0 92238 98.0 end 

end 

end 

end

0.8357 2 end

read parm 

arp=17x17 

axp= 1 end 

nuc= u-234 u-235 u-236 u-238 pu-238 pu-239 pu-240 

pu-241 pu-242 am-241 am-242m am-243 np-237 end 

end parm 

read hist 

power=35.00 burn=100 nlib=2 end 

power=28.7 burn=23 down=100 end 

power=24. burn=300 nlib=4 down=365.25 end 

end hist 

read keno 

infinite pin cell lattice 

* materials 

"* 101 = uo2, uniform axial region 

2 = Zircaloy 

* 3 = Water 

keno model for infinite pin cell lattice 

read param tme=10000 gen=510 nsk=10 npg=1000 end parm 

read geom I Fuel Pin 

unit 1 

cylinder 101 1 0.4782 50.0 -50.0 

cylinder 2 1 0.5588 50.0 -50.0 

cuboid 3 1 4p0.7366 50.0 -50.0

end geom 

read bounds 

end data 

end keno 

end

all=reflect end bounds

Sample Problems Section 9

34
Sample Problems

Table 10 STARBUCS input listing for sample problem 2 

=starbucs 

w17x17 pwr pin cell model - 18 axial zones 

44groupndf5 latticecell 

uo2 1 den=10.96 0.95 293.0 92235 2.0 92238 98.0 end

'zircalloy 

cr 2 

fe 2 

zr 2 

'water

0 

0 

0

7.5891E-05 end 

1.4838E-04 end 

4.2982E-02 end

h2o 3 1 end 

end comp

' --- fuel-pin-cell geometry: 

squarepitch 1.2598 0.8192 1 3 0.8357 2 end

read control 

arp=17x17 nax=-18 

nuc= u-234 u-235 u-236 u-238 pu-238 pu-240 

pu-241 pu-242 am-241 am-242m am-243 np-237 end 

end control 

read hist 

power=35.001 burn=100 nlib=l end 

power=28.5 burn=230 down=100 nlib=2 end 

power=24.001 burn=300 nlib=2 down=1826 end 

end hist 

read keno 

* infinite pin cell lattice 

' materials 

' 101-118 = uo2, 18-axial zone model 

2 = Zircaloy 

'* 3 = Water 

keno model for infinite pin cell lattice 

read param tme=10000 gen=510 nsk=10 npg=1000 end param

read geom I 

unit 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder

Fuel Pin 

1

1 

1 

1 

1 

1 

1 

1 

1 

1 

1 

1 

1 

1 

1

101 

102 

103 

104 

105 

106 

107 

108 

109 

110 

111 

112 

113 

114

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782

-162.53 

-142.22 

-121.90 

-101.58 

-81.27 

-60.95 

-40.63 

-20.32 

0.00 

21.32 

40.63 

60.95 

81.27 

101.58

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85

35

Section 9
Sample Problems

cylinder 115 1 0.4782 121.90 -182.85 

cylinder 116 1 0.4782 142.22 -182.85 

cylinder 117 1 0.4782 162.53 -182.85 

cylinder 118 1 0.4782 182.85 -182.85 

cylinder 2 1 0.5588 182.85 -182.85 

cuboid 3 1 4p0.7366 182.85 -182.85 

end geom 

read bounds all=reflect end bounds 

end data 

end keno 

end 

9.3 Sample Problem 3 

Sample problem 3, listed in Table 11, performs a burnup credit criticality safety calculation using the 238-group 

ENDF/B-V cross-section library for an array of Combustion Engineering (CE) 14 x 14 spent fuel assemblies in 

water. A subset of burnup credit actinides and fission products are included in the criticality calculation. A user

supplied 18-axial-region burnup profile of the assemblies is input. This profile was obtained from the axial 

bumup profile database' 7 for Maine Yankee assembly N863. Note that the axial profile will be normalized by the 

code (sum to a value equal to the number of axial zones) using NPR=yes (default). The 3.3 wt % enriched U0 2 

fuel is assumed to achieve a discharge burnup of 37626 MWd/t in three cycles of approximately 12.5 GWd/t per 

cycle and a downtime per cycle of 80 days, followed by a cooling time after irradiation of 5 years (1826 days). An average assembly power level of 32 MW/t is used for the depletion calculation. Two libraries per cycle are 

requested during the depletion. The nominal CE 14 x 14 assembly design specifications were obtained from 

Ref. 18. The assembly pitch in the criticality calculations is 22.78 cm. A cross-section view of the assembly 

geometry, a 2 x 8 array of water reflected assemblies, is illustrated in Figure 4.

