MCNPX EXTENSIONS

Version 2.5.0

by

John S. Hendricks

Gregg W. McKinney

Laurie S. Waters

Teresa L. Roberts

Harry W. Egdorf

Joshua P. Finch

Holly R. Trellue

Eric J. Pitcher

Douglas R. Mayo

Martyn T. Swinhoe

Stephen J. Tobin

Joe W. Durkee

Los Alamos National Laboratory

Franz X. Gallmeier

Oak Ridge National Laboratory

Jean-Christophe David

CEA-Saclay, Gif-sur-Yvette, France

William B. Hamilton

HQC Professional Services

Julian Lebenhaft

Paul Scherrer Institute, Switzerland

April 2005

CONTENTS

Page

1.0. Introduction [1](#introduction)

1.1. MCNPX Extensions [1](#mcnpx-extensions)

1.2. Guarantee [3](#guarantee)

1.3. Availability [3](#availability)

2.0. PHYSICS extensions [4](#physics-extensions)

2.1. Neutrons: PHYS:N [4](#neutrons-physn)

2.1.1. Mix and Match (TABL) [5](#mix-and-match-tabl)

2.1.2. Fission Multiplicity (FISM) [6](#fission-multiplicity-fism)

2.1.3. Light-Ion Recoil (RECL) [7](#light-ion-recoil-recl)

2.2. Photons: PHYS:P [8](#photons-physp)

2.2.1. Photonuclear Model Physics (ISPN)
[8](#photonuclear-model-physics-ispn)

2.2.2. Photon Doppler Broadening (NODOP)
[9](#photon-doppler-broadening-nodop)

2.3. Protons: PHYS:H [10](#protons-physh)

2.4. Charged Particles: PHYS:x [11](#charged-particles-physx)

2.5. Mix-and-Match Nuclide Replacement: MX
[11](#mix-and-match-nuclide-replacement-mx)

2.6. Physics Models: LCA [14](#physics-models-lca)

2.6.1. Inline Generation of Double Differential Cross Sections and
Residual Nuclei
[14](#inline-generation-of-double-differential-cross-sections-and-residual-nuclei)

