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.. _11-1:
AMPX Library Utility Modules
============================
*D. Wiarda, L. M. Petrie*
Abstract
The purpose of this section is to document selected AMPX modules that
can benefit the analyst interested in editing, converting, or combining
cross-section libraries normally used by the SCALE system modules. The
input description for these codes is provided in the documentation of
the AMPX nuclear data processing code system that is distributed with
SCALE package.
.. _11-1-1:
Introduction
------------
AMPX is a modular system [1]_ that generates continuous energy (CE) and
multigroup (MG) cross section data from evaluated nuclear data files
such as ENDF/B. All the nuclear data libraries distributed with SCALE
have been processed using AMPX. In addition to data processing modules,
AMPX also includes a number of useful utility modules for checking,
manipulating, and editing the libraries in SCALE. This section lists and
briefly describes some of the AMPX utility codes that may be useful to
SCALE users. Input instructions for these codes can be found the AMPX
code documentation, which is distributed with the SCALE code package.
Additional AMPX modules of interest may also found in the documentation.
.. _11-1-2:
AJAX: MODULE TO MERGE, COLLECT, ASSEMBLE, REORDER, JOIN, AND/OR COPY SELECTED DATA FROM AMPX MASTER LIBRARIES
-------------------------------------------------------------------------------------------------------------
AJAX (**A**\ utomatic **J**\ oining of **A**\ MPX **X**-Sections) is a
module to combine data from different AMPX libraries. Options are
provided to allow merging from any number of files.
.. _11-1-3:
ALPO: MODULE TO CONVERT AMPX LIBRARIES INTO ANISN FORMAT
--------------------------------------------------------
ALPO (**A**\ NISN **L**\ IBRARY **P**\ RODUCTION **O**\ PTION) is a
module for converting AMPX working libraries into the library format
used by the legacy discrete ordinates transport codes ANISN and
DORT/TORT contained in the DOORS package. [2]_
.. _11-1-4:
CADILLAC: MODULE TO MERGE MULTIPLE COVARIANCE DATA FILES
--------------------------------------------------------
CADILLAC (**C**\ ombine **A**\ ll **D**\ ata **I**\ dentifiers
**L**\ isted in **L**\ ogical **A**\ MPX **C**\ overx-format) is a
module that can be used to combine multiple covariance data files in
COVERX format into a single covariance data file. The material IDs can
be changed as needed by the user.
.. _11-1-5:
COGNAC: MODULE TO CONVERT COVARIANCE DATA FILES IN COVERX FORMAT
----------------------------------------------------------------
COGNAC (**C**\ onversion **O**\ perations for **G**\ roup-dependent
**N**\ uclides in **A**\ MPX **C**\ overx-format) is a module that can
be used to convert a single COVERX-formatted data file from bcd format
to binary. Also, COGNAC can be used to convert from binary to bcd,
binary to binary, and bcd to bcd.
.. _11-1-6:
LAVA: MODULE TO MAKE AN AMPX WORKING LIBRARY FROM AN ANISN LIBRARY
------------------------------------------------------------------
LAVA (**L**\ et **A**\ NISN **V**\ isit **A**\ MPX) is a module that can
convert an ANISN formatted library (neutron, gamma, or coupled
neutron-gamma) to an AMPX working library that can be used in XSDRNPM.
.. _11-1-7:
MALOCS: MODULE TO COLLAPSE AMPX MASTER CROSS-SECTION LIBRARIES
--------------------------------------------------------------
MALOCS (**M**\ iniature **A**\ MPX **L**\ ibrary **O**\ f **C**\ ross
**S**\ ections) is a module to collapse AMPX master cross-section
libraries. The module can be used to collapse neutron, gamma-ray, or
coupled neutron-gamma master libraries.
.. _11-1-8:
PALEALE: MODULE TO LIST INFORMATION FROM AMPX LIBRARIES
--------------------------------------------------------
PALEALE lists selected data by nuclide, reaction, data-type from AMPX
master and working libraries.
