.. _3-1:
TRITON: A Multipurpose Transport, Depletion, And Sensitivity and Uncertainty Analysis Module
============================================================================================
*M. A. Jessee, D. Wiarda, K. T. Clarno, U. Mertyurek, K. Bekar*
ABSTRACT
The TRITON computer code is a multipurpose SCALE control module for
transport, depletion, and sensitivity and uncertainty analysis. TRITON
can be used to provide automated, problem-dependent cross section
processing followed by multigroup transport calculations for one-, two-,
and three-dimensional (1D, 2D, and 3D) configurations. Additionally,
this functionality can be used in tandem with the ORIGEN depletion
module to predict isotopic concentrations, source terms, and decay heat,
as well as to generate few-group homogenized cross sections for nodal
core calculations.
TRITON has been designed using the modular approach that is a hallmark
of SCALE functionality. TRITON provides the capability to perform
deterministic transport analysis for 1D geometries using XSDRNPM and for
2D geometries using NEWT. TRITON also includes 3D Monte Carlo depletion
capabilities using KENO V.a and KENO-VI. For Monte Carlo depletion
calculations, TRITON supports both multigroup and continuous-energy
options.
The TSUNAMI-2D sequence in TRITON provides calculation of 2D forward and
adjoint transport solutions in NEWT, calculation of sensitivity
coefficients, and calculation of the uncertainty in *k*\ :sub:`eff` and other
responses due to cross section covariance data. The SAMS module is used
to determine the sensitivity of the calculated value of responses to the
nuclear data used in the calculation as a function of nuclide, reaction
type, and energy. The uncertainty in the calculated value of the
response, resulting from uncertainties in the basic nuclear data used in
the calculation, is estimated using energy-dependent cross section
covariance matrices. The implicit effects of the cross section
processing calculations are also treated.
ACKNOWLEDGMENTS
The authors express gratitude to B. T. Rearden for supervision of the
SCALE project and review of the manuscript. The authors acknowledge R.
Y. Lee and M. Aissa of the U.S. Nuclear Regulatory Commission (NRC) for
their support of this project and Mark DeHart, as the original developer
of TRITON.
.. _3-1-1:
Introduction
------------
TRITON (Transport Rigor Implemented with Time-dependent Operation for
Neutronic depletion) is a multipurpose SCALE control module for
transport, depletion, and sensitivity and uncertainty analysis for
reactor physics applications. TRITON can be used to provide automated,
problem-dependent cross section processing followed by multigroup (MG)
neutron transport calculations for one-, two-, and three-dimensional
(1D, 2D, and 3D) configurations. Additionally, this functionality can be
used in tandem with the ORIGEN depletion module to predict isotopic
concentrations, source terms, and decay heat.
Most notable improvements to TRITON in this latest release are the
reduced run-time due to the integration of the XSProc module for cross
section processing coupled with improvements in the NEWT transport
module and the capability to perform continuous energy (CE) transport
calculations using Monte Carlo transport codes KENO-V.a and KENO-VI (MG
Monte Carlo depletion existed in previous versions).
TRITON has been designed using the modular approach that is a hallmark
of SCALE functionality. TRITON provides the capability to perform
deterministic transport analysis for 1D geometries using XSDRNPM and for
2D geometries using NEWT. TRITON also includes 3D Monte Carlo depletion
capabilities using KENO V.a and KENO-VI.
The sensitivity and uncertainty (S/U) analysis sequence in TRITON,
TSUNAMI-2D, provides calculation of the 2D forward and adjoint transport
solutions in NEWT; calculation of sensitivity coefficients; and
calculation of the uncertainty in *k*\ :sub:`eff` and other responses due to
cross section covariance data. The SAMS module is used to determine the
sensitivity of the calculated value of the response to the nuclear data
used in the calculation as a function of nuclide, reaction type, and
energy. The uncertainty in the calculated value of the response,
resulting from uncertainties in the basic nuclear data used in the
calculation, is estimated using energy-dependent cross section
covariance matrices. The implicit effects of the cross section
processing calculations are predicted using SENLIB and BONAMIST. The
energy-dependent sensitivity data computed with TSUNAMI-2D are stored in
a sensitivity data file (.sdf file) that is suitable for assessing
system similarity for code validation purposes using TSUNAMI-IP or for
data assimilation using TSURFER.
As a SCALE control module, TRITON automates execution of SCALE
functional modules and manages data transfer and input/output processes
for multiple analysis sequences. Each of TRITON’s nine calculational
sequences is provided in :numref:`tab3-1-1`, which lists the sequence name
keyword, the sequence description, and the function modules invoked
within each sequence. The method for cross section processing is
selected using a separate “\ *parm=*\ ” keyword, which is described in
more detail in the next section.'
.. _tab3-1-1:
.. table::
:align: center
+-----------------+-----------------+-----------------+-----------------+
| **Sequence | **Primary SCALE | **parm= | **Sequence |
| keyword** | modules** | options** | function** |
+=================+=================+=================+=================+
| **Cross section | | | |
| processing | | | |
| sequences** | | | |
+-----------------+-----------------+-----------------+-----------------+
| ``=T-XSEC`` | XSProc | bonami | Preparation of |
| | | | multigroup (MG) |
| | | centrm\ :sup:`a`| cross section |
| | | | library. |
| | | xslevel=1/2/3/4 | |
+-----------------+-----------------+-----------------+-----------------+
| **Transport | | | |
| sequences** | | | |
+-----------------+-----------------+-----------------+-----------------+
| ``=T-XSDRN`` | XSProc, XSDRNPM | bonami | 1D MG |
| | | | deterministic |
| | | centrm\ :sup:`a`| transport |
| | | | calculation. |
| | | xslevel=1/2/3/4 | |
| | | | |
| | | weight\ :sup:`b`| |
+-----------------+-----------------+-----------------+-----------------+
| ``=T-NEWT`` | XSProc, NEWT | | 2D MG |
| | | | deterministic |
| | | | transport |
| | | | calculation. |
+-----------------+-----------------+-----------------+-----------------+
| **Depletion | | | |
| sequences** | | | |
+-----------------+-----------------+-----------------+-----------------+
| ``=T-DEPL-1D`` | XSProc, | bonami | 1D MG |
| | XSDRNPM, | | deterministic |
| | ORIGEN, OPUS | centrm | transport, |
| | | | coupled with |
| | | xslevel=1/2/3\ | ORIGEN |
| | | *a*/4 | depletion. |
| | | | |
| | | addnux=0/1/2\ | |
| | | :sup:`a`/3/4 | |
| | | | |
| | | weight\ :sup:`b`| |
+-----------------+-----------------+-----------------+-----------------+
| ``=T-DEPL`` | XSProc, NEWT, | | 2D MG |
| | ORIGEN, OPUS | | deterministic |
| | | | transport, |
| | | | coupled with |
| | | | ORIGEN |
| | | | depletion. |
+-----------------+-----------------+-----------------+-----------------+
| ``=T5-DEPL`` | XSProc\ :sup:`c`| | 3D, Monte Carlo |
| | KENO-V.a, | | transport |
| | ORIGEN, OPUS | | (KENO-V.a), |
| | | | coupled with |
| | | | ORIGEN |
| | | | depletion. |
+-----------------+-----------------+-----------------+-----------------+
| ``=T6-DEPL`` | XSProc\ *c*, | | 3D, Monte Carlo |
| | KENOVI, ORIGEN, | | transport |
| | OPUS | | (KENO-VI), |
| | | | coupled with |
| | | | ORIGEN |
| | | | depletion. |
+-----------------+-----------------+-----------------+-----------------+
| **Sensitivity | | | |
| and Uncertainty | | | |
| Analysis | | | |
| sequences** | | | |
+-----------------+-----------------+-----------------+-----------------+
| ``=TSUNAMI-2D`` | XSProc, NEWT, | bonamist | 2D forward and |
| | SAMS, BONAMIST | | adjoint |
| ``=TSUNAMI-2DC``| | bonami | transport |
| | | | calculations, |
| | | centrm\ *a* | followed by S/U |
| | | | analysis with |
| | | | SAMS |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`a`\ Defau\| | | |
| lt parm value. | | | |
| Note that | | | |
| centrm is | | | |
| equivalent to | | | |
| xslevel=4. See | | | |
| :ref:`3-1-3-7-2`| | | |
| for details. | | | |
| | | | |
| :sup:`b`\ pa\| | | |
| rm=weight | | | |
| is used to | | | |
| generate a | | | |
| broad group | | | |
| cross | | | |
| section | | | |
| library. | | | |
| This parm | | | |
| option is | | | |
| only | | | |
| available | | | |
| for the | | | |
| T-DEPL | | | |
| sequence. | | | |
| | | | |
| *c*\ T5-DEPL | | | |
| and T6-DEPL | | | |
| is also | | | |
| available in | | | |
| CE-mode, | | | |
| which does | | | |
| not invoke | | | |
| XSProc for | | | |
| cross | | | |
| section | | | |
| processing. | | | |
+-----------------+-----------------+-----------------+-----------------+
.. _3-1-2:
Overview of TRITON Sequences
----------------------------
The TRITON control module supports nine calculational sequences, each
with its own design and applications. Each of these sequences is
described in the following subsections.
The first subsection covers the basic cross section processing sequence
T-XSEC. The T-XSEC sequence prepares problem-dependent multigroup cross
sections for subsequent transport analysis. The second subsection covers
TRITON’s transport analysis sequences, while the third subsection
discusses TRITON’s depletion analysis sequences. The final subsection is
dedicated to the TSUNAMI-2D sensitivity and uncertainty analysis
sequences in TRITON.
.. _3-1-2-1:
Cross section processing sequence (T-XSEC)
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
The T-XSEC sequence provides the ability to prepare a problem-dependent
multigroup cross section library using SCALE cross section processing
modules to appropriately account for spatial and energy self-shielding
effects. The problem-dependent cross section library contains
microscopic cross sections for each nuclide for each material
composition defined in the TRITON input. SCALE provides several unit
cell types (e.g., a lattice of pins, an infinite medium, a multiregion
problem, or a doubly heterogeneous cell) to correct the cross sections
for spatial and energy self-shielding. Multiple cell calculations can be
used in the same calculation. The calculation of multigroup cross
sections is executed by XSProc (:ref:`7-1`).
.. _3-1-2-2:
Transport sequences (T-XSDRN, T-NEWT)
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
The TRITON transport sequences build upon the cross section processing
sequence by automating a transport calculation after cross section
processing. Both 1D and 2D discrete-ordinates transport calculations can
be performed using XSDRNPM and NEWT, respectively. The T-XSDRN sequence
calls XSDRNPM for transport analysis in slab, sphere, or cylindrical
geometries, while the T-NEWT sequence calls NEWT for analyses in 2D
*xy-*\ geometries. In addition to the input necessary for cross section
processing, an XSDRN or NEWT input model is also required. The XSDRN
model input is discussed in Appendix A of TRITON; the NEWT model input
requirements are described in the NEWT chapter. Similar capabilities and
applications for KENO-V.a and KENO-VI are handled through the CSAS5 and
CSAS6 sequences, respectively.
.. _3-1-2-3:
Depletion sequences (T-DEPL, T-DEPL-1D, T5-DEPL, T6-DEPL)
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
The TRITON depletion sequences build upon the transport sequences by
automating depletion/decay calculations after the transport calculations
for each material designated for depletion. One or more materials in the
model can be designated for depletion. Each designated material is
depleted using region-averaged reaction rates, accounting for all
regions in the model associated with a given depletion material. The
TRITON depletion calculation procedure is described further in the next
subsection. TRITON automates the various computational processes—cross
section processing, transport, and depletion—over a series of depletion
and decay intervals supplied by the user. The depletion procedure is
discussed in :ref:`3-1-2-3-1`. The 2D TRITON depletion sequence (T-DEPL),
which uses NEWT for the transport calculations, also provides the
capability to generate lattice-physics data for nodal core calculations.
These lattice physics capabilities are discussed in :ref:`3-1-2-3-2`.
Within TRITON depletion calculations, TRITON invokes the ORIGEN
depletion module for the time-dependent transmutation of each
user-defined material. TRITON provides ORIGEN the neutron flux
space-energy distribution, the multigroup cross sections, material
concentrations, and material volumes. ORIGEN performs the flux
normalization, cross section collapse, and multi-material
depletion/decay operations to determine new isotopic concentrations for
the next calculation.
.. _3-1-2-3-1:
Predictor-corrector depletion process
^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^
For all depletion sequences, TRITON automates cross section processing,
transport, and depletion calculations over a series of depletion-decay
intervals supplied by the user. A **depletion interval** represents a
time interval in which the model power level is assumed constant. A
depletion model that exhibits various power level changes will require
multiple depletion intervals to accurately model the changes in power.
Each depletion interval can be followed by a decay calculation over a
user-specified **decay interval**.
Within a given depletion interval (e.g., an LWR operating at constant
power for a 12-month fuel cycle), the isotope concentrations of
different depletion materials change, which induces changes in the
problem-dependent multigroup cross sections (through spatial and energy
self-shielding effects) as well as the neutron flux distribution,
leading to different power distributions and transmutation rates in
depletion materials. This requires TRITON to represent each depletion
interval as a series of smaller time intervals in which cross section
processing and transport solutions are recomputed to accurately model
these time-dependent effects. A **depletion subinterval** represents a
time interval in which TRITON performs cross section processing and
transport calculations to determine cross sections and flux
distributions used in the depletion calculations. All depletion
subintervals for a given depletion interval have the same length—for
example, one 12-month depletion interval can be represented as a series
of 12 one-month depletion subintervals, or as 6 two-month depletion
subintervals. Alternatively, the 12-month depletion interval can be
modeled as two consecutive 6-month depletion intervals, each one having
a different number of subintervals. Therefore the formulation of a
**depletion scheme** in TRITON is highly flexible. A depletion scheme is
the set of user-defined depletion and decay intervals with associated
power levels and number of subintervals. *TRITON does not provide
automated means to determine the appropriate depletion scheme for a
given application. The user must determine the accurate depletion scheme
specific to his or her application.*
TRITON uses a predictor-corrector approach to process the user-defined
depletion scheme. The predictor-corrector approach performs cross
section processing and transport calculations based on anticipated
isotope concentrations at the *midpoint* of a depletion subinterval.
Depletion calculations are then performed over the full subinterval
using cross sections and flux distributions predicted at the midpoint.
Depletion calculations are then extended to the midpoint of the next
subinterval (possibly through a decay interval and into a new depletion
interval), followed by cross section processing and transport
calculations at the new midpoint. The iterative process is repeated
until all depletion subintervals are processed. In order to start the
calculation, a “bootstrap case” is required using initial isotope
concentrations for the initial cross section processing and transport
calculation. The bootstrap calculation is used to determine the
anticipated isotope concentrations at the midpoint of the first
depletion subinterval.
The predictor-corrector approach is best explained by an example.
:numref:`fig3-1-1` illustrates the predictor-corrector process for a
hypothetical depletion scheme with two depletion intervals. The first
depletion interval contains two subintervals, followed by a decay
interval. The second depletion interval contains one subinterval and is
also followed by a decay interval. In :numref:`fig3-1-1`, cross section
processing and transport calculations are represented by the “T” label,
and depletion calculations are represented by the “D" label. For this
example, four sets of calculations would be necessary: one for each of
the three depletion subintervals, and one for the initial “bootstrap
case.” These calculations are represented in the following eight steps.
Step 1 T\ :sub:`0`: Cross section processing and transport calculation
using initial (i.e., time-zero) isotope concentrations.
Step 2 D\ :sub:`1`: Depletion calculation from time-zero to the midpoint
of the first depletion subinterval. The dashed horizontal arrow denotes
a “predictor” depletion step.
Step 3 T\ :sub:`1`: Cross section processing and transport calculation
at the midpoint of the first depletion subinterval.
Step 4 D\ :sub:`1`: Depletion calculation for the first depletion
subinterval. The solid horizontal arrow across the subinterval denotes a
“corrector” depletion step. *Corrector steps use cross sections and flux
distribution computed at the subinterval midpoint.* This is represented
by a solid arrow from T\ :sub:`1` to D\ :sub:`1`.
D\ :sub:`2`: Predictor depletion calculation for the second depletion
subinterval. *Predictor steps use cross sections and flux distribution
computed at the*\ **previous**\ *subinterval midpoint.* This is
represented as the dashed arrow from T\ :sub:`1` to D\ :sub:`2`.
Step 5 T\ :sub:`2`: Cross section processing and transport calculation
at the midpoint of the second depletion subinterval.
Step 6 D\ :sub:`2`: Corrector depletion calculation for the second
depletion subinterval, followed by the decay calculation at the end of
the first depletion interval.
D\ :sub:`3`: Predictor depletion calculation for the third depletion
subinterval. The third depletion subinterval is the first and only
subinterval associated with the second depletion interval.
Step 7 T\ :sub:`3`: Cross section processing and transport calculation
at the midpoint of the third depletion subinterval.
Step 8 D\ :sub:`3`: Corrector depletion calculation for the third
depletion subinterval. This calculation is followed by a second decay
calculation.
.. _fig3-1-1:
.. figure:: figs/TRITON/fig1.png
:align: center
:width: 500
Predictor/corrector depletion algorithm used by TRITON.
The depletion calculations are performed by ORIGEN and span either the
first half of a subinterval (predictor step) or the full subinterval
(corrector step). ORIGEN performs these depletion calculations and
possible decay calculations over a series of smaller time intervals. The
**ORIGEN time intervals** are automatically determined by TRITON
depending on the length of the depletion subinterval and decay interval.
Additionally, TRITON will automatically adjust the number of
subintervals per depletion interval if the time length of the
user-defined subinterval is large (i.e., >400 days). TRITON writes the
utilized depletion scheme near the top of the output file. The depletion
scheme output edit is further described in :ref:`3-1-5-4-1`.
.. _3-1-2-3-2:
Lattice physics analysis
^^^^^^^^^^^^^^^^^^^^^^^^
The 2D depletion sequence (T-DEPL) may be used to generate lattice
physics data for subsequent core analysis calculations using core
simulator software. Core simulators typically employ few-group nodal
diffusion theory for neutronic calculations, coupled with other
calculation methods for thermal hydraulics, fuel performance, and plant
operation (e.g., soluble boron letdown or control rod movement). Core
simulation requires the use of pretabulated **lattice physics data** for
the neutronic calculations—that is, few-group homogenized cross
sections, with appropriate discontinuity factors, pin powers, and
kinetic parameters, functionalized in terms of burnup and other system
conditions such as fuel temperature and moderator density.
To support lattice physics database preparation, the NEWT transport
module contains flexible input options to define the few-group energy
structure, spatial homogenization regions, and discontinuity factors.
After the transport calculation at the midpoint of each depletion
subinterval, NEWT computes the lattice physics data and stores this data
on a temporary file. TRITON reads the temporary file and archives the
lattice physics data onto a separate database file. In addition, the
*T-DEPL* sequence supports branch calculations in which perturbations
may be applied to certain system conditions such as fuel temperatures
and moderator density. TRITON automates the cross section processing and
transport calculations for each branch condition at the midpoint of the
depletion subinterval. NEWT computes the lattice physics data for the
branch calculations, and TRITON archives this data onto the lattice
physics database file.