36

Section 9
Section 9 Sample Problems 

NNV 

n•..n.rW*Of 

Z 

ii~I~i 

~rn21, 

-~st~~nn* †††† 

~Iii~ii~ ~h*. 6 

0 F37
Sample Problems Section 9 

Table 11 STARBUCS input listing for sample problem 3 

=starbucs 

ce 14x14 asmby 

238groupndf5 1atticecell 

'initial fuel 

uo2 1 den=10.045 1 273 92234 0.0294 92235 3.3 92236 0.0152 92238 96.6554 end 'zircalloy 

arbmzirc 6.44 4 0 0 1 40000 97.91 26000 0.5 50116 0.86 50120 0.73 2 1 620 end 'water 

h2o 3 1 end 

end comp 

fuel-pin-cell geometry: squarepitch 1.4732 0.9563 1 3 1.1176 2 0.9855 0 end 

read control 

arp=14x14 nax=18 

axp= 

0.67053 0.93322 1.02433 1.05329 1.06026 1.06185 

1.06215 1.06249 1.06312 1.06408 1.06541 1.06702 

1.06836 1.06760 1.05918 1.02515 0.92262 0.66935 end 

nuc= 

u-234 u-235 u-236 u-238 pu-238 pu-239 pu-240 pu-241 pu-242 am-241 np-237 

mo-95 tc-99 ru-101 rh-103 ag-109 cs-133 nd-143 

nd-145 sm-147 sm-149 sm-150 sm-151 eu-151 sm-152 

eu-153 gd-155 end 

end control 

read hist 

power=32.00 burn=391.937 nlib=2 down=80 end 

power=32.00 burn=391.937 nlib=2 down=80 end 

power=32.00 burn=391.937 nlib=2 down=1826 end 

end hist 

read keno 

ce 14x14 assembly model, 2x8 array of assemblies 

* materials 

* 101 = uo2, lower axial region (0.67053) 

* 118 = uo2, upper axial region (0.66935) 

2 = Zircaloy 

* 3 = Water 

read param 

tme=10000 gen=510 nsk=10 npg=1000 

lib=40 

end parm 

read geom ' Fuel Pin 

unit 1 

cylinder 101 1 0.4782 -162.53 -182.85 

cylinder 102 1 0.4782 -142.22 -182.85 

cylinder 103 1 0.4782 -121.90 -182.85 

cylinder 104 1 0.4782 -101.58 -- 182.85 

cylinder 105 1 0.4782 -81.27 -182.85 

cylinder 106 1 0.4782 -60.95 -182.85

38
Section 9 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cuboid

107 

108 

109 

110 

111 

112 

113 

114 

115 

116 

117 

118 

2 3

1 

1 

1 

1 

1 

1 

1 

1 

1 

1 

1 

1 

1 1

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.5588 

4 p0.7366

-40.63 

-20.32 

0.00 

21.32 

40.63 

60.95 

81.27 

101.58 

121.90 

142.22 

162.53 

182.85 

182.85 

182.85

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85

2 x 2 Array of Fuel Pins

unit 

array 1 3*0

2

' Large Water Hole 

unit 3 

cylinder 3 1 1.3140 

cylinder 2 1 1.4160 

cuboid 3 1 4p1.4732

182.85 

182.85 

182.85

-182.85 

-182.85 

-182.85

' Assembly Unit 

unit 4 

array 2 -10.3124 -10.3124 -182.85 

cuboid 3 1 

4 pll.390 182.85 -182.85

I Assembly Array 

global 

unit 5 

array 3 3*0 

reflector 3 1 6r. 

end geom 

read array 

ara=l nux=2 nuy=2

2x8)

30).0 1 

nuz=1 fill

11 

1 1 end fill 

ara=2 nux=7 nuy=7 nuz=l fill 

2222222 

2322232 

2222222 

2223222 

2222222 

2322232 

2 2 2 2 2 2 2 end fill 

ara=3 nux=2 nuy=8 nuz=l fill 

16r4 end fill 

end array 

read bounds all=void end bounds 

end data 

end keno 

end

39

Sample Problems
Sample Problems Section 9

9.4 Sample Problem 4 

Sample problem 4, listed in Table 12, illustrates the application of STARBUCS for a criticality safety analysis of 

a bumup credit cask. The cask geometry in this example is based on a 32-assembly generic burnup credit cask 

model and is illustrated in Figure 5. 

The assemblies are assumed to be Westinghouse 17 x 17 OFA assemblies with an initial enrichment of 4.98 wt %. 

The standard composition description for this problem includes the fuel assembly and all cask structural material 

definitions. The analysis applies built-in 18-axial zone profiles, and actinide-only burnup credit (i.e., only a 

subset of actinides and no fission products). The assembly is irradiated to an average burnup of about 

50 GWd/MTU. The axial burnup profile is automatically selected by the code based on the average assembly burnup. Isotopic correction factors are applied to the calculated actinide inventories. The correction factors were 

obtained from Ref. 19. An axial moderator density is also applied. Note that entries in the MOD= array are not 

realistic for a PWR and are only intended to illustrate the use of this feature. Since the ARP library applied in this 

calculation does not have variable moderator density, the values in the MOD= array have not effect on the 

calculation. The criticality evaluation of the cask is performed following a cooling time of 1826 days (5 years). The calculations apply the 44-group ENDF/B-V cross-section library.