2.6.2. CEM2k Physics Model (ICEM) [16](#cem2k-physics-model-icem)

2.6.3. INCL4/ABLA (Cugnon/Schmidt) Physics Models (ICEM)
[17](#incl4abla-cugnonschmidt-physics-models-icem)

2.7. INCL4 Options: LCC [18](#incl4-options-lcc)

2.8. ABLA Fission-Evaporation Model: LEA
[18](#abla-fission-evaporation-model-lea)

2.9. Improved S(α,β) Treatment: MT [18](#improved-sαβ-treatment-mt)

2.10. User Specification of Multiplicity Constants: FMULT
[19](#user-specification-of-multiplicity-constants-fmult)

2.10.1. Nuclides without Transport Cross Sections
[20](#__RefHeading___Toc100926598)

2.10.2. Nuclides without Multiplicity Data
[21](#nuclides-without-multiplicity-data)

3.0. SOURCE features [21](#source-features)

3.1. Particle-Type Specification: SDEF
[22](#particle-type-specification-sdef)

3.2. Positron Sources: SDEF [22](#positron-sources-sdef)

3.3. Spontaneous Fission Sources: SDEF
[22](#spontaneous-fission-sources-sdef)

3.3.1. Spontaneous Fission Physics [22](#spontaneous-fission-physics)

3.3.2. Normalization of Summary and Tally Information
[23](#normalization-of-summary-and-tally-information)

3.3.3. Fission Multiplicity Output [23](#fission-multiplicity-output)

3.4. Multiple Particles on SDEF Card
[25](#multiple-particles-on-sdef-card)

3.4.1. Source Particle Type as the Independent Variable
[25](#source-particle-type-as-the-independent-variable)

3.4.2. Source Particle Type as a Dependent Variable
[25](#source-particle-type-as-a-dependent-variable)

3.4.3. Additional Capabilities and Cautions
[25](#additional-capabilities-and-cautions)

3.4.4. Tally and Summary Table Normalizations
[26](#tally-and-summary-table-normalizations)

3.5. Sources on Cylindrical Surfaces: SDEF
[26](#sources-on-cylindrical-surfaces-sdef)

3.5.1. Examples [27](#examples)

3.5.2. Additional Information and Cautions
[27](#additional-information-and-cautions)

3.6. Repeated Structures Source Specifications: SI
[28](#repeated-structures-source-specifications-si)

3.7. Translated Sources with Dependence
[28](#translated-sources-with-dependence)

4.0. TALLY features [28](#tally-features)

4.1. Pulse-Height Tallies with Variance Reduction
[29](#pulse-height-tallies-with-variance-reduction)

4.1.1. Future Work [29](#future-work)

4.2. Default Dose Functions (DF) [29](#default-dose-functions-df)

4.3. Lattice Tally Speedup (F) [31](#lattice-tally-speedup-f)

4.4. Radiography Tally (FIP, FIR, FIC)
[32](#radiography-tally-fip-fir-fic)

4.4.1. Pinhole Radiography [32](#pinhole-radiography)

4.4.2. Transmitted Image Projection [32](#transmitted-image-projection)

4.5. Photonuclear and Proton Reaction Multipliers (FM)
[32](#photonuclear-and-proton-reaction-multipliers-fm)

4.6. Expanded Cosine Specification (C)
[33](#expanded-cosine-specification-c)

4.7. Pulse-Height Light Tally with Anticoincidence: FT8 PHL
[33](#pulse-height-light-tally-with-anticoincidence-ft8-phl)

4.8. Coincidence Capture Tally (FT8 CAP)
[35](#coincidence-capture-tally-ft8-cap)

4.8.1. Restrictions [36](#restrictions)

4.8.2. Interpreting Capture Tally Output
[36](#interpreting-capture-tally-output)

4.8.3. FT8 Capture Tallies with Time Gating
[39](#ft8-capture-tallies-with-time-gating)

4.8.4. Possible Future Extensions of FT8 CAP Option
[42](#possible-future-extensions-of-ft8-cap-option)

4.9. Residual Nuclei Tally: FT8 RES [43](#residual-nuclei-tally-ft8-res)

5.0. Other input file extensions [44](#other-input-file-extensions)

5.1. Logarithmic Interpolation [44](#logarithmic-interpolation)

5.2. Auxiliary Input File and Encryption
[45](#auxiliary-input-file-and-encryption)

5.2.1. Reading from an Auxiliary File
[45](#reading-from-an-auxiliary-file)

5.2.2. Suppressing Input Printing [45](#suppressing-input-printing)

5.2.3. Encryption of Input [46](#encryption-of-input)

5.3. HISTP File Size Control [46](#histp-file-size-control)

5.4. DXTRAN Improvements [47](#dxtran-improvements)

5.4.1. Detectors with Model Physics (Isotropic Approximation)
[47](#detectors-with-model-physics-isotropic-approximation)

5.4.2. Detector/DXTRAN Underflow Control (DBCN)
[47](#detectordxtran-underflow-control-dbcn)

5.5. PTRAC for All Particles and for Coincidence Counting
[47](#ptrac-for-all-particles-and-for-coincidence-counting)

5.5.1. PTRAC for All Particles (PTRAC)
[47](#ptrac-for-all-particles-ptrac)

5.5.2. PTRAC Capture File for Coincidence Counting (PTRAC)
[48](#ptrac-capture-file-for-coincidence-counting-ptrac)

6.0. graphics [49](#graphics)

6.1. Color Contour Plots for Mesh Tallies, Radiography Tallies, and
Lattice Tallies
[49](#color-contour-plots-for-mesh-tallies-radiography-tallies-and-lattice-tallies)

6.1.1. User Interface for Contour Tally Plots
[50](#user-interface-for-contour-tally-plots)

6.1.2. Mesh Tally Contour Plot Example
[52](#mesh-tally-contour-plot-example)

6.1.3. Radiography Tally Coutour Plot Example
[58](#radiography-tally-coutour-plot-example)

6.1.4. Lattice Tally Plot Examples [60](#lattice-tally-plot-examples)

6.1.5. Other Notes and Future Work [61](#other-notes-and-future-work)

6.2. Enhanced Color Geometry Plots: 64-Color Plotting and Shading by
Cell Parameters
[61](#enhanced-color-geometry-plots-64-color-plotting-and-shading-by-cell-parameters)

6.3. i,j,k Lattice Index Labeling [62](#ijk-lattice-index-labeling)

6.4. Weight-Window-Generator Superimposed Mesh Plots
[63](#weight-window-generator-superimposed-mesh-plots)

6.4.1. Geometry Plot of Weight-Window-Generator Superimposed Mesh
[63](#geometry-plot-of-weight-window-generator-superimposed-mesh)