The TRITON input options for branch calculations are described in :ref:`3-1-3-3-2`,
and the file format of the lattice physics database is
provided in the Appendix B of TRITON.
.. note:: The TRITON input options for
branch calculations are designed to be highly flexible to support a
large range of core analyses; therefore, TRITON does not provide
automated means to determine the branch calculations. The user must
determine the necessary branch calculations for his or her core analysis
and be knowledgeable of the capabilities and limitations of the cross
section treatment of the core simulator. The TRITON Lattice Physics
Primer has been developed to provide guidance on appropriate TRITON
branch calculations for LWR core analysis (NUREG/CR-7041) and in “Cross
Section Generation Guidelines for TRACE-PARCS” (NUREG/CR-7164).
.. _3-1-2-4:
S/U analysis sequences (TSUNAMI-2D, TSUNAMI-2DC)
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
TRITON supports a 2D sequence (TSUNAMI-2D) to support cross section
sensitivity and uncertainty (S/U) analysis. The TSUNAMI-2D sequence is
similar in function to the 2D transport sequence T-NEWT, except that
TRITON sets up additional calculations to perform S/U analysis. After
the initial transport calculation, a second transport calculation is
performed to compute the adjoint flux solution. Both the forward and
adjoint fluxes are saved to different files that are read by the SAMS
module to compute sensitivity coefficients and the uncertainty in
*k*\ :sub:`eff`. In addition to S/U analysis for *k*\ :sub:`eff`, the TSUNAMI-2D
sequence allows for S/U analysis of user-defined ratios of forward flux
responses, such as flux-weighted cross sections, reaction rate ratios,
and power-peaking factors. For each user-defined response ratio, TRITON
automates additional fixed-source adjoint transport calculations in
NEWT, followed by S/U calculations in SAMS. TRITON iteratively calls
NEWT and SAMS for each response ratio definition. TRITON also
automatically sets up the fixed-source input for NEWT, which is
implicitly defined by the response ratio specification.
For TSUNAMI-2D calculations, the TRITON input is similar to 2D transport
calculations (T-NEWT) with some additional input options. Input data
blocks are available to define response ratios for S/U analysis.
Additional input data blocks are available for the SAMS module to
control various aspects of the sensitivity and uncertainty calculations
and output formatting. The S/U input options are further discussed in
the TSUNAMI-1D manual. Examples of TSUNAMI-2D models are provided in
:ref:`3-1-6`.
Although TSUNAMI-2D is similar to the T-NEWT sequence, enhanced versions
of the cross section processing codes are used to compute sensitivity
data necessary for the SAMS calculation. In place of the BONAMI code
used by T-NEWT, TSUNAMI-2D utilizes a sensitivity version called
BONAMIST. This enhanced code computes the problem-dependent multigroup
cross sections along with their sensitivities to the input data, the
so-called “implicit sensitivities.” Implicit sensitivity effects from
ENDF/B-VII cross sections are now treated with full-range Bondarenko
factors present on the multigroup library with BONAMIST.
The NEWT-based TSUNAMI-2D sequence is functionally similar to the
TSUNAMI-1D and TSUNAMI-3D_K5/TSUNAMI-3D_K6 sequences of SCALE, in which
cross section processing, forward and adjoint transport calculations,
and S/U calculations are automated using XSDRN and KENO V.a/KENO-VI,
respectively. Like these S/U sequences, TRITON supports an auxiliary
sequence, TSUNAMI-2DC\ *,* which does not perform the additional adjoint
transport calculations and SAMS calculations. The TSUNAMI-2DC sequence
does not employ the sensitivity version of the cross section processing
code, BONAMIST, as the sensitivity data from this code is not used in
the calculation sequence. The TSUNAMI-2DC sequence is identical to the
T-NEWT sequence with one exception: if user-defined response ratios are
provided in the input, then TSUNAMI-2DC will compute the response ratios
and edit their values in the output file.
.. note:: During the development of SCALE 6.2, the cross section processing
routines were significantly modernized to use XSProc for all of the
TRITON sequence except TSUNAMI-2D; therefore, the run times will be
slower for each T-NEWT calculation and the solution will be somewhat
different. However, the other TSUNAMI sequences do use XSProc.
.. _3-1-3:
Input Description
-----------------
TRITON input is free-form and keyword based, similar in form to many
other modules in SCALE. With a few exceptions, the following formatting
rules apply:
- Data is limited to 255 columns but may wrap into as many lines as are
needed.
- Comment lines start with a tick mark (') in the first column of a
line and may be placed anywhere in the input.
- The keyword-based input is case insensitive.
- TRITON input is organized into blocks of data. Each data block begins
with *read blockname* and terminates with *end blockname*.
- Blocks of data may appear in any order. Each block of data may appear
only once in the input.
- Input can be redirected from an auxiliary file by using the open
angle bracket (<) and the name of the file—for example,
*0, TRITON updates the appropriate parameters and
re-executes the cross section processing and transport calculations.
Responses of interest are saved to a database file (i.e., the txtfile16
file) for both the nominal and perturbed-state conditions, and TRITON
reverts to cross sections and fluxes from the reference branch 0 to
proceed with the depletion calculation. The process repeats following
each depletion subinterval, until all depletion subintervals are
simulated. Responses of interest are added to the database file for all
branches at each depletion subinterval.
.. _fig3-1-8:
.. figure:: figs/TRITON/fig8.png
:align: center
:width: 500
Schematic illustration of T-DEPL branch calculations during depletion.
Branch perturbations may be applied to any of the following five
parameters: fuel temperature, moderator temperature, moderator density,
moderator soluble boron concentration, and control rod insertion. These
properties may be varied individually or simultaneously. Branch
calculations are specified in the TRITON *BRANCH* data block. The
*BRANCH* data block has the following form.
::
READ branch
define deftype I1 I2 ... In end
...
tf=fueltemp tm=modtemp dm=moddens sb=boronconc cr=inout end
...
END branch
where
*deftype* = 'fuel,' 'mod,' 'crout', or 'crin',
*I\ i* = list of materials associated with type definition *deftype*,
*fueltemp* = branch fuel temperature (K),
*modtemp* = branch moderator temperature (K),
*moddens* = branch moderator density (g/cm\ :sup:`3`),
*boronconc* = soluble boron concentrations (ppm),
*inout* = control rod/blade state (out = 0, in = 1).
The type definitions must come first within the *BRANCH* block, and at
least one definition is always required. The 'fuel' type definition is
used to specify which of the problem materials are considered to be fuel
during branch calculations; similarly, the 'mod' type definition
specifies the material or materials that are to be considered moderator.
The 'crout' definition specifies the materials that are in place in the
transport model when control structures are withdrawn, while the 'crin'
definition specifies the materials that are present in the transport
model when a control structure is inserted. The 'fuel' definition must
be present if any fuel temperature branches are performed. The 'mod'
type definition must be present whenever moderator temperature,
moderator density, or soluble boron branches are performed. Both the
'crout' and 'crin' definitions must be present if control rod branches
are requested. Definitions may not be repeated—for example, 'define
fuel' may occur only once.
Type definitions are followed by branch specifications. For each branch,
one or more branch specifications may be given; if a particular property
is omitted, then the reference conditions of the original model and
material specifications are used. **The first branch specification must
describe the nominal conditions,** and all parameters must be specified
for this branch. Each branch specification can optionally define up to
five branch keywords before terminating with the *end* keyword. The five
branch keywords are as follows.
`` tf`` = fuel temperature (K),
`` tm`` = moderator temperature (K),
`` dm`` = moderator density (g/cm\ :sup:`3`),
`` sb`` = soluble boron concentration (ppm boron), and
`` cr`` = control rod state (out = 0, in = 1).
The format of a *BRANCH* block is best illustrated by an example.
:numref:`fig3-1-9` shows a complete branch data block for a five-branch
calculation, with embedded descriptions of each branch. Note that there
are six entries; the first branch is the reference or branch 0 state.
In this example, materials 11 and 12 are specified as 'fuel', and fuel
temperature perturbations will be applied to only these materials. The
nominal temperature for both materials is determined from the branch 0
input (901 K). The nominal fuel temperature must be the same for all
materials in the definition and must be consistent with the initial
standard composition input. Similarly, materials 13 and 14 are defined
as the moderator materials. The temperature (559 K), density (0.76
g/cm\ :sup:`3`), and soluble boron concentrations (655 ppm) for the
reference state must be identical to those of the initial material
specifications and must be identical for all materials defined as
moderator.
.. _fig3-1-9:
.. figure:: figs/TRITON/fig9.svg
:align: center
:width: 500
In a reactor core, when a control structure (rod, blade, etc.) is
withdrawn, the volume occupied by the structure is replaced by something
else. Thus, in a branch calculation with rod insertion and withdrawal,
the material(s) present for both states must be specified. If the
reference condition is defined as control rods withdrawn (i.e., cr = 0),
the NEWT geometry model must contain the materials defined by 'crout'.
For a control rod insertion branch (cr = 1), TRITON exchanges the
materials specified in the 'crin' definition (30, 31) with corresponding
materials in the 'crout' definition (20, 21). Conversely, if the
reference condition is defined as control rods inserted (i.e., cr = 1),
the NEWT geometry model must contain the materials defined by 'crin'.
For a control rod withdrawal branch (cr = 0), TRITON exchanges the
materials specified in the 'crout' definition with corresponding
materials in the 'crin' definition. For this reason, unique material
numbers must be paired between crin and crout definitions. For example,
consider a zirc-clad B\ :sub:`4`\ C control rod inserted during a
control rod insertion branch, with materials 30 and 31 representing the
clad and rod materials, respectively. In the withdrawn position, both
the clad and poison materials are replaced by the moderator. To have
consistent definitions of 'crin' and 'crout', two moderator materials
must be defined for the withdrawn state: one corresponding to the clad
material and one corresponding to the rod material.
As mentioned earlier, only one condition keyword is required per branch,
but all five may be used. However, the reference state (branch 0) entry
must specify all five conditions. For subsequent branches, when a
specific branch state is not specified, the reference state is used. In
the above example, the first entry, branch zero, specifies the reference
state with a fuel temperature of 901 K, moderator temperature of 559 K,
moderator density of 0.4 g/cm\ :sup:`3`, control rod withdrawn, and a
soluble boron concentration of 655 ppm. The second entry (branch 1)
specifies a moderator density of 0.80 g/cm\ :sup:`3` and the control rod
state as withdrawn. Since the reference state is for a withdrawn control
rod, the statement cr = 0 is redundant (but completely acceptable). The
next branch is identical to the previous branch, except that in this
case the control rod is inserted. For both cases, reference fuel and
moderator temperatures were used. In the following branch, the soluble
boron concentration is changed to 20 ppm, and the moderator density is
again set to a value of 0.8 g/cm\ :sup:`3`. In fact, this moderator
density is applied to all five branches. Along with the moderator
density change, the soluble boron concentration is changed to 1300 ppm
for the next branch. And finally, in the last branch, in addition to the
moderator density change, the fuel temperature is changed to 559 K. For
this case, reference conditions are used for boron concentration,
moderator temperature, and control rod state.
Note that TRITON compares the reference values of fuel temperature,
moderator temperature, moderator density, and soluble boron
concentration with the data entered in the *COMPOSITION* block. TRITON
prints warning messages if the data in the *COMPOSITION* block and
*BRANCH* block are inconsistent. Also note that each branch calculation
is independent of other branch calculations. Thus, the order in which
branch calculations are computed is not important.
Branch calculations are usually requested for lattice physics analysis,
where the objective is to generate a database of few-group homogenized
cross sections for nodal core calculations. Thus, BRANCH blocks are used
in tandem with the NEWT’s *COLLAPSE, HOMOGENIZATION,* and *ADF* blocks.
With these blocks of data, TRITON will archive lattice physics
data—few-group homogenized cross sections, assembly discontinuity
factors (ADFs), homogenized kinetic parameters, pin powers, and form
factors—to a binary file called xfile016 in the SCALE temporary working
directory. An auxiliary text-formatted data file called txtfile16 is
also created in the SCALE temporary working directory. This file format
is documented in Appendix A (:ref:`3-1a` of TRITON.
.. _3-1-3-3-3:
BRANCH block with user-defined Dancoff factors
^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^
As previously mentioned in :ref:`3-1-3-1-2`, TRITON uses Dancoff factors
as part of its cross section processing calculations. Dancoff factors
play an important role in characterizing spatial self-shielding effects.
The XSProc module computes the Dancoff factors based on the *CELLDATA*
input. For a square-pitched lattice cell example, Dancoff factors are
computed by DANCOFF by assuming that the fuel pin is within an infinite
lattice of identical fuel pins. The assumption of an infinite uniform
lattice of fuel pins may lead to inaccurate Dancoff factors for certain
configurations such as BWR assembly designs, leading to inappropriate
problem-dependent multigroup cross sections. Moreover, the Dancoff
factors may change significantly for certain branch conditions, such as
changing the in-channel moderator density in a BWR assembly.
The TRITON BRANCH block allows the user to specify material-dependent
Dancoff factors for various branch conditions. Branch-specific Dancoff
factors may be utilized by defining a new set of material-dependent
Dancoff factors using the *d2pset* type definition. The set of Dancoff
factors may be included in a branch specification by using the *d2p=*
keyword. The *BRANCH* block now has the following format.
::
READ branch
define deftype I1 I2 ... In end
define d2pset id M1 D1 M2 D2 ... Mn Dn end
...
tf=fueltemp tm=modtemp dm=moddens sb=boronconc cr=inout d2p=d2pID end
...
END branch
In the type definition section, the *d2pset* keyword is followed by a
positive integer identifier, which is subsequently followed by pairs of
material identifiers and their user-defined Dancoff factor value.
Multiple material/Dancoff pairs may be entered for a particular set
definition, as long as the material identifiers are unique. Multiple set
definitions are allowed, as long as the set identifiers are unique.
The *d2p=* keyword in the branch specification can be assigned to any
set identifier defined in the branch definition section. If *d2p=* is
utilized, the material/Dancoff pairs in the set definition are applied
for the given branch condition. The values *d2p=0* and *d2p=-1* have
special meaning. If *d2p=* is set to 0, the material/Dancoff pairs
defined in the *CELLDATA* block are utilized. If *d2p=* is set to -1,
the default MIPLIB-computed Dancoff factors will be utilized, even if
material/Dancoff pairs are defined in the *CELLDATA* block using the
*dan2pitch* keyword available there. The nominal (branch 0) condition
**must** use the material/Dancoff pairs (if defined) in the *CELLDATA*
block; therefore, the first branch specification **must** not set the
*d2p* keyword to anything other than zero. (Note: *d2p=0* need not be
defined for the first branch condition since this is always the case.)
In :numref:`fig3-1-10`, the *BRANCH* block from the previous example has been
modified to use branch-specific Dancoff factors. In this example, the
nominal branch defines the reference moderator density to be
0.4 g/cm\ :sup:`3`, and five branches use a higher moderator density of
0.8 g/cm\ :sup:`3`. The Dancoff factors for the higher moderator density
condition are different from the reference moderator density. To account
for the different Dancoff factors at the higher moderator density
condition, a set of material/Dancoff pairs are defined with the set
identifier of 400. In the set, fuel material 11 has a Dancoff factor of
0.4, and fuel material 12 has a Dancoff factor of 0.5. The set of
Dancoff factors is used for the five branch states through the
specification of the *d2p=* keyword to 400.
.. _fig3-1-10:
.. figure:: figs/TRITON/fig10.svg
:align: center
:width: 500
*Example BRANCH block input with Dancoff factors.*
.. _3-1-3-3-4:
DE\ **P**\ LETION block
^^^^^^^^^^^^^^^^^^^^^^^
The *DEPLETION* block, used by the four depletion sequences, is simple
in concept but performs four important functions. First, this block
specifies the materials for which depletion calculations are to be
performed. In general, it is desirable to perform depletion calculations
only for fuel and target materials of interest. Calculating the
depletion of gas gaps, cladding, moderator, or coolant is usually of
little value unless the material contains components that will be
significantly depleted with burnup. Additionally, it is not usually
desirable to deplete soluble poisons in reactor coolants. Therefore, the
*DEPLETION* block requires that the user specify the materials to be
depleted. There are no defaults; hence, the block is required for all
depletion sequences.
The second function of the *DEPLETION* block is to specify the basis to
which the model power is normalized. In general, the average
time-dependent power to which an irradiated object is exposed is known.
For example, an LWR fuel assembly discharged from a reactor is known to
have operated at certain power levels for one or more time periods. The
individual pins in the assembly will have varying power levels depending
on position and assembly design. In such a case, the basis for the input
power is the full assembly. Fluxes computed in the transport solution
will be normalized by TRITON based on reaction rates and energies **in
all problem materials** (depleted and nondepleted materials) such that
the assembly-wide power will match the power given in *BURNDATA* block.
However, it is often the case in radiochemical assay analysis that the
burnup history of a specific pin is known and isotopic concentrations
for that pin are desired. It is still necessary to model the full
assembly in order to properly characterize the fluxes in that pin. In
such a case, it would be advantageous to specify the operating history
for the pin instead of the full assembly. When this is done, the average
specific power of the full assembly will be different from that of the
pin and will be computed automatically based on power distributions
calculated for the assembly. In other words, powers for other materials
in the assembly will be normalized such that the power in the pin of
interest matches that specified in the *BURNDATA* block. The material of
that pin becomes the *basis* for power normalization.
:ref:`3-1-3-3-4-1` below describes the general format of the
*DEPLETION* block that is available to all four depletion sequences. The
third function of the *DEPLETION* block is an optional function used to
specify ORIGEN solver options and ORIGEN depletion mode for each
depletion material. These options are further described in Sect.
:ref:`3-1-3-3-4-2`. The fourth function of the *DEPLETION* block is to define
optional deletion instructions used to simplify cross section processing
using the ASSIGN function. Special provisions have been made in the 1D
and 2D depletion sequence (T-DEPL-1D and T-DEPL) to reduce the number of
cross section processing calculations in order to decrease calculation
run-time. The ASSIGN functionality is further described in Sect.
:ref:`3-1-3-3-4-3`.
.. _3-1-3-3-4-1:
Basic DEPLETION block format
''''''''''''''''''''''''''''
The basic format of the DEPLETION block is as follows:
::
READ depletion M1 M2 M3... Mn END depletion
where M\ :sub:`i` represents the SCALE material numbers for materials to
be depleted. As discussed above, the *DEPLETION* block can also be used
to specify the basis for the input power. Power normalization is
accomplished by prefixing the material number(s) with a negative sign
(–). For example, consider a problem in which materials 1, 2, and 3 are
being depleted, but the power for material 1 is known. The *DEPLETION*
block for this case is
::
READ depletion -1 2 3 END depletion
In this case, powers for all materials will be normalized such that the
power in material 1 matches the input power specification in the
*BURNDATA* block.
