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.. _STARBUCS:

STARBUCS: A Scale Control Module for Automated Criticality Safety Analyses Using Burnup Credit
==============================================================================================

*G. Radulescu and I. C. Gauld*

STARBUCS is an analysis sequence in SCALE for automating criticality
safety and burnup loading curve analyses of spent fuel systems employing
burnup credit. STARBUCS requires only the fresh fuel composition, an
irradiation history, and the KENO model for a spent fuel configuration
to be provided in an input file. It automatically performs all necessary
calculations to determine spent fuel compositions, self-shielded cross
sections, and the *k*\ :sub:`eff` of the spent fuel configuration. In addition,
for burnup loading curve analyses, STARBUCS performs iterative
calculations to search for initial fuel enrichments that result in an
upper subcritical limit. STARBUCS allows the user to simulate 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. A depletion analysis calculation for each of the
burnup-dependent regions of a spent fuel assembly, or any other system
containing spent nuclear fuel, is performed using the ORIGEN-ARP
sequence of SCALE. For criticality safety calculations employing
multigroup cross section data, the spent fuel compositions are used to
generate resonance self-shielded cross sections for each region of the
problem. The region dependent nuclide concentrations and cross sections
are applied in a three-dimensional criticality safety calculation using
the KENO code. Both KENO V.a and KENO-VI criticality codes are supported
for single criticality safety calculations using burnup credit, but only
KENO V.a can be used in criticality calculations for burnup loading
curve analyses. Although STARBUCS was developed specifically 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
UO\ :sub:`2` spent nuclear fuel.

Introduction
------------

The U.S. Nuclear Regulatory Commission (NRC) issued Revision 3 of the
Interim Staff Guidance 8 (ISG-8) (:cite:`us_nuclear_regulatory_commission_burnup_2012`) on burnup credit in
September, 2012. ISG-8 provides guidance on the application of
burnup-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 assuming
unirradiated fuel compositions), credit for burnup requires the
prediction of both fissile material and absorber nuclide concentrations
in spent nuclear fuel (SNF) and consideration of many burnup-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 use of spent fuel nuclide concentrations in the
criticality evaluation also necessitates consideration of many
additional sources of uncertainty associated with fuel depletion. ISG-8
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 nuclide 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 control
sequence called STARBUCS (**St**\ andardized **A**\ nalysis of
**R**\ eactivity for **Bu**\ rnup **C**\ redit using **S**\ CALE) was
developed in SCALE 5. STARBUCS automates the generation of
spatially-varying nuclide 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 modules in the sequence, executes the
modules through the SCALE driver, and performs all flow control, module
interface, and data management functions. The STARBUCS sequence uses
well-established code modules currently available in SCALE. STARBUCS
also performs iterations over a range of initial fuel enrichments to
determine the initial enrichments below which UO\ :sub:`2` commercial
spent fuel may be loaded in a transport/storage cask for specified
burnup values. With this capability, STARBUCS assists in generating
burnup loading curves for criticality safety analyses of spent fuel in
transport and storage casks.

The STARBUCS sequence automates the depletion calculations using the
ORIGEN-ARP methodology to perform a series of cross section preparation
and depletion calculations to generate a comprehensive set of spent fuel
isotopic inventories for each spatially-varying burnup region of an
assembly. The spent fuel nuclide concentrations are subsequently input
to either CSAS5 or CSAS6 to and perform a criticality calculation of the
system using the KENO V.a or KENO-VI code, respectively, to determine
the neutron multiplication factor (*k*\ :sub:`eff`) for the system. Only
minimal input is required by the user to perform a typical burnup-credit
analysis. The user can specify the assembly-average irradiation history,
the axial density variation of the reactor moderator, the axial- and
horizontal-burnup profile, and the nuclides that are to be applied in
the criticality safety analysis. Nuclide correction factors may also be
applied to the predicted concentrations to account for known bias and/or
uncertainty in the predicted SNF compositions.

Methodology
-----------

The STARBUCS control module is a burnup-credit sequence designed to
perform 3-D Monte Carlo criticality safety calculations that include the
effects of spatially-varying burnup in SNF configurations. STARBUCS
offers two options: either perform a single criticality safety
calculation with burnup credit or perform iterative calculations for
burnup loading curve analyses of commercial UO\ :sub:`2` spent fuels.
The sequence contains a set of instructions designed to automatically
process input data, execute code modules currently available in SCALE
for depletion, resonance cross section, and criticality calculations. In
addition, for burnup loading curve analyses, STARBUCS checks whether
*k*\ :sub:`eff` converges to a user-provided upper subcritical limit, adjusts
the initial fuel enrichment using the least squares method, and repeats
the sequence until either convergence is achieved or determine that no
solution can be found. The overall program structures and flow for a
single criticality calculation and for burnup loading curve calculations
are illustrated in :numref:`fig2-3-1` and :numref:`fig2-3-2`, respectively.

