LA-UR-08-07182

*Approved for public release;*

*distribution is unlimited.*

*Title:*

*Authors:*

*Intended for:*

MCNPX 2.7.A Extensions

Denise B. Pelowitz

John S. Hendricks

Joe W. Durkee

Michael R. James

Michael L. Fensin

Gregg W. McKinney

Stepan G. Mashnik

Laurie S. Waters

Los Alamos National Laboratory

[http://mcnpx.lanl.gov](http://mcnpx.lanl.gov/)

![](media/image1.png){width="2.4069444444444446in" height="0.975in"}

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**MCNPX 2.7.A Extensions**

by

Denise B. Pelowitz

John S. Hendricks

Joe W. Durkee

Michael L. Fensin

Michael R. James

Gregg W. McKinney

Stepan G. Mashnik

Laurie S. Waters

Los Alamos National Laboratory

November 6, 2008

Copyright 2008. Los Alamos National Security, LLC. All rights reserved.

This material was produced under U.S. Government contract
DE-AC52-06NA25396 for Los Alamos National Laboratory, which is operated
by Los Alamos National Security, LLC for the U.S. Department of Energy.
The Government is granted for itself and others acting on its behalf a
paid-up, nonexclusive, irrevocable worldwide license in this material to
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acting on its behalf a paid-up, nonexclusive, irrevocable worldwide
license in this material to reproduce, prepare derivative works,
distribute copies to the public, perform publicly and display publicly,
and to permit others to do so.\
\
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NOR LOS ALAMOS NATIONAL SECURITY, LLC, NOR ANY OF THEIR EMPLOYEES, MAKES
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**TABLE OF CONTENTS**

[1 Introduction 1](#introduction)

[1.1 Summary of New MCNPX Features 1](#summary-of-new-mcnpx-features)

[1.2 Availability 1](#availability)

[1.3 Documentation 2](#documentation)

[1.4 Cash Awards Quality Assurance Program
2](#cash-awards-quality-assurance-program)

[2 Description of new MCNPX Features
3](#description-of-new-mcnpx-features)

[2.1 Tally Tagging 3](#tally-tagging)

[2.1.1 User Interface 3](#user-interface)

[2.1.2 Usage and Special Cases 4](#usage-and-special-cases)

[2.1.3 Examples 5](#examples)

[2.2 CEM 3.02 Upgrade^,^ 6](#cem-3.02-upgrade)

[2.3 MCPLOT Graphics and "HELP" Command Upgrades
8](#mcplot-graphics-and-help-command-upgrades)

[2.3.1 Graphics Upgrades 8](#graphics-upgrades)

[2.3.2 "HELP" Command Upgrade 10](#help-command-upgrade)

[2.4 Embedded Sources 12](#embedded-sources)

[2.4.1 User Interface 12](#user-interface-1)

[2.4.2 Examples 12](#examples-1)

[2.5 Stopping Power Energy Table Control
15](#stopping-power-energy-table-control)

[3 MCNPX 2.7.A Feature EXTENSIONS and Enhancements
15](#mcnpx-2.7.a-feature-extensions-and-enhancements)

[3.1 Spontaneous Photon Automatic Weight Adjustment
16](#spontaneous-photon-automatic-weight-adjustment)

[3.2 Burnup Enhancements 18](#burnup-enhancements)

[3.3 Lattice FILL Error Trap 19](#lattice-fill-error-trap)

[3.4 LINUX Static Build 19](#linux-static-build)

[3.5 Compilation without Graphics 19](#compilation-without-graphics)

[3.6 Minor Enhancements: 20](#minor-enhancements)

[4 MCNPX Corrections 20](#mcnpx-corrections)

[4.1 Significant Problem Corrections
20](#significant-problem-corrections)

[4.1.1 Minor Burnup Reaction Rates Off
20](#minor-burnup-reaction-rates-off)

[4.1.2 Spontaneous Fission Normalization Error
20](#spontaneous-fission-normalization-error)

[4.1.3 DXTRAN Fail with Repeated Structures
20](#dxtran-fail-with-repeated-structures)

[4.1.4 Heavy Ion Errors 21](#heavy-ion-errors)

[4.1.5 S(α,β) File Reading Failure 21](#sαβ-file-reading-failure)

[4.1.6 Burnup Continue Runs 21](#burnup-continue-runs)

[4.2 Minor Problem Corrections 21](#minor-problem-corrections)

[4.2.1 INTEL Compiler Workaround 21](#intel-compiler-workaround)

[4.2.2 LAQGSM Corrections 21](#laqgsm-corrections)

[4.2.3 Compiler Workaround 21](#compiler-workaround)

[4.2.4 Detector Tallies in Lattice Geometries
22](#detector-tallies-in-lattice-geometries)

[4.2.5 Burnup Duplicate ZAIDs 22](#burnup-duplicate-zaids)

[4.2.6 Source Particle Types Unrecognized in Dependent Distributions
22](#source-particle-types-unrecognized-in-dependent-distributions)

[4.2.7 Incorrect Warning Error for Detector Tallies
22](#incorrect-warning-error-for-detector-tallies)

[4.2.8 Choice of High-Energy Mode 22](#choice-of-high-energy-mode)

[4.2.9 Lost Particles Cause Crash 22](#lost-particles-cause-crash)

[5 Future work 22](#future-work)

**MCNPX 2.7.A Extensions**

**ABSTRACT**

> MCNPX is a Fortran90 Monte Carlo radiation transport computer code
> that transports nearly all particles at nearly all energies for nearly
> all applications. The new capabilities of the latest version,
> MCNPX 2.7.A, beyond those in MCNPX 2.6.0, are described.

# Introduction

MCNPX (MCNP eXtended) is a Fortran90 (F90) Monte Carlo radiation
transport computer code that transports nearly all particles at nearly
all energies. It is a superset of MCNP4C3 and has many capabilities
beyond MCNP4C3. MCNPX is a production computer code that models the
interaction of radiation with matter. Its quality is highly regarded and
cash awards (described in Section 1.4) are offered for bugs.