.. _11-1-9:
RADE: MODULE TO CHECK AMPX CROSS-SECTION LIBRARIES
--------------------------------------------------
RADE (**R**\ ancid **A**\ MPX **D**\ ata **E**\ xposer) is provided to
check AMPX- and ANISN-formatted multigroup libraries. It will check
neutron, gamma, or coupled neutron-gamma libraries.
.. _11-1-10:
TOC: MODULE TO PRINT AN AMPX LIBRARY TABLE OF CONTENTS
------------------------------------------------------
Program TOC is a utility program to print a sorted table of contents of
an AMPX cross section library. It is designed to be run interactively,
with the cross section library specified as the argument.
References
~~~~~~~~~~
.. [1]
D. Wiarda, M. L. Williams, C. Celik, and M. E. Dunn, “AMPX: A
Modern Cross Section Processing System For Generating Nuclear Data
Libraries,” *Proceedings of International Conference on Nuclear
Criticality Safety,* Charlotte, NC, Sept 13-17 2015.
.. [2]
**“**\ \ DOORS3.2a:   One, Two- and Three-Dimensional Discrete
Ordinates Neutron/Photon Transport Code System”, Radiation Shielding
Information Center package CCC-650, Oak Ridge National Laboratory
(2003).
......@@ -848,7 +848,7 @@ The limits on the above integral correspond to:
\alpha_{\mathrm{L}}\left(\mathrm{E}^{\prime}, \mathrm{E}\right)=\alpha\left(\mathrm{E}^{\prime}, \mathrm{E}, \mu_{0}=-1\right) \quad \text { and } \quad \alpha_{\mathrm{H}}\left(\mathrm{E}^{\prime}, \mathrm{E}\right)=\alpha\left(\mathrm{E}^{\prime}, \mathrm{E}, \mu_{0}=1\right) .
The alpha moments for n > 0 can be evaluated very efficiently using a
recursive relation :cite:`illiams_submoment_2000`:
recursive relation :cite:`williams_submoment_2000`:
.. math::
......
.. _11-3:
COMPOZ Data Guide
=================
*J. R. Knight* [1]_ *and L. M. Petrie* 
ABSTRACT
The COMPOZ program used to create the Standard Composition Library is
described. Of particular importance is documentation of the COMPOZ input
data file structure. Knowledge of the file structure allows users to
edit the data file and subsequently create their own site-specific
composition library.
ACKNOWLEDGMENT
This work was originally funded by the Office of Nuclear Material Safety
and Safeguards of the U.S. Nuclear Regulatory Commission.
.. _11-3-1:
Introduction
------------
COMPOZ is the program that creates (writes) the SCALE Standard
Composition Library. Data are input in free form. A text data file
containing the input to COMPOZ (and the Standard Composition Library) is
available with the SCALE system. Execution of COMPOZ using this data
file creates the Standard Composition Library currently available with
the SCALE package. This section provides documentation of the data file
structure. Knowledge of the data file structure allows users to edit the
data file and subsequently create their own site-specific or
user-specific composition library.
COMPOZ is intended to create or make *permanent* *changes* to and/or to
print the composition library and should not be used for any other
purpose. To avoid confusion with the Standard Composition Library
provided with SCALE, it is strongly recommended that only *new* keywords
and compositions be used in any site-specific or user-specific library.
.. _11-3-2:
Input Data Description
----------------------
COMPOZ input data are entered in free form. All data must be followed by
at least one blank. The COMPOZ input data file contains *five* data
records or blocks:
1. COMPOZ mode flag selects whether a new standard composition library
will be created, or an old standard composition library will be
listed. Only if a new library is being created are the following data
records entered. A new library is created with a filename of
“xfile089”. If an old library is being dumped as an ASCII file, it
will be named “_sclN…N” where N…N is an 18 digit sequence number that
is incremented starting from 0 for each library dumped in the same
directory.
2. The header record contains the library identification, a set of
parameters describing the size of the library, and library title with
80 characters per line.
3. The standard composition table contains the name, theoretical
density, number of elements, and other information about each
standard composition. Individual nuclides, mixtures, and compounds
are all included in the table.