Figure 5 Plot of the generic 32-assembly burnup credit cask geometry

40
Sample Problems

Table 12 STARBUCS input listing for sample problem 4

=starbucs 

w17x17 pwr 18-axial zones, gbc-32 cask model 

44groupndf5 latticecell 

uo2 I den=10.96 0.95 293.0 92235 4.98 92238 95.02 end 

'zircalloy 

cr 2 0 7.5891E-05 end 

fe 2 0 1.4838E-04 end 

zr 2 0 4.2982E-02 end 

.water 

h2o 3 1 end 

'Stainless steel 

ss304 4 1 end

' boral center - b-1O loading of 

b-10 5 0 6.5795E-03 293.0 

b-li 5 0 2.7260E-02 293.0 

c 5 0 8.4547E-03 293.0 

al 5 0 4.1795E-02 293.0 

'Stainless steel 

ss304 6 1 end 

' aluminum 

al 7 0 0.0602 293.0 

end comp

--- fuel-pin-cell geometry: 

squarepitch 1.25984 0.819

read control 

arp=17x17 nax=-18 

nuc= u-234 0.635 

u-235 1.085 

u-236 0.910 

u-238 0.992 

pu-238 0.856 

pu-239 1.076 

pu-240 0.945 

pu-241 1.087 

pu-242 0.848 

am-241 0.609 

am-243 0.804 

np-237 0.697 end 

mod= 0.720 0.709 0.699 

0.646 0.635 0.625 

0.572 0.562 0.551 

end control

read hist 

power=32.89 burn=100 

power=32.89 burn=200 

power=32.89 burn=900 

power=32.89 burn=320

0.688 

0.614 

0.540

0.0225 g/cm3 

end 

end 

end 

end

end

20 1 3 0.83570 2 end 

0.678 0.667 0.657 

0.604 0.593 0.583 

end

end 

end 

nlib=3 end 

down=-1826 end

end hist 

read keno 

keno model for Generic 32-Assembly Burnup Credit Cask (GBC-32)

41

Section 9
Sample Problems 

* Assembly Type: Westinghouse 17x17 OFA/V5 

* Materials 

* 101 - 118 = uo2, axial regions 1 through 18 

2 = Zircaloy 

* 3 = Water 

* 4 = Stainless Steel 

5 = Boral 

* 6 = Stainless Steel 

7 = Al 

read param tme=10000 gen=510 nsk=10 npg=1000 end param

Fuel Pin 

1

101 

102 

103 

104 

105 

106 

107 

108 

109 

110 

111 

112 

113 

114 

115 

116 

117 

118 

0 

2 

3

0.40958 -162.53 -182.85 

0.40958 -142.22 -182.85 

0.40958 -121.90 -182.85 

0.40958 -101.58 -182.85 

0.40958 -81.27 -182.85 

0.40958 -60.95 -182.85 

0.40958 -40.63 -182.85 

0.40958 -20.32 -182.85 

0.40958 0.00 -182.85 

0.40958 21.32 -182.85 

0.40958 40.63 -182.85 

0.40958 60.95 -182.85 

0.40958 81.27 -182.85 

0.40958 101.58 -182.85 

0.40958 121.90 -182.85 

0.40958 142.22 -182.85 

0.40958 162.53 -182.85 

0.40958 182.85 -182.85 

0.41783 182.85 -182.85 

0.47498 182.85 -182.85 

2p0.6299 2pO.6299 182.88

read geom I 

unit 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cuboid 

unit 

cylinder 

cylinder 

cuboid 

unit 

cuboid 

cuboid

Guide Thimble/Instrument Tube (assumed to be same) 

2

3 

2 

3

1 0.5613 

1 0.6020 

1 0.6299 -0.6299

Top Half Horizontal Boral Panel 

4

7 1 9.5250 

5 1 9.5250

-9.5250 

-9.5250

365.76 

365.76 

0.6299 -0.6299 365.76

0.02540 -0.0 

0.128270 -0.0

0. 

0. 

0.

Right-Hand Side Half Vertical Boral Panel 

5

7 1 0.02540 -0.0 

5 1 0.128270 -0.0

Bottom Half Horizontal Boral Panel 

6

7 1 9.5250 

5 1 9.5250

-9.5250 

-9.5250

9.5250 -9.5250 365.76 0. 

9.5250 -9.5250 365.76 0.

0.0 -0.02540 365.76 0. 

0.0 -0.128270 365.76 0.

Left-Hand Side Half Vertical Boral Panel 

7

42

Section 9

-182.88

unit 

cuboid 

cuboid 

unit 

cuboid 

cuboid

365.76 0. 

365.76 0.

unit
Sample Problems

1 

1

0.0 

0.0

Assembly Basket Cell 

101

cuboid 

cuboid 

unit 

array 

cuboid 

cuboid 

cuboid 

hole 

hole 

hole 

hole 

unit 

cuboid 

cuboid 

hole 

I

5 11.75000 

6 0. 

7 -11.75000

-0.02540 9.5250 -9.5250 

-0.128270 9.5250 -9.5250

-10.7086 

-11.0000 

-11.7500 

-11.87827 

11.75000 

0. 

-11.75000 

0.

Top Boral/Basket Plate 

110

1 

1

11.7500 -11.7500 

11.7500 -11.7500 

0. -0.7500

0. 

11.0000 

11.7500 

11.87827 

0. 

0. 

0. 

0.

0.0000 

0.0000 

0.

-11.0000 

-11.7500 

-11.87827

-0.7500 

-0.8783

Bottom Boral/Basket Plate 

111

4 

3 

4

1 

1

11.7500 -11.7500 

11.7500 -11.7500 

0. 0.7500

0.7500 

0.8783 

0.

Left-Hand Side Boral/Basket Plate 

112

4 1 0.0000 

3 1 0.0000 

7 -0.75000

-0.7500 

-0.8783 

0.

10.9999 

10.9999 

0.