The sequence uses well-established code modules currently available in
the SCALE code system. These modules include ARP and ORIGEN to perform
the depletion analysis phase of the calculations. ORIGEN-ARP is a
sequence within the SCALE system that serves as a faster alternative to
the TRITON depletion sequence of SCALE to perform point-irradiation
calculations with the ORIGEN code using problem-dependent cross
sections. ARP uses an algorithm that enables the generation of cross
section libraries for the ORIGEN code by interpolation over pregenerated
cross section libraries. The ORIGEN code performs isotopic generation
and depletion calculations to obtain the spent fuel nuclide
compositions. For criticality safety calculations using multigroup cross
section data, problem dependent cross sections are processed with the
resonance self-shielding capabilities of XSProc using the
region-dependent compositions from the depletion analyses. Finally, the
region dependent nuclide concentrations and cross sections are applied
in a 3-D criticality calculation for the system using either KENO V.a or
KENO-VI to calculate the *k*\ :sub:`eff` value.

The ORIGEN-ARP depletion analysis methodology represents a significant
increase in computational speed as compared to equivalent calculations
performed using the SCALE depletion analysis sequences that use
two-dimensional transport methods, with virtually no sacrifice in
accuracy. ARP uses an algorithm that enables the generation of cross
sections for the ORIGEN code by interpolating on cross sections
available in pre-generated data libraries. For uranium-based fuels the
interpolation parameters available are initial fuel enrichment, burnup
and, optionally, moderator density. STARBUCS creates input files for ARP
and ORIGEN for each burnup-dependent region of an assembly and
calculates the spent fuel nuclide concentrations for the region using a
user-specified assembly irradiation history, cooling time, and burnup
profiles. The ORIGEN libraries must be available in advance of a
STARBUCS burnup-credit calculation. These libraries may be created using
TRITON. The libraries include the effects of assembly design and
operating conditions on the neutron cross sections used in the burnup
analysis. Several ORIGEN libraries are distributed in the SCALE code
system and can be applied in a STARBUCS analysis. Alternatively, a user
may create a specific ORIGEN library for other assembly types or
operating conditions not available in the default libraries. The
generation of ORIGEN reactor libraries is discussed in the ORIGEN
Reactor Libraries chapter.

The depletion phase of the analysis is performed using ARP and ORIGEN to
calculate the compositions of each discrete 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 nuclide inventories generated by ORIGEN, saves them, prepares
the ARP and ORIGEN input files for the next burnup 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 MTHM (10:sup:`6` g heavy metal).

After all depletion calculations are completed, STARBUCS reads the spent
fuel nuclide inventories for all regions and prepares input for the
criticality calculation. The concentrations of all nuclides in the
ORIGEN depletion analysis are converted from gram-atom units (per MTU)
to units of atoms/b-cm applied in the criticality calculation. The
criticality calculation is performed using the capabilities in the CSAS5
or CSAS6 control module of SCALE. Specifically, STARBUCS prepares input
for the CSAS5 module when criticality calculations are to be performed
using KENO V.a, and for the CSAS6 sequence when using KENO-VI. Note that
only the criticality safety sequence CSAS5 of SCALE can be used for
burnup loading curve calculations.

For burnup loading curve iterative calculations, STARBUCS employs the
search algorithm described in CSAS5 section on *Optimum
(Minimum/Maximum) Search* to determine initial fuel enrichments that
satisfy a convergence criterion for the k\ :sub:`eff` of the spent fuel
configuration. If convergence is not achieved in a search pass, the
initial fuel enrichment is automatically adjusted. This sequence repeats
until either k\ *eff* converges to an upper subcritical limit or until
the algorithm determines that a solution is not possible. The procedure
is repeated for each requested burnup value. The maximum allowable
iterations, upper subcritical limit, tolerance for convergence, and a
range of initial fuel enrichments can be set by the user. The lower and
upper enrichment bounds as well as the burnup values for spent fuel
regions must be contained within the range of enrichment and burnup
values used to generate the applicable ORIGEN library. The control
module prepares a STARBUCS input file for each search pass requesting a
single criticality calculation using the calculated spent fuel
compositions. In this input file, the burnup history data block and/or
the fuel mixture compositions are updated based on the outcome of the
search sequence. The pre-burnup compositions for the two minor uranium
isotopes, :sup:`234`\ U and :sup:`236`\ U, are updated in the STARBUCS
input file for a new pass only if they were included in the initial
input file prepared by the user. Their updated weight percentages are
based on the assumption that the mass ratios
:sup:`234`\ U/\ :sup:`235`\ U and :sup:`236`\ U/\ :sup:`235`\ U do not
change with fuel enrichment.

.. _fig2-3-1:
.. figure:: figs/STARBUCS/fig1.png
  :align: center
  :width: 600

  Modules and flow of STARBUCS sequence for criticality calculations.

.. _fig2-3-2:
.. figure:: figs/STARBUCS/fig2.png
  :align: center
  :width: 600

  Modules and flow of STARBUCS sequence for burnup loading curve calculations.

.. _cap-and-lim:

Capabilities and Limitations
----------------------------

STARBUCS is designed to facilitate criticality safety analyses employing
burnup credit by automating and linking the depletion and criticality
calculations. The STARBUCS 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 current regulatory guidance on burnup credit. However, STARBUCS
is sufficiently general to allow virtually any configuration involving
irradiated nuclear material to be analyzed. Limitations and some of the
key capabilities of the STARBUCS sequence are described below.