## Summary of New MCNPX Features

The new capabilities and enhancements of MCNPX 2.7.A beyond MCNPX 2.6.0
are listed below. (Where applicable, the initials of the principal
developers are shown in parentheses.[^1]) Note that the delayed-gamma
emission and criticality source convergence acceleration capabilities
are developmental. They are not backed by the MCNPX cash awards quality
assurance program and must be used cautiously. However, these
capabilities are considered of such significance that they are included.
Recall also that as of version 2.6.0 total ν (not prompt ν) is the
default for fission neutrons. The new MCNPX 2.7.A capabilities include
the following:

-   Tally tagging (GWM)

-   CEM 3.02 upgrade (MRJ/SGM)

-   Plot appearance upgrades (JWD)

-   Embedded sources (JSH)

-   Stopping power energy table control (GWM)

## Availability

The latest general-release version is MCNPX 2.6.0, which is available
from the Radiation Safety Information Computational Center (RSICC)[^2]
and the Office of Economic Cooperation and Development/Nuclear Energy
Agency (OECD/NEA) in Paris, France.[^3]

For approved users, beta test program versions may be downloaded from
the MCNPX website.[^4] Approximately 2000 MCNPX beta test users are an
essential part of the MCNPX software quality assurance program. In
general, to become a new beta test user usually requires being an MCNPX
sponsor, collaborator, or U.S. citizen participant in an MCNPX workshop
training course. For more information, contact us at the following
address: mcnpx@lanl.gov.

MCNPX is U.S. export controlled. Generally, MCNPX is easy to obtain for
legitimate purposes, but restrictions protect the code, data, and
documentation once it is acquired.

## Documentation

*Restricted Documentation:*

Complete MCNPX documentation is available in the **"**MCNPX™ User's
Manual,"[^5] which comes with MCNPX from RSICC or OECD/NEA and is
available to MCNPX beta testers from the Worldwide Web (WWW).[^6]

*Unrestricted Documentation:*

Documentation of MCNPX features in MCNPX 2.6.0---that is, all MCNPX
features beyond MCNPX 2.3.0 and MCNP4C---is provided in "MCNPX
Extensions, Version 2.6.0."[^7]

MCNPX capabilities developed since MCNPX 2.6.0 are documented in this
report.

All of the extended MCNPX capabilities are summarized in the one-page
MCNPX features list.[^8]

## Cash Awards Quality Assurance Program

We are so confident in the quality of MCNPX that John Hendricks offers
cash awards to the first person who finds any bug in the code. He will
pay \$20 to the first person finding anything that does not meet or
exceed the capabilities of MCNPX 2.3.0 and MCNP4C3. European users will
be awarded €20. He also will pay a brand new \$2 bill for any error in
MCNPX that has been inherited from its constituent codes.[^9] A list of
all winners since June 2001 can be found on the MCNPX WWW site.[^10]

# Description of new MCNPX Features

Note that the delayed-gamma emission and criticality source convergence
acceleration capabilities are developmental. These capabilities are not
backed by the MCNPX cash awards program and must be used cautiously.
Nevertheless, they are considered of such significance that they are
included.

The principal new capabilities of MCNPX beyond MCNPX 2.6.0 described in
detail in this section are

-   Tally tagging

-   CEM 3.02 upgrade

-   Plot appearance upgrades

-   Embedded Sources

-   Stopping power energy table control

## Tally Tagging 

Tally tagging, a new feature in v.2.7.a, provides the ability to
separate a tally into components based on how and where the scoring
particle was produced. This feature is available for both standard and
detector tallies; however, it is only implemented for neutron, photon,
and electron tallies.

### User Interface

Tally tagging is specified by using the new keyword, TAG, with its
associated parameter, *a*, on the FT*n* tally special treatment card,
which is associated with a standard F*n* tally. The format is

FT*n* TAG *a*

where *n* is the neutron or photon tally number, and *a* specifies how
scatter is to be treated (i.e., whether the creation tag on a particle
should be retained or a separate scatter tag be invoked):

> *a*=1 indicates that all collided particles will lose their tag and
> that bremsstrahlung and annihilation photons will be included in the
> bin of collided particles;
>
> *a*=2 indicates that all collided particles will lose their tag, but
> that bremsstrahlung and annihilation photons will be given special
> tags that allow them to be segregated; and
>
> *a*=3 indicates that all collided particles will retain their
> production tag.

Binning specifications for the tagged tally must be provided on the FU
special tally card. Each *bin*~i~ (described below) requests three
distinct pieces of tagging information:

> 1\) a cell of interest where particles are produced;
>
> 2\) a target nuclide from which the particle is emitted; and
>
> 3\) a reaction or, in the case of spallation, a residual nuclide of
> interest.

The format on the FU card when used in association with the tagging
treatment is

FU*n* *bin*~1~ *bin*~2~ \... *bin*~N~

where each tagging *bin*~i~ has the form CCCCCZZAAA.RRRRR. A user cell
number is designated by CCCCC (leading zeros are not required); ZZAAA
represents a five-digit isotope identifier for a target nuclide, where
ZZ represents the atomic number and AAA represents the atomic mass
number and RRRRR specifies a reaction identifier for library
interactions or a residual nuclide ZZAAA identifier for high-energy
model interactions. Special cases for the CCCCCZZAAA.RRRRR FU card
entries are provided in Section 2.2.2. Note that the ZZ atomic number
specification is limited to two characters; therefore nuclides with
Z\>99 cannot be tagged. By default, a total over all specified bins is
provided automatically for the FU special tally. To suppress the
generation of a total, add the "NT" parameter after the last specified
bin.

### Usage and Special Cases

If cell tagging is not desired, the CCCCC portion of the tag should be
omitted or, alternatively, set to "00000". In either case, tally
contributions will be accumulated for all cells for that FU bin,
provided the ZZAAA.RRRRR portion of the tag is satisfied. In the case of
particle production from electrons, which are material based (not
nuclide specific), the CCCCC input should be used to identify the cell
and the ZZAAA input should be set to "00000". The suffix RRRRR refers to
a standard ENDF reaction number for library interactions (e.g., "00102"
stipulates \[n,gamma\]) or, in the case of high-energy model
interactions, RRRRR refers to a residual nuclide ZZAAA identifier (e.g.,
"06012" for ^12^C). Additional library-related reactions numbers can be
found in Appendix G of the MCNPX User's Manual.