Note that multiple materials can be used as a power basis. Consider a
fuel assembly with three fuel types represented by materials 1, 2, and
3, and also containing cladding as material 4 and water as material 5.
:numref:`tab3-1-2` illustrates multiple ways that the power basis for this
assembly might be specified and describes the effect of each
specification.
.. _tab3-1-2:
.. table:: Effects of different power basis specifications.
:align: center
+-----------------------------------+-----------------------------------+
| ``READ depletion`` | The assembly-averaged power is |
| ``1 2 3`` | normalized to match the input |
| ``END depletion`` | specific power |
+===================================+===================================+
| ``READ depletion`` | The assembly-averaged power is |
| | normalized such that the power of |
| ``-1 2 3`` | material 1 matches the input |
| | specific power |
| ``END depletion`` | |
+-----------------------------------+-----------------------------------+
| ``READ depletion`` | The assembly-averaged power is |
| | normalized such that the average |
| ``-1 -2 3`` | power in materials 1 and 2 |
| | matches the input specific power |
| ``END depletion`` | |
+-----------------------------------+-----------------------------------+
| ``READ depletion`` | The assembly-averaged power is |
| | normalized to match input |
| ``1 2 3 4 5`` | specific powers. TRITON will |
| | attempt to do depletion in |
| ``END depletion`` | cladding and moderator materials |
| | too. (Note that cladding and |
| | moderator materials should be |
| | depleted using the |
| | deplete-by-flux option described |
| | in the next subsection) |
+-----------------------------------+-----------------------------------+
| ``READ depletion`` | The assembly-averaged power is |
| | normalized such that the average |
| ``-1 -2 -3`` | power in materials 1–3 matches |
| | the input specific power. This is |
| ``END depletion`` | not the same as the normalizing |
| | specification for an assembly |
| | average, because it neglects |
| | contributions of n-γ sources in |
| | moderator and cladding materials |
+-----------------------------------+-----------------------------------+
.. _3-1-3-3-4-2:
ORIGEN depletion options
''''''''''''''''''''''''
ORIGEN provides two input options for the flux used in the depletion
calculation: direct specification of fluxes (i.e., deplete by flux) or
indirect specification of fluxes in terms of power (i.e., deplete by
power). The ORIGEN depletion is based on a known flux; however, it is
more often the case that one knows the specific power in a depletion
region rather than the actual flux. When ORIGEN is used in
deplete-by-power mode, ORIGEN will internally determine the
corresponding flux from the input-specific power and internal tables of
fission and capture energy releases for the nuclides present and the
macroscopic cross sections of those nuclides. Additionally, at each
ORIGEN time interval, ORIGEN recalculates the material power density as
nuclide inventories change. Hence, the deplete-by-power mode will result
in a time-varying flux, whereas the deplete-by-flux mode will result in
a constant flux over the calculation time interval. Since reactors
typically operate at a constant (or nearly so) power level, with varying
local fluxes, the deplete-by-power option is closer to reality. However,
the choice of approach is generally not an issue. Significant
differences between calculation results between the two depletion modes
would indicate that the TRITON depletion subintervals are too large.
By default, all TRITON depletion materials use the deplete-by-power
mode. However, there exist some circumstances where deplete-by-flux is
more appropriate. In deplete-by-power mode, ORIGEN will often halt when
an attempt to maintain constant power results in a large change in flux
between ORIGEN time intervals. Large changes in flux can occur in media
where isotope contents are changing rapidly with time, such as in a
gadolinium-bearing burnable absorber rod, where gadolinium is being
rapidly depleted with time. Another circumstance pertains to activation
analysis of nonfuel materials. The flux for these materials is typically
governed by external power sources (i.e., fuel materials located
elsewhere in the problem domain) rather than by internal power sources.
Therefore, the deplete-by-flux option is recommended for these
materials.
TRITON provides the option to specify deplete-by-flux mode for selected
depletion materials using a modified form of the depletion
specification:
::
READ depletion M1 M2 M3...Mi-1 flux Mi Mi+1... Mn END depletion
Materials preceding the *flux* keyword are depleted using the
deplete-by-power mode; materials following the flux keyword are depleted
using the deplete-by-flux mode. For example, consider a problem in which
materials 1–6 are to be depleted, but materials 3 and 4 represent
nonfuel materials that do not contribute significantly to the total
power and are therefore to be depleted assuming constant flux. The
*DEPLETION* block for this situation could be specified as follows.
::
READ depletion 1 2 5 6 flux 3 4 END depletion
The DEPLETION block also supports the specification of the ORIGEN
calculation method. The default option is solver=matrex, which
represents the matrix exponential option. The other option is
solver=cram, which represents the new CRAM solver option in ORIGEN. An
example depletion specification for the cram solver is as follows.
::
READ depletion 1 2 5 6 flux 3 4 solver=cram END depletion
.. _3-1-3-3-4-3:
Cross section processing simplification using ASSIGN
''''''''''''''''''''''''''''''''''''''''''''''''''''
When depleting a large number of fuel materials, considerable time may
be spent in the cross section processing calculations prior to the
multigroup transport calculation. Fuel assembly designs may require
20-200 unique depletion materials across the different fuel pins in the
assembly. In such cases, an assembly model may require hours of run-time
for each pin-wise cross section processing calculation in order to
perform a 10-minute transport solution.
Although highly rigorous, such a cross section processing process is
extremely burdensome for depletion calculations, especially if branch
calculations are requested. To reduce run-time, the 2D depletion
sequence (*T-DEPL*) provides the option to group depletion materials
together such that they are tracked independently in the depletion
calculation but use a common set of microscopic cross sections. The
microscopic cross sections for a given depletion group are computed
using the average composition of all the depletion materials within the
group. Typically, this grouping is applied to fuel pins of identical
initial composition. Although the nuclide number densities of such pins
will diverge with burnup as a function of location within an assembly,
the cross sections of these pins are well represented by a single pin
cell calculation with an average composition representative of all these
pins.
Although the material grouping option introduces approximations in the
cross section processing calculations, which in turn affects the
transport and depletion calculations, internal investigations have shown
that solution accuracy can be maintained for a wide range of assembly
designs while significantly improving the run-time.
The alternate format of the *DEPLETION* block for simplified cross
section processing is as follows.
::
READ depletion M1 M2 M3... Mz END
assign N1 Ma Mb ... Mx end
assign N2 Mf Mg ... My end
...
assign Nn Mj Mk ... Mz end
END depletion
Similar to the basic format, each material designated for depletion
(M\ :sub:`i`) is listed after *READ* depletion and before the *END*
keyword. Each designated depletion material must be present in the 2D
NEWT model. After the first *END* keyword, the alternate format contains
a list of material “assignments” used to simplify cross section
processing for a group of depletion materials. The material assignments
begin with the *assign* keyword and terminate with the *end* keyword.
After the *assign* keyword, a unique representative material identifier
(N\ :sub:`j`) is defined. The representative material is associated with
the group of depletion materials that immediately follows in the
*assign* definition. The representative material identifier is used in
the *COMPOSITION* and *CELLDATA* blocks to define the initial
composition, temperature, and cell definition for the group of depletion
materials. Thus, the *assign* definitions in *TRITON* are currently
constrained such that each depletion material group must have the same
initial composition. After the last *assign* definition, the depletion
block is terminated with *END depletion*.
Only depletion materials may be assigned to representative materials.
The group of depletion materials assigned to a particular representative
material must **not** appear in the *COMPOSITION* and *CELLDATA* blocks.
The use of material assignments is best illustrated by an example.
:numref:`fig3-1-11` shows a complete T‑DEPL input that uses material
assignments. A 2D plot of the model is shown in :numref:`fig3-1-12`. In this
example, two fuel materials are defined as materials 1 and 2 in the
*COMPOSITON* block. In the *DEPLETION* block, the list of depletion
materials includes materials 1, 20, 30, and 40. Depletion materials 20,
30, and 40 are “assigned” to representative material 2. Material 2 does
not appear in the depletion list or the transport model; materials 20,
30, and 40 do. But only material 2 is defined in the *COMPOSITION* and
*CELLDATA* blocks. In the transport model, four units are defined, one
for each material. An array is used to place each unit in its own
location.
The initial calculation uses material 2 to define the compositions of
materials 20, 30, and 40, since all are initially identical. Microscopic
cross sections computed for material 2 are used for each of the three
assigned depletion materials during the transport calculation and the
depletion calculation. After the first depletion calculation, materials
20, 30, and 40 will have different isotopic concentrations because of
different locations in the nonsymmetric transport model. At this time,
the number densities in each of these three materials are averaged and
used to update the concentration of representative material 2. A new set
of cell calculations will be performed for materials 1 and 2; this will
be followed by a transport calculation that uses the microscopic cross
sections for material 2 along with local nuclide number densities for
materials 20, 30, and 40 to calculate new and unique macroscopic cross
sections for each. The transport and subsequent depletion calculation
are then run. The iterative process will continue until all depletion
steps have been completed.
.. _fig3-1-11:
.. figure:: figs/TRITON/fig11.svg
:align: center
:width: 800
Example input with material assignments.
.. _fig3-1-12:
.. figure:: figs/TRITON/fig12.png
:align: center
:width: 500
Example 2D model plot of material assignments.
The use of assignments can make a considerable difference in run-time
performance with minimal sacrifice in accuracy. The above example ran
1.8 times faster using the assignment of three similar pins to one
initial specification. A larger BWR calculation, in which 41 pin
positions were depleted independently, was run in an assessment of the
accuracy of the method. Using this approach, the simplified
representation ran 20 depletion steps in 20% of the time required for
the explicitly modeled cells. :numref:`fig3-1-13` shows a comparison of the
eigenvalues using the simplified (with assignments) and explicit
(without assignments) models. Also shown on the figure is the percent
difference between the approximate and explicit models. For this model,
the error in *k*\ :sub:`eff` remains well below 0.05%.
Note that one can combine depletion mode control with material
assignments, as follows.
::
READ depletion 1 2 5 6
flux 3 4 end
assign 11 1 2 end
assign 12 3 4 end
assign 13 5 6 end
END depletion
.. _fig3-1-13:
.. figure:: figs/TRITON/fig13.png
:align: center
:width: 500
Eigenvalue comparison of simplified cross section processing example.
.. _3-1-3-3-5:
TIMETABLE block
^^^^^^^^^^^^^^^
In many depletion analyses, material properties can change due to
influences outside the depletion process (e.g., boron letdown in
pressurized water reactors [PWRs], the insertion or removal of poisons
during or between fuel cycles, or changes in temperatures of materials
with time). The *TIMETABLE* block has been provided to allow
modification of properties during a depletion calculation. Timetables
may be entered for any material or for select nuclides within a material
and allow changes in number densities or temperatures. Timetables may
also be entered to swap a material in and out of the geometry during
depletion.
The *TIMETABLE* block takes the following general format.
::
READ timetable
[time dependent specifications for a given material]
[time dependent specifications for a given material]
[time dependent specifications for a given material]
END timetable
Three different material specifications are allowed to modify
temperature, density, or swap materials.
Temperature timetable entries are specified in the format
::
temperature I t1 K1 t2 K2 t3 K3...tC KC end
where
*I* = material ID number;
*t*\ :sub:`i` = time (days) in calculation where temperature *K*\ :sub:`i` is set, *i*
= 1 to C;
*K*\ :sub:`i` = temperature (in K) of specified materials at time *t*\ :sub:`i`, *i* =
1 to C;
*C* = number of time steps.
Density entries have an analogous specification, with the addition of a couple of extra terms:
::
density I M N1 N2 N3 ... NM t1 D1 t2 D2 t3 D3...tC DC end
where
*I* = material ID number;
*M* = number of nuclides to which this change is applied;
*N*\ :sub:`i` = nuclide ID for the *i*\ :sub:`th` nuclide in the list, *i* = 1 to M;
*t*\ :sub:`j` = time (days) in calculation where density multiplier D\ :sub:`j`
is set, *i* = 1 to C;
*D*\ :sub:`j` = density multiplier (fractional change) of specified nuclides at
time *t*\ :sub:`j`, *i* = 1 to C;
*C* = number of time steps.
In both formats, time and data (temperature or density multiplier) must
be entered in pairs. Note that density changes may be applied to
specific nuclides, while for temperature the change must be applied to
all nuclides within the material simultaneously. If *M* (the number of
nuclides for which the density is to be modified) is specified as 0 and
no nuclide IDs are entered, then the timetable values are applied to all
nuclides in the material.
Note that timetable entries are specified at distinct times in the
calculation. These times are measured relative to the beginning of the
calculation and are continuous (as opposed to *BURNDATA* entries, which
give burn times or down times in increments per depletion interval). The
initial timetable entry should always begin at t=0 days. To allow for
time-dependent changes in properties, TRITON applies linear
interpolation between data pairs. To hold a parameter constant over a
time interval, that parameter should be specified at the same value at
both the beginning and the end of this time interval.
The application of timetable entries is best illustrated by example.
Consider the depletion scheme described by the *BURNDATA* block of
:numref:`fig3-1-7`, which contains three depletion intervals. Assume that the
moderator, material 3, has temperatures and boron concentrations that
vary over the three depletion intervals in the following manner:
+--------------+-------------------------+-----------------+-----+
| **Interval** | **Boron concentration** | **Temperature** | |
| | | | |
| | **(ppm)** | **(K)** | |
+==============+=========================+=================+=====+
| | **BOC** | **EOC** | |
+--------------+-------------------------+-----------------+-----+
| 1 | 1000 | 100 | 615 |
+--------------+-------------------------+-----------------+-----+
| 2 | 1250 | 130 | 685 |
+--------------+-------------------------+-----------------+-----+
| 3 | 980 | 100 | 610 |
+--------------+-------------------------+-----------------+-----+
.. _fig3-1-14:
.. figure:: figs/TRITON/fig14.png
:align: center
:width: 500
Example temperature and density *TIMETABLE* block input.
It is important to note that time-dependent changes to temperatures and
number densities are not applied continuously over the depletion
calculation but instead are applied only at the times at which cross
section processing and transport calculations are performed—that is, the
midpoint of the depletion subintervals. *The user must determine the
accurate depletion scheme specific to his or her application to
accurately model time-dependent changes in system properties.*
Density timetable specifications can be used to effectively exchange
compositions of a single material. One may construct a compound material
comprised of two distinct materials at their design densities; a
timetable specification can be used to set the density multiplier to 1.0
for the nuclides initially present and to use a multiplier of 0.0 for
all nuclides in materials that are not intended to be present at time
zero. The timetable can then affect the exchange by changing the
multipliers from 0 to 1, and from 1 to 0, at the time of the material
exchange. One must bear in mind that timetable processing within TRITON
performs linear interpolation between time points; if the exchange is
intended to occur at a specific moment in time, then the timetable
should be set up with the exchange occurring within a very short period.
Moreover, it is important to note that material exchanges for two
materials that have common nuclides are more difficult to model. For
example, a B\ :sub:`4`\ C absorber material and borated H\ :sub:`2`\ O
moderator material both contain boron nuclides in common. In order to
exchange the B\ :sub:`4`\ C absorber material and the borated
H\ :sub:`2`\ O moderator material, the carbon, oxygen, and hydrogen
density multipliers would be 0 or 1, but the boron density multipliers
would need to be derived from the boron concentrations in both
materials.
Material exchange timetables offer another option to users to exchange
one material with another material during depletion calculations.
Material exchange timetable has a similar format to temperature
timetables:
::
swap I1 I2 t1 S1 t2 S2 t3 S3...tC SC end
where
*I1* = first material ID
*I2* = second material ID
*t*\ :sub:`j` = time (days) in calculation where swap ID is set, *i* = 1 to C;
*S*\ :sub:`j` = swap value 0/1 at time *t*\ :sub:`j`, *i* = 1 to C;
*C* = number of time steps.
The first two entries in the timetable specify the material IDs for swap
materials. The remaining entries are entered in pairs: the first pair
value is a time value, the second pair value is either “0” or “1”. “0”
instructs TRITON to model the swap materials as defined in the nominal
model. “1” instructs TRITON to swap the materials (swap every *I1* for
every *I2* and swap every *I2* for every *I1*). The swap state persists
until the next time entry in the timetable. For the last time entry in
the timetable, the swap state persists for the duration of the
calculation. For the example:
::
read timetable
swap 5 6 0 0 100 1 200 0 end
read timetable
- Do not perform the material swap on the interval [0, 100],
- Perform the material swap on the interval [100, 200], and
- Return to the nominal state at time 200 days until the duration of
the calculation.
Depending on the *BURNDATA* specification, there may be one or more
depletion/decay steps between timetable entries. Moreover, for accurate
depletion modeling, material exchanges must not occur during a depletion
subinterval. If a material exchange occurs during a depletion interval,
TRITON will subdivide the depletion subinterval at the time of the
material exchange. Extending the example above, assume the *BURNDATA*
block is as follows:
::
read timetable
swap 5 6 0 0 100 1 200 0 end
read timetable
read burndata
power=40 burn=300 nlib=4 end
end burndata
Without the material exchange table, the depletion subintervals are [0,
75], [75, 150], [150, 225], and [225, 300]. With the material exchange
table, the subintervals are:
- [0, 75] – Swap value is 0
- [75, 100] – Swap value is 0
- [100-150] – Swap value is 1, i.e. materials 5 and 6 swap definitions
- [150-200] – Swap value is 1, i.e. materials 5 and 6 swap definitions
- [200-225] – Swap value is 0, materials 5 and 6 return to their
original definitions
- [225-300] – Swap value is 0
As a limitation of the material exchange timetable, if a depletion
material is removed from the geometry, the isotope concentrations at the
time of removal are stored in‑memory, and then reused upon re-entry into
the geometry. In other words, the depletion material does not undergo
radioactive decay for the period of time outside the problem geometry.
.. _3-1-3-3-6:
OPUS block
^^^^^^^^^^
The OPUS module of SCALE is fully documented in the OPUS chapter of the
SCALE manual. OPUS provides the ability to extract specific data from
ORIGEN output libraries, perform unit conversions, and generate plot
data for post-calculation analysis. In essence, OPUS is an ORIGEN
post-processor that provides data in the desired form for a desired
subset of nuclides. TRITON by default calls OPUS to extract nuclide
concentrations for selected nuclides for all depletion materials and for
the most important nuclides. TRITON provides the capability to specify
the full set of OPUS commands to tailor OPUS calculations to obtain
specific information. TRITON allows a stacked set of OPUS calculations
in order to retrieve selected data for selected nuclides.
The content of the *OPUS* block is based on standard OPUS input
parameters, as described in the OPUS chapter; the details of OPUS
control and use are not repeated here. However, additional input is
necessary to support TRITON operations with OPUS, because TRITON enables
additional capabilities beyond those provided for in standard OPUS
input. For example, OPUS is designed to process the output file from a
single ORIGEN calculation; because ORIGEN is a point depletion solver,
the output represents data from a single material. TRITON is typically
used to perform multiple depletion calculations at each depletion
step—one calculation for each material being depleted. Hence, multiple
OPUS calculations are needed to obtain results from multiple materials.
The OPUS calculations are performed automatically by TRITON but require
the user to specify the materials for which OPUS processing is desired.