Right-Hand Side Boral/Basket Plate 

113

4 1 0.7500 

3 1 0.8783 

5 0.75000

Cask Inner Volume 

unit 200 

3 1 87.500 

Assemblies 

101 -35.634840 

101 -11.878270 

101 11.878270 

101 35.634840

101 

101 

101 

101 

101 

101

-59.391400 

-35.634840 

-11.878270 

11.878270 

35.634840 

59.391400

101 -59.391400 

101 -35.634840

-0.0000 

-0.0000 

0.

10.9999 

10.9999 

0.

-0.0000 

-0.0000 

-10.9999 

-10.9999

-10.9999 

-10.9999

365.76 0. 

365.76 0. 

365.76 0. 

365.76 0.

365.76 0. 

365.76 0.

395.76 -30.00

59.391400 

59.391400 

59.391400 

59.391400 

35.634840 

35.634840 

35.634840 

35.634840 

35.634840 

35.634840

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0.

11.878270 0. 

11.878270 0.

43

7 

5

365.76 0. 

365.76 0.

1 

3 

4 

3 

4

-10.7086 

1 11.0000 

1 11.7500 

1 11.87827 

0.

365.76 

365.76 

365.76

0. 

0. 

0.

4 

3 

6

365.76 0. 

365.76 0.

unit 

cuboi d 

cuboid 

hole 

unit 

cuboid 

cuboid 

hole 

unit 

cuboid 

cuboid 

hole 

global 

cylinder 

hol e 

hole 

hole 

hole 

hole 

hole 

hole 

hole 

hole 

hol e 

hol e

hole 

hole

Section 9
Sample Problems Section 

9

hole 

hole 

hole 

hole 

hole 

hole 

hole 

hole 

hole 

hole 

l 

hole 

hole 

hole 

hole 

hole 

hole hol e 

hol e hole 

hole 

hole 

hole 

hole 

hole 

hol e 

hol e 

hole 

hol e 

hole 

hol e 

hol e 

hol e 

I 

hole 

hole 

hol e 

hol e 

hol e 

hol e 

hol e 

hol e 

hol e 

hole 

hole 

holie 

hol e 

hol e 

hol e 

hol e 

hol e 

hol e 

hol ner

101 

101 

101 

101 

101 

101 

101 

101 

101 

101 

101 

101 

101 

101 

101 

101 

101 

101 

101 

101

Exterior Half Boral Panels 

Top Plates 

110 -35.634840 72.14; 

110 -11.878270 72.14; 

110 11.878270 72.14 

110 35.634840 72.14; 

110 59.391400 48.39: 

110 -59.391400 48.39: 

Bottom Plates 

111 -35.634840 -72.14 

111 -11.878270 -72.14 

111 11.878270 -72.14 

111 35.634840 -72.14; 

111 59.391400 -48.391 

111 -59.391400 -48.391 

Left-Hand Side Plates 

113 -48.391420 59.391 

113 -72.147980 35.63z 

113 -72.147980 11.878 

113 -72.147980 -11.878 

113 -72.147980 -35.634 

113 -48.391420 -59.391 

Right-Hand Side Plates 

112 48.391420 59.391 

112 72.147980 35.634 

112 72.147980 11.878 

112 72.147980 -11.87E 

112 72.147980 -35.634 

112 48.391420 -59.391

7980 

7980 

7980 

7980 

1420 

1420 

7980 

7980 

7980 

1980 

1420 

1420 

1400 

4840 

3270 

3270 

•840 

.400 

.400 

•840 

3270 

1270 

,840 

.400

Steel Cask/Overpack 

6 1 107.5

-11.878270 

11.878270 

35.634840 

59.391400 

-59.391400 

-35.634840 

-11.878270 

11.878270 

35.634840 

59.391400 

-59.391400 

-35.634840 

-11.878270 

11.878270 

35.634840 

59.391400 

-35.634840 

-11.878270 

11.878270 

35.634840

Cube of Water Surrounding Cask

44

11.878270 

11.878270 

11.878270 

11.878270 

-11.878270 

-11.878270 

-11.878270 

-11.878270 

-11.878270 

-11.878270 

-35.634840 

-35.634840 

-35.634840 

-35.634840 

-35.634840 

-35.634840 

-59.391400 

-59.391400 

-59.391400 

-59.391400

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

0. 

415.76 -60.
Section 9 Sample Problems 

cuboid 0 1 108 -108 108 -108 415.76 -60. 

end geom 

Assembly Type: Westinghouse 17x17 OFA/V5 

read array 

ara=1 nux=17 nuy=17 nuz=1 

fill 

11111111111111111 

11111111111111111 

11111211211211111 

11121111111112111 

11111111111111111 

11211211211211211 

11111111111111111 

11111111111111111 

11211211211211211 

11111111111111111 

11111111111111111 

11211211211211211 

11111111111111111 

11121111111112111 

11111211211211111 

11111111111111111 

11111111111111111 

end fill1 

end array 

read bounds xyf=specular end bounds 

end data 

end keno 

end 

9.5 Sample Problem 5 

Sample problem 5, listed in Table 13, uses the CE 14 x 14 assembly design from problem 3, and performs a 

burnup credit calculation using the horizontal burnup profile option. The assembly configuration is taken to be a 

simple 2 x 2 assembly array with water reflection. This problem is only designed to illustrate the basic features of 

the horizontal profile option. In this example, it is assumed that there is a burnup gradient across the assemblies, 

such that half the fuel pins have a burnup exceeding the average assembly burnup by 10% and half the pins have a 