In general, a zero input for any portion of the tag results in the sum
of all contributions related to the entry. For example, the tag
"0000092000.00000" will collect all tally contributions where any
isotope of uranium (z=92) had produced the particle making the tally.
However, the tag "0000000000.00000" is reserved for scattered particles.
Note that each tally contribution is made only to the first FU bin that
satisfies the tag description (i.e., those that have not already been
tallied). If no appropriate FU bin is found, the tally contribution is
not made; however a special "everything else" *bin*~i~ (i.e., "1e10")
can be specified to collect any portion of the tally that falls into no
other bin. When the "everything else" bin is used, then the user is
assured that the "user-bin total" bin will reproduce the original tally
as if the FT*n* TAG option has not been used.

Special designations for CCCCCZZAAA:

-0000000001 or -1 source particle tag for all cells

-CCCCC00001 source (i.e., uncollided) particle tag for cell CCCCC

0000000000 or 0 scattered particle tag

10000000000 or 1e10 everything else tag

Photon tally special designations for ZZAAA.RRRRR:

00000.00001 bremsstrahlung from electrons

ZZ000.00003 fluorescence from nuclide ZZ

00000.00003 K x-rays from electrons

00000.00004 annihilation photons from e^-^

ZZ000.00005 Compton photons from nuclide ZZ

ZZAAA.00006 muonic x-rays from nuclide ZZAAA

Electron tally special designations for ZZAAA.RRRRR:

ZZ000.00001 photoelectric from nuclide ZZ

ZZ000.00003 Compton recoil from nuclide ZZ

ZZ000.00004 pair production from nuclide ZZ

ZZ000.00005 Auger electron from nuclide ZZ

00000.00005 Auger electron from electrons

00000.00006 knock-on electrons

Neutron and photon tally special designations for ZZAAA.RRRRR:

ZZAAA.99999 delayed particles from fission of ZZAAA

### Examples

Example 1:

F1:N 10

FT1 TAG 1

FU1 0000092235.00016 0000092235.0000 1e10

If an \[n,2n\] neutron that is produced from an interaction with ^235^U
contributes to the F1 tally, then its contribution will be included only
in the first FU bin even though its tag also will satisfy the criteria
for the 2^nd^ FU bin. Thus, the order of the FU bin tags is important
for segregating the tally. Note in this example that neutrons produced
by some other reaction with ^235^U will be placed in the 2^nd^ FU bin
and neutrons produced by reactions with other target nuclides will be
placed in the last ("everything else") bin. The sum of these three bins
should preserve the value of the original F1:N tally.

Example 2:

F1:P 1

FT1 TAG 1

FU1 0.0 01001.00102 01001.00000

26056.00102 26056.00051 26056.00052

26056.24052 26056.26053 26056.26054 26056.26055

26056.00000

In this case, all collided photons will be put into the FU "0.0" bin.
All capture gammas from ^1^H will be put into the 01001.00102 bin; all
remaining gammas from ^1^H interactions will be put into the 01001.00000
bin. All capture gammas from ^56^Fe will go into the 26056.00102 bin;
all \[n,n'\] 1^st^ level gammas will go into the 26056.00051 bin; all
\[n,n'\] 2^nd^ level gammas will go into the 26056.00052 bin; all
deexcitation gammas from the spallation of ^56^Fe into ^52^Cr will go
into the 26056.24052 bin; etc. All remaining gammas produced from ^56^Fe
interactions will go into the 26056.00000 bin.

Example 3:

F5:P 0 0 0 1

FT5 TAG 3

FU5 -1.0 0000106012.00005 0000106012.00000

0000026056.00102 0000026056.00000

0000000000.00051

10000000000.00000

In this case, all collided photons will retain their original creation
tag. All source photons will go into the -1.0 bin. All Compton photons
from ^12^C in cell 1 will be put into the 2^nd^ bin; all remaining
photons produced from interactions with ^12^C in cell 1 will go into the
3^rd^ bin. All capture gammas from ^56^Fe will go into the 4^th^ bin;
all remaining photons/gammas produced from interactions with ^56^Fe will
go into the 5^th^ bin. All \[n,n'\] 1^st^ level gammas will be put into
the 6^th^ bin, and all remaining photons/gammas that were not included
in any of the previous bins will be placed in the last bin.

## CEM 3.02 Upgrade[^11]^,^[^12]

In versions 03.02 of CEM and LAQGSM, the Fermi breakup model is used at
all stages of a reaction for an excited nucleus with atomic mass A\<13.
Previously, in versions 03.01 of these codes, the Fermi breakup model
was used to calculate the disintegration of light nuclei but was not
used at the preequilibrium, evaporation, and fission stages for an
excited nucleus or a fission fragment with A\<13.

Also, the computer routines that describe the Fermi breakup model,
written by the original authors in Russia twenty years ago, historically
have had some occasional odd behavior. These problems have been
corrected in version 03.02. First, no longer do these routines allow the
rare production of some light unstable fragments like ^5^He, ^5^Li, 8Be,
and ^9^B as a result the breakup of some light excited nuclei. Also no
longer permitted is the very rare production of "neutron stars" or
"proton stars," that is, the production of residual nuclei produced via
Fermi breakup that consist of only neutrons or only protons. Code
crashes caused by an occasional division by zero within the Fermi
breakup routines have been eliminated as have several other coding bugs.

CEM 3.02 was integrated into MCNP6 and MCNPX by H. Grady Hughes and
Richard E. Prael. The 03.02 versions of CEM and LAQGSM more correctly
describe nuclear reactions on intermediate and light nuclei and the
production of fragments heavier than ^4^He from heavy targets than did
their predecessors. Additionally, these new versions produce very few
unstable unphysical final products and are free of the corrected bugs.

The file mollinix.tbl is now in the MCNPX data files and is required for
CEM 03.02.

A graphical comparison of Bertini, CEM 03.01, and CEM 03.02 appears in
Figures 1 and 2. The plots are results of neutron flux near a proposed
tungsten target at the LANL LANSCE facility. The driver for the plotted
flux was an 800‑MeV proton beam. Overall, the energy spectrum show very
good agreement between the CEM versions, but is not as good with
Bertini.

![](media/image2.png){width="5.572916666666667in"
height="3.748611111111111in"}

Figure 1. Neutron flux generated by an 800‑MeV proton beam in LANSCE
material samples. (Note that CEM 03.03 is mislabeled as "cem303' in the
plot.)

![](media/image3.png){width="5.572916666666667in"
height="3.754861111111111in"}

Figure 2. Neutron flux generated by an 800‑MeV proton beam in LANSCE
fuel sample. (Note that CEM 03.03 is mislabeled as "cem303' in the
plot.)