1. STARBUCS limitations include the use of a single UO\ :sub:`2` fuel
   type and, for analyses employing multigroup cross section data, the
   use of geometry configurations consisting of spent fuel rod arrays.
   However, the type of spent fuel configurations that can be analyzed
   is entirely 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
   UO\ :sub:`2` irradiated nuclear fuel.

2. Only the criticality safety sequence CSAS5 of SCALE can be used for
   burnup loading curve calculations; therefore KENO V.a geometry
   description must be available in a STARBUCS input file for burnup
   loading curve calculations.

3. Burnup calculations can incorporate any desired operating history.
   The user may enter the specific power, cycle lengths, cycle down
   time, post-irradiation cooling time, etc. The axial-water-moderator
   density variation may also be specified in the depletion analysis,
   provided the ORIGEN cross section library contains such data.

4. The effects of assembly design, soluble boron concentrations,
   burnable poison exposure, reactor operating conditions, etc., are
   accounted for in the ORIGEN cross section libraries used in the
   ORIGEN depletion calculations. Libraries for several fuel assembly
   designs are distributed with SCALE. These libraries can also be
   readily created for any reactor and fuel assembly design that can be
   represented in the depletion analysis sequences of the SCALE 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.

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 burnup-dependent 18-axial-zone
profiles taken from :cite:`lancaster_actinide-only_1998`, or the user may input an arbitrary
user-defined burnup distribution with up to 100-axial zones and up to
7-horizontal zones. The depletion analysis calculations for each zone
are performed for all nuclides (the ORIGEN data libraries contain cross
section and decay data for more than 1000 unique actinides, fission
products, and structural activation products). The specific nuclides to
be considered in the *k*\ :sub:`eff` analysis may be input by the user. If no
nuclide set is explicitly selected, then all nuclides that have cross
section data in the ORIGEN library are 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 ORIGEN cross section library for UO\ :sub:`2` fuel must
be available for the depletion analysis using STARBUCS. The user may use
the libraries distributed with SCALE (e.g., ge7×7-0, ge8×8-4, ce14×14,
w15×15, w17×17_ofa) or the user may generate their own problem-specific
libraries using the TRITON depletion analysis sequence available in
SCALE. A complete list of ORIGEN libraries distributed with SCALE and
methods for generating ORIGEN libraries are both described in the ORIGEN
Reactor Libraries chapter. The range of initial fuel enrichment and
requested burnup values to be used in the STARBUCS calculations must be
contained within the range of the enrichments and burnups used to
generate the applicable ORIGEN library.

The user is required to provide a complete KENO V.a model of the spent
fuel configuration for burnup loading curve calculations and a complete
KENO V.a or KENO-VI model of the spent fuel configuration for single
criticality calculations using burnup credit. 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 burnup-dependent regions or zones in the problem. The nuclide
concentrations after irradiation and decay are automatically applied to
the KENO criticality analysis. The mixture numbers for each of the fuel
regions are identified by unique mixture numbers assigned automatically
by STARBUCS based on the axial and horizontal regions in the problem
(see :numref:`fig2-3-3`). The user is required to specify the geometry/extent
of the axial and horizontal zones in the KENO model and apply the
appropriate mixture numbers for the desired configuration based on the
mixture identifying scheme. STARBUCS performs no checking of the
criticality model to verify that all mixtures in the problem have been
used or that the order of the mixture numbers in the KENO model
corresponds to the corresponding order of the input burnup 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 zone mixtures are assigned to the
correct KENO V.a or KENO-VI geometry regions. For instance, the user
could (intentionally) reverse the order of the axial-material
identifiers in the KENO model to simulate inverted fuel, or zone
mixtures could be omitted to simulate a problem using only a subset of
the available fuel zones that were simulated in the depletion analysis.

.. _fig2-3-3:
.. figure:: figs/STARBUCS/fig3.png
  :align: center
  :width: 600

  Fuel and material mixture numbering convention used in STARBUCS.

.. _fig2-3-4:
.. figure:: figs/STARBUCS/fig4.png
  :align: center
  :width: 600

  Example of mixture numbering scheme used in STARBUCS.

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. The maximum number of horizontal zones is restricted to seven if
   there is no gap or second moderator mixture, six if a gap or second
   moderator mixture is defined, and five if both a gap and a second
   moderator are defined. The number of axial-fuel zones is limited such
   that the product of horizontal zones ∗ axial zones is less than or
   equal to 100. These limits constrain the maximum mixture number used
   for burned fuel in the KENO criticality calculation to less than 1000
   and assign unique mixture numbers to clad, moderator, and gap
   mixtures for lattice cell descriptions. The convention used to number
   the depleted fuel zones is to start at mixture 101 and increment by 1
   for each axial-burnup region. Thus, for a case with 10 axial-burnup
   regions, the fuel mixtures used in the criticality analysis would
   range from 101 to 110. For a similar case having two horizontal zones
   in addition to the axial zones, the mixture numbers would also
   include mixtures 201 to 210.