bumup of 10% less than the average, with the two burnup regions separated by the assembly diagonal. The input 

card required to simulate the two horizontal burnup regions in an assembly is 

hzp= 0.9 1.1 end 

STARBUCS applies these factors to calculate compositions for each of the horizontally-varying burnup regions in 

each zone of the problem. It is important to note that the option inherently assumes that there is an equal 

volume/mass of fuel in each of the horizontal (an axial) zones since the code weights all regions equally when 

determining the average assembly burnup. To illustrate this, consider modeling an assembly with only one 

quadrant having a burnup that is 10% higher than the other three quadrants. The user would enter data for each 

of the four horizontal assembly quadrants or zones, e.g.,

45
Sample Problems

hzp= 0.9766 0.9766 0.9766 1.0700 end 

such that the average of the HZP array entries is unity. This ensures that the average assembly burnup will be that 

specified in the power history data block. Note that this array is automatically normalized in NPR=yes (default). 

However, the user could substantially reduce the computational time involved by specifying only two fuel 

regions, e.g., 

hzp= 0.9766 1.0700 end 

and turning off the normalization option (e.g., NPR=no). The normalization option must be turned off to prevent the profile from being altered (since the sum is not equal to 2). This allows the user to account for the fact that, in 

this scenario, there are three quadrants having a lower bumup (and consequently three times the mass) and just 

one quadrant having an elevated burnup compared to the average. However, it is the responsibility of the user to 

ensure that the profiles and the KENO V.a problem description produce the desired average burnup. 

In this sample problem the four assemblies are aligned so the lower burnup regions of the assemblies are adjacent 

to one another to maximize the system reactivity. The assembly geometry showing the different burnup regions of the assemblies is illustrated in Figure 6. The criticality calculation is performed using the 238-group ENDF/B

V cross-section library.

46

Section 9
Sample Problems

Table 13 STARBUCS input listing for sample problem 5 

=starbucs 

ce 14x14 asmby 

238groupndf5 latticecell 

'initial fuel 

uo2 1 den=10.045 1 273 

92234 0.027 92235 3.038 92236 0.014 92238 96.921 end 'zircalloy 

arbmzirc 6.44 4 0 0 1 40000 97.91 26000 0.5 50116 0.86 50120 0.73 2 1 620 end 'water 

h2o 3 1 end 

end comp 

S--fuel-pin-cell geometry: 

squarepitch 1.4732 0.9563 1 3 1.1176 2 0.9855 0 end 

read control 

arp=14x14 

nax=18 

axp= 

0.67053 0.93322 1.02433 1.05329 1.06026 1.06185 

1.06215 1.06249 1.06312 1.06408 1.06541 1.06702 

1.06836 1.06760 1.05918 1.02515 0.92262 0.66935 end 

nhz= 2 

hzp= 0.9 1.1 end 

nuc= 

u-234 u-235 u-236 u-238 pu-238 pu-239 pu-240 

pu-241 pu-242 am-241 np-237 

mo-95 tc-99 ru-101 rh-103 ag-109 cs-133 nd-143 

nd-145 sm-147 sm-149 sm-150 sm-151 eu-151 sm-152 

eu-153 gd-155 end 

end control 

read hist 

power=28.00 burn=520.833 nlib=2 down=80 end 

power=28.00 burn=520.833 nlib=2 down=80 end 

power=28.00 burn=520.833 nlib=2 down=-1865 end 

end hist 

read keno 

ce 14x14 assembly model, 2x8 array of assemblies 

'* materials 

* 101 = uo2, lower axial region, low burnup region 

* 118 = uo2, upper axial region, low burnup region 

* 201 = uo2, lower axial region, high burnup region 

"* 218 = uo2, upper axial region, high burnup region 

2 = Zircaloy 

* 3 = Water 

read param 

tme=10000 gen=510 nsk=10 npg=1000 

lib=40 

end parm 

read geom ' Fuel Pin, Low Burnup Region 

unit 1

47

Section 9
Sample Problems

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cuboid

101 

102 

103 

104 

105 

106 

107 

108 

109 

110 

111 

112 

113 

114 

115 

116 

117 

118 

2 3

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.5588 

4 p0.7366

-162.53 

-142.22 

-121.90 

-101.58 

-81.27 

-60.95 

-40.63 

-20.32 

0.00 

21.32 

40.63 

60.95 

81.27 

101.58 

121.90 

142.22 

162.53 

182.85 

182.85 

182.85

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

r182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85

Fuel Pin, High Burnup Region

unit 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cylinder 

cuboid 

2 x 2 

uni t

array 1 3*0

2 

201 

202 

203 

204 

205 

206 

207 

208 

209 

210 

211 

212 

213 

214 

215 

216 

217 

218 

2 3

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.4782 

0.558E 

4 p0.7366

-162.53 

-142.22 

-121.90 

-101.58 

-81.27 

-60.95 

-40.63 

-20.32 

0.00 

21.32 

40.63 

60.95 

81.27 

101.58 

121.90 

142.22 

162.53 

182.85 

182.85 

182.85

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85 

-182.85

Array of Lower Burnup Fuel Pins 3

2 x 2 Array of Higher Burnup Fuel Pins

unit 

array 2 3*0

' Large 

unit 

cylinder 

cylinder 

cuboid

4

Water Hole 5

3 

2 3

1 

1 1

1.3140 

1.4160 

4 pl .4732

182.85 

182.85 

182.85

-182.85 

-182.85 

-182.85

Assembly 1 Unit

48

Section 9
Section 9 

unit 

array 

cuboid

Sample Problems

6 

3 -10.3124 -10.3124 -182.85 

3 1 

4 p11.3

9 0 182.85 -182.85

Assembly 2 Unit

7 

4 -10.3124 -10.3124 -182.85 

3 1 4p11.390 182.85 -182.85

' Assembly 3 Unit 

unit 8 

array 5 -10.3124 -10.3124 -182.85 

cuboid 3 1 

4 pl1.