Additionally, TRITON supports stacked OPUS cases within the *READ OPUS*
data block; hence, keywords are introduced to separate stacked cases.
There are two alternatives available to SCALE users that are
complimentary to the OPUS block within TRITON. First, standalone OPUS
case(s) can be used to post process the ORIGEN binary concentration file
(.f71 extension). This file is automatically saved in the output
directory with the file name ``${OUTBASENAME}.f71``. (e.g. if the input file
is reactor.inp, the concentration file is saved in the output directory
as reactor.f71) Second, the user may also open the concentration file
within Fulcrum to enable similar post-processing capabilities.
.. _3-1-3-3-6-1:
Selection of materials for OPUS processing
''''''''''''''''''''''''''''''''''''''''''
Beyond standard OPUS input keywords (see OPUS chapter), TRITON reads a
*matl=* keyword to allow specification of ID number(s) for the
material(s) in the problem for which outputs are desired. The
*matl=…end* input keyword accepts one or more materials from the
*DEPLETION* data block for which OPUS processing is desired. If omitted,
OPUS processing will be performed for all materials in the *DEPLETION*
block. For example, consider the following *DEPLETION* and *OPUS* data
blocks:
::
READ depletion 1 2 3 4 5 6 END depletion
READ opus
units=gram symnuc=u-234 u-235 u-236 u-238 pu-238 pu-239
pu-240 pu-241 pu-242 pu-243 np-237 end
time=year
END opus
In this example, OPUS processing will be performed for all depletion
materials, 1–6. Adding a subset of materials using the *matl=* keyword,
for example.
::
READ depletion 1 2 3 4 5 6 END depletion
READ opus
units=gram symnuc=u-234 u-235 u-236 u-238 pu-238 pu-239
pu-240 pu-241 pu-242 pu-243 np-237 end
time=year
matl=1 2 3 end
END opus
will result in OPUS calculations for materials 1, 2, and 3 only.
Although ORIGEN calculations are performed only for individual
materials, TRITON provides the capability of combining the results of
all or a subset of all depletion materials to get a multimaterial
average set of ORIGEN responses. TRITON provides two special ID numbers
for combining material results. Specification of material ID 0 will
return system-averaged results for the entire set of depletion materials
(typically, all fuel elements in a depletion model). Specification of
material –1 returns the average of only those materials with ID > 0
present in the *matl=* list. Again, this is best illustrated by example.
Specification of the data blocks
::
READ depletion 1 2 3 4 5 6 END depletion
READ opus
units=gram symnuc=u-234 u-235 u-236 u-238 pu-238 pu-239
pu-240 pu-241 pu-242 pu-243 np-237 end
time=year
matl=1 2 3 0 -1 end
END opus
will result in five OPUS calculations and five sets of results—one for
each of materials 1, 2, and 3, one for the average of materials 1–6 (due
to input of material ID 0), and one for the average of materials 1–3
(due to input of material ID –1).
.. _3-1-3-3-6-2:
Specification of stacked OPUS cases
'''''''''''''''''''''''''''''''''''
In a given calculation, multiple output units may be desired (e.g.,
grams, curies, and watts), or multiple time scales (e.g., seconds and
years), or a combination of these or other parameters. TRITON provides
the ability to stack inputs such that multiple cases may be run within a
single TRITON calculation. In order to stack cases, the keywords *new
case* are entered in the input stream. Any parameters following these
keywords are used to define a new OPUS case.
There is no limit on the number of stacked cases that may be input;
however, the *matl=* specification may be used only in the first case.
OPUS calculations are run for each of the materials in this list, for
all cases.
Consider a depletion calculation where gadolinium pins are present in
the assembly design. One may wish to determine the quantities of
gadolinium nuclides from the initial poison rods (tracked as a light
element by ORIGEN within TRITON) and from fission (tracked as a fission
product by ORIGEN). One may also need masses of selected actinides as
well as the total (α,n) reaction rate. :numref:`fig3-1-15` shows how the *new
case* keyword set is used to define unique OPUS calculations. In this
example, the *new case* keywords are shown in upper case and on a line
by themselves, but this has been done only for readability. The text may
be entered in lower case and on the same line as other keywords. Note,
however, that the *matl=* specification is given only in the first case.
All OPUS calculations will be performed for materials 1, 2, and 3 and
for the average of these three materials.
.. _fig3-1-15:
.. figure:: figs/TRITON/fig15.svg
:align: center
:width: 500
Example OPUS block input.
.. _3-1-3-4:
S/U analysis input
~~~~~~~~~~~~~~~~~~
An example input structure for the sensitivity and uncertainty (S/U)
analysis sequences is provided in :numref:`fig3-1-16`.
.. _fig3-1-16:
.. figure:: figs/TRITON/fig16.svg
:align: center
:width: 500
Structure of S/U analysis input.
The TRITON S/U sequences support the blocks of data highlighted in red:
the *SAMS, HTML, COVARIANCE, DEFINITIONS,* and *SYSTEMRESPONSES* data
blocks. These data blocks, along with the *KEEP_OUTPUT* block, may
appear only once, in any order, and must follow the *COMPOSITION* and
*CELLDATA* blocks and must precede the *MODEL* block. If specified, the
*HTML* and *COVARIANCE* blocks must follow the *SAMS* block.
These five blocks of data serve the same function as described in the
TSUNAMI-1D manual, and the information is not repeated here.
The *MODEL* block contains a full NEWT transport model input description
and is required for both S/U analysis sequences. *The* MODEL *block must
be the last block of data in the input file.* The *MODEL* block provides
the physical layout of the configuration for which the transport
calculation is to be performed, along with general control parameters.
The NEWT input is described fully in the NEWT chapter and is not
repeated here.
.. _3-1-3-5:
*ALIAS* block
~~~~~~~~~~~~~
The optional *ALIAS* block may be used to simplify model development
within TRITON by defining a set of material numbers that will be
inserted in place of the alias when that alias is used in subsequent
data blocks. Aliases function as variables for which a user-defined set
of materials are inserted; they are identified by a dollar character ($)
preceding a single-word alphanumeric label. The *ALIAS* block is used to
preprocess an input, creating a new, modified input deck with all alias
variable substitutions included. TRITON then processes the modified
input deck before proceeding with the calculation.
The use of an alias variable is best illustrated by a brief example.
Assume that the alias ``$fuel`` is defined as materials 1, 2, and 3, and
``$mod`` as materials 4, 5, and 6. (The input format for defining aliases is
described below.) The user wishes to create three identical sets of
materials and use them in three identical pin cell specifications. In
the *COMPOSITION* data block, specifications could be written in the
following form
::
h2o $mod den=0.6616 1.0 595 end
wtpt-boron $mod 0.6616 1 5000 100 655e-6 595 end
TRITON would create a modified input with the alias expanded as follows:
::
uo2 1 den=10.29 0.9322 920 92235 3.0 92238 97.0 end
uo2 2 den=10.29 0.9322 920 92235 3.0 92238 97.0 end
uo2 3 den=10.29 0.9322 920 92235 3.0 92238 97.0 end
h2o 4 den=0.6616 1.0 595 end
h2o 5 den=0.6616 1.0 595 end
h2o 6 den=0.6616 1.0 595 end
wtpt-boron 4 0.6616 1 5000 100 655e-6 595 end
wtpt-boron 5 0.6616 1 5000 100 655e-6 595 end
wtpt-boron 6 0.6616 1 5000 100 655e-6 595 end
Similarly, if the alias were used in the *CELLDATA* block as
::
latticecell squarepitch pitch=1.26 $mod fuelr=0.4095 $fuel end
then TRITON would expand the aliases to
::
latticecell squarepitch pitch=1.26 4 fuelr=0.4095 1 end
latticecell squarepitch pitch=1.26 5 fuelr=0.4095 2 end
latticecell squarepitch pitch=1.26 6 fuelr=0.4095 3 end
In a depletion calculation, one may wish to deplete a large number of
fuel rods independently because of different geometric locations in a
fuel assembly. Even though each fuel rod may have the same initial
composition, each must be specified as a unique material composition in
order to be depleted independently. Furthermore, multiple cell
specifications must all use unique material identifiers for each cell
component. Thus, if one desired to deplete 25 fuel materials in a
fuel/clad/moderator pin cell, one would need to set up material
composition definitions for 25 fuels, 25 moderators, and 25 clads. Then
one would need to provide 25 pin cell specifications. By using aliases,
one need only specify the material identifiers corresponding to each
alias and then provide only one material composition specification for
each alias type, and then one pin cell specification. TRITON will
automatically expand the aliases and create a revised input with all
materials and cell specifications explicitly defined.
.. note:: Note that
although this will simplify the pin cell input in the CELLDATA, 25 pin
cell calculations would still be required. The number of pin cell
calculations can be reduced by using the ASSIGN function described in
:ref:`3-1-3-3-4-3`.
The purpose of the *ALIAS* block is to define a set of alias variables
to be used in subsequent data blocks. The *ALIAS* block is optional, but
aliases may not be used in other blocks if an *ALIAS* block is not
present to define the aliases. An *ALIAS* block may contain as many
aliases as desired. Each alias specification consists of three parts:
the alias name, consisting of a dollar sign followed by up to 11
alphanumeric characters with no embedded spaces; the material number or
numbers; and an *end* keyword. Material numbers may be entered in any
order and may be separated by spaces or commas (or both). Material
numbers may also be separated by a dash (-), but this represents an
inclusive list. In other words, a material specification of 1-3 (or 1 -
3) indicates materials 1, 2, and 3. The example *ALIAS* block below
illustrates the various means for assigning a set of materials for an
alias definition.
::
read alias
$fueltype1 1 2 3 end
$fueltype2 4,5,6, 31-33 end
$clad1 21,22,23 end
$clad2 24 25 26 34-36 end
$mod1 11 - 13 end
$mod2 14-16, 37-39 end
end alias
The *ALIAS* block simply serves to assign material identifiers to
specific variables, and the variables are used in subsequent data
blocks. The same material identifier can be used in more than one alias
if desired. As indicated earlier, TRITON will preprocess any input deck
containing an *ALIAS* block and replace instances of alias variables
with the appropriate material identifiers. The following subsections
describe how aliases are implemented in TRITON’s various input blocks,
as the form of alias variable substitution is block dependent. Aliases
are processed only in these input blocks; aliases used in other blocks
will result in an error.
.. _3-1-3-5-1:
*COMPOSITION* block aliases
^^^^^^^^^^^^^^^^^^^^^^^^^^^
The *COMPOSITION* block uses aliases to create multiple copies of each
standard composition specification, replacing the alias variable with
each material identifier associated with the alias definition. For
example, consider the following alias definition in an *ALIAS* block:
::
read alias
$fuel 1 2 10 end
end alias
and the standard composition specification:
::
uo2 $fuel den=10.045 1 800 92235 2.5 92238 97.5 end
A modified TRITON input would be created with the standard composition
specification replaced by
::
uo2 1 den=10.045 1 800 92235 2.5 92238 97.5 end
uo2 2 den=10.045 1 800 92235 2.5 92238 97.5 end
uo2 10 den=10.045 1 800 92235 2.5 92238 97.5 end
.. _3-1-3-5-2:
*CELLDATA* block aliases
^^^^^^^^^^^^^^^^^^^^^^^^
*CELLDATA* block *latticecell* specifications typically contain more
than one material; therefore, multiple aliases are permitted in each
cell specification. However, this constrains the set of aliases used in
the cell specification to have the same number of material identifiers
associated with it.
Consider the *ALIAS* block:
::
read alias
$fuel 1-3 10 end
$clad 4,5,6,11 end
$mod 7 8-9 12 end
end alias
All three aliases contain four materials each. One could then create a
single cell specification that uses one or more of these alias
variables, such as
::
latticecell squarepitch pitch=1.26 $mod fuelr=0.41 $fuel cladr=0.50 $clad end
This would result in the following alias expansion by TRITON:
::
latticecell squarepitch pitch=1.26 7 fuelr=0.41 1 cladr=0.50 4 end
latticecell squarepitch pitch=1.26 8 fuelr=0.41 2 cladr=0.50 5 end
latticecell squarepitch pitch=1.26 9 fuelr=0.41 3 cladr=0.50 6 end
latticecell squarepitch pitch=1.26 12 fuelr=0.41 10 cladr=0.50 11 end
Material identifiers are substituted according to their position in the
alias definition (i.e., the first substitution will use the first
material associated with each alias, and the second expansion will use
the second material associated with each alias, etc.)
Material numbers should not be entered manually in a cell specification; for example,
::
latticecell triangpitch pitch=1.26 $mod fuelr=0.4095 1 end
TRITON would allow this to occur and would create a set of cell specifications as follows:
::
latticecell triangpitch pitch=1.26 2 fuelr=0.4095 1 end
latticecell triangpitch pitch=1.26 3 fuelr=0.4095 1 end
where $mod was defined as materials 2 and 3. However, SCALE does not
allow the same material identifier to occur in two different cell
specifications, and the fact that material 1 occurs in two different
cell specifications would result in TRITON ending with an error. *Note
that alias expansions for*\ **multiregion**\ *and*\ **doublehet**\ *cell
specifications are not supported. Also note that TRITON will not
copy*\ **centrmdata**\ *and*\ **moredata**\ *specifications that follow
a cell specification that uses an alias variable.*
.. _3-1-3-5-3:
*DEPLETION* block aliases
^^^^^^^^^^^^^^^^^^^^^^^^^
Aliases in the TRITON *DEPLETION* are simply replaced by the set of
materials associated with the alias. For example, the *ALIAS* block
::
read alias
$fuel 1 2 10 end
end alias
and DEPLETION block
::
read depletion 7 8 9 $fuel end depletion
would be expanded to
::
read depletion 7 8 9 1 2 10 end depletion
Aliases may be mixed with actual material numbers in the depletion
block, along with the flux and assign keywords. *However, the negative
sign—used to define the basis for power normalization—cannot precede an
alias definition.*
.. _3-1-3-5-4:
*TIMETABLE* block aliases
^^^^^^^^^^^^^^^^^^^^^^^^^
*TIMETABLE* block alias expansion is similar to that of the
*COMPOSITION* block: TRITON will create a new timetable entry for each
material associated with the alias used in the *TIMETABLE* definition.
For the *TIMETABLE* block below, using the alias *$allmod*, unique
timetables will be created for each material identifier associated with
this alias variable.
.. note:: Note that alias expansion
of **density** timetable entries is not yet supported.
::
read timetable
temperature $allmod
0.0 615
121.0 615
121.01 685
322.5 685
352.5 610
738.75 610 end
end timetable
.. _3-1-3-5-5:
*BRANCH* block aliases
^^^^^^^^^^^^^^^^^^^^^^
Aliases may be used within the *define* keyword definitions of the
*BRANCH* block. Aliases are simply replaced by the list of materials
associated with the alias, as is done for the *DEPLETIO*\ N block.
Hence,
::
read alias
$fuel 1 2 10 end
end alias
used with
::
read branch
define fuel $fuel end
md=0.75 tm=559 tf=880 sb=0.0 cr=0 end
tf=1600 end
end branch
would be expanded to
::
read branch
define fuel 1 2 10 end
md=0.75 tm=559 tf=880 sb=0.0 cr=0 end
tf=1600 end
end branch
.. _3-1-3-5-6:
NEWT *MATERIAL* block aliases
^^^^^^^^^^^^^^^^^^^^^^^^^^^^^
The *MATERIAL* block within the NEWT model section of a TRITON input can
also use aliases. As with *COMPOSITION* and *TIMETABLE* entries, TRITON
will create a new material specification for each material represented
by an alias. For the sample material block below, using the alias
*$fuel*, unique material block entries will be created for each material
associated with the alias variable.
::
read materials
mix=$fuel pn=1 com=“3.25 wo uo2 fuel” end
mix=21 pn=1 com=“zirc cladding” end
mix=31 pn=1 com=“water” end
end materials
If an alias were defined as
::
$fuel 10 11 12 end
then the *MATERIAL* block would be expanded to
::
read materials
mix=10 pn=1 com=“3.25 wo uo2 fuel” end
mix=11 pn=1 com=“3.25 wo uo2 fuel” end
mix=12 pn=1 com=“3.25 wo uo2 fuel” end
mix=21 pn=1 com=“zirc cladding” end
mix=31 pn=1 com=“water” end
end materials
.. _3-1-3-6:
*KEEP_OUTPUT* block
^^^^^^^^^^^^^^^^^^^
When performing depletion calculations for a number of different
materials, TRITON output can become quite voluminous. Often, much of
that output is not needed for calculations that seek only eigenvalues,
sources, or concentrations as a function of irradiation history. TRITON
provides the ability to trim output to only those portions for which
output is desired. Output produced directly by the TRITON module is
always provided and cannot be disabled, but output from any other code
in the sequence can be automatically removed from the output listing.
Retaining certain output is accomplished using the *KEEP_OUTPUT* data
block.
The *KEEP_OUTPUT* data block provides the ability to preserve only
selected outputs. The format of this data block is
::
read keep_output
module_1 module_1 ... module_i ... module_N
end keep_output
where ``module_i`` represents any valid module name from the list of modules
invoked by TRITON, as listed here:
xsproc xsdrn newt kenova kenovi couple origen
By default, the output from all these modules is retained with the
exception of XSProc COUPLE and ORIGEN. SAMS and OPUS output is always
retained.
.. _3-1-3-7:
TRITON control parameters
~~~~~~~~~~~~~~~~~~~~~~~~~
TRITON supports the following of control parameter options:
parm=
CHECK,
CENTRM, 2REGION, XSLEVEL=N,
WEIGHT, WEIGHT=N,
ADDNUX=N,
INFDCUTOFF=X,
CXM=N,
MAXDAYS=N
If an invalid control parameter option is specified, including
misspelled keywords, an error message will be generated and execution
terminated. TRITON also provides the ability to nest several control
parameter keywords together; to combine keywords (where appropriate), a
list may be entered, enclosed in parentheses, and separated by commas.
For example, to specify CHECK, 2REGION, and ADDNUX=1 at the same time,
input would begin with
::
=t-depl parm=(check, 2region,addnux=1)
The following subsections provide more detail on each of the control parameters listed above.
.. _3-1-3-7-1:
Check mode: *parm=check*
^^^^^^^^^^^^^^^^^^^^^^^^
Specification of *parm=check* will request that TRITON read all input
and ensure that no input errors are present, without running additional
calculations. In this mode, all input is set up as if a full calculation
will be run, but the sequence exits without any functional module
execution. The check mode is useful for debugging or obtaining processed
standard composition data, without actually running a calculation. It
can also be used to generate plot files for embedded NEWT and KENO
inputs for additional review and checking of input specifications. Of
course, some errors may be uncovered only by dynamically executing the
functional modules; hence, there are rare occasions where a *parm=check*
run will complete with no errors but will fail when run outside of check
mode as the problem begins to run.
.. _3-1-3-7-2:
Multigroup cross section processing options
^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^
The most common use of *parm=* sequence control is in the selection of
an alternate multigroup cross section processing mode.