## MCPLOT Graphics and "HELP" Command Upgrades 

MCPLOT has been upgraded, providing improved appearance of graphics and
an enhanced help package. MCPLOT plots table-range cross-section data,
tallies, and other quantities such as convergence criteria and
criticality parameters. Among these improvements are the following:

-   Significantly improved help package with alphabetized commands and
    the capability of retrieving short descriptions of each command with
    its options

-   Capitalized fonts and capitalization of code-supplied titles and
    quantities. Among other things, "mev" is now "MeV"

-   Horizontally displayed y-axis labels

-   Enable 70-character plot titles

-   Increase the limit of orders of magnitude for tally and other data
    from 17 to 35. This allows time phenomena from nanoseconds to years
    can be placed on a log plot.

-   Allow log10 axes for contour plots

### Graphics Upgrades 

Among the graphical improvements are the following:

*Improved readability and flexibility of axis values*: Ordinate values
are now displayed horizontally and the default print density of the
numbering on each axis was adjusted to eliminate overcrowding. This
default density of the axes tick values can be overridden by the user
using the *nsteps* parameter on the XLIMS and/or YLIMS command. For
logarithmic plots, larger tick marks now signify major decades while
smaller marks indicate values between decades. The number of decades
that can be displayed on a logarithmic plot has been increased from 17
to 35. This increase allows time phenomena from nanoseconds to years to
be placed on a logarithmic plot. Also, use of the "E" format to
represent scientific notation decreases the tick label length and
consequently enhances readability.

*Axes and legend labels*: For energy plots, the default units now read
"MeV" instead of "mev" and the first letter of each word in default axis
labels is now capitalized. Character limits for labels have been
increased: up to 23 characters can now be displayed in the legend label
and 70‑character-long main titles are now permitted for both screen and
PDF plots.

*Information and legend blocks*: The first letter of the left-aligned
tally bin type is now capitalized and the *probid* is included in the
information block to explicitly indicate the date and time stamp of the
runtpe or mctal file instead of the build date of the executable.

Figure 3 illustrates several plot upgrades, including (1) longer titles,
(2) legend problem identification (*probid*:), (3) alignment of legend
tally type labels, (4) increased length of legend curve identifiers,
(5) capitalized letters for axes and legend labels, (6) horizontal
display of ordinate numbers, and (7) wider major ticks for logarithmic
axes.

![](media/image4.jpeg){width="5.329861111111111in"
height="3.5006944444444446in"}

Figure 3. Upgraded MCPLOT 2-D tally graphics.

Figure 4 illustrates (1) 200-decade logarithmic-axis plot capability,
(2) wider ticks at major logarithmic-axis decades, (3) automated
refinement of the number of minor ticks, (4) automate refinement of the
number density on logarithmic axes.

![](media/image5.png){width="4.691666666666666in"
height="3.497916666666667in"}

Figure 4. Upgraded MCPLOT cross-section plot graphics.

*Contour plots*: Contour plots may now be created using logarithmic
axes. In addition, contour-plot legend information now ranges from low
to high as one moves up the page.

Figure 5 illustrates (1) the use of logarithmic-axis plotting for
contour plots and (2) legend decade identifiers ranging from low to
high, moving up the page.

![](media/image6.png){width="4.754166666666666in"
height="3.4972222222222222in"}

Figure 5. MCPLOT log-scale contour plot.

### "HELP" Command Upgrade 

The MCNPX MCPLOT "HELP" command has been upgraded to improve its
functionality. The three-column listing of HELP commands is now
alphabetized. In addition, tips appear at the bottom of the three-column
list of HELP commands to indicate how to

1)  invoke a listing of all HELP commands with an explanation of their
    function and use syntax ("HELP ALL"),

2)  provide a listing of function and syntax for a single HELP command
    ("HELP \<*command*\>"),

3)  request an overview of the MCPLOT capability ("OVERVIEW"), and

4)  summarize input and execution-line options ("EXECUTE").

To further aid the user, some commands are now cross-referenced to guide
the user toward related commands that may be of interest.

Example 1: After invoking the MCPLOT utility, if you enter the command
"HELP", you will receive the following response:

> mcplot\>
>
> help
>
> Available commands:
>
> bar loglog subtitle
>
> below mt tally
>
> contour noerrbar term
>
> coplot factor tfc
>
> dump options thick
>
> end par thin
>
> factor pause title
>
> file pert view
>
> fixed plinear wmctal
>
> free plot xlims
>
> freq printal xs
>
> help printpts xtitle
>
> hist reset ylims
>
> iptal return ytitle
>
> kcode rmctal ztitle
>
> label runtpe &
>
> legend scales ?
>
> linlin set 3d
>
> linlog spline
>
> loglin status
>
> Type \"help all\" for a verbose list of all help commands
>
> \"help \<command\>\" to list a specific help command,
>
> \"help overview\" for an overview of MCPLOT,
>
> or \"help execute\" for MCPLOT input & execution-line
>
> options.
>
> mcplot\>

Example 2: After invoking the MCPLOT utility, if you enter the command
"HELP legend", you will receive the following response:

> mcplot\>
>
> help legend
>
> legend \> Syntax: legend \[x y\]
>
> Include or omit the legend according to the values of
>
> optional parameters x and y.
>
> If no x and no y, put the legend in its
>
> normal place. (DEFAULT)
>
> If x=0 and no y, omit the legend.
>
> If both x and y are defined, for 2-D plots only,
>
> Put most of the legend in its usual place,
>
> but put the part that labels the plot lines at
>
> location x,y.
>
> Units for x and y are the units used for
>
> the x and y axes.
>
> The maximum legend curve label length is
>
> 23 characters.
>
> See \"label\".
>
> mcplot\>

## Embedded Sources 

Source distributions now can be embedded within each other to describe
accelerator micro-pulses and other sources that vary regularly.

### User Interface

The format to specify an embedded source is

SDEF TME = ( D11 \< D12 \< D13 )

or, for distributions of distributions, the following form may be used:

SDEF TME = D41

SI41 S 51 ( D11 \< D12 \< D13 ) 52

In both cases, distributions 11, 12, 13 are all for the same variable,
time. Distribution 11 covers a small time range that is repeated as
often as needed to exactly fill the larger time range of
distribution 12. Similarly, distribution 12 is repeated as often as
needed to exactly fill the even larger time range of distribution 13.