2. Mixture numbers for the clad, gap (if applicable), and moderator may
   also be used directly in the KENO model. Additional unique mixture
   numbers are required by the code for the lattice cell descriptions
   for each separate fuel zone (except for mixture 0 for void). These
   additional mixtures are assigned automatically by the code and are
   shown in :numref:`fig2-3-3` for a lattice cell consisting of fuel, gap,
   clad, and moderator. The additional mixture numbers may also be used
   directly in the KENO model. Mixture number allocation is illustrated
   in :numref:`fig2-3-4` for an example case where the number of different
   horizontal zones is four and the maximum number of axial zones is
   limited to 25.

3. All structural materials in the problem must have mixture numbers
   different from the numbers automatically generated by the code (see
   :numref:`fig2-3-4` for an example of available mixture numbers). For the
   example shown in :numref:`fig2-3-4`, mixtures 5–100, 126–200, 226–300,
   326–400, 501, 601, 701, 426–500, and 801–2147 are not allocated by
   STARBUCS and may be defined by the user in the composition data block
   and used in the geometry model. If the constraints in paragraph 1 are
   followed, mixture numbers less than 100 that were not used for fuel,
   gap, clad, moderator and mixture numbers from 1001 to 2147 are always
   available for structural materials. Note that STARBUCS does not
   provide a warning or stop program execution if a mixture number
   assigned to a structural material has also been generated internally
   by the computer code. The mixture numbers for structural materials
   are not changed and are thus applied in the KENO model in a
   one-to-one correspondence with the standard composition mixture as
   done for typical CSAS calculations. Therefore, the use of a mixture
   number for structural materials that is identical to one of the
   mixture numbers automatically generated by the code results in the
   combination of both materials in the composition for the mixture
   number.

4. Not all SCALE standard composition alphanumeric names (see the
   Standard Composition Library chapter) are currently recognized by
   STARBUCS. The use of special materials (e.g., C-GRAPHITE, NIINCONEL,
   H-POLY), particularly as fuel materials, that have nuclide
   identifiers that are not readily translated to ORIGEN ZA numbers
   should be avoided since these materials cannot be depleted.

5. A single STARBUCS calculation is limited to a single initial fuel
   type (composition, enrichment, assembly design, etc.). Configurations
   involving multiple fuel types may be solved by running a separate
   STARBUCS case for each type, saving the corresponding CSAS cases
   generated by STARBUCS that contain the irradiated fuel nuclide
   compositions, and manually merging the cases in such a way that all
   required fuel types are represented in the final case.


Input Description
-----------------


STARBUCS 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 and
cross section processing module of SCALE (XSProc) and conforms to the
standard input conventions (see
Chapter \ 7 (SECTIONREFERENCE)
In addition to the standard composition data, three more input data
blocks are required by STARBUCS. The data blocks are entered in the form

.. highlight:: scale

::

  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 or
search parameter data, and the KENO V.a or KENO-VI input specifications.
The valid block keywords for a single criticality safety calculation
using burnup credit and for burnup loading curve calculations are listed
in :numref:`tab2-3-1` and :numref:`tab2-3-2`, respectively. A minimum of four
characters is required for most keywords. The exception is the
criticality model input data block READ KENOVA or READ KENOVI in which
case the code must check additional character positions to determine the
CSAS control sequence to be executed. 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 activates the data block input. All input data pertinent to block
**XXXX** are then entered. Entering **END XXXX** followed by two or more
blanks terminates data block **XXXX**.

.. _tab2-3-1:
.. table:: Valid data block keywords for a single criticality safety calculation using burnup credit
  :align: center

  +---------------------+---------------------+
  | **Data block type** | **Block keyword**   |
  +---------------------+---------------------+
  | Control parameters  | CONTROL             |
  +---------------------+---------------------+
  | Burnup history      | HISTORY or BURNDATA |
  +---------------------+---------------------+
  | KENO V.a input      | KENOVA or KENO5     |
  +---------------------+---------------------+
  | KENO-VI input       | KENOVI or KENO6     |
  +---------------------+---------------------+

.. _tab2-3-2:
.. table:: Valid data block keywords for burnup loading curve calculations.
  :align: center

  +---------------------+-------------------+
  | **Data block type** | **Block keyword** |
  +---------------------+-------------------+
  | Control parameters  | CONTROL           |
  +---------------------+-------------------+
  | Search parameters   | SEARCH            |
  +---------------------+-------------------+
  | KENO V.a input      | KENOVA or KENO5   |
  +---------------------+-------------------+

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.

A single data entry may be entered anywhere on a line but cannot be
divided between two lines; however, array data entries may be divided
over many lines. The code identifies data 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 in 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 or a comma.