By default, XSProc enables both the BONAMI and CENTRM modules for cross
section processing. BONAMI-only XSProc calculations can be performed
using the control parameter *parm=bonami*.
TRITON also supports the control parameter *parm=(xslevel=N)*. The
*xslevel* parameter option initializes various CENTRM options for the
XSProc calculations. The *xslevel* option is equivalent to initializing
all unit cell calculations with the following *centrmdata*
specifications:
::
parm=(xslevel=1):
centrmdata
npxs=5 nfst=0 nthr=3 nmf6=-1 alump=0.3 demin=0.125 pmc_omit=1 pmc_dilute=5.0e5
end centrmdata
parm=(xslevel=2):
centrmdata npxs=5 nfst=0 nthr=3 nmf6=-1 end centrmdata
parm=(xslevel=3):
centrmdata alump=0.3 demin=0.125 pmc_omit=1 pmc_dilute=5.0e5 end centrmdata
parm=(xslevel=4):
[no centrmdata statement]
The option *parm=(xslevel=4)* is equivalent to *parm=centrm*. The option
*parm=(xslevel=3)* is the default for depletion sequences and is
equivalent to *parm=centrm* but with some minor approximations to
decrease run time. The option *parm=(xslevel=2)* is equivalent to
*parm=2region* for all sequences.
Note that the *xslevel=1* and *xslevel=3* options have additional
specifications for keywords *alump*, *demin*, *pmc_omit*, and
*pmc_dilute*. These keywords are further discussed in the XSProc
chapter. The additional keyword specifications are used to decrease
run-time for the CENTRM and PMC calculations. Internal investigations
have shown that the approximations introduced by the additional keyword
specifications have minimal impact on solution accuracy for a wide range
of calculations. Therefore the additional keyword specifications are
used by default for depletion calculations, where several CENTRM and PMC
calculations are necessary. The additional keyword values are not used
by default for nondepletion calculations to be consistent with the SCALE
CSAS5 and CSAS6 criticality sequences.
The TSUNAMI-2D sequence also invokes the BONAMIST module, a modified
version of BONAMI to support sensitivity calculations. TSUNAMI-2D
calculations may use parm=centrm, 2region, bonami, or xslevel. However,
these cross section processing options will not utilize the
sensitivity-enabled version of BONAMI.
.. _3-1-3-7-3:
Creating a broad group library: *parm=weight, parm=(weight=N)*
^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^
Used in tandem with the TRITON T-NEWT sequence, the specification
*parm=weight* extends the sequence by setting up and executing the
MALOCS2 module to generate a weighted broad-group cross-library (AMPX
master format). The spectrum generated in the NEWT transport calculation
is used as the weighting function for the collapse. Additionally, the
broad-group library energy structure is defined by the NEWT *COLLASPE*
block.
The *parm=weight* option uses the problem-averaged flux spectrum for the
weighting function in the collapse. The problem may be a simple pin cell
or a full assembly. However, there may be cases where the flux in a
specific region or material is most appropriate for the spectral
collapse. TRITON allows identification of a specific material from which
the collapsing spectrum should be used. When specified in the form
*parm=(weight=N)*, the average flux determined for material N is used in
place of the total domain spectrum to perform the collapse.
TRITON sample problem 1 (:ref:`3-1-6-1`) provides an example of the use
of T-NEWT to produce a new broad-group library. Note that the
broad-group library produced in this calculation will reside in the
SCALE temporary working directory with the name *newxnlib* at the end of
the calculation. If the library will be needed for future calculations,
the user should use a shell script to copy the library back to a more
permanent location, and perhaps give it a more meaningful name. In
sample problem 1, the SCALE 252-group master library is collapsed to 56
energy groups.
The process for creating a broad-group master library is also supported
in the 2D depletion sequence T‑DEPL. When *parm=weight* or
*parm=(weight=N)* is specified in a depletion calculation, the input
cross section library must be one of the SCALE 238-group or 252-group
libraries, which will automatically be collapsed to the SCALE 49-group
or 56-group structure, respectively. An initial fine group calculation
is performed for the input configuration, and the flux from the solution
is used to create the broad group library. The initial calculation is
then repeated with the new broad group library, followed by the
remainder of the depletion calculation. *Note that for lattice physics
calculations, the NEWT* COLLAPSE *block will be based on the 49-group
(or 56-group) energy structure, not the fine group structure.*
It is important to note that the 252-group library contains intermediate
resonance parameters and other data that cannot be accurately collapsed
into 56-group data with the collapsing procedures available in MALOCS2.
These parameters are important for bonami-only cross section processing
calculations, i.e., *parm=bonami*. Therefore, the *parm=centrm* option
is recommended for follow-on application of the collapsed 56-group
collapsed library. The 238-group and 49-group libraries do not contain
intermediate resonance parameter data, and bonami-only processing is
available, provided that this cross section processing option and group
structure is suitable for the intended application.
.. _3-1-3-7-4:
Inclusion of additional nuclides for depletion: *parm=(addnux=N)*
^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^
For depletion calculations, it is important to add trace quantities (1 ×
10\ :sup:`–20` at/b-cm) of certain nuclides to the inventories of
depletion materials in order to accurately track the nuclides’ impact on
cross section processing and transport calculations as a function of
burnup. By default, TRITON automatically adds to all fuel materials
trace quantities of a set of nuclides that have been determined to be
important in the characterization of spent fuel. TRITON recognizes fuel
materials as any material containing quantities of heavy metals (Z > 89)
in the standard composition specification.
TRITON provides user control of the set of nuclides added to a fuel
material through the *parm=(addnux=N)* control parameter, where N is an
integer value. For N = 0, no nuclides are added, which is generally a
very poor approximation and should only be used when the ramifications
are fully understood. For N = 1, a bare minimum set of 15 nuclides
(actinides) are added; this will generate improved number density
estimates for actinides in low-burnup fuels but will not update cross
sections for fission products of primary importance. Again, use of this
option is discouraged unless it addresses special modeling needs. For
N = 2, the default setting for the TRITON depletion sequences, 95
nuclides are added. N = 3 and N = 4 add 231 and 388 nuclides,
respectively. Note that in previous versions of TRITON, N = 2 would add
64 nuclides. The set of 64 nuclides is still supported by specifiying
*parm=(addnux=-2)* in the input. The default in the SCALE 6.1 release
remains *parm=(addnux=2).* :numref:`tab3-1-3` through :numref:`tab3-1-7` list the set
of nuclides added in trace quantities for each value of *addnux*.
.. _tab3-1-3:
.. table:: Additional nuclides added in trace quantities for *parm=(addnux=1)*.
:align: center
+----------------------+----------------+----------------+----------------+
| | :sup:`234`\ U | :sup:`235`\ U | :sup:`236`\ U |
+----------------------+----------------+----------------+----------------+
| :sup:`238`\ U | :sup:`237`\ Np | :sup:`238`\ Pu | :sup:`239`\ Pu |
+----------------------+----------------+----------------+----------------+
| :sup:`240`\ Pu | :sup:`241`\ Pu | :sup:`242`\ Pu | :sup:`241`\ Am |
+----------------------+----------------+----------------+----------------+
| :sup:`242`\ Am | :sup:`243`\ Am | :sup:`242`\ Cm | :sup:`243`\ Cm |
+----------------------+----------------+----------------+----------------+
| \*15 nuclides total. | | | |
+----------------------+----------------+----------------+----------------+
.. _tab3-1-4:
.. table:: Additional nuclides added in trace quantities for *parm=(addnux= -2)*.
:align: center
+-----------------+-----------------+-----------------+-----------------+
| :sup:`1`\ H | :sup:`10`\ B | :sup:`11`\ B | |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`14`\ N | :sup:`16`\ O | :sup:`83`\ Kr | :sup:`93`\ Nb |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`94`\ Zr | :sup:`95`\ Mo | :sup:`99`\ Tc | :sup:`103`\ Rh |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`105`\ Rh | :sup:`106`\ Ru | :sup:`109`\ Ag | :sup:`126`\ Sn |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`135`\ I | :sup:`131`\ Xe | :sup:`135`\ Xe | :sup:`133`\ Cs |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`134`\ Cs | :sup:`135`\ Cs | :sup:`137`\ Cs | :sup:`143`\ Pr |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`144`\ Ce | :sup:`143`\ Nd | :sup:`145`\ Nd | :sup:`146`\ Nd |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`147`\ Nd | :sup:`147`\ Pm | :sup:`148`\ Pm | :sup:`149`\ Pm |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`148`\ Nd | :sup:`147`\ Sm | :sup:`149`\ Sm | :sup:`150`\ Sm |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`151`\ Sm | :sup:`152`\ Sm | :sup:`151`\ Eu | :sup:`153`\ Eu |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`154`\ Eu | :sup:`155`\ Eu | :sup:`152`\ Gd | :sup:`154`\ Gd |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`155`\ Gd | :sup:`156`\ Gd | :sup:`157`\ Gd | :sup:`158`\ Gd |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`160`\ Gd | :sup:`244`\ Cm | | |
+-----------------+-----------------+-----------------+-----------------+
| \*49 additional | | | |
| nuclides in | | | |
| addition to the | | | |
| 15 nuclides | | | |
| added in | | | |
| addnux=1, for a | | | |
| total of 64. | | | |
+-----------------+-----------------+-----------------+-----------------+
.. _tab3-1-5:
.. table:: Additional nuclides added in trace quantities for *parm=(addnux=2)*.
:align: center
+-----------------+-----------------+-----------------+-----------------+
| :sup:`91`\ Zr | :sup:`93`\ Zr | :sup:`95`\ Zr | :sup:`96`\ Zr |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`95`\ Nb | :sup:`97`\ Mo | :sup:`98`\ Mo | :sup:`99`\ Mo |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`100`\ Mo | :sup:`101`\ Ru | :sup:`102`\ Ru | :sup:`103`\ Ru |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`104`\ Ru | :sup:`105`\ Pd | :sup:`107`\ Pd | :sup:`108`\ Pd |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`113`\ Cd | :sup:`115`\ In | :sup:`127`\ I | :sup:`129`\ I |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`133`\ Xe | :sup:`139`\ La | :sup:`140`\ Ba | :sup:`141`\ Ce |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`142`\ Ce | :sup:`143`\ Ce | :sup:`141`\ Pr | :sup:`144`\ Nd |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`153`\ Sm | :sup:`156`\ Eu | :sup:`242m`\ Am | |
+-----------------+-----------------+-----------------+-----------------+
| \*31 additional | | |
| nuclides in | | |
| addition to the | | |
| 15 nuclides in | | |
| :numref:`tab3-1-3` and | | |
| 49 nuclides in | | |
| :numref:`tab3-1-4`, | | |
| for a total of | | |
| 95. | | |
+-----------------------------------+-----------------+-----------------+
.. _tab3-1-6:
.. table:: Additional nuclides added in trace quantities for *parm=(addnux=3)*.
:align: center
+-----------------+-----------------+-----------------+-----------------+
| :sup:`72`\ Ge | :sup:`73`\ Ge | :sup:`74`\ Ge | :sup:`76`\ Ge |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`75`\ As | :sup:`79`\ Br | :sup:`76`\ Se | :sup:`77`\ Se |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`78`\ Se | :sup:`80`\ Se | :sup:`82`\ Se | :sup:`81`\ Br |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`80`\ Kr | :sup:`82`\ Kr | :sup:`84`\ Kr | :sup:`85`\ Kr |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`86`\ Kr | :sup:`85`\ Rb | :sup:`86`\ Rb | :sup:`87`\ Rb |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`84`\ Sr | :sup:`86`\ Sr | :sup:`87`\ Sr | :sup:`88`\ Sr |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`89`\ Sr | :sup:`90`\ Sr | :sup:`89`\ Y | :sup:`90`\ Y |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`91`\ Y | :sup:`90`\ Zr | :sup:`92`\ Zr | :sup:`92`\ Mo |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`94`\ Mo | :sup:`96`\ Mo | :sup:`94`\ Nb | :sup:`96`\ Ru |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`98`\ Ru | :sup:`99`\ Ru | :sup:`100`\ Ru | :sup:`105`\ Ru |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`102`\ Pd | :sup:`104`\ Pd | :sup:`106`\ Pd | :sup:`110`\ Pd |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`107`\ Ag | :sup:`111`\ Ag | :sup:`106`\ Cd | :sup:`108`\ Cd |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`110`\ Cd | :sup:`111`\ Cd | :sup:`112`\ Cd | :sup:`114`\ Cd |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`115m`\ Cd | :sup:`116`\ Cd | :sup:`140`\ Ce | :sup:`113`\ In |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`140`\ La | :sup:`112`\ Sn | :sup:`114`\ Sn | :sup:`115`\ Sn |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`116`\ Sn | :sup:`117`\ Sn | :sup:`118`\ Sn | :sup:`119`\ Sn |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`120`\ Sn | :sup:`122`\ Sn | :sup:`123`\ Sn | :sup:`124`\ Sn |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`125`\ Sn | :sup:`121`\ Sb | :sup:`123`\ Sb | :sup:`124`\ Sb |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`125`\ Sb | :sup:`126`\ Sb | :sup:`120`\ Te | :sup:`122`\ Te |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`123`\ Te | :sup:`124`\ Te | :sup:`125`\ Te | :sup:`126`\ Te |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`127m`\ Te | :sup:`128`\ Te | :sup:`129m`\ Te | :sup:`130`\ Te |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`132`\ Te | :sup:`130`\ I | :sup:`131`\ I | :sup:`124`\ Xe |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`126`\ Xe | :sup:`128`\ Xe | :sup:`129`\ Xe | :sup:`130`\ Xe |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`132`\ Xe | :sup:`134`\ Xe | :sup:`136`\ Xe | :sup:`134`\ Ba |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`135`\ Ba | :sup:`136`\ Ba | :sup:`137`\ Ba | :sup:`138`\ Ba |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`136`\ Cs | :sup:`142`\ Pr | :sup:`142`\ Nd | :sup:`150`\ Nd |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`151`\ Pm | :sup:`144`\ Sm | :sup:`148`\ Sm | :sup:`154`\ Sm |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`152`\ Eu | :sup:`157`\ Eu | :sup:`232`\ U | :sup:`233`\ U |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`159`\ Tb | :sup:`160`\ Tb | :sup:`160`\ Dy | :sup:`161`\ Dy |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`162`\ Dy | :sup:`163`\ Dy | :sup:`164`\ Dy | :sup:`165`\ Ho |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`166`\ Er | :sup:`167`\ Er | :sup:`175`\ Lu | :sup:`176`\ Lu |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`181`\ Ta | :sup:`182`\ W | :sup:`183`\ W | :sup:`184`\ W |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`186`\ W | :sup:`185`\ Re | :sup:`187`\ Re | :sup:`197`\ Au |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`231`\ Pa | :sup:`233`\ Pa | :sup:`230`\ Th | :sup:`232`\ Th |
+-----------------+-----------------+-----------------+-----------------+
| \*136 | | |
| additional | | |
| nuclides in | | |
| addition to the | | |
| 15 nuclides in | | |
| :numref:`tab3-1-3`, 49 | | |
| nuclides in | | |
| :numref:`tab3-1-4`, | | |
| and 31 nuclides | | |
| in Table 3.1.5, | | |
| for a total of | | |
| 231. | | |
+-----------------------------------+-----------------+-----------------+
.. _tab3-1-7:
.. table:: Additional nuclides added in trace quantities for *parm=(addnux=4)* (continued in the following table).
:align: center
+-----------------+-----------------+-----------------+-----------------+
| :sup:`2`\ H | :sup:`3`\ H | :sup:`3`\ He | :sup:`4`\ He |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`6`\ Li | :sup:`7`\ Li | :sup:`7`\ Be | :sup:`9`\ Be |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`15`\ N | :sup:`17`\ O | :sup:`19`\ F | :sup:`23`\ Na |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`24`\ Mg | :sup:`25`\ Mg | :sup:`26`\ Mg | :sup:`27`\ Al |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`28`\ Si | :sup:`29`\ Si | :sup:`30`\ Si | :sup:`31`\ P |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`32`\ S | :sup:`33`\ S | :sup:`34`\ S | :sup:`36`\ S |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`35`\ Cl | :sup:`37`\ Cl | :sup:`36`\ Ar | :sup:`38`\ Ar |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`40`\ Ar | :sup:`39`\ K | :sup:`40`\ K | :sup:`41`\ K |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`40`\ Ca | :sup:`42`\ Ca | :sup:`43`\ Ca | :sup:`44`\ Ca |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`46`\ Ca | :sup:`48`\ Ca | :sup:`45`\ Sc | :sup:`46`\ Ti |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`47`\ Ti | :sup:`48`\ Ti | :sup:`49`\ Ti | :sup:`50`\ Ti |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`50`\ Cr | :sup:`52`\ Cr | :sup:`53`\ Cr | :sup:`54`\ Cr |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`55`\ Mn | :sup:`54`\ Fe | :sup:`56`\ Fe | :sup:`57`\ Fe |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`58`\ Fe | :sup:`58`\ Co | :sup:`58m`\ Co | :sup:`59`\ Co |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`58`\ Ni | :sup:`59`\ Ni | :sup:`60`\ Ni | :sup:`61`\ Ni |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`62`\ Ni | :sup:`64`\ Ni | :sup:`63`\ Cu | :sup:`65`\ Cu |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`70`\ Ge | :sup:`69`\ Ga | :sup:`71`\ Ga | :sup:`74`\ As |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`74`\ Se | :sup:`79`\ Se | :sup:`78`\ Kr | :sup:`110m`\ Ag |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`113`\ Sn | :sup:`123`\ Xe | :sup:`130`\ Ba | :sup:`132`\ Ba |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`133`\ Ba | :sup:`136`\ Ce | :sup:`138`\ Ce | :sup:`139`\ Ce |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`138`\ La | :sup:`148m`\ Pm | :sup:`153`\ Gd | :sup:`156`\ Dy |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`158`\ Dy | :sup:`166m`\ Ho | :sup:`162`\ Er | :sup:`164`\ Er |
+-----------------+-----------------+-----------------+-----------------+
.. table::
:align: center
+-----------------+-----------------+-----------------+-----------------+
| :sup:`168`\ Er | :sup:`170`\ Er | :sup:`174`\ Hf | :sup:`176`\ Hf |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`177`\ Hf | :sup:`178`\ Hf | :sup:`179`\ Hf | :sup:`180`\ Hf |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`182`\ Ta | :sup:`191`\ Ir | :sup:`193`\ Ir | :sup:`196`\ Hg |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`198`\ Hg | :sup:`199`\ Hg | :sup:`200`\ Hg | :sup:`201`\ Hg |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`202`\ Hg | :sup:`204`\ Hg | :sup:`204`\ Pb | :sup:`206`\ Pb |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`207`\ Pb | :sup:`208`\ Pb | :sup:`209`\ Bi | :sup:`223`\ Ra |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`224`\ Ra | :sup:`225`\ Ra | :sup:`225`\ Ac | :sup:`226`\ Ac |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`227`\ Ac | :sup:`226`\ Ra | :sup:`227`\ Th | :sup:`228`\ Th |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`229`\ Th | :sup:`233`\ Th | :sup:`234`\ Th | :sup:`232`\ Pa |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`235`\ Np | :sup:`236`\ Np | :sup:`238`\ Np | :sup:`239`\ Np |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`237`\ U | :sup:`239`\ U | :sup:`240`\ U | :sup:`241`\ U |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`236`\ Pu | :sup:`237`\ Pu | :sup:`243`\ Pu | :sup:`244`\ Pu |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`246`\ Pu | :sup:`244`\ Am | :sup:`244m`\ Am | :sup:`241`\ Cm |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`245`\ Cm | :sup:`246`\ Cm | :sup:`247`\ Cm | :sup:`248`\ Cm |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`249`\ Cm | :sup:`250`\ Cm | :sup:`249`\ Bk | :sup:`250`\ Bk |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`249`\ Cf | :sup:`250`\ Cf | :sup:`251`\ Cf | :sup:`252`\ Cf |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`253`\ Cf | :sup:`254`\ Cf | :sup:`253`\ Es | :sup:`254`\ Es |
+-----------------+-----------------+-----------------+-----------------+
| :sup:`255`\ Es | | | |
+-----------------+-----------------+-----------------+-----------------+
| \*158 | | |
| additional | | |
| nuclides in | | |
| addition to the | | |
| 15 nuclides in | | |
| :numref:`tab3-1-3`, 49 | | |
| nuclides in | | |
| :numref:`tab3-1-4`, 30 | | |
| nuclides in | | |
| :numref:`tab3-1-5`, | | |
| and 136 | | |
| nuclides in | | |
| :numref:`tab3-1-6`, | | |
| for a total of | | |
| 388. | | |
+-----------------------------------+-----------------+-----------------+
.. _3-1-3-7-5:
Few-group reaction cross section calculation control for continuous energy depletion: *parm=(cxm=N)*
^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^
In continuous energy depletion calculations, few group reaction cross
sections are computed by KENO directly rather than using a
post-processing approach that TRITON uses for multigroup mode. In
addition to these region averaged multigroup reaction cross sections,
KENO also provides problem-dependent region-averaged multigroup fluxes
to TRITON that will be used by COUPLE to generate one-group cross
section library for each depletion material.