4. The nuclide information table contains the nuclide identification
number, atomic mass, and resonance energy cross sections.
5. The isotopic distribution table contains the nuclide identification
number and the atom percent of each isotope used in specifying the
default enrichment.
.. note:: For executing COMPOZ via SCALE, an =COMPOZ is required in the
first eight columns of a record preceding the mode flag, and an END is
required in the first three columns of a record inserted after the last
data record. If debug output is desired, then use =COMPOZ PRINTDEBUG to
execute compoz.
.. _11-3-2-1:
COMPOZ mode selector
~~~~~~~~~~~~~~~~~~~~
1. LGEN =
0 – create a new library and list it
1 – list an existing library
>1 – list an existing library and write an ASCII input file.
If LGEN is 0, then input the following data to create a new standard
composition library.
.. _11-3-2-2:
Library heading information
~~~~~~~~~~~~~~~~~~~~~~~~~~~
1. IDT – library identification number
2. TITLE – 1 line of 80 characters used to identify the library
.. _11-3-2-3:
Standard composition table
~~~~~~~~~~~~~~~~~~~~~~~~~~
1. SCID – Composition name, maximum of 12 characters.
2. ROTH – Theoretical density, gm/cm\ :sup:`3`.
3. ICP –
0 for a mixture,
1 for a compound.
4. NCZA – Element or nuclide ID
5. ATPM – Weight percent if ICP = 0. Number of atoms per molecule if ICP = 1.
6. END – Keyword END to terminate this standard composition.
For each composition, items 4 and 5 are repeated until all components of
the composition are described. Items 1 through 6 are entered in a
similar fashion for all compositions. After all the standard
compositions are read, terminate the table with an END [label], where
[label] represents an optional label.
.. _11-3-2-4:
Nuclide information table
~~~~~~~~~~~~~~~~~~~~~~~~~
1. NZA – Nuclide ID. This should be the mass number + 1000 \* the atomic
number.
2. AM – Atomic mass, C-12 scale.
3. SIGS – Resonance energy scattering cross section, barns.
4. SIGT – Resonance energy total cross section, barns.
5. NU*SIGF – Resonance energy nu*sigf cross section, barns.
The resonance energy cross sections are averaged over the appropriate
energy range for the nuclide. Items 1–5 are repeated for all nuclides.
After all nuclides are entered, terminate the nuclide table with an END
[label].
.. _11-3-2-5:
Isotopic distribution table
~~~~~~~~~~~~~~~~~~~~~~~~~~~
1. NZN – 1000 \* atomic number of variable isotope elements.
2. ISZA – Isotope ID.
3. ABWP – Default abundance, atom percent.
4. END – Keyword END to terminate this isotopic specification.
The default abundance is generally the naturally occurring abundance.
For each element, items 2 and 3 are repeated until 100% total abundance
is described, making a set for this element. The next element is
described in the same fashion in the next set, etc. After all isotopic
distributions are entered, terminate the isotopic distribution table
with an END [label].
.. _11-3-3:
Sample Problem
--------------
The following sample problem first lists the SCALE standard composition
library, then creates a new, short standard composition library, then
lists and outputs an ASCII copy of this new library, and finally copies
this new copy back to the output directory.