39 0 182.85 -182.85 

Assembly 4 Unit 

unit 

9

6 -10.3124 -10.3124 -182.85 

3 1 4pll.390 182.85 -182.85

' Assembly 

global 

unit

Array (2 x 2)

10

array 7 3*0 

reflector 3 1 6r: 

end geom 

read array 

ara=1 nux=2 nuy=2 

1 1

1 1 

ara=2 

22 

22 

ara=3 

33 

35 

33 

33 

33 

35 

44

end fill 

nux=2 nuy=2 

end fill 

nux=7 nuy=7 

33333 

33354 

33444 

35444

3444 

4445 

4444

4 

4 4

ara=4 nux=7 nu 

3333333 

4533353 

4443333 

4445333 

4444333 

4544453 

4444444 

ara=5 nux=7 

n

44 

45 

44 

44 

44 

45 

43 

ara=6

end 

iy=7

30.0

1

nuz=1 fill 

nuz=1 fill 

nuz=1 fill 

fill 

nuz=1 fill

end fill 

uy=7 nuz=l fill

44443 

44453 

44433 

45333 

33333 

33353 

3 3 3 3 3 end fill 

nux=7 nuy=7 nuz=1 fill

3444444

unit 

array 

cuboid

array 

cuboid

49
Sample Problems Section 9 

3544454 

3344444 

3335444 

3333344 

3533354 3333334 end fill 

ara=7 nux=2 nuy=2 nuz=l fill 

67 

8 9 end fill 

end array 

read bounds all=void end bounds 

end data 

end keno 

end 

Following the STARBUCS calculation, the KENO V.a geometry model could be readily altered to simulate other 

assembly configurations (e.g., shuffle the fuel assembly locations). The KENO V.a case could subsequently be 

executed as a standalone case since all of the material cross sections have already been created during the initial 

STARBUCS run and are available on logical unit 40. This facilitates the rapid evaluation of different fuel 

configuration without the need to regenerate the material cross sections using STARBUCS.

50
Figure 6 Plot of the 2 x 2 array of CE 14 x 14 assemblies with burnup gradient 

51
10 FUTURE DEVELOPMENTS 

The STARBUCS prototype described in this report represents the initial implementation of an automated bumup 

credit computational methodology in SCALE. The bumup credit sequence uses existing, well-established 

modules of the SCALE code system. It is anticipated that the sequence will be further developed to provide 

additional enhancements, and remove some of the present limitations and restrictions of the program. Some of 

the potential future STARBUCS development areas are listed below. 

1. Provide greater integration of the STARBUCS program within SCALE. This involves incorporating routines 

in the CSAS control module that are used to generate input for the resonance cross-section processing codes 

directly into STARBUCS. This would allow STARBUCS to execute each of the function modules (e.g., 

BONAMI and NITAWL) via the SCALE driver directly, instead of calling the CSAS control module as is 

performed in the current version of the code. This capability would allow STARBUCS to function in a way 

that is very similar to CSAS and would allow a search capability to be incorporated within STARBUCS. 

Such a capability could be used to automatically generate a loading curve by searching on enrichment and 

burnup combinations that meet a specified subcritical safety criterion. 

2. Upgrade STARBUCS to run within the SCALE-5 system. This would allow STARBUCS to use the standard 

SCALE-5 versions of the ARP and ORIGEN-S codes (i.e., special versions would no longer be required). 

3. Modify the format of the criticality cross-section library generated by the sequence to a Monte Carlo format 

instead of the AMPX working format library currently used. This would remove the need to have 

STARBUCS generate a MIXTURE data block for the KENO V.a input. 

4. Upgrade STARBUCS to directly access the SCALE Standard Composition Library to map nuclide identifiers 

to their alphanumeric names. This would expand the number of standard compositions that could be used in 

the burnup credit sequence. 

5. Provide enhanced error checking and diagnosis of input data, and monitoring of program execution for 

successful completion. 

6. Extend STARBUCS to use either KENO V.a or KENO-VI as the criticality module. 

7. Develop additional standard ARP cross-section libraries for a wide range of assembly classes and design 

characteristics that can be used with the STARBUCS bumup credit sequence.

53
11 REFERENCES 

1. "Spent Fuel Project Office Interim Staff Guidance - 8, Rev. 1 - Limited Burnup Credit," USNRC, July 30, 

1999. 

2. SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, 

NUREG/CR-0200, Rev. 6 (ORNL/NUREG/CSD-2/R6), Vols. I, H, and III, May 2000. Available from the 

Radiation Safety Information Computational Center at Oak Ridge National Laboratory as CCC-545. 