Option *parm=(cxm=N)* is used to setup continuous-energy depletion
calculation with different modes of calculation, which tells KENO the
details of the tallying process for the reaction cross sections and
mixture fluxes. Available calculations modes and their descriptions are
presented in :numref:`tab3-1-8`.
.. _tab3-1-8:
.. list-table:: cxm values and their descriptions.
:align: center
* - .. image:: figs/TRITON/tab8.svg
:width: 800
..
:sup:`1` Energy group structure in KENO and associated number of
energy groups, NGP, should be consistent with those from the ORIGEN
library used in the problem.
:sup:`2` cxm=2 is the default mode for reaction cross section
calculations.
.. _3-1-3-7-6:
Infinite dilution cutoff control: *parm=(infdcutoff=X)*
^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^
The addition of nuclides to depletion materials as described in the
previous section can lead to increased run-times for CENTRM-based XSProc
calculations. However, many nuclides (e.g., low-density nuclides) are
effectively infinitely dilute and can be treated as such and omitted
from the expensive point-wise cross section collapse operation. For the
option *parm=(infdcutoff=sigma0)* sequence option, XSProc will compute
an effective background microscopic cross section for each nuclide. If
the computed background cross section is greater than the cutoff value
*sigma0*, recommended as 5 × 10\ :sup:`5` \ barns, then the nuclide is
considered infinitely dilute and the infinitely dilute multigroup cross
section is utilized from the cross section library.
In general, a *sigma0* cutoff value of 5 × 10\ :sup:`5` barns will be
acceptable for most applications. However, TRITON and the centrmdata
card in the *CELLDATA* block provide a means for the user to control the
cutoff value. The cutoff value may be assigned in either of two ways. A
single global value may be assigned to all cells using the TRITON
*parm=* specifier with the keyword *infdcutoff*, for example,
*parm=(infdcutoff=1e10)*. Addition of the specifier with a value of
1 × 10\ :sup:`10` will set the cutoff value to 1 × 10\ :sup:`10` for all
cells in the problem, which is generally appropriate for most
calculations. However, a provision is made to specify a unique cutoff
value to each cell using the *pmc_dilute* keyword in a *centrmdata*
specification. An example of this is shown in the description of
*parm=xslevel* in :ref:`3-1-3-7-2`.
The default value of *sigma0* depends on the sequence and cross section
processing option. For nondepletion sequences that use *parm=centrm*,
the default is 0. The default value of 0 instructs PMC to include all
nuclides for PMC processing. For depletion sequences that use
*parm=centrm* or for any sequence that uses *parm=2region*, the default
value is 5 × 10\ :sup:`5` barns.
.. _3-1-3-7-7:
Override of the maximum number of days per depletion subinterval: *PARM=(MAXDAYS=N)*
^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^
TRITON is set to limit ORIGEN time intervals to no more than 40 days to
avoid potential numerical error that would be introduced if depletion
were performed over a long time interval. For depletion subintervals of
more than 400 days (10 time intervals of 40 days), TRITON will
automatically increase the number of depletion subintervals in a
depletion interval. The depletion subinterval is based on a rule of
thumb for ORIGEN depletion. However, the rule breaks down when burning
at very low powers for extended time intervals. Thus, TRITON allows the
user to override the default behavior by specifying a new value for the
maximum number of days per ORIGEN time interval. A 100-day limit per
ORIGEN time interval may be set using *parm=(maxdays=100)*. In
overriding the default behavior, the user must be aware of any potential
errors introduced in the approximation.
.. _3-1-4:
Output Files Ccreated by TRITON
-------------------------------
TRITON produces a variety of output files that may be of use in related
calculations. Of those files, only certain files are copied back to the
return directory: the TRITON output file (.out); plot files generated by
NEWT, KENO, or OPUS (.plt); SAMS sensitivity data files (.sdf), in the
case of an S/U calculation; ORIGEN binary concentration files (.f71) and
HTML-formatted output (.html), where available. The TRITON output file
is a concatenated listing of outputs from TRITON and all modules for
which output is kept. Other files of potential interest are not copied,
and the user should be aware of these files and their names so that they
may be retrieved using a SHELL script after TRITON execution is
complete. The following subsections list those files and their purposes.
.. _3-1-4-1:
Standard composition restart files
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
At the end of all depletion calculations, standard composition files are
automatically produced for each material, listing the nuclides and
number densities of the materials at the time the transport calculation
(i.e., XSDRN, NEWT, KENO) is performed. Only nuclides for which cross
section data are available in the master cross section library are saved
in these files. Files are saved using the file naming convention
StdCmpMix\ *NNNNN*, where *NNNNN* is the material identifier. The file
contains compositions at the final time of the calculation. Additional
files are saved with the file naming convention
StdCmpMix\ *NNNNN_MMMMM*, where *MMMMM* is an index to a particular time
step in the depletion calculation. For example, if a calculation is
completed with materials 1 and 40 for two depletion steps, then the
following files will be created in the temporary working directory.
.. code:: none
StdCmpMix00001_00000 (t=0)
StdCmpMix00001_00001 (midpoint of 1st depletion step)
StdCmpMix00001_00002 (midpoint of 2nd depletion step)
StdCmpMix00001_00003 (final compositions, end of 2nd depletion step)
StdCmpMix00001 (same as StdCmpMix00001_00003)
StdCmpMix00040_00000 (t=0)
StdCmpMix00040_00001 (midpoint of 1st depletion step)
StdCmpMix00040_00002 (midpoint of 2nd depletion step)
StdCmpMix00040_00003 (final compositions, end of 2nd depletion step)
StdCmpMix00040 (same as StdCmpMix00040_00003)
The contents of these files will be a standard composition description
of each material by atomic contents—that is, SCALE standard nuclide IDs
(e.g., U-235), number density, and temperature (using the temperature of
the original material). Using SCALE’s external file read capability,
these outputs may be automatically included in a follow-on calculation
that relies on depleted/decayed number densities. TRITON sample problem
7 (:ref:`3-1-6-7`) provides an example of the use of these restart files.
.. important:: Standard composition restart files should be used only for follow-on
criticality or shielding calculations.
.. _3-1-4-2:
Lattice physics parameters
~~~~~~~~~~~~~~~~~~~~~~~~~~
During T-DEPL depletion calculations that use branch states and
homogenization, a database of few-group cross sections is saved for each
branch state and at each depletion step containing homogenized cross
section data and other lattice physics parameters (e.g., discontinuity
factors, pin power peaking factors, diffusion coefficients, etc.). The
*xfile016* file is intended for post-processing, to be read and written
in the desired format for subsequent nodal diffusion core simulator
calculations. The *xfile016* file is a binary-formatted file, which is
described in detail in Appendix A of TRITON. An auxiliary text-formatted
database file (*txtfile16*) is also created that contains the same data
as the binary-formatted file.
.. _3-1-4-3:
ORIGEN binary library files
~~~~~~~~~~~~~~~~~~~~~~~~~~~
During depletion calculations, ORIGEN binary library files are created
to archive cross sections for each depletion material at each depletion
subinterval. These files can be used in future depletion calculations in
ORIGEN, ORIGAMI, and ARP. For each depletion material, the ORIGEN binary
library file is named *ft33f001.mix*\ NNNN, where *NNNN* is the material
number for each depleted material. Additionally, the combined cross
section file is saved with the name *ft33f001.cmbined*. Here is an
example of a script to retrieve cross section files.
POSIX:
::
=shell
cp ft33f001.mix0001 ${OUTDIR}/pwr_mix01.arp
cp ft33f001.cmbined ${OUTDIR}/pwr_asmb.arp
end
Windows:
::
=shell
copy ft33f001.mix0001 %OUTDIR%\pwr_mix01.arp
copy ft33f001.cmbined %OUTDIR%\pwr_asmb.arp
end
.. _3-1-4-4:
ORIGEN binary concentration file (.f71)
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
During depletion calculations, TRITON creates the ORIGEN binary
concentration file (.f71). This file is created in the temporary
directory as *ft71f001* and is copied back at the end of the SCALE
calculation to the return directory with the name ${OUTBASENAME}.f71.
TRITON archives computed concentrations for each depletion material at
the beginning and end of each depletion subinterval or decay interval.
These files include concentrations and also decay heat term, photon and
neutron data, and other quantities or interest computed by ORIGEN. These
data may be post-processed by the OPUS module.
The .f71 file contains concentrations for each individual material, and
it also contains the combined concentrations of the individual material
results (i.e., the net response for the entire system). The TRITON
output contains an index of the contents of this file (see
:ref:`3-1-5-4-5`).
.. _3-1-5:
Output Description
------------------
This section contains a brief description and explanation of TRITON
output. As with any SCALE module, TRITON output begins with the SCALE
header, the job information, the input file, and the program
verification information. These outputs are common to all SCALE modules.
Likewise, all SCALE calculations report a run-time summary at the end of
the output file.
.. _3-1-5-1:
Control parameter edit
~~~~~~~~~~~~~~~~~~~~~~
When TRITON control parameters are specified using the parm= command
(see :ref:`3-1-3-7`), all specified parameters are echoed following the
above output, with an explanation of the meaning of the parameter, as
shown below. If no parameters are specified, no edit is provided.
.. code:: none
The following TRITON control parameters were requested:
WEIGHT - Weighted collapsed master library
option selected for t-newt calculation, based
on system-averaged flux.
ADDNUX - specifies the set of additional nuclides added
in trace quantities for depletion
calculations. Set 1 was selected.
See TRITON manual for more information.
.. _3-1-5-2:
T-XSEC output
~~~~~~~~~~~~~
The T-XSEC sequence performs only cross section processing functions.
The XSProc output is written to the output file as the calculation
proceeds.
.. _3-1-5-3:
T-NEWT and T-XSDRN output
~~~~~~~~~~~~~~~~~~~~~~~~~
By default, the T-NEWT and T-XSDRN outputs include only the NEWT and
XSDRN output respectively. The XSProc output can be included by using
the *KEEP_OUTPUT* block (see :ref:`3-1-3-6`).
.. _3-1-5-4:
Depletion sequence output
~~~~~~~~~~~~~~~~~~~~~~~~~
The output of TRITON depletion sequences contains several depletion
edits. The edits are described in the following subsections. These
output edits are written to the output file in the order in which they
are computed during the calculation.
.. _3-1-5-4-1:
Burnup history summary (all depletion sequences)
^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^
TRITON generates the burnup history summary table after processing the
*BURNDATA* block. An example of this table is as follows.
.. list-table::
:align: center
* - .. image:: figs/TRITON/tab3-1-5-4-1.svg
:width: 700
This table shows the results of a burnup history using one depletion
interval with seven depletion subintervals. Column 1 is the cumulative
depletion subinterval number. Column 2 is the depletion interval number,
and column 3 is the depletion subinterval number within the current
depletion interval. Columns 4–6 echo the specific power, depletion
interval, and decay interval specified in the *BURNDATA* block. The
final column shows the cumulative burnup at the midpoint of each
depletion subinterval.
.. _3-1-5-4-2:
Embedded transport model output
^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^
The output from the initial transport calculation follows the burnup
history edit. The output edits for NEWT, XSDRN, KENO-V.a, and KENO-VI
are described in their respective manual sections.
.. _3-1-5-4-3:
System mass balance table
^^^^^^^^^^^^^^^^^^^^^^^^^
After the initial transport calculation output, a summary of system mass
information is printed, an example of which is provided as follows.
.. list-table::
:align: center
* - .. image:: figs/TRITON/tab3-1-5-4-3.svg
:width: 800
This table provides mass and density data for each material used in the
transport model. Column 1 provides the material identifier, and columns
5 and 6 provide the material density and material heavy metal density,
respectively, in units of grams per cubic centimeter. Heavy metal mass
is determined from masses of all nuclides with an atomic number greater
than 89. The final column provides the depletion mode for each material
(see :ref:`3-1-3-3-4-2`). Column 2 provides the “prenormalized” heavy
metal mass of each material. The units for this mass value depend on the
transport model. For 2D *xy* NEWT models, the units are grams per
centimeter since there is no *z-* dimension in the model. Similarly, the
units are grams per centimeter for 1D cylinder XSDRN models, grams per
square centimeter for 1D slab XSDRN models, and grams for 1D spherical
XSDRN models and 3D KENO models. The total prenormalized heavy metal
mass is printed in the final row of the table as well as in the table
header (highlighted in red). The heavy metal mass is normalized such
that a total system mass of 1 MTHM is present. The volume scaling factor
used to normalize the system mass is also printed in the table banner
(highlighted in red). The units of the volume scaling factor depend on
the transport model. Column 3 prints the normalized material heavy metal
mass in units of grams, which is equal to the prenormalized material
heavy metal mass in column 2 multiplied by the volume scaling factor in
the table header. The total normalized mass is printed in the final row
and also in the table header (highlighted in red). The fourth column
shows the fractional heavy metal mass of all materials, which is equal
to the normalized heavy metal mass in column 3, divided by the total
normalized system heavy metal mass in the table header.
.. _3-1-5-4-4:
Power balance tables
^^^^^^^^^^^^^^^^^^^^
As the TRITON calculation proceeds, the results of the cross section
processing and transport calculations are used to calculate fluxes and
powers in each material. The output segment listed below shows the
results for the first calculation based on the initial material
compositions. The total power (column 2) represents the
material-specific power in units of MW/MTHM of initial **system** mass.
The fractional power (column 3) is equal to the total power for a
material divided by the total system power (highlighted in red). The
mixture power (column 4) represents material-specific power in units of
MW/MTHM of initial **material** mass. The mixture power is equal to the
total power of the material divided by the fractional heavy metal mass
of the material, which is provided in the system mass balance table
(:ref:`3-1-5-4-3`). If the material does not contain heavy metal, then
“N/A” is printed in the mixture power column. Columns 5 and 6 show the
material thermal and total flux values, respectively, in units of
neutrons/cm\ :sup:`2`-sec. The thermal flux is determined by integrating
multigroup flux values for energy groups below 0.625 eV. If the specific
power is normalized to the total system power, the summation of the
material powers in column 1 should match the input specification in the
*BURNDATA* block (in the example given here, 20.86 MW/MTHM).
.. list-table::
:align: center
* - .. image:: figs/TRITON/tab3-1-5-4-4.svg
:width: 800
The form of the output changes if the specific power is normalized to
the power to one or more specific materials. For the case above, if
depletion was performed with input power normalized to material 7, the
power output table would have the following form.
.. list-table::
:align: center
* - .. image:: figs/TRITON/tab3-1-5-4-4(2).svg
:width: 800
In this example, material 7 has the input-specified power (20.86
MW/MTHM, highlighted in red), and the power in the remainder of the
model materials is normalized according to this basis material.
.. _3-1-5-4-5:
ORIGEN binary concentration file listing
''''''''''''''''''''''''''''''''''''''''
After all depletion calculations are completed, TRITON creates an ORIGEN
binary concentration file (.f71) with isotopic concentrations for each
depletion material. The order and content of the .f71 file is provided
in the TRITON output. An example of this edit is shown below. For each
depletion material, the output gives the location in the file, the
ORIGEN time interval number, the depletion interval time in days, the
cumulative time in years, and a title. After all materials are added to
the library, the system average of all libraries is computed and added
to the library. In this case, with only one depletion material, the
system average would be the same as the single material.
.. code:: none
*** Depletion calculations completed. Processing ORIGEN libraries. ***
49 time dumps found on this set of libraries.
File ft71f001 contains origen/opus-formatted binary data for 49 time dumps
from each of 1 depletion materials, plus a final set for the sum of all
depletion materials. Isotopic data locations are listed according to the following table.
(Note that there are two data records present for each time step.)
Position Time Step Cycle Time (d) Cumulative Time (y) Case Name
1 1 0.0000E+00 0.0000E+00 Pass no. 1 First depletion calculation, mix no. 1
2 2 3.8273E+01 1.0479E-01
3 3 7.6545E+01 2.0957E-01
4 4 1.1482E+02 3.1436E-01
...
12 12 4.2100E+02 1.1526E+00
13 13 1.1111E+00 1.1557E+00 Pass no. 1 Decay calculation, mix no. 1
14 14 3.3333E+00 1.1618E+00
15 15 1.0000E+01 1.1800E+00
16 16 3.0000E+01 1.2348E+00
17 17 3.8273E+01 1.3396E+00 Pass no. 2 First depletion calculation, mix no. 1
...
27 27 4.2100E+02 2.3874E+00
28 28 1.1111E+00 2.3904E+00 Pass no. 2 Decay calculation, mix no. 1
29 29 3.3333E+00 2.3965E+00
30 30 1.0000E+01 2.4148E+00
31 31 3.0000E+01 2.4695E+00
32 32 3.8273E+01 2.5743E+00 Pass no. 3 First depletion calculation, mix no. 1
33 33 7.6545E+01 2.6791E+00
...
42 42 4.2100E+02 3.6222E+00
43 43 2.5050E+00 3.6290E+00 Pass no. 3 Decay calculation, mix no. 1
...