Overview of input structure
~~~~~~~~~~~~~~~~~~~~~~~~~~~

An overview of the input to the STARBUCS sequence is given in
:numref:`tab2-3-3`. This table provides an outline of the input data block
structure. The input data in positions 1 to 5 (see :numref:`tab2-3-3`) are read
and processed by the material and cross section processing module of
SCALE (XSProc). These are the first data read by the code and must be in
the order indicated. Data positions 6, 7 or 8, and 9 are read directly
by STARBUCS and may be entered in any order.

.. _tab2-3-3:
.. table:: Outline of input data for the STARBUCS sequence
  :align: center

  +-----------------+-----------------+-----------------+-----------------+
  | **Data**        | **Type of       | **Data entry**  | **Comments**    |
  |                 | data**          |                 |                 |
  | **position**    |                 |                 |                 |
  +-----------------+-----------------+-----------------+-----------------+
  |                 | Sequence name   | =STARBUCS       | Start in column |
  |                 |                 |                 | one             |
  +-----------------+-----------------+-----------------+-----------------+
  | 1               | TITLE           | Enter a title   | 80 characters   |
  +-----------------+-----------------+-----------------+-----------------+
  | 2               | Standard SCALE  | Library name    | The currently   |
  |                 | pointwise or    |                 | available       |
  |                 | multigroup      |                 | standard SCALE  |
  |                 | cross section   |                 | cross section   |
  |                 | library name or |                 | libraries are   |
  |                 |                 |                 | listed in the   |
  |                 | the name of a   |                 | SCALE Cross     |
  |                 | user-supplied   |                 | Section         |
  |                 | multigroup      |                 | Libraries       |
  |                 | cross section   |                 | chapter, table  |
  |                 | library         |                 | *Standard SCALE |
  |                 |                 |                 | Cross-Section   |
  |                 |                 |                 | Libraries*.     |
  |                 |                 |                 |                 |
  |                 |                 |                 | STARBUCS allows |
  |                 |                 |                 | a non-standard  |
  |                 |                 |                 | SCALE           |
  |                 |                 |                 | multigroup      |
  |                 |                 |                 | cross section   |
  |                 |                 |                 | library to be   |
  |                 |                 |                 | used in a       |
  |                 |                 |                 | criticality     |
  |                 |                 |                 | calculation.    |
  +-----------------+-----------------+-----------------+-----------------+
  | 3               | Standard        | Enter the       | Begins this     |
  |                 | Composition     | appropriate     | data block with |
  |                 | specification   | data            | READ COMP and   |
  |                 | data            |                 | terminate with  |
  |                 |                 |                 | END COMP. See   |
  |                 |                 |                 | Standard        |
  |                 |                 |                 | Composition     |
  |                 |                 |                 | section for     |
  |                 |                 |                 | details.        |
  +-----------------+-----------------+-----------------+-----------------+
  | 4               | Type of         | LATTICECELL     | Begins this     |
  |                 | calculation     |                 | data block with |
  |                 |                 |                 | READ CELL and   |
  |                 |                 |                 | terminates with |
  |                 |                 |                 | END CELL. Only  |
  |                 |                 |                 | regular unit    |
  |                 |                 |                 | cells may be    |
  |                 |                 |                 | used. See       |
  |                 |                 |                 | XSProc section  |
  |                 |                 |                 | for details.    |
  +-----------------+-----------------+-----------------+-----------------+
  | 5               | Unit cell       | Enter the       | Each dimension  |
  |                 | geometry        | appropriate     | may be entered  |
  |                 | specification\  | data            | as a diameter.  |
  |                 | :sup:`a`        |                 | See XSProc      |
  |                 |                 |                 | section for     |
  |                 |                 |                 | LATTICECELL.    |
  +-----------------+-----------------+-----------------+-----------------+
  | 6               | Control         | Enter the       | Begins this     |
  |                 | parameter data  | desired data    | data block with |
  |                 |                 |                 | READ CONT and   |
  |                 |                 |                 | terminate with  |
  |                 |                 |                 | END CONT.       |
  |                 |                 |                 | See Conntrol pa\|
  |                 |                 |                 | rameter data sec|
  +-----------------+-----------------+-----------------+-----------------+
  | 7\ :sup:`b`     | Burnup history  | Enter the       | Begins this     |
  |                 | specification   | desired data    | data block with |
  |                 |                 | for each cycle  | READ HISTORY    |
  |                 |                 |                 | (or BURNDATA)   |
  |                 |                 |                 | and terminate   |
  |                 |                 |                 | with            |
  |                 |                 |                 | END HISTORY (or |
  |                 |                 |                 | BURNDATA).      |
  |                 |                 |                 | See Burnup hist/|
  |                 |                 |                 | ory data sec.   |
  +-----------------+-----------------+-----------------+-----------------+
  | 8\ :sup:`b`     | Search          | Enter the       | Begins this     |
  |                 | parameter data  | desired data    | data block with |
  |                 |                 |                 | READ SEARCH and |
  |                 |                 |                 | terminate with  |
  |                 |                 |                 | END SEARCH.     |
  |                 |                 |                 | See Search para/|
  |                 |                 |                 | meter data sec. |
  +-----------------+-----------------+-----------------+-----------------+
  | 9               | KENO data       | Enter KENO      | Begins this     |
  |                 |                 | criticality     | data block with |
  |                 |                 | model           | READ KENOVA (or |
  |                 |                 |                 | KENO5) and      |
  |                 |                 |                 | terminate with  |
  |                 |                 |                 | END KENOVA (or  |
  |                 |                 |                 | KENO5).         |
  |                 |                 |                 |                 |
  |                 |                 |                 | For KENO-VI use |
  |                 |                 |                 | block keyword   |
  |                 |                 |                 | KENOVI (or      |
  |                 |                 |                 | KENO6) in place |
  |                 |                 |                 | of KENOVA       |
  |                 |                 |                 | (or KENO5). See |
  |                 |                 |                 | Keno Input Data.|
  +-----------------+-----------------+-----------------+-----------------+
  |                 | Terminate input | END             | Must begin in   |
  |                 |                 |                 | column 1.       |
  +-----------------+-----------------+-----------------+-----------------+
  | :sup:`a` \Input |                 |                 |                 |
  | data required o\|                 |                 |                 |
  | nly for critica\|                 |                 |                 |
  | lity calculatio\|                 |                 |                 |
  | ns employing    |                 |                 |                 |
  | multigroup      |                 |                 |                 |
  | cross section   |                 |                 |                 |
  | libraries. Only |                 |                 |                 |
  | one unit cell   |                 |                 |                 |
  | may be defined  |                 |                 |                 |
  | in the cell     |                 |                 |                 |
  | data block for  |                 |                 |                 |
  | STARBUCS.       |                 |                 |                 |
  |                 |                 |                 |                 |
  | :sup:`b` Either |                 |                 |                 |
  | burnup history  |                 |                 |                 |
  | specification   |                 |                 |                 |
  | or search       |                 |                 |                 |
  | parameter data  |                 |                 |                 |
  | may be defined  |                 |                 |                 |
  | in a STARBUCS   |                 |                 |                 |
  | input.          |                 |                 |                 |
  +-----------------+-----------------+-----------------+-----------------+

Sequence specification card
~~~~~~~~~~~~~~~~~~~~~~~~~~~