3. L. C. Leal et al., "Automatic Rapid Processing for the Generation of Problem-Dependent 

SAS2H/ORIGEN-S Cross-Section Libraries," Nuclear Technology, Vol. 127, No. 1, pp. 1-23, July 1999. 

4. S. M. Bowman and L. C. Leal, "ORIGEN-ARP: Automatic Rapid Process for Spent Fuel Depletion, Decay, 

And Source Term Analysis," Vol. I, Sect. D1 of SCALE: A Modular Code System for Performing 

Standardized Computer Analyses for Licensing Evaluation, NUREG/CR-0200, Rev. 6 

(ORNL/NUREG/CSD-2/R6), Vols. I, II, and mI, May 2000. Available from the Radiation Safety 

Information Computational Center at Oak Ridge National Laboratory as CCC-545. 

5. N. F. Landers and L. M. Petrie, "CSAS: Control Module for Enhanced Criticality Safety Analysis 

Sequences," Vol. I, Sect. C4 of SCALE: A Modular Code System for Performing Standardized Computer 

Analyses for Licensing Evaluation, NUREG/CR-0200, Rev. 6 (ORNL/NUREG/CSD-2/R6), Vols. I, II, and 

III, May 2000. Available from the Radiation Safety Information Computational Center at Oak Ridge 

National Laboratory as CCC-545. 

6. N. F. Landers and L. M. Petrie, "KENO V.a: An Improved Monte Carlo Criticality Program with 

Supergrouping," Vol. 1, Sect. F1 1 of SCALE: A Modular Code System for Performing Standardized 

Computer Analyses for Licensing Evaluation, NUREG/CR-0200, Rev. 6 (ORNL/NUREG/CSD-2/R6), Vols. 

I, II, and III, May 2000. Available from the Radiation Safety Information Computational Center at Oak 

Ridge National Laboratory as CCC-545. 

7. 0. W. Hermann and R. M. Westfall, "ORIGEN-S: SCALE System Module to Calculate Fuel Depletion, 

Actinide Transmutation, Fission Product Buildup and Decay, and Associated Radiation Source Terms," 

Vol. 11, Sect. F7 of SCALE: A Modular Code System for Performing Standardized Computer Analyses for 

Licensing Evaluation, NUREG/CR-0200, Rev. 6 (ORNL/NUREG/CSD-2/R6), Vols. I, IL and III, May 

2000. Available from the Radiation Safety Information Computational Center at Oak Ridge National 

Laboratory as CCC-545. 

8. 0. W. Hermann and R. M. Westfall, "SAS2H: A Coupled One-Dimensional Depletion and Shielding 

Analysis Module," Vol. I, Sect. S2 of SCALE: A Modular Code System for Performing Standardized 

Computer Analyses for Licensing Evaluation, NUREG/CR-0200, Rev. 6 (ORNL/NUREG/CSD-2/R6), Vols. 

I, II, and Ill, May 2000. Available from the Radiation Safety Information Computational Center at Oak 

Ridge National Laboratory as CCC-545. 

9. N. M. Greene, "BONAMI: Resonance Self-Shielding by the Bondarenko Method," Vol. II, Sect. F1 of 

SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing 

Evaluation, NUREG/CR-0200, Rev. 6 (ORNL/NUREG/CSD-2/R6), Vols. I, 11, and III, May 2000. 

Available from the Radiation Safety Information Computational Center at Oak Ridge National Laboratory as 

CCC-545. 

10. N. M. Greene, L. M. Petrie, R. M. Westfall, "NITAWL-II: SCALE System Module for Performing 

Resonance Shielding and Working Library Production," Vol. H, Sect. F2 of SCALE: A Modular Code 

System for Performing Standardized Computer Analyses for Licensing Evaluation, NUREG/CR-0200, 

Rev. 6 (ORNL/NUREG/CSD-2/R6), Vols. I, H, and III, May 2000. Available from the Radiation Safety 

Information Computational Center at Oak Ridge National Laboratory as CCC-545. 

11. N. M. Greene, L. M. Petrie, and S. K. Fraley, "ICE: Module to Mix Multigroup Cross Sections," Vol. II, 

Sect. F8 of SCALE: A Modular Code System for Performing Standardized Computer Analyses for Licensing

55
References

Evaluation, NUREG/CR-0200, Rev. 6 (ORNL/NUREG/CSD-2/R6), Vols. I, II, and III, May 2000. 

Available from the Radiation Safety Information Computational Center at Oak Ridge National Laboratory as 

CCC-545. 

12. N. M. Greene, "User's Guide for AMPX Utility Modules," Vol. III, Sect. M15 of SCALE: A Modular Code 

System for Performing Standardized Computer Analyses for Licensing Evaluation, NUREG/CR-0200, 

Rev. 6 (ORNL/NUREG/CSD-2/R6), Vols. I, II, and mI1, May 2000. Available from the Radiation Safety 

Information Computational Center at Oak Ridge National Laboratory as CCC-545. 

13. Office of Civilian Radioactive Waste Management, Topical Report on Actinide-Only Burnup Credit for PWR 

Spent Nuclear Fuel Packages, DOE/RW-0472, Rev. 2, U.S. Department of Energy, Office of Civilian 

Radioactive Waste Management, September 1998. 

14. T. A. Parish and C. H. Chen, "Bounding Axial Profile Analysis for the Topical Report Database," Nuclear 

Engineering Dept., Texas A&M University, March 1997. 