.. code:: scale
:class: long
=compoz
' print the current standard composition library
1
end
=compoz
' create a new standard composition library
0
' library identification number
101
' library title
scale-X standard composition library
' standard composition table
' all nuclide IDs here must be in the nuclide table
h 1.0000 0 1001 100.0000 end
o 1.0000 0 8016 100.0000 end
u 19.0500 0 92000 100.0000 end
h2o 0.9982 1 1001 2
8016 1 end
uo2 10.9600 1 92000 1
8016 2 end
' end of standard composition table
end stdcmp
' nuclide table
' ID AWR SigmaS SigmaT nuSigmaF
1001 1.00783 20.38087 20.38782 0.00000
1002 2.01410 3.39486 3.39487 0.00000
8016 15.99491 3.88696 3.88696 0.00000
8017 16.99913 3.74000 3.74501 0.00000
8018 17.99916 3.79000 3.79000 0.00000
92233 233.03964 12.46693 37.62292 100.78482
92234 234.04095 12.18716 16.09542 2.66969
92235 235.04393 11.90249 35.22383 90.23152
92236 236.04556 12.27302 14.93351 1.33334
92237 237.04874 14.24581 24.68619 1.93695
92238 238.05080 12.32636 14.62708 0.65970
' end of nuclide table
end nuclides
' isotope distribution table
' all nuclide IDs here must be in the nuclide table
1000 1001 99.9885
1002 0.0115 end
8000 8016 99.7570
8017 0.0380
8018 0.2050 end
92000 92234 0.0054
92235 0.7204
92238 99.2742 end
' end of isotope distribution table
end isotopes
' end of compoz input
end
=compoz
' print and create an ASCII copy of the current standard composition library
' (created in the previous step)
2
end
=shell
# copy the ASCII copy back to the output directory
copy_file _scl000000000000000000 ${OUTBASE}.stdcmplib
end
.. [1]
Formerly with Oak Ridge National Laboratory.
This source diff could not be displayed because it is too large. You can view the blob instead.
.. _10-2a:
COVLIB Appendix A: Cross section plots for U, Pu, TH, B, H, He, and Gd Nuclides
===============================================================================
Plots of cross section differences between various evaluations are shown
below. The legend below applies to all plots shown in this appendix.
.. image:: figs/COVLIBAppA/img1.png
:align: center
:width: 500
.. _fig10-2a-1:
.. figure:: figs/COVLIBAppA/fig1.png
:align: center
:width: 600
:sup:`239`\ Pu fission and capture comparison between ENDF/B-VI,
JENDL 3.3, and JEF 3.1.
.. _fig10-2a-2:
.. figure:: figs/COVLIBAppA/fig2.png
:align: center
:width: 600
:sup:`240`\ Pu fission and capture comparison between ENDF/B-VI,
JENDL 3.3 and JEF 3.1.
.. _fig10-2a-3:
.. figure:: figs/COVLIBAppA/fig3.png
:align: center
:width: 600
:sup:`241` fission and capture comparison between ENDF/B-VI,
JENDL 3.3 and JEF 3.1.
.. _fig10-2a-4:
.. figure:: figs/COVLIBAppA/fig4.png
:align: center
:width: 600
:sup:`233` fission and capture comparison between ENDF/B-VI,
JENDL 3.3 and JEF 3.1.
.. _fig10-2a-5:
.. figure:: figs/COVLIBAppA/fig5.png
:align: center
:width: 600
:sup:`235` fission and capture comparison between ENDF/B-VI,
JENDL 3.3 and JEF 3.1.
.. _fig10-2a-6:
.. figure:: figs/COVLIBAppA/fig6.png
:align: center
:width: 600
:sup:`238`\ U capture comparison between ENDF/B-VI, JENDL 3.3
and JEF 3.1.
.. _fig10-2a-7:
.. figure:: figs/COVLIBAppA/fig7.png
:align: center
:width: 600
:sup:`232`\ Th capture comparison between ENDF/B-VII
(beta2), ENDF/B‑VI, JENDL 3.3 and JEF 3.1.
.. _fig10-2a-8:
.. figure:: figs/COVLIBAppA/fig8.png
:align: center
:width: 600
:sup:`10`\ B capture and :sup:`3`\ He elastic comparison
between ENDF/B-VII (beta2), ENDF/B-VI, JENDL 3.3 and JEF 3.1.
.. _fig10-2a-9:
.. figure:: figs/COVLIBAppA/fig9.png
:align: center
:width: 600
:sup:`1`\ H and :sup:`2`\ H elastic comparison between
ENDF/B-VII (beta2), ENDF/B-VI, JENDL 3.3 and JEF 3.1.
.. _fig10-2a-10:
.. figure:: figs/COVLIBAppA/fig10.png
:align: center
:width: 600
:sup:`152`\ Gd and :sup:`154`\ Gd capture comparison
between ENDF/B-VII (beta2), ENDF/B-VI, JENDL 3.3 and JEF 3.1.