Note that the parentheses are optional and that the designator "D" on
the SI card with "S" option is also optional. Thus

SDEF TME = ( D11 \< D12 \< D13 ) and

SDEF TME = D11 \< D12 \< D13 are equivalent.

Also,

SI41 S 51 ( D11 \< D12 \< D13 ) 52 ,

SI41 S 51 D11 \< D12 \< D13 52 ,

SI41 S 51 ( 11 \< 12 \< 13 ) 52 , and

SI41 S 51 11 \< 12 \< 13 52 are all equivalent.

The embedded distributions must start at zero or a fatal error message
is issued. For (D11\<D12\<D13) the lowest value on the SI11 and SI12
cards must be zero. The embedding distribution, D13, can have any range.

The embedded distributions should fit within each other (nearly)
exactly. If they do not there is a fatal error message, "embedded
distribution *nn* has improper range" and the distribution will spill
into the next bin and have a strange normalization for values in its
last bin.

Only continuous source distributions such as ERG, TME, X, Y, Z, DIR, RAD
and EXT may use embedded distributions.

### Examples

Example 1:

> sdef tme= d11\<d12\<d13
>
> si11 0 1 2
>
> sp11 0 1 0
>
> si12 0 100 150
>
> sp12 0 1 0
>
> si13 -150 900
>
> sp13 0 1

Figure 6 shows the MCPLOT plot (MCNPX  Z option) of the tally of this
source in a vacuum during the time regime ranging from -200 to 1000
shakes. The seven repetitions of distribution 12 (which is described in
SI12 as ranging from 0 to 150 shakes) are clearly seen embedded within
distribution 13 (which describes the time period from -150 to 900
shakes.)

Figure 7 shows the MCPLOT plot from -5 to 25 shakes. The 1^st^ 13 of a
total 50 repetitions of distribution 11 (which are each two shakes in
duration) are seen in the first part of the 2^nd^ repetition of
distribution 12.

![](media/image7.png){width="5.134722222222222in" height="3.5in"}

Figure 6. MCPLOT plot of tally from -200 to 1000 shakes shows embedded
pulses defined by distribution 12.

![](media/image8.png){width="5.134722222222222in" height="3.5in"}

Figure 7. MCPLOT plot of tally from -5 to 25 shakes, displaying embedded
pulses defined by distribution 11.

Example 2 (embedding a distribution of distributions):

> sdef erg d21 cel=d11 tme=d41
>
> si41 S 52\<51 (D31\<32\<d33) 61
>
> sp41 .1 .8 .1
>
> si51 A -26 -16
>
> sp51 0 1
>
> si52 H 0 1 2
>
> sp52 0 1 0
>
> si61 A 32 40
>
> sp61 1 0
>
> si31 0 1 2
>
> sp31 0 1 0
>
> si32 0 16
>
> sp32 -41 8 8
>
> si33 -16 32
>
> sp33 0 1

This more complex example illustrates how embedded distributions can
reside within distributions of distributions (D41), and can use built-in
functions (D32 uses a Gaussian centered at t=8 with FWHM=8) and
interpolated distributions (D51 and D61 use the SI  A option).
Distribution D52 is embedded in distribution D51; distribution D31 is
embedded in distribution D32, which is embedded in distribution D33. The
tally plot is in Figure 8.

![](media/image9.png){width="5.121527777777778in" height="3.5in"}

> Figure 8. MCPLOT plot of tally from -30 to 50 shakes.

## Stopping Power Energy Table Control 

A new 11th entry on the PHYS:\<*pl*\> card, where \<*pl*\> is any
charged particle type, now controls the stopping power table energy
spacing. This 11^th^ entry is the parameter *efac*. When *efac* is
specified, the energy fraction (*efac*) for stopping power tables is the
following:

E~n-1~ = E~n~ \* *efac*

This applies to charged particles only. The value of *efac* is limited
to 0.8≤*efac* ≤0.99. A larger *efac* produces more points in the
stopping power tables. Default is *efac* =0.917.

# MCNPX 2.7.A Feature EXTENSIONS and Enhancements 

Some MCNPX features have been extended and have included changes or
additions to the user interface. Where appropriate, initials of the
principal developer(s) are provided in parentheses:[^13]

-   spontaneous photon automatic weight adjustment (MRJ/LSW/GWM)

-   burnup enhancements (MLF)

-   lattice FILL error trap (JSH)

-   LINUX static build (GWM)

-   Compilation without graphics (GWM/MLF)

-   Minor enhancements

## Spontaneous Photon Automatic Weight Adjustment

The decay (or spontaneous) gamma source is a feature that allows the
user to easily create a gamma spectra source associated with the decay
of specified isotopes. With the implementation of the CINDER database
and a gamma-line database, MCNPX can now lookup the appropriate activity
and decay lines associated with a large number of isotopes. For many
other isotopes, a multi-group treatment is available.

*Spontaneous material source*: The user can build a material that
includes unstable isotopes and specify their appropriate fractions. In
this case---analogous to the "SF" particle identification feature---the
"SP" particle type can be specified on the SDEF card. The decay gammas
will then be created based the unstable isotopes found in the material
card at the sampled source location. If multiple unstable isotopes are
present, they will be sampled based on their relative activities
*within* that material.

Example 1:

> m1 1001 1
>
> 8016 1
>
> 7016 1e-4 \$ Unstable isotope N-16
>
> 25054 1e-2 \$ Unstable isotope Mn-54
>
> sdef par=sp pos= 0 0 0 \$ Location of material 1
>
> phys:p 5j -102

When the SF source is specified, the code will create an optional
table 44 that summarizes the activities of the materials in the problem.

*Spontaneous gamma particle specification*: The other way to start decay
gammas builds upon the heavy ion feature in MCNPX. The user can specify
a Z\*1000+A as a particle type on the SDEF card and set the energy of
that particle to zero. This will signal the code not to transport the
heavy ion but to start its decay gammas. Note that in this case, the '#'
(heavy ion) *must* appear on the mode card.

Example 2:

> mode p \#
>
> sdef par=7016 erg=0 pos= 0 0 0
>
> phys:p 5j -102

*Features and defaults*: The decay gammas that are emitted are assumed
to arise from the instantaneous activity of a large pool of decaying
isotopes. So the time behavior will assume either whatever is specified
on the SDEF card or the default of TME=0.