The STARBUCS analytical sequence is initiated with “=STARBUCS” beginning
in column 1 of the input. This instructs the SCALE driver module to
execute the STARBUCS sequence. The input data are then entered in
free-format. The input is terminated with the word “END” 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.

Optional sequence parameters
~~~~~~~~~~~~~~~~~~~~~~~~~~~~

To check the input data, run STARBUCS and specify PARM=CHECK or PARM=CHK
after the analytical sequence specification as shown below.

::

  =STARBUCS PARM=CHK

Other optional input for the PARM field to control multigroup resonance
self-shielding calculations are described in the XSProc section of this
manual.

XSProc
~~~~~~

The XSProc 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. A detailed description of this portion of the input can be
found in the section on XSProc (Chapter 7 (SECTIONREFERENCE)). Only one UO\ :sub:`2` fuel
type is permitted in STARBUCS. Therefore, a single fuel mixture defining
the fresh fuel composition and, for criticality safety calculations
employing multigroup cross sections, the geometry description of a
single fuel lattice cell are required in a STARBUCS input file. Only the
regular unit cells SQUAREPITCH, TRIANGPITCH, SPHSQUAREP, SPHTRIANGP, and
SYMMSLACELL may be specified for the LATTICECELL entry. Outside
diameters of the fuel, gap, and clad mixtures (i.e., not the radii) are
required.