.. _fig10-2a-11:
.. figure:: figs/COVLIBAppA/fig11.png
:align: center
:width: 500
:sup:`155`\ Gd capture comparison between ENDF/B-VII
(beta2), ENDF/B-VI, JENDL 3.3 and JEF 3.1.
......@@ -10,7 +10,7 @@ ACKNOWLEDGMENTS
The authors would like to acknowledge the important contributions to
CRAWDAD made by former ORNL staff N. M. Greene and D. F. Hollenbach.
.. _7-1:
.. _7-7-1:
Introduction
------------
......
.. _5-0:
Depletion, Activation, and Spent Fuel Source Terms Overview
===========================================================
*Introduction by W. A. Wieselquist*
SCALE’s general depletion, activation, and spent fuel source terms analysis
capabilities are enabled through a family of modules related to the main ORIGEN
depletion/irradiation/decay solver. The nuclide tracking in ORIGEN is based on
the principle of explicitly modeling all available nuclides and transitions in
the current fundamental nuclear data for decay and neutron-induced transmutation
and relies on fundamental cross section and decay data in ENDF/B VII. Cross
section data for materials and reaction processes not available in ENDF/B-VII
are obtained from the JEFF-3.0/A special purpose European activation library
containing 774 materials and 23 reaction channels with 12,617 neutron-induced
reactions below 20 MeV. Resonance cross section corrections in the resolved and
unresolved range are performed using a continuous-energy treatment by data
modules in SCALE. All nuclear decay data, fission product yields, and gamma-ray
emission data are developed from ENDF/B-VII.1 evaluations. Decay data include
all ground and metastable state nuclides with half-lives greater than 1
millisecond. Using these data sources, ORIGEN currently tracks 174 actinides,
1149 fission products, and 974 activation products. The purpose of this chapter
is to describe the stand-alone capabilities and underlying methodology of
ORIGEN—as opposed to the integrated depletion capability it provides in all
coupled neutron transport/depletion sequences in SCALE, as described in other
chapters. Through the stand-alone capabilities, there is generality to handle
arbitrary systems (e.g., fast reactor fuel depletion or structural activation)
by providing arbitrary flux spectra and arbitrary one-group cross sections to
the module COUPLE, which in turn creates ORIGEN library (.f33) files containing
the problem-dependent, one-group reaction coefficients required to solve the
actual equations governing depletion/decay. These libraries are required input
for the ORIGEN module, along with the initial isotopics and irradiation history,
in terms of either a time-dependent power or flux level. Two high-performance
equation solvers are available: the hybrid linear chains and matrix exponential
method and a new Chebyshev Rational Approximation Method (CRAM). Typical
execution times are on the order of a few seconds for a multi-step solution,
with each individual solution (step) taking approximately 10 milliseconds.
ORIGEN also includes capabilities for continuous feed and removal by element.
Output capabilities include isotopics (moles or grams), source spectra (alpha,
beta, gamma, and neutron), activity (becquerels or curies), decay heat (total
watts or gamma only), and radiological hazard factors (maximum permissible
concentrations in air or water). These results can be displayed in the output
file (.out extension) and/or archived in an ORIGEN binary results file (.f71
extension). The use of current, fundamental data resources is a key feature of
ORIGEN, including nuclear decay data, multigroup neutron reaction cross
sections, neutron-induced fission product yields, and decay emission data for
photons, neutrons, alpha particles, and beta particles. The nuclear decay data
are based primarily on ENDF/B-VII.1 evaluations. The multigroup nuclear reaction
cross section libraries now include evaluations from the JEFF 3.0/A neutron
activation file containing data for 774 target nuclides, more than 12,000
neutron-induced reactions, and more than 20 different reaction types below 20
MeV, provided in various energy group structures. Energy-dependent
ENDF/B-VII.0-based fission product yields are available for 30 fissionable
actinides. Gamma-ray and x-ray emission data libraries are based on
ENDF/B-VII.1. The photon libraries contain discrete photon line energy and
intensity data for decay gamma-ray and x-rays emission for 1,132 radionuclides,
prompt and delayed continuum spectra for spontaneous fission, (α,n) reactions in
oxide fuel, and bremsstrahlung from decay beta (electron and positron) particles
slowing down in either a UO2 fuel or water matrix. Methods and data libraries
used to calculate the neutron yields and energy spectra for spontaneous fission,
(α,n) reactions, and delayed neutron emission are adopted from the SOURCES4C
code. Capabilities to calculate the beta and alpha particle emission source and
spectra have also been added.