The sampled heavy ion isotopes will appear in Table 110 as the source
particle although the created gamma lines do not.

It is also possible to do distributed sources with either of these
options, with both options together or in combination with other sources
in MCNPX.

As is the case with other source particles, be aware that if an isotope
emits multiple gamma lines, the emissions will not necessarily be
correlated.

Also note that isotopes with half-lives longer than 1x10^18^ seconds
(\~3.17x10^10^ years) are treated as stable.

*Exponential decay function*: A new source probability function has been
added that allows users to create a source with an exponential decay
shape. The special function in invoked on the associated source
probability (SP*n*) card by a "-7" followed by the half-life specified
in units of shakes. The default half-life is 1 shake.

Example 3:

> sdef tme=d1
>
> sp1 -7 2e8 \$ 2e8 shakes = 2 seconds

*SPn W source probability option*: A new source probability option has
been added. The "SP*n* W" is a distribution specification that can only
be applied to particle distributions. This distribution allows the user
to specify actual intensities for a mix of particle sources. The
intensities will be normalized per standard operation of the source,
however the factor used to renormalize the intensities will be applied
to the source weight to give the tallies the correct magnitude.

Example 4:

> si1 n p h
>
> sp1 w 3e9 5e9 2e9

The source shown here will create three sources and sample them by the
relative magnitudes (neutrons 30%, photons 50%, protons 20%). The weight
assigned to each particle will be the sum of the non-normalized values,
3e9+5e9+2e9=1e10.

The second aspect of the "SP*n* W" option is to allow SP (spontaneous
photon) and SF (spontaneous fission) sources to be assigned magnitudes
automatically in the code. A negative number corresponding to one of
these two source types tells MCNPX to use the cell volume represented by
the cell number that follows for the computation of the intensity.

Example 5:

> si1 sp sf n
>
> sp1 w -10 -15 3e9
>
> In the source described here, the spontaneous photon source will look
> to cell 10 and use the material and volume to calculate the overall
> activity that will be substituted into the SP1 distribution.
> Correspondingly, the SF source will look to the material and volume in
> cel 15 for the intensity of the spontaneous fission source (note that
> a +SF normalizes per spontaneous fission neutron, a ‑SF normalizes per
> spontaneous fission). The neutron source is unchanged.

Once the overall activity is computed, the renormalization will be done
as before and the value passed into the weight parameter.

This option will also work with the SB (source biasing) option where
sampling frequencies can be adjusted with the weights to bias the
distributions.

If the cell specified in the SP*n* W option is a lattice cell, then the
code may not know the correct volume for this cell. A volume can be
supplied by a VOL card or with a VOL keyword on the cell card. If the
user does not wish to correct the volume in these ways, a WGT keyword
can be used with the source as a multiplicative factor.

Example 6:

> sdef par=d1 wgt=264
>
> si1 L sf
>
> sp1 w -35
>
> In this example, the spontaneous fission source is weighted by the
> activity from cell 35 which has been duplicated 264 times in the
> geometry. The final source weight will be the activity from cell 35 \*
> 264.

## Burnup Enhancements 

*Parallelizing the CINDER90 Interface*: The burnup capability in
previous versions of the code required that MCNPX calculate steady-state
reaction rates (when transport cross sections were available) and
63‑group fluxes to generate depletion coefficients for the
time-dependent CINDER90 burnup calculation. Because MCNPX was already
parallelized, the steady-state reaction-rate calculation was also
parallelized in MCNPX 2.6.B to execute in multiprocessor mode. However,
at the end of each steady-state reaction-rate calculation each slave
processor had to rendezvous with the master processor to perform the
interface/CINDER90 subroutines in a serial execution.

Several independent serial looping structures within the burnup
interface allow the assemblage of depletion coefficients and isotopic
information into a format suitable for CINDER90 execution. The
computational expense associated with this serial execution increases
with the number of burn materials requested in the burnup calculation.
For example, burning a single material may only require one minute of
interface/CINDER90 execution time; however, burning 150 materials may
require 300 minutes.

Because the depletion of each burn material is an independent
calculation, the computational cost associated with the
interface/CINDER90 execution was remedied by parallelizing the
interface/CINDER90 routines for multi-processor execution. The
interface/CINDER90 calculation is now distributed across all available
processors including the master processor. Each processor executes the
interface/CINDER90 routines for a subset of burn materials and then
rendezvous at the end of the interface/CINDER90 routines to determine
total system burnup as well as atom densities for the subsequent
reaction-rate calculation. Because the interface/CINDER90 routines are
independent for each burn material and minimal communication across
processors is necessary to execute the interface/CINDER90 subroutines,
the computational enhancement scales almost linearly with the number of
processors.

*Variable Dimensioning for Burnup and Spontaneous Photons*: The
dimensions of spontaneous photon and burnup arrays was made variable so
that these arrays only take up computer memory space when they are used
for these features. Certain arrays for the burnup and spontaneous photon
capability were automatically dimensioned by the mix variable for any
type of calculation. This unnecessary dimensioning wasted memory, thus
limiting the size of calculations that could be examined by the code.
All the arrays for the burnup capability are now dimensioned by unity if
a BURN card is not present; similarly, all arrays for the spontaneous
photon capability are dimensioned by unity if the spontaneous-photon
capability is not used (MLF).

*Unlimited CONC Changes*: Additional dynamic dimensioning now permits an
unlimited number of user-specified concentration changes (CONC option on
BURN card.) Users implementing the MATMOD keyword in MCNPX 2.6.A(0)
received strange results when trying to manually modify more than three
nuclides per time step. Users may now manually modify as many nuclides
as exist in the burn material. Nuclides that make up the burn material
are included from the material card, fission product tier, and isotope
generator algorithm (MLF)

## Lattice FILL Error Trap

Errors in a fully specified lattice FILL are now captured with a "Bad
Trouble" error. This is not a bug fix because it only happens with user
errors when setting up geometries. Previously, an erroneous lattice fill
would go undetected and usually cause a crash or lost particles, but
sometimes cause wrong answers.

Consider the following:

FILL 0:2 0:1 0:0 1 1 1 1 1 1 2

The fully specified FILL describes a 3 x 2 x 1 lattice, so it should be
followed by six universes. Instead there are seven. The new lattice fill
error trap now detects either too few or too many universes.