50 1 0.0000E+00 0.0000E+00 Weighted sum of concentrations of all depleted mat
51 2 3.8273E+01 1.0479E-01
...
96 47 2.0292E+02 4.1777E+00
97 48 6.0875E+02 5.2888E+00
98 49 1.8262E+03 8.6222E+00
---------------------------------------------------------------------------------------------
The requested OPUS output edits follow the .f71 file summary edit.
.. _3-1-5-5:
TSUNAMI-2D output
~~~~~~~~~~~~~~~~~
Output from the TSUNAMI-2D and TSUNAMI-2DC sequence is similar to that
of the T-NEWT sequence, with the addition of the SAMS module output. If
the user requests sensitivity and uncertainty analysis for
non-\ *k*\ :sub:`eff` responses via the *DEFINITIONS* and *SYSTEMRESPONSES*
blocks, the TRITON output will edit out the computed values of these
responses. An example of this is shown here for sample problem 9 in
:ref:`3-1-6-9`.
::
Computed Response Ratio value for 'U235-abs' is: 1.428612E+01
Computed Response Ratio value for 'U235-fis' is: 1.155715E+01
Computed Response Ratio value for 'U238-abs' is: 3.036017E-01
Computed Response Ratio value for 'U238-fis' is: 3.393816E-02
In addition to the text-formatted output, TRITON generates
HTML-formatted output for TSUNAMI-2D calculations. The HTML output is
fully described in the SAMS chapter.
.. _3-1-6:
TRITON Sample Cases
-------------------
This section provides descriptions of the 13 TRITON sample problems
included with SCALE. Note that all of these problems (along with all
other SCALE sample problems) are typically executed in the initial SCALE
installation to test the performance of various codes and options, for
validation of the installation process. Because of the number of
problems that are executed, these sample problems are adjusted to run as
fast as possible so that all test problems may be completed in
relatively short order. To accomplish this, crude modeling
approximations (reduced convergence, few histories, simplified cross
section processing, low-order quadrature and scattering approximations,
coarse computational grids, reduced numbers of libraries per depletion
cycle, etc.) may be used. Hence, although these problems provide
guidance in setting up and executing TRITON problems, it is generally a
good idea to review all control settings to ensure sufficient accuracy
in one’s own calculations.
Additional TRITON input files for several reactor types can be generated
with the SCALE/ORIGEN Library Generator (SLIG). The SLIB documentation
is available as Appendix B of the ORIGEN chapter.
.. _3-1-6-1:
TRITON sample problem 1: triton1.inp
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
Sample problem triton1.inp is an example of a T-NEWT transport
calculation sequence. Input begins (as with all SCALE sequences) with a
title card and cross section library specification; this calculation is
performed using the 252-group ENDF/B-7.1 library. After the library
specification, three materials are defined in the composition block,
followed by a cell specification and the NEWT transport model.
This example includes an axial height of 37.1 cm and will therefore do a
buckled calculation based on this height. The geometric model consists
of a simple pin cell, with cylindrical fuel and clad regions inside a
square moderator region, with a 6 × 6 base grid. The NEWT *BOUNDS* block
specifies that periodic boundary conditions are used for this model.
This simple problem also demonstrates the use of TRITON’s automatic
cross section collapse capability—\ *parm=weight*. For *T-NEWT*
calculations, TRITON uses the NEWT *COLLAPSE* block to define the
broad-group energy structure. For this sample problem, the cross
sections are collapsed to a 56‑group format. The new broad-group library
will be identified as filename *newxnlib* in the temporary working
directory, which can be used in follow-up SCALE calculations.
..
::
' THIS SAMPLE PROBLEM TEST THE FOLLOWING:
' ** t-newt sequence
' ** v7-252 group library
' ** centrm cross section processing (default for t-newt calculations)
' ** parm=weight option for the t-newt sequence, which uses the NEWT collapse block to specify a 252 -> 56 group collapse.
' ** latticecell cross section processing option
=t-newt parm=weight
Buckled pin-cell transport calculation
v7-252
read comp
u-234 1 0 6.74213e-6 296.15 end
u-235 1 0 7.65322e-4 296.15 end
u-236 1 0 3.68820e-6 296.15 end
u-238 1 0 2.20912e-2 296.15 end
o 1 0 4.57338e-2 296.15 end
b-10 1 0 3.64042e-9 296.15 end
b-11 1 0 1.46531e-8 296.15 end
cr 25 0 6.67242e-5 296.15 end
fe 25 0 1.25922e-4 296.15 end
sn 25 0 4.17642e-4 296.15 end
o 25 0 2.63724e-4 296.15 end
zr 25 0 3.78392e-2 296.15 end
h 26 0 6.68559e-2 296.15 end
o 26 0 3.34279e-2 296.15 end
end comp
read celldata
latticecell squarepitch pitch=1.2600 26 fuelr=0.4095 1 cladr=0.4750 25 end
end celldata
read model
238 group solution
read parm
dz=37.1
end parm
read materials
mix=1 com="3.0 enriched fuel, pin location 1" end
mix=25 com="cladding" end
mix=26 com="water" end
end materials
read geom
global unit 1
cylinder 10 0.4095
cylinder 20 0.4750
cuboid 30 4p0.63
media 1 1 10
media 25 1 20 -10
media 26 1 30 -20
boundary 30 6 6
end geom
read collapse
8r1 2r2 3 3r4 5 5r6 6r7 2r8 3r9 4r10 4r11 12 13 10r14 3r15 16 6r17
3r18 18r19 2r20 6r21 22 3r23 24 7r25 26 16r27 2r28 11r29 30 31 14r32
33 2r34 35 3r36 35r37 5r38 7r39 11r40 4r41 2r42 43 44 3r45 2r46 2r47 2r48
2r49 2r50 51 52 2r53 54 3r55 10r56
' OLD 238G collapse to 49G
' 7r1 2 3 2r4 5 6 7 8 8 8r9 14r10 6r11 10r12 13 7r14 11r15 12r16 30r17 16r18 2r19
' 6r20 3r21 6r22 14r23 3r24 5r25 4r26 5r27 5r28 5r29 10r30 5r31 32 33 34 2r35
' 36 37 38 2r39 2r40 3r41 2r42 43 44 45 46 47 3r48 9r49 end collapse
read bounds
all=periodic
end bounds
end model
end
.. _3-1-6-2:
TRITON sample problem 2: triton2.inp
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
Sample problem triton2.inp is an example of a T-XSDRN transport
calculation sequence. In this case, the parameter specification
*parm=2region* instructs TRITON to perform cross section processing
using the CENTRM-based two-region option in place of the default
CENTRM-based S\ :sub:`N` option (see :ref:`3-1-2-1`). As in sample
problem 1, a simple square-pitched pin cell is modeled but in this case
using an XSDRN model block rather than the NEWT model block. The
moderator radius was defined in order to preserve the volume of the
moderator region.
::
' THIS SAMPLE PROBLEM TEST THE FOLLOWING:
' ** t-xsdrn sequence
' ** v7-238 group library
' ** 2region cross section processing
' ** latticecell cross section processing option
=t-xsdrn parm=2region
pin-cell model with MOX
v7-238
read comp
' Fuel
u-234 1 0 2.5952E-7 900 end
pu-238 1 0 4.6610E-5 900 end
pu-241 1 0 1.7491E-4 900 end
pu-242 1 0 1.3201E-4 900 end
o-16 1 0 4.6586E-2 900 end
pu-240 1 0 4.8255E-4 900 end
pu-239 1 0 1.0156E-3 900 end
u-235 1 0 5.4287E-5 900 end
u-238 1 0 2.1387E-2 900 end
' zirc
zr-90 2 0 3.8657E-2 620 end
fe 2 0 1.3345E-4 620 end
cr 2 0 6.8254E-5 620 end
' h2o
h-1 3 0 4.8414E-2 575 end
o-16 3 0 2.4213E-2 575 end
b-10 3 0 4.7896E-6 575 end
b-11 3 0 1.9424E-5 575 end
end comp
read cell
latticecell squarepitch pitch=1.3127 3 fueld=0.8200 1 cladd=0.9500 2 end
end cell
read model
pin-cell model with MOX
read parm
sn=16
end parm
read materials
mix=1 com='fuel' end
mix=2 com='clad' end
mix=3 com='moderator' end
end materials
read geom
geom=cylinder
rightBC=white
zoneIDs 1 2 3 end zoneids
zoneDimensions 0.41 0.475 0.7406117 end zoneDimensions
zoneIntervals 3r10 end zoneIntervals
end geom
end model
end
.. _3-1-6-3:
TRITON sample problem 3: triton3.inp
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
Sample problem 3 illustrates the input format for a T-DEPL-1D depletion
calculation. In this case, a single square-pitched pin-cell model is
depleted, where the fuel composition is comprised of UO\ :sub:`2` fuel
homogenized with aluminum and B\ :sub:`4`\ C. Although this is not
representative of real fuel, it does allow one to observe the effect of
boron depletion during burnup; results will show an increasing
multiplication factor as boron is depleted, followed by a decreasing
eigenvalue after the fuel depletion becomes the dominant contributor to
reactivity change. Three depletion intervals are specified with the same
power and no decay intervals. Two depletion subintervals are specified
for the first two depletion intervals, with only one depletion
subinterval for the final depletion interval. Note that this may be
insufficient to capture the effect of boron depletion early in life;
fewer depletion subintervals are used here only to reduce the run-time
for this sample problem. In this model, power is normalized such that
material 1 has a power density of 21.22 MW/MTHM (or MT/MTU for
UO\ :sub:`2` fuel), and OPUS output is requested for 35 nuclides. The
problem is run using the addnux=3 option set to add trace quantities of
231 nuclides to depletion materials.
..
::
' THIS SAMPLE PROBLEM TEST THE FOLLOWING:
' ** t-depl-1d sequence
' ** v7-238 group library
' ** Sn centrm cross section processing (default for t-depl-1d calculations)
' ** latticecell cross section processing option
' ** parm=addnux=3 option to add 231 nuclides to fuel material
' ** deplete-by-constant power
' ** mixture power normalization
' ** opus block
=t-depl-1d parm=(addnux=3)
Infinite lattice depletion model for a single pincell.
v7-238
read comp
' Fuel/AL2O3-B4C
uo2 1 den=10.045 1 841 92234 0.022 92235 2.453 92236 0.011 92238 97.514 end
b-10 1 0 8.5900E-4 841.0 end
b-11 1 0 3.4400E-3 841.0 end
c 1 0 1.0700E-3 841.0 end
al 1 0 3.9000E-2 841.0 end
' Clad
wtptzirc 4 6.44 4 40000 97.91 26000 0.5 50116 0.86 50120 0.73 1.0 620 end
' Moderator
h2o 5 den=0.7573 1 557 end
end comp
read celldata
latticecell squarepitch pitch=1.4732 5 fuelr=0.47250 1 cladr=0.5588 4 end
end celldata
read depletion
-1
end depletion
read burndata
power=21.220 burn=750 down=0 nlib=2 end
power=21.220 burn=750 down=0 nlib=2 end
power=21.220 burn=375 down=0 nlib=1 end
end burndata
read opus
units=gram
symnuc=u-234 u-235 u-236 u-238 pu-238 pu-239
pu-240 pu-241 pu-242 pu-243 np-237
cs-133 cs-134 cs-135 cs-137 nd-143 nd-144 nd-145 nd-146
nd-148 nd-150 pm-147 sm-147 sm-148 sm-149 sm-150 sm-151
sm-152 eu-153 sm-154 eu-154 gd-154 eu-155 gd-155 o-16 end
matl=0 1 end
end opus
read model
Infinite-lattice pin model (one-fourth)
read parm
sn=16
end parm
read materials
mix=1 com='fuel' end
mix=4 com='clad' end
mix=5 pn=2 com='water' end
end materials
read geom
geom=cylinder
rightBC=white
zoneIDs 1 4 5 end zoneids
zoneDimensions 0.47250 0.5588 0.83116409 end zoneDimensions
zoneIntervals 3r10 end zoneIntervals
end geom
end model
end
.. _3-1-6-4:
TRITON sample problem 4: triton4.inp
^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^^
Sample problem *triton4.inp* performs a large-scale depletion
calculation for a one-fourth PWR assembly, taking advantage of symmetry
to reduce the problem size. The same fuel material is used in each fuel
rod, which will result in assembly-averaged isotopic compositions for
all fuel rods. If one wanted to obtain an isotopic estimate for one or
more unique fuel rod locations, then different materials would be
specified for different rod positions. Even though all fuel is identical
at the beginning of life, unique materials must be specified if one
desires to perform tracking of the unique response of each unique fuel
position.
The problem parameter specification *parm=(weight)* instructs TRITON to
perform an automated cross section library collapse. For library
collapse automation within depletion calculations (see :ref:`3-1-3-7-3`),
TRITON will perform a single 252-group calculation at t = 0 to generate
the 56-group cross section library. TRITON will restart the depletion
calculation at t = 0 using the broad-group library after it is created.
Because *parm=weight* is specified, the *NEWT COLLAPSE* block must
comply with the 56-group energy structure and not the 252-group energy
structure. The *COLLAPSE* block input is slightly different for the
library collapse automation for *T-NEWT* calculations, where the *NEWT
COLLAPSE* block must comply with the 252-group energy structure.
Problem 4 also uses a timetable to specify boron letdown in the
moderator. The initially specified boron concentration in the *COMP* (or
*COMPOSITION*) data block is multiplied by a density multiplier at the
time of each cross section processing and transport calculation (i.e.,
the midpoint of depletion subinterval). Linear interpolation is
performed between values on the timetable to obtain the multiplier for a
given time. Typically a multiplier of 1.0 is used for t = 0, and the
beginning-of-life boron concentration is input in the *COMPOSITION*
block, but this example demonstrates that this is not necessary. For
this calculation, a 500 ppm boron concentration is specified in the
standard composition description, and a concentration of (500
ppm)*(1.832), or 916 ppm, would be used in the t = 0 transport
calculation.
Problem 4 is also an example of a lattice physics calculation for a full
fuel assembly. The NEWT model employs coarse-mesh finite-difference
acceleration, whole-assembly homogenization, 2-energy-group collapse,
and a pin-power print, and computes assembly discontinuity factors.
Although this sample problem will create the cross section database file
for core calculations, this sample problem does not contain branching
calculations, nor do lattice physics calculations typically use boron
letdown curves. Additional guidance for TRITON lattice physics
calculations can be found in the lattice physics primer.
Because only one fuel material is used, only one cell specification is
necessary. If multiple fuel materials were used, then a corresponding
cell specification would be required for each fuel, with a unique clad
and moderator material identifier for each cell. To apply boron letdown
properly, the moderator present in each cell specification would need to
have the same letdown curve specified. Hence, a letdown timetable would
need to be specified for each moderator (even if the moderators are not
all used in the NEWT *model* block). If multiple fuel materials are
used, requiring corresponding multiple clad, moderation, cell, and
timetable specifications, the use of an *alias* specification can
simplify input. Aliases are described in :ref:`3-1-3-5`; sample problems
triton6.inp (:ref:`3-1-6-6`), triton8.inp (:ref:`3-1-6-8`), and
triton12.inp (:ref:`3-1-6-12`) demonstrate the use of aliases.
This case also illustrates the use of stacked OPUS cases within a single
TRITON input file. Here, an OPUS calculation is requested to obtain the
mass in grams of 26 actinides and fission products for material 1 and
for the entire system; since material 1 is the entire set of depletion
materials, the system output will be identical to the material 1 output.
A second OPUS calculation is also specified, which requests a ranked
output of the top 20 nuclides in terms of decay heat (in watts).),
TRITON will perform a single 252-group calculation at t = 0 to generate
the 56-group cross section library. TRITON will restart the depletion
calculation at t = 0 using the broad-group library after it is created.
Because *parm=weight* is specified, the *NEWT COLLAPSE* block must
comply with the 56-group energy structure and not the 252-group energy
structure. The *COLLAPSE* block input is slightly different for the
library collapse automation for *T-NEWT* calculations, where the *NEWT
COLLAPSE* block must comply with the 252-group energy structure.
Problem 4 also uses a timetable to specify boron letdown in the
moderator. The initially specified boron concentration in the *COMP* (or
*COMPOSITION*) data block is multiplied by a density multiplier at the
time of each cross section processing and transport calculation (i.e.,
the midpoint of depletion subinterval). Linear interpolation is
performed between values on the timetable to obtain the multiplier for a
given time. Typically a multiplier of 1.0 is used for t = 0, and the
beginning-of-life boron concentration is input in the *COMPOSITION*
block, but this example demonstrates that this is not necessary. For
this calculation, a 500 ppm boron concentration is specified in the
standard composition description, and a concentration of (500
ppm)*(1.832), or 916 ppm, would be used in the t = 0 transport
calculation.
Problem 4 is also an example of a lattice physics calculation for a full
fuel assembly. The NEWT model employs coarse-mesh finite-difference
acceleration, whole-assembly homogenization, 2-energy-group collapse,
and a pin-power print, and computes assembly discontinuity factors.
Although this sample problem will create the cross section database file
for core calculations, this sample problem does not contain branching
calculations, nor do lattice physics calculations typically use boron
letdown curves. Additional guidance for TRITON lattice physics
calculations can be found in the lattice physics primer.
Because only one fuel material is used, only one cell specification is
necessary. If multiple fuel materials were used, then a corresponding
cell specification would be required for each fuel, with a unique clad
and moderator material identifier for each cell. To apply boron letdown
properly, the moderator present in each cell specification would need to
have the same letdown curve specified. Hence, a letdown timetable would
need to be specified for each moderator (even if the moderators are not
all used in the NEWT *model* block). If multiple fuel materials are
used, requiring corresponding multiple clad, moderation, cell, and
timetable specifications, the use of an *alias* specification can
simplify input. Aliases are described in :ref:`3-1-3-5`; sample problems
triton6.inp (:ref:`3-1-6-6`), triton8.inp (:ref:`3-1-6-8`), and
triton12.inp (:ref:`3-1-6-12`) demonstrate the use of aliases.
This case also illustrates the use of stacked OPUS cases within a single
TRITON input file. Here, an OPUS calculation is requested to obtain the
mass in grams of 26 actinides and fission products for material 1 and
for the entire system; since material 1 is the entire set of depletion
materials, the system output will be identical to the material 1 output.
A second OPUS calculation is also specified, which requests a ranked
output of the top 20 nuclides in terms of decay heat (in watts).