The ORIGEN reactor libraries distributed with SCALE include a set of
pre-calculated ORIGEN libraries (with TRITON) for a variety of fuel assembly
designs:
- BWR 7×7, 8×8-1, 8×8-2, 9×9-2, 9×9-9, 10×10-9, 10×10-8, SVEA-64,
SVEA-96, and SVEA-100;
- PWR 14×14, 15×15, 16×16, 17×17, 18×18;
- CANDU reactor (19-, 28-, and 37-element bundle designs);
- Magnox graphite reactor;
- Advanced Gas-Cooled Reactor (AGR);
- VVER 440 and VVER 1000;
- RBMK;
- IRT;
- MOX BWR 7×7, 8×8-1, 8×8-2, 9×9-2, 9×9-9, 10×10-9, 10×10-8, SVEA-64, SVEA-96, and
SVEA-100;
- MOX PWR 14×14, 15×15, 16×16, 17×17, 18×18.
These libraries may be
used to rapidly assess spent fuel isotopics and source terms in these systems
for arbitrary burnups and decay times. For UO2-based assembly isotopics, the new
ORIGAMI sequence provides a very convenient, easy-to-use interface. The most
general capability, and requiring more user input, is available using the ARP
interpolator module in conjunction with the ORIGEN solver module. Finally, with
regards to user interfaces, ORIGEN has a new keyword-based input in SCALE 6.2
but also maintains the ability to read SCALE 6.1 input. Both ORIGEN and ORIGAMI
are tightly integrated with the SCALE graphical user interface, Fulcrum, which
includes syntax highlighting, input checking with immediate feedback, and (.f71)
output viewing. Additionally, Fulcrum provides an ORIGAMI Automator project
interface to characterize the fuel inventory for an entire reactor site and
generate data needed for severe accident analysis. ORIGAMI Automator is not
documented in this chapter, but a primer is available with step by step
instructions on its use.
Deterministic Transport Overview
================================
*Introduction by S. M. Bowman*
SCALE deterministic transport capabilities enable criticality safety,
depletion, sensitivity, and uncertainty analysis, as well as hybrid
approaches to Monte Carlo analysis. SCALE provides a one-dimensional
(1D) transport solver for eigenvalue neutronics and fixed source
neutron-gamma analysis with XSDRN, two-dimensional (2D) eigenvalue
neutronics with NEWT, and a three-dimensional (3D) transport solver for
hybrid acceleration of Monte Carlo fixed source and eigenvalue
calculations with Denovo. Generally, the use of these transport solvers
in SCALE is best accessed through the capability specific sequences:
CSAS and Sourcerer for criticality safety, TRITON for 1D and 2D
depletion, TSUNAMI‑1D and TSUNAMI-2D for sensitivity and uncertainty
analysis, and MAVRIC for 3D fixed source hybrid Monte Carlo analysis.
XSDRN
-----
XSDRN is a multigroup discrete-ordinates code that solves the 1D
Boltzmann equation in slab, cylindrical, or spherical coordinates.
Alternatively, the user can select P1 diffusion theory, infinite medium
theory, or Bn theory. A variety of calculational types is available,
including fixed source, eigenvalue, or search calculations. In SCALE,
XSDRN is used for several purposes: eigenvalue (*k*\ :sub:`eff`) determination;
cross section collapsing; and computation of fundamental-mode or
generalized adjoint functions for sensitivity analysis.
NEWT
----