15. L. M. Petrie, P. B. Fox, K. Lucius, "Standard Composition Library," Vol. III, Sect. M8 of SCALE: A 

Modular Code System for Performing Standardized Computer Analyses for Licensing Evaluation, 

NUREG/CR-0200, Rev. 6 (ORNL/NUREG/CSD-2/R6), Vols. I, II, and Il, May 2000. Available from the 

Radiation Safety Information Computational Center at Oak Ridge National Laboratory as CCC-545. 

16. N. F. Landers, L. M. Petrie, and J. A. Bucholz, "The Material Information Processor for SCALE," Vol. III, 

Sect. M7 of SCALE: A Modular Code System for Performing Standardized Computer Analyses for 

Licensing Evaluation, NUREG/CR-0200, Rev. 6 (ORNL/NUREG/CSD-2/R6), Vols. I, II, and III, May 

2000. Available from the Radiation Safety Information Computational Center at Oak Ridge National 

Laboratory as CCC-545. 

17. R. J. Cacciapouti and S. Van Volkinburg, Axial Burnup Profile Database for Pressurized Water Reactors, 

YAEC-1937, Yankee Atomic Electric Company, May 1997. 

18. 0. W. Hermann, S. M. Bowman, M. C. Brady, and C. V. Parks, Validation of the Scale System for PWR 

Spent Fuel Isotopic Composition Analyses, ORNL/TM- 12667, Lockheed Martin Energy Systems, Inc., 

Oak Ridge National Laboratory, March 1995. 

19. M. D. DeHart and 0. W. Hermann, An Extension of the Validation of SCALE (SAS2H) Isotopic Predictions 

for PWR Spent Fuel, ORNL/TM-13317, Lockheed Martin Energy Research Corp., Oak Ridge National 

Laboratory, September 1996.

56

Section 11
NRC FORM 335 

(2-89) 

NRCM 1102 

3201,3202

2. TITLE AND SUBTITLE 

STARBUCS: A Prototypic SCALE Control Module for Automated Criticality 

Safety Analyses Using Bumup Credit

1. REPORT NUMBER 

(Assigned by NRC, Add Vol., Supp., 

Rev., and Addendum Numbers, 

If any.) 

NUREG/CR- 6748 

ORNL/TM-2001/33

3. DATE REPORT PUBLISHED 

MONTH YEAR 

October 2001 

4. FIN OR GRANT NUMBER 

W6479

5. AUTHOR(S) 6. TYPE OF REPORT 

I. C. Gauld and S. M. Bowman Technical 

7. PERIOD COVERED (inclusive Dates) 

8. PERFORMING ORGANIZATION - NAME AND ADDRESS (If NRC, provide Division, Office or Region, U.S. Nuclear Regulatory Commission, and mailing 

address; iN contractor, provide name and mailing address.) 

Oak Ridge National Laboratory 

Managed by UT-Battelle, LLC 

Oak Ridge, TN 37831-6370

9. SPONSORING ORGANIZATION -NAME AND ADDRESS (If NRC, type "Same as above"; f contractor, provide NRC Division, Office or Region, U.S. Regulatory 

Commission, and mailing address.) 

Division of Systems Analysis and Regulatory Effectiveness 

Office of Nuclear Regulatory Research 

U.S. Nuclear Regulatory Commission 

Washington, DC 20555-0001 

10. SUPPLEMENTARY NOTES 

R. Y. Lee, NRC Project Manager

11. ABSTRACT (200 words or less) 

STARBUCS is a new prototypic analysis sequence for performing automated criticality safety analyses of spent fuel 

systems employing burnup credit. A depletion analysis calculation for each of the burnup-dependent regions of a 

spent fuel assembly, or other system containing spent fuel, is performed using the ORIGEN-ARP sequence of 

SCALE. The spent fuel compositions are then used to generate resonance self-shielded cross sections for each 

region of the problem, which are applied in a three-dimensional criticality safety calculation using the KENO V.a code. 

This prototypic burnup credit analysis sequence allows the user to simulate the axial and horizontal burnup gradients 

in a spent fuel assembly, select the specific actinides and/or fission products that are to be included in the criticality 

analysis, and apply isotopic correction factors to the predicted spent fuel nuclide inventory to account for calculational 

bias and uncertainties. Although STARBUCS was developed to address the burnup credit analysis needs for spent 

fuel transport and storage applications, it provides sufficient flexibility to allow criticality safety assessments involving 

many different potential configurations of spent nuclear fuel to be simulated.

12. KEY WORDSIDESCRIPTORS (List words or phrases that will assist researchers In locating the report.) 13. AVAILABILITY STATEMENT 

unlimited 

criticality safety, bumup credit, depletion, STARBUCS, SCALE 

14. SECURITY CLASSIFICATION 

(This Page) unclassified 

(Thls Report) unclassified 

15. NUMBER OF PAGES 

16. PRICE

NRC FORM 335 12-89)

U.S. NUCLEAR REGULATORY COMMISSION 

BIBLIOGRAPHIC DATA SHEET 

(See instructions on the reverse)
Federal Recycling Program
SAFETY ANALYSES USING BURNUP CREDIT

UNITED STATES 

NUCLEAR REGULATORY COMMISSION 

WASHINGTON, DC 20555-0001 

OFFICIAL BUSINESS 

PENALTY FOR PRIVATE USE, $300