## LINUX Static Build 

The MCNPX executables provided for LINUX and built with the Intel
compiler have been compiled so that users can use executables without
having to install a compiler. To make LINUX Intel static builds simply
add

> \--with-LDFLAGS=-static-libcxa

to the configure line when compiling.

## Compilation without Graphics 

MCNPX may now be built without graphics on Windows (using both Intel 9.1
and 10.1) for MPI parallel versions. The MCNPX C code was revised to
omit the use of TIMEVAL and GETTIMEOFDAY() and to use instead the C++
intrinsic function FTIME.

## Minor Enhancements:

-   *E* \> *1 TeV / nucleon Warning:* The warning for source particles
    with E \> 1 TeV/nucleon was improved. Suggested by Robert
    Singleterry (NASA) (MRJ/GWM)

-   Removed DBCN(20) which caused tracking with MCNPX 2.5.0 (JSH)

-   Removed DBCN(21) which caused tracking with MCNPX 2.5.E (JSH)

-   Removed DBCN(25) which turned off the Eric Pitcher / Bob Little
    S(α,β) broadening enhancement in MCNPX2.6.0 (JSH)

-   Deleted the screen print of contour levels during geometry plots
    unless contour levels were set by the user. (JSH)

# MCNPX Corrections

## Significant Problem Corrections

The following problems could cause incorrect answers. Fortunately, they
occurred only in very special situations and affected only a few MCNPX
users.

*MCNPX v.2.7.A corrections to MCNPX v.2.6.0*

### Minor Burnup Reaction Rates Off

Data were loaded incorrectly (off by one energy bin) into the CINDER90
63‑group flux bin structure for determining the contribution of a
specific track traverse for a given burn material cell. The error only
affected burnup/depletion for minor reactions that are treated with the
63‑group fluxes -- not fission, capture, and other major reactions that
are treated continuously or non-burnup problems. The effect is generally
so small as to be unnoticed in benchmark calculations. The error was a
bug in the binary search algorithm in the burn_trk.F subroutine and has
been in all MCNPX versions with burnup. A \$20 award was made to Charles
Whitmer (Whitmer Consulting, North Bend, WA) (X-3:08-77). (MLF)

### Spontaneous Fission Normalization Error

When there is a spontaneous fission source (PAR=SP) and there is more
than one spontaneous fission nuclide in a cell, then the relative
weighting of source strengths was done based upon spontaneous fission
yield data. However, MCNPX uses yield data in gram units and did not
convert to atom units for the relative weighting. This error only
affects spontaneous fission when there is more than one spontaneous
fission nuclide in the cell. Whereas spontaneous fission is only done
with actinides whose mass and atom fractions are similar, the effect is
small and generally within the uncertainty of the yield. The error has
been in all MCNP and MCNPX 2.6.0 and earlier versions. (MRJ)

### DXTRAN Fail with Repeated Structures

Contributions from repeated structure / lattice cells to DXTRAN spheres
are incorrect because the lattice level is not saved. This applies to
all previous code versions including MCNPX 2.6.0 and MCNP5. There is no
problem with only one DXTRAN sphere, but with additional DXTRAN spheres
or point detectors, the lattice level index is wrong and particles can
get lost, cause a crash, or cause silent wrong answers. A \$2 award was
made to Richard Olsher, LANL RP-2 (X-3:JSH-09-007). (GWM)

### Heavy Ion Errors 

The heavy ion ipt is set to the ZA in bankit for banking heavy ions, but
it was not being set back to mipt. This could lead to out-of-bounds
array calls and other errors if the banking routine expected to use the
ipt value after the bankit call. In bankit, the ipt is now restored to
mipt after a heavy ion is banked. (MRJ)

### S(α,β) File Reading Failure 

If S(α,β) data files are mixed in with other types of data files (like
neutron or photon data tables) and if they are read before the other
tables, some cross-section pointers can be incorrect causing wrong
answers. The problem affects MCNPX 2.6.0 and all previous versions. In
one case, *keff* was wrong by a factor of two. Note that none of the
data libraries distributed with MCNP/MCNPX mixes S(α,β) and other data
table types. Therefore, this is only a problem for users who mix data
type when constructing their own libraries of nuclear data tables. (GWM)

### Burnup Continue Runs 

Continue runs of burnup problems that included the MATVOL or MATMOD
keywords did not track and were wrong because arrays *bvol* and *conb*
were not written to the RUNTPE file. These arrays are now written to the
RUNTPE file and continue runs now work with all burnup capabilities.
(MLF)

## Minor Problem Corrections

The following problems do not cause wrong answers, but they may cause
crashes when encountered. In any case, they occurred only in very
special situations and affected only a few MCNPX users.

*MCNPX v. 2.6.0 corrections to MCNPX v.2.6.F*

### INTEL Compiler Workaround 

Large problems failed with the INTEL 10 compiler requiring a workaround.
A \$2 award was made to Manuel Bardies (INSERM, Nantes, France)
(X-3:JSH-2008-050). (GWM)

### LAQGSM Corrections 

*LAQGSM Standalone*: The LAQGSM module compiled as a standalone code
wrote to the wrong unit number. (GWM)

*LAQGSM Infinite Loops*: Several instances of infinite loops in LAQGSM
have been identified and are now resampled rather than having MCNPX
hang. (MRJ)

*Coulomb Barrier Physics*: Coulomb barrier physics has been added for
heavy ions. (MRJ)

*Heavy Ion SPABI*: Secondary particle production biasing, the SPABI
input card option, failed for heavy ions and has been corrected. (MRJ)

### Compiler Workaround 

The floating point variable TDS was used in a 'do loop' in routine
ITALLY. (This is a harmless error or compile error, depending upon the
system.) An award of €20 was made to Frank Gunsing, (CEA/Saclay, France)
(X-3:JSH-2008-045). (JSH)

### Detector Tallies in Lattice Geometries 

If a detector tally contribution is from a cell filled with a lattice,
then the mesh type 4 or point detector tallies did not score. A \$2
award was made to Róbert Kákonyi (University of Szeged, Hungary)
(X-3:JSH-2008-049). (GWM)