Control parameter data
~~~~~~~~~~~~~~~~~~~~~~

The control parameter data block allows the user to specify control
parameters and array data related to many of the burnup-credit analysis
parameters to be used in the problem. All input is by keyword entry. 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 are summarized in
Table 2.3.4. The individual keyword entries are described below.

1.  ARP= NAME OF THE ORIGEN LIBRARY TO BE USED. A character string with
    the name of the ORIGEN 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. A description of
    an ARP input and the location of the ORIGEN cross section libraries
    are provided in *ARP Input Description* located in the ORIGEN ARP
    Module chapter. STARBUCS calculations are limited to UO\ :sub:`2`
    spent fuels.

2.  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 must be chosen such that due product of NAX \* NHZ is
    less than or equal to 100 (i.e., NAX:sub:`max` is 100, 50, 33, 25,
    20, 16, or 14 when the number of horizontal zones is 1, 2, 3, 4, 5,
    6, or 7, respectively). By default, the profile is automatically
    normalized to unity by the code unless NPR=no. Built-in
    burnup-dependent 18‑axial-zone profiles may be selected with an
    entry of –18. These built-in profiles and the burnup range over
    which they are applied, are listed in :numref:`tab2-3-5`. These profiles
    have been proposed elsewhere (Ref. 2) as bounding axial profiles and
    are included as options for convenience only. The default value of
    NAX is –18 (use built-in profiles).

3.  NHZ= NUMBER OF HORIZONTAL ZONES. This is the number of
    horizontal-burnup subdivisions in the assembly. An optional entry if
    no horizontal profile is requested. The maximum value is seven
    zones. The exact limit is determined by the number of mixtures
    defined in the lattice cell description. If a gap and second
    moderator type are used the number of horizontal zones is limited to
    five.

4.  NUC= BURNUP-CREDIT NUCLIDES used in the criticality calculation. A
    list of actinides and/or fission products that are to be included in
    the KENO criticality safety calculation. This is an array entry
    keyword and is delimited by the keyword END. The nuclides are
    entered using their standard composition alphanumeric names, as
    listed in the Standard Composition Library chapter of the SCALE
    manual. Isotopic correction factors may be entered, optionally,
    immediately following the nuclide name. The isotopic correction
    factors will be multiplied times the spent fuel nuclide
    concentrations to account for isotopic composition bias.
    The concentration of any nuclide that does not have a correction
    factor is not adjusted. To select all available actinide and fission
    product nuclides (with cross section data and atom densities greater
    than 1.0E−29) 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. Note that the set of
    nuclides tracked by ORIGEN in any decay or irradiation calculation,
    documented in the ORIGEN Reaction Resource Contents chapter, is much
    larger than the set of nuclides with available cross sections for
    neutron transport calculations, documented in the SCALE Cross
    Section Libraries chapter. Only nuclides with available cross
    sections for neutron transport calculations are included in the
    irradiated fuel compositions for criticality calculations.

5.  FLE= FUEL LIGHT ELEMENT NUCLIDES. A user-provided list of light
    element nuclides that are to be included in the irradiated fuel
    compositions for a CSAS5 or a CSAS6 calculation. This is an array
    entry keyword and is delimited by the keyword END. The nuclides are
    entered using their standard composition alphanumeric names, as
    listed in Standard Composition Library chapter of the SCALE manual.
    To select all available light element nuclides (with cross section
    data and atom densities greater than 1.0E−29) for the criticality
    calculation, the user may specify FLE= ALL, without an END
    terminator. This is the only situation where an array entry does not
    require an END. The use of the keyword FLE is not required if only
    o-16 is to be included in the composition of irradiated uranium
    oxide fuel pellets. For these material mixtures, o-16 will be
    automatically included in irradiated fuel compositions due to its
    significant concentration. Isotopic correction factors are not
    allowed for light element nuclides. Note that the set of nuclides
    tracked by ORIGEN in any decay or irradiation calculation,
    documented in the ORIGEN Reaction Resource Contents chapter, is much
    larger than the set of nuclides with available cross sections for
    neutron transport calculations, documented in the SCALE Cross
    Section Libraries chapter. Only nuclides with available cross
    sections for neutron transport calculations are included in the
    irradiated fuel compositions for criticality calculations.

6.  AXP= AXIAL-BURNUP PROFILE. The user-supplied axial-burnup profile of
    the assembly to be used in the analysis. This entry is required
    unless use of the built-in burnup-dependent axial profiles shown in
    :numref:`tab2-3-5` is requested (NAX= −18). If NAX is set to anything other
    than −18, 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; this
    may be disabled by setting 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 fuel region subdivisions of unequal volume
    are used, normalization should not be applied and the user must
    ensure a correct correspondence between the axial-profile input and
    the axial regions specified in the criticality calculation. AXP** is
    an array entry and must be delimited by an END that must not start
    in the first column.

7.  HZP= HORIZONTAL-BURNUP PROFILE. An optional array entry used to
    specify a burnup gradient across assemblies. The elements of the
    array are the ratios of the burnups of horizontal subdivisions in
    the assembly to average assembly burnup (entry for the POWER=
    keyword described in :ref:`burnup-history-data`). If NHZ is input, the HZP array
    must contain NHZ entries delimited by an END that must not start in
    the first column. Otherwise, the value of NHZ is determined
    automatically by the code. The profile will be normalized if NPR=yes
    (default). Sample problem 5 illustrates use of this option.

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). If
    fuel region subdivisions of unequal volume are used, NPR=NO should
    be specified.

10. MOD= AXIAL MODERATOR DENSITY. This is an array entry keyword and is
    delimited by the keyword END. The array dimension is equal to the
    number of axial zones (NAX entry) and the array values are provided
    in the same order as the AXP array elements. This input array is
    required only if the applicable ORIGEN library contains variable
    moderator density cross sections.

11. BUG= DEBUG PRINT OPTION. BUG=yes will print program debugging
    variables and arrays in STARBUCS. The default is BUG=no.

.. _tab2-3-4:
.. table:: Table of control parameter data.
  :align: center

  +-----------------+-----------------+-----------------+-----------------+
  | **Keyword**     | **Data**        | **Default**     | **Comments**    |
  |                 |                 |                 |                 |
  | **name**        | **type**        | **value**       |                 |
  +-----------------+-----------------+-----------------+-----------------+
  | READ CONTROL    |                 | Initiate        |                 |
  |                 |                 | reading the     |                 |
  |                 |                 | control         |                 |
  |                 |                 | parameter block |                 |
  |                 |                 | of data         |                 |
  +-----------------+-----------------+-----------------+-----------------+
  | ARP=            | Character       | None            | Name of the     |
  |                 |                 |                 | ORIGEN library  |
  |                 |                 |                 | to be used.     |