::
' THIS SAMPLE PROBLEM TEST THE FOLLOWING:
' ** t-depl sequence
' ** v7-252 group library
' ** 2region cross section processing
' ** parm=weight option for the t-depl sequence, which uses builtin 49-group collapse
' ** latticecell cross section processing option
' ** deplete-by-constant power
' ** system power normalization
' ** timetable block using density multiplier
' ** opus block defining multiple plots
=t-depl parm=(2region,weight)
Large scale 2-D depletion model with a boron letdown curve
v7-252
read comp
uo2 1 den=10.412 1 900 92234 0.04 92235 4.11 92238 95.85 end
wtptzirc 25 6.44 4 40000 97.91 26000 0.5 50116 0.86 50120 0.73 1.0 600 end
h2o 26 den=0.6798 1 593 end
wtptbor 26 0.6798 1 5000 100 500e-6 593 end
end comp
read celldata
latticecell squarepitch pitch=1.2600 26 fuelr=0.4025 1 cladr=0.4750 25 end
end celldata
read depletion
1
end depletion
read timetable
densmult 26 2 5010 5011
0.0 1.832
106 1.419
205 1.033
306 0.641
385 0.611
473 1.797
592 1.371
704 0.973
817 0.568
875 0.362 end
end timetable
::
read burndata
power=37.883 burn=385 down=88 nlib=1 end
power=32.215 burn=402 down=158 nlib=1 end
end burndata
read opus
units=gram
symnuc=u-234 u-235 u-236 u-238 pu-238 pu-239
pu-240 pu-241 pu-242 np-237 am-241 am-243 cm-242 cm-243
cs-134 cs-137 nd-143 nd-144 nd-145 nd-146 cm-244 cm-245
cm-246 cm-247 ru-106 am-242m end
matl=0 1 end
newcase
units=watts sort=yes nrank=20 time=years
end opus
read model
One-fourth fuel assembly
read parm
drawit=yes cmfd=yes xycmfd=0 echo=yes collapse=yes sn=4 inners=3 outers=200 epsilon=1e-3
end parm
read materials
mix=1 com='4.11 wt % enriched fuel' end
mix=25 com='cladding' end
mix=26 com='water' end
end materials
read collapse
40r1 16r2
end collapse
read homog
500 whole_assm 1 25 26 end
end homog
read adf
1 500 n=10.71 e=10.71 end adf
read geom
' unit 25 is a right-half water hole
unit 25
cylinder 10 .4500 chord +x=0.0
cylinder 20 .4950 chord +x=0.0
cuboid 30 0.63 0.0 0.63 -0.63
media 26 1 10
media 25 1 20 -10
media 26 1 30 -20
boundary 30 2 4
' unit 45 is top-half water hole
unit 45
cylinder 10 .4500 chord +y=0.0
cylinder 20 .4950 chord +y=0.0
cuboid 30 0.63 -0.63 0.63 0.0
media 26 1 10
media 25 1 20 -10
media 26 1 30 -20
boundary 30 4 2
::
' unit 46 is a 1/4 water hole
unit 46
cylinder 10 .4500 chord +x=0 chord +y=0
cylinder 20 .495 chord +x=0 chord +y=0
cuboid 30 0.63 0. 0.63 0.
media 26 1 10
media 25 1 20 -10
media 26 1 30 -20
boundary 30 2 2
' unit 1 is a full material #1 rod
unit 1
cylinder 10 .4025
cylinder 20 .4950
cuboid 30 0.63 -0.63 0.63 -0.63
media 1 1 10
media 25 1 20 -10
media 26 1 30 -20
boundary 30 4 4
' unit 2 is a top-half material #1 rod
unit 2
cylinder 10 .4025 chord +y=0
cylinder 20 .4950 chord +y=0
cuboid 30 0.63 -0.63 0.63 0.0
media 1 1 10
media 25 1 20 -10
media 26 1 30 -20
boundary 30 4 2
' unit 3 is a right-half material #1 rod
unit 3
cylinder 10 .4025 chord +x=0
cylinder 20 .4950 chord +x=0
cuboid 30 0.63 0.0 0.63 -0.63
media 1 1 10
media 25 1 20 -10
media 26 1 30 -20
boundary 30 2 4
global unit 100
cuboid 1 10.71 0.0 10.71 0.0
array 10 1
media 26 1 1
boundary 1
end geom
read array
ara=10 nux=9 nuy=9 pinpow=yes typ=cuboidal
fill 46 2 2 45 2 2 45 2 2
3 1 1 1 1 1 1 1 1
3 1 1 1 1 1 1 1 1
25 1 1 1 1 1 1 1 1
3 1 1 1 1 1 1 1 1
3 1 1 1 1 1 1 1 1
25 1 1 1 1 1 1 1 1
3 1 1 1 1 1 1 1 1
3 1 1 1 1 1 1 1 1 end fill
end array
read bounds
all=refl
end bounds
end model
end
.. _3-1-6-5:
TRITON sample problem 5: triton5.inp
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
Sample problem *triton5.inp* is similar to *triton4.inp*, except that it
is a T5-DEPL case; thus, a KENO V.a transport model is used in place of
the NEWT model of the earlier case. The KENO V.a model, although 3D, is
axially uniform with reflecting boundary conditions, so it is
effectively the same model as the 2D model of *triton4.inp*. Moreover,
the KENO V.a model represents the full assembly rather than a one-fourth
model. Hence, both cases will generate similar results. In the KENO
model, only 300,000 neutron histories are retained, which is somewhat
low to obtain good statistics on fluxes. The 238 ENDF/B-VII library is
used for this sample problem compared to the 252 ENDF/B-VII.1 library
utilized in *triton4.inp*.
::
' THIS SAMPLE PROBLEM TEST THE FOLLOWING:
' ** t-depl sequence
' ** v7-238 group library
' ** 2region cross section processing
' ** latticecell cross section processing option
' ** deplete-by-constant power
' ** system power normalization
' ** timetable block using density multiplier
=t5-depl parm=2region
Large scale 2-D depletion model with boron density change.
V7-238
read comp
uo2 1 den=10.412 1 900 92234 0.04 92235 4.11 92238 95.85 end
wtptzirc 25 6.44 4 40000 97.91 26000 0.5 50116 0.86 50120 0.73 1.0 600 end
h2o 26 den=0.6798 1 593 end
wtptbor 26 0.6798 1 5000 100 500e-6 593 end
end comp
read celldata
latticecell squarepitch pitch=1.2600 26 fuelr=0.4025 1 cladr=0.4750 25 end
end celldata
read depletion
1
end depletion
read timetable
densmult 26 2 5010 5011
0.0 1.832
106 1.419
205 1.033
306 0.641
385 0.611
473 1.797
592 1.371
704 0.973
817 0.568
875 0.362 end
end timetable
read burndata
power=37.883 burn=385 down=88 nlib=1 end
power=32.215 burn=402 down=158 nlib=1 end
end burndata
read model
read parm
cfx=yes gen=620 nsk=20 npg=500 plt=no htm=no
end parm
::
read geom
' unit 2 is a water hole
unit 2
cylinder 26 1 .4500 10.0 0.0
cylinder 25 1 .4950 10.0 0.0
cuboid 26 1 0.63 -0.63 0.63 -0.63 10.0 0.0
' unit 1 is a material #1 rod
unit 1
cylinder 1 1 .4025 10.0 0.0
cylinder 25 1 .4950 10.0 0.0
cuboid 26 1 0.63 -0.63 0.63 -0.63 10.0 0.0
global unit 100
array 10 0.0 0.0 0.0
end geom
read array
ara=10 nux=17 nuy=17 nuz=1 typ=cuboidal
fill 17r1
17r1
8r1 2 8r1
17r1
17r1
8r1 2 8r1
17r1
17r1
2r1 2 2r1 2 2r1 2 2r1 2 2r1 2 2r1
17r1
17r1
8r1 2 8r1
17r1
17r1
8r1 2 8r1
17r1
17r1 end fill
end array
read bounds
all=refl
end bounds
end data
end model
end
.. _3-1-6-6:
TRITON sample problem 6: triton6.inp
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
Sample problem *triton6.inp* performs T-DEPL depletion in a pin cell
model; however, the pin is discretized into five equal-volume rings of
fuel. Thus, CENTRM-based S\ :sub:`N` cross section processing is
necessary to capture the radial burnup of the pin cell. A *multiregion*
cell specification is given to allow specification of the varying radii
for the fuel regions. Because the multiregion cell is cylindrical, the
moderator volume is represented in terms of a radius that corresponds to
the volume associated with the pin pitch. The right boundary condition
for the cell is set to *white*; this is important, as the default right
boundary condition for a multiregion cylinder is vacuum. In this case,
addnux=1 is also requested in the parameter specification, simply for a
faster (but less accurate) calculation. Material aliases are used to
simplify input. The calculation is performed with the 238 ENDF/B-VII
library. The TRITON *TIMETABLE* block is used to demonstrate
time-dependent temperature changes to the moderator material.
::
' THIS SAMPLE PROBLEM TEST THE FOLLOWING:
' ** t-depl sequence
' ** v7-238 group library
' ** centrm cross section processing
' ** multiregion cross section processing option
' ** deplete-by-constant power
' ** parm=addnux=1 option to add 15 nuclides to fuel material
' ** system power normalization
' ** timetable block using temperature change
' ** alias block definition
' ** opus block
=t-depl parm=(centrm,addnux=1)
Pin-cell depleted in rings
v7-238
read alias
$fuel 1-5 end
end alias
read comp
uo2 $fuel den=9.459 1 829.0 92234 0.027 92235 3.038 92236 0.014 92238 96.921 end
wtptzirc 10 6.44 4 40000 97.91 26000 0.5 50116 0.86 50120 0.73 1.0 620 end
h2o 11 den=0.7575 1 557 end
wtptbor 11 0.7575 1 5000 100 654e-6 557 end
end comp
read celldata
multiregion cylindrical right=white end
1 0.16425
2 0.28449
3 0.36727
4 0.43456
5 0.49275
10 0.55880
11 .83120 end zone
end celldata
read depletion
$fuel
end depletion
read timetable
temperature 11
' cycle 1
0.0 557.0
306.0 557.0
' cycle 2
377.0 540.0
838.1 557.0 end
end timetable
::
read burndata
power=27.24 burn=306.0 down=71 nlib=1 end
power=34.57 burn=461.1 down=1870 nlib=1 end
end burndata
read opus
units=gram symnuc=u-235 u-238 pu-239 pu-241 nd-148 end matl=0 1 2 3 4 5 end
end opus
read model
Infinite lattice PWR pin cell
read parm
drawit=yes prtbroad=yes epsilon=1e-3 soln=b1 converg=matl
end parm
read materials
mix=$fuel com='3.038 wt % enriched fuel' end
mix=10 pn=0 com='cladding' end
mix=11 com='water' end
end materials
read geom
global unit 1
cylinder 1 .16425 chord +x=0 chord +y=0
cylinder 2 .28449 chord +x=0 chord +y=0
cylinder 3 .36727 chord +x=0 chord +y=0
cylinder 4 .43456 chord +x=0 chord +y=0
cylinder 5 .49275 chord +x=0 chord +y=0
cylinder 20 .5588 chord +x=0 chord +y=0
cuboid 30 0.7366 0.0 0.7366 0.0
media 1 1 1
media 2 1 2 -1
media 3 1 3 -2
media 4 1 4 -3
media 5 1 5 -4
media 10 1 20 -5
media 11 1 30 -20
boundary 30 4 4
end geom
read bounds
all=refl
end bounds
end model
end
.. _3-1-6-7:
TRITON sample problem 7: triton7.inp
~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~~
Sample problem triton7.inp is an example of a T-DEPL depletion
calculation for a full PWR fuel assembly model. Depletion is performed
on the basis of material 7, which is located in a single fuel pin for
which destructive assay measurements were performed. All other fuel is
modeled as a single (average) material, material 1. The parameter
specification *parm=(2region,addnux=1,weight)* was chosen to reduce the
run-time of the sample problem.
This sample problem also demonstrates the use of TRITON’s standard
composition restart files and SCALE external file reading capabilities
to represent the time-dependent behavior of an assembly in which
burnable poisons are removed after the first cycle of operation.
This problem consists of two TRITON 2D depletion cases. In the first
case, the full assembly model contains borosilicate glass burnable
poison rods (BPRs), material 4, which are included in the list of
materials to be depleted. The calculation is run for the entirety of the
first operational cycle, which included a 40-day mid-cycle decay
interval. The model also includes a 64-day decay interval after the end
of the operational cycle. When this calculation is completed, TRITON
creates in the temporary working directory a standard composition file
for each material containing the isotopic inventories for each depletion
material at the end of the 64-day decay interval. The second TRITON
calculation reads the standard composition specifications for materials
1 and 7 as part of the input to provide the fuel state for the second
calculation. In the second model, the BPRs are removed and replaced with
the moderator in the embedded NEWT model. The initial depletion
calculation uses the 252 ENDF/B-VII.1 library. With the
*parm=(…,weight)* option, a 56 group library is created in the temporary
directory called *newxnlib*. This library is used for the second
*T-NEWT* calculation.
::
' THIS SAMPLE PROBLEM TEST THE FOLLOWING:
' ** t-depl sequence
' ** v7-252 group library
' ** 2region cross section processing
' ** latticecell cross section processing option
' ** deplete-by-constant flux
' ** parm=addnux=1 option to add 15 nuclides to fuel material
' ** mixture power normalization
' ** timetable block using density multiplier
' ** composition restart files.
' ** weight used to collapse library for reuse in restart calculation
=t-depl parm=(2region,addnux=1,weight)
ASSEMBLY model with BPRs with depletion
v7-252
read comp
uo2 1 den=9.550 1 743 92234 0.023 92235 2.561 92236 0.013 92238 97.403 end
wtptzirc 2 6.44 4 40000 97.91 26000 0.5 50116 0.86 50120 0.73 1.0 620 end
h2o 3 den=0.7544 1 559 end
wtptbor 3 0.7544 1 5000 100 652.5e-6 559 end
wtptbpr 4 2.081 6 8016 53.58 11000 2.82 13027 1.758 14000 37.63 19000 0.33 5000 3.882 1 559 end
wtptair 5 0.00129 2 7000 78.0 8016 22.0 1 559.0 end
ss304 6 1 559.0 end
uo2 7 den=9.550 1 743 92234 0.023 92235 2.561 92236 0.013 92238 97.403 end
wtptzirc 8 6.44 4 40000 97.91 26000 0.5 50118 0.64 50120 0.95 1 595 end
h2o 9 den=0.7544 1 559 end
wtptbor 9 0.7544 1 5000 100 652.5e-6 559 end
end comp
read celldata
latticecell squarepitch pitch=1.43 3 fueld=0.9484 1 cladd=1.0719 2 end
latticecell squarepitch pitch=1.43 9 fueld=0.9484 7 cladd=1.0719 8 end
end celldata
read depletion
1 -7 flux 4
end depletion
read timetable
density 3 2 5010 5011
0.00 1.000
243.5 1.000
283.5 0.379
527.0 0.379 end
density 9 2 5010 5011
0.00 1.000
243.5 1.000
283.5 0.379
527.0 0.379 end
end timetable
::
read burndata
power=20.86 burn=243.5 down=40.0 nlib=1 end
power=20.15 burn=243.5 down=64.0 nlib=1 end
end burndata
read model
ASSEMBLY model with BPRs with depletion
read parm
drawit=yes inners=2 epsilon=-5e-2 cmfd=1 xycmfd=0 echo=yes solntype=b1 timed=yes
end parm
read materials
mix=6 pn=1 com="SS-304 - BPR clad" end
mix=5 pn=1 com="air in BPRs" end
mix=4 pn=1 com="borosilicate glass" end
mix=3 pn=2 com="water" end
mix=2 pn=1 com="cladding" end
mix=1 pn=1 com="2.561 wt % enriched fuel " end
mix=7 pn=1 com="rod N-9" end
end materials
read geom
global unit 10
cuboid 13 10.725 0.0 10.725 0.0
array 101 13 place 1 1 -0.715 -0.715
media 3 1 13
boundary 13 15 15
unit 1
cuboid 13 1.43 0.0 1.43 0.0
cylinder 12 0.53595 origin x=0.715 y=0.715
cylinder 11 0.4742 origin x=0.715 y=0.715
media 3 1 13 -12
media 2 1 12 -11
media 1 1 11
boundary 13 2 2
unit 2
cuboid 13 1.43 0.0 1.43 0.0
cylinder 14 0.28385 origin x=0.715 y=0.715
cylinder 15 0.30035 origin x=0.715 y=0.715
cylinder 16 0.50865 origin x=0.715 y=0.715
cylinder 17 0.55755 origin x=0.715 y=0.715
media 3 1 13 -17
media 6 1 17 -16
media 4 1 16 -15
media 6 1 15 -14
media 5 1 14
boundary 13 2 2
unit 3
cuboid 13 1.43 0.0 1.43 0.0
cylinder 12 0.6934 origin x=0.715 y=0.715
cylinder 11 0.6502 origin x=0.715 y=0.715
media 3 1 13 -12
media 2 1 12 -11
media 3 1 11
boundary 13 2 2
unit 4
cuboid 13 1.43 0.715 1.43 0.0
cylinder 12 0.53595 origin x=0.715 y=0.715 chord +x=0.715
cylinder 11 0.4742 origin x=0.715 y=0.715 chord +x=0.715
media 3 1 13 -12
media 2 1 12 -11
media 1 1 11
boundary 13 1 2
unit 5
cuboid 13 1.43 0.0 1.43 0.715
cylinder 12 0.53595 origin x=0.715 y=0.715 chord +y=0.715
cylinder 11 0.4742 origin x=0.715 y=0.715 chord +y=0.715
media 3 1 13 -12
media 2 1 12 -11
media 1 1 11
boundary 13 2 1
unit 6
cuboid 13 1.43 0.715 1.43 0.0
cylinder 12 0.6934 origin x=0.715 y=0.715 chord +x=0.715
cylinder 11 0.6502 origin x=0.715 y=0.715 chord +x=0.715
media 3 1 13 -12
media 2 1 12 -11
media 3 1 11
boundary 13 1 2
::
unit 7
cuboid 13 1.43 0.0 1.43 0.715
cylinder 12 0.6934 origin x=0.715 y=0.715 chord +y=0.715
cylinder 11 0.6502 origin x=0.715 y=0.715 chord +y=0.715
media 3 1 13 -12
media 2 1 12 -11
media 3 1 11
boundary 13 2 1
unit 8
cuboid 13 1.43 0.715 1.43 0.715
cylinder 12 0.6934 origin x=0.715 y=0.715 chord +x=0.715 chord +y=0.715
cylinder 11 0.6502 origin x=0.715 y=0.715 chord +x=0.715 chord +y=0.715
media 3 1 13 -12
media 2 1 12 -11
media 3 1 11
boundary 13 1 1
unit 9
cuboid 13 1.43 0.0 1.43 0.0
cylinder 12 0.53595 origin x=0.715 y=0.715
cylinder 11 0.4742 origin x=0.715 y=0.715
media 3 1 13 -12
media 2 1 12 -11
media 7 1 11
boundary 13 2 2
end geom
read array
ara=101 nux=8 nuy=8 typ=cuboidal fill
8 5 5 5 7 5 5 5
4 1 1 1 1 1 1 1
4 1 1 1 1 2 1 1
4 1 1 3 1 1 1 1
6 1 1 1 1 1 1 1
4 9 2 1 1 2 1 1
4 1 1 1 1 1 1 1
4 1 1 1 1 1 1 1 end fill
end array
end model
end
::
=t-newt parm=(2region)
ASSEMBLY model without BPRs
newxnlib
read comp