*MCNPX v.2.7.A corrections to MCNPX v.2.6.0*

### Burnup Duplicate ZAIDs 

Duplicate ZAIDs on a materials card caused burnup to crash. \$20 to
Gregory A. Johnson (Hamilton Sunstrand--Rocketdyne, Canoga Park, CA)
(X-3:08-60). (MLF)

### Source Particle Types Unrecognized in Dependent Distributions

Specification of different source particles in dependent distributions
is unrecognized. For example, the combination

SDEF PAR=FPOS=D99

and

DS99 S D12 D13

fails and gives spurious warning messages. MCNPX thinks D12 and D13 are
particle types rather than distributions (where the particle types would
then be specified). An award of \$20 was made to Gregory Thoreson (LANL
D-5) (X-3:08-78). (LSW/JSH)

### Incorrect Warning Error for Detector Tallies 

The error message "tally not scored beyond last time bin" for detector
tallies printed the time at the collision rather than the time at the
detector for point/ring detectors. The bug was in all previous MCNP,
MCNP5, and MCNPX versions. A \$2 award was made to Bill Sailor (LANL,
ISR-4) (X-3:08-76). (GWM)

### Choice of High-Energy Mode 

Neutrons, protons, or deuterium colliding with neutrons protons, or
deuterium at high energies should go to FLUKA, not LAQGSM. A \$2 award
was made to Sukesh Aghara (Prairie View A&M University, Texas)
(X-3:JSH-2008-048). (MRJ)

### Lost Particles Cause Crash

See Section 4.1.3 -- DXTRAN fails with repeated structures.

# Future work

The following projects are at least partially funded and are actively
being developed:

-   Additional CINDER90 capabilities:

    -   Burnup / depletion enhancements;

    -   Fixed source depletion;

    -   Automatic production of background activation; and

    -   Automatic production of radioactive sources;

-   Correlated secondary particle production;

-   Improvements in photonuclear cross sections and yields.

The following projects are on our wish list. Some have been partially
developed but await further funding.

-   Improved convergence of eigenfunctions in criticality problems;

-   Integration of HTAPE tallies directly into MCNPX, including continue
    runs;

-   Electric and magnetic field tracking;

-   Detectors and DXTRAN for all neutral particles at all energy ranges
    with anisotropic scattering (currently approximated as isotropic for
    models);

-   CAD link;

-   Secondary particle angle biasing for isotropic distributions;

-   Neutral particle perturbation techniques extended to physics model
    region;

-   Plotting of physics model total and absorption cross sections; and

-   Forced collisions for neutral particles extended to physics models.

[^1]: John S. Hendricks (JSH, LANL, X-3-MCC); Gregg W. McKinney (GWM,
    LANL, X-3-MCC), Joseph W. Durkee (JWD, LANL, D-5), Jay S. Elson
    (JSE, LANL, D-5), Michael L. Fensin (MLF, LANL, N-4), Michael R.
    James (MRJ, LANL, D-5), Russell C. Johns (RCJ, LANL, D-5), Stepan G.
    Mashnik (SGM, LANL, X-3-MCC), Denise B. Pelowitz (DBP, LANL, D-5),
    Laurie S. Waters (LSW, LANL, D-5), Franz X. Gallmeier (FXG, ORNL)

[^2]: <http://rsicc.ornl.gov/>

[^3]: [http://www.nea.fr](http://www.nea.fr/)/

[^4]: <http://mcnpx.lanl.gov/beta/>

[^5]: Denise B. Pelowitz, ed., "MCNPX User's Manual, Version 2.6.0," Los
    Alamos National Laboratory report, LA‑CP-07-1473 (April 2008),
    <http://mcnpx.lanl.gov/docs/MCNPX_2.6.0_Manual.pdf/>.

[^6]: <http://mcnpx.lanl.gov/>.

[^7]: <http://mcnpx.lanl.gov/opendocs/reports/Interface.pdf/>.

[^8]: <http://mcnpx.lanl.gov/opendocs/misc/FeaturesList.pdf/>

[^9]: Cash Award Fine Print: This offer is subject to cancellation or
    modification without notice. A bug is defined as an error we choose
    to correct in the source code. We make awards even for the most
    trivial or insignificant of problems, but not for proposed code
    enhancements or proposed extended capabilities. Awards are given
    only to the first MCNPX user reporting a problem. MCNP/X development
    team members are ineligible. Reported problems must be reproducible,
    and awards are paid when the correction is integrated into a
    forthcoming MCNPX version. We endeavor to make MCNPX the most
    error-free and robust Monte Carlo radiation-transport code possible,
    and we back this code with a cash award.

[^10]: <http://mcnpx.lanl.gov/opendocs/misc/Cashbugs.pdf/>.

[^11]: S. G. Mashnik, R. E. Prael, and K. K. Gudima, \"Implementation of
    CEM03.01 into MCNP6 and its Verification and Validation Running
    through MCNP6. CEM03.02 Upgrade,\" research note X-3-RN(U)-07-03,
    Los Alamos National Laboratory (2007).

[^12]: S. G. Mashnik, K. K. Gudima, R. E. Prael, A. J. Sierk, M. I.
    Baznat, and N. V. Mokhov, \"CEM03.03 and LAQGSM03.03 Event
    Generators for the MCNP6, MCNPX, and MARS15 Transport Codes,\"
    invited lectures presented at the Joint ICTP-IAEA Advanced Workshop
    on Model Codes for Spallation Reactions, February 4--8, 2008, ICTP,
    Trieste, Italy, Los Alamos National Laboratory report LA-UR-08-2931
    (2008); E-print: arXiv:0805.0751v2 \[nucl-th\]; IAEA Report
    INDC(NDS)-0530, Vienna, Austria, August 2008, p. 53.

[^13]: John S. Hendricks (JSH, LANL, X-3-MCC); Gregg W. McKinney (GWM,
    LANL, X-3-MCC), Joseph W. Durkee (JWD, LANL, D-5), Jay S. Elson
    (JSE, LANL, D-5), Michael L. Fensin (MLF, LANL, N-4), Michael R.
    James (MRJ, LANL, D-5), Russell C. Johns (RCJ, LANL, D-5), Stepan G.
    Mashnik (SGM, LANL, X-3-MCC), Denise B. Pelowitz (DBP, LANL, D-5),
    Laurie S. Waters (LSW, LANL, D-5), Franz X. Gallmeier (FXG, ORNL)