  |                 |                 |                 | Required.       |
  |                 |                 |                 | Library must be |
  |                 |                 |                 | defined in      |
  |                 |                 |                 | SCALE text file |
  |                 |                 |                 | ARPDATA.TXT.    |
  +-----------------+-----------------+-----------------+-----------------+
  | NAX=            | Integer         | −18             | 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 |
  |                 |                 |                 | -burnup         |
  |                 |                 |                 | profile.        |
  +-----------------+-----------------+-----------------+-----------------+
  | NHZ=            | Integer         | 1               | Number of       |
  |                 |                 |                 | horizontal-burn |
  |                 |                 |                 | up              |
  |                 |                 |                 | subdivisions.   |
  |                 |                 |                 | Maximum value   |
  |                 |                 |                 | of              |
  |                 |                 |                 | 5–7 zones (see  |
  |                 |                 |                 | Sect. 2.3.4.5). |
  |                 |                 |                 | No entry is     |
  |                 |                 |                 | required if     |
  |                 |                 |                 | horizontal      |
  |                 |                 |                 | profile is not  |
  |                 |                 |                 | used.           |
  +-----------------+-----------------+-----------------+-----------------+
  | NUC=            | Character and   | None            | List of         |
  |                 | real mixed      |                 | burnup-credit   |
  |                 | array\ :sup:`a` |                 | nuclides, and   |
  |                 |                 |                 | optionally the  |
  |                 |                 |                 | corresponding   |
  |                 |                 |                 | isotopic        |
  |                 |                 |                 | correction      |
  |                 |                 |                 | factors, to be  |
  |                 |                 |                 | included in the |
  |                 |                 |                 | criticality     |
  |                 |                 |                 | calculation.\   |
  |                 |                 |                 | :sup:`b`        |
  |                 |                 |                 | Array entry     |
  |                 |                 |                 | generally       |
  |                 |                 |                 | delimited by    |
  |                 |                 |                 | END, unless ALL |
  |                 |                 |                 | is selected.    |
  |                 |                 |                 | Nuclides are    |
  |                 |                 |                 | input using     |
  |                 |                 |                 | their standard  |
  |                 |                 |                 | composition     |
  |                 |                 |                 | alphanumeric    |
  |                 |                 |                 | identifiers.    |
  +-----------------+-----------------+-----------------+-----------------+
  | FLE=            | Character       | o-16            | List of light   |
  |                 | array\ :sup:`a` |                 | element         |
  |                 |                 |                 | nuclides to be  |
  |                 |                 |                 | included in the |
  |                 |                 |                 | criticality     |
  |                 |                 |                 | calculation.\   |
  |                 |                 |                 | :sup:`b`        |
  |                 |                 |                 | Array entry     |
  |                 |                 |                 | generally       |
  |                 |                 |                 | delimited by    |
  |                 |                 |                 | END, unless ALL |
  |                 |                 |                 | is selected.    |
  |                 |                 |                 | Nuclides are    |
  |                 |                 |                 | input using     |
  |                 |                 |                 | their standard  |
  |                 |                 |                 | composition     |
  |                 |                 |                 | alphanumeric    |
  |                 |                 |                 | identifiers.    |
  +-----------------+-----------------+-----------------+-----------------+
  | AXP=            | Real array\     | See NAX         | Axial-burnup-pr |
  |                 | :sup:`a`        |                 | ofile           |
  |                 |                 |                 | array. Required |
  |                 |                 |                 | if NAX > 0. NAX |
  |                 |                 |                 | entries that    |
  |                 |                 |                 | define the      |
  |                 |                 |                 | axial-burnup    |
  |                 |                 |                 | shape. The      |
  |                 |                 |                 | profile is      |
  |                 |                 |                 | automatically   |
  |                 |                 |                 | normalized if   |
  |                 |                 |                 | NPR=YES         |
  |                 |                 |                 | (default).      |
  |                 |                 |                 | Delimited by    |
  |                 |                 |                 | END.            |
  +-----------------+-----------------+-----------------+-----------------+
  | HZP=            | Real array\     | None            | Horizontal-burn |
  |                 | :sup:`a`        |                 | up-profile      |
  |                 |                 |                 | array. Required |
  |                 |                 |                 | if NHZ > 1.     |
  |                 |                 |                 | Array containin |
  |                 |                 |                 | g               |
  |                 |                 |                 | NHZ entries     |
  |                 |                 |                 | that define the |
  |                 |                 |                 | horizontal,     |
  |                 |                 |                 | or radial,      |
  |                 |                 |                 | burnup profile  |
  |                 |                 |                 | for the         |
  |                 |                 |                 | analysis. Array |
  |                 |                 |                 | is              |
  |                 |                 |                 | automatically   |
  |                 |                 |                 | normalized by   |
  |                 |                 |                 | the code.       |
  |                 |                 |                 | Delimited by    |
  |                 |                 |                 | END.            |
  +-----------------+-----------------+-----------------+-----------------+
  | MOD=            | Real array\     | None            | Axial-moderator |
  |                 | :sup:`a`        |                 | density,        |
  |                 |                 |                 | applied in the  |
  |                 |                 |                 | fuel depletion  |
  |                 |                 |                 | analysis.       |
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