6.4.2. Example [64](#example)

6.5. Proton and Photonuclear Cross-Section Plots
[65](#proton-and-photonuclear-cross-section-plots)

6.5.1. Photonuclear Cross-Section Plots
[65](#photonuclear-cross-section-plots)

6.5.2. Proton Cross-Section Plots [67](#proton-cross-section-plots)

6.6. Pause Command [67](#pause-command)

7.0. Parallel processing [67](#parallel-processing)

7.1. Distributed Memory Multitasking for the Entire Energy Range of All
Particles
[67](#distributed-memory-multitasking-for-the-entire-energy-range-of-all-particles)

7.2. MPI Multiprocessing [67](#mpi-multiprocessing)

7.3. MPI Speedup for Criticality Calculations
[68](#mpi-speedup-for-criticality-calculations)

7.3.1. DXTRAN and Detector Tracking Differences
[68](#dxtran-and-detector-tracking-differences)

7.3.2. User Interface Changes for Both Sequential and Parallel KCODE
Problems
[69](#user-interface-changes-for-both-sequential-and-parallel-kcode-problems)

7.3.3. User Interface Changes for Parallel KCODE Problems
[69](#user-interface-changes-for-parallel-kcode-problems)

7.3.4. Compiler and Operating System Problems for Parallel KCODE
Problems
[69](#compiler-and-operating-system-problems-for-parallel-kcode-problems)

8.0. Compiling, configuration, and installation
[70](#compiling-configuration-and-installation)

8.1. F90 Modularity and Dynamic Memory Allocation
[70](#f90-modularity-and-dynamic-memory-allocation)

8.2. F90 Autoconfiguration [70](#f90-autoconfiguration)

8.3. 64-Bit-Integer Support [70](#bit-integer-support)

8.3.1. Recompilation Requirements for 64-Bit Integers
[71](#recompilation-requirements-for-64-bit-integers)

8.3.2. Additional Notes [71](#additional-notes)

8.4. NAG/IBM/INTEL F90 and Compiler Extensions
[71](#nagibmintel-f90-and-compiler-extensions)

9.0. Summary of MCNPX Capabilities Since MCNP4B (by version)
[72](#summary-of-mcnpx-capabilities-since-mcnp4b-by-version)

10.0. Future work [74](#future-work-1)

11.0. BIBLIOGRAPHY [75](#bibliography)

11.1. Code Manuals [75](#code-manuals)

11.2. Code Version Release Announcements
[75](#code-version-release-announcements)

11.3. Reports [76](#reports)

11.4. Papers [76](#papers)

12\. AcknowledgmentS [79](#acknowledgments)

MCNPX Extensions, Version 2.5.0

**ABSTRACT**

MCNPX is a Fortran 90 Monte Carlo radiation transport computer code
capable of tracking 34 particle types at nearly all energies. It is a
superset of MCNP4C and MCNPX2.3.0. This document describes the
capabilities beyond these constituent codes.

# Introduction

MCNPX is a Fortran 90 (f90) Monte Carlo radiation transport computer
code that transports 34 particle types, including four light ions, at
nearly all energies. MCNPX stands for MCNP eXtended. It is a superset of
MCNP4C3 and MCNPX2.3.0 and has many capabilities beyond these
constituent codes. These capability extensions enhance physics, sources,
tallies, graphics, and infrastructure.

MCNPX is a production computer code for modeling the interaction of
radiation with matter, and its quality is guaranteed. See Section 1.2
regarding the MCNPX guarantee and cash awards. MCNPX is used widely
around the world. See Section 1.3 for availability.

MCNPX is backward compatible with MCNP4C3 and older MCNPX (2.3.0 and
earlier) versions. Old input files still work and usually track, giving
identical results. The new MCNPX capabilities involve some changes, but
mostly extensions, to the input cards.

## MCNPX Extensions

Each successive version of MCNPX adds new capabilities and modernizes
the code for new hardware, operating systems, and compilers. The
extensions of MCNPX beyond MCNP4C and MCNPX2.3.0 now are listed with the
initials of the principal developers shown in parentheses.[^1] (Section
9 contains a complete list of new capabilities since MCNP4B grouped
according to version release.)

Physics

-   Seamless transport of 34-particle types at nearly all energies by
    mixing and matching of nuclear data table and model physics (JSH),

-   CEM2k physics (SGM/AJS/FXG),

-   INCL4/ABLA physics models (JCD/JSH),

-   Extension of neutron model physics below 20 MeV (JSH),

-   Fission multiplicity (JSH),

-   Light-ion recoil (JSH),

-   Inline generation of double differential cross sections and
    residuals (JSH),

-   Photon Doppler broadening (from MCNP5) (AS/GWM),

-   Weight-window generator (WWG) and exponential transform for model
    physics (FXG/ JSH), and

-   Improved S(α,β) physics (RCL/EJP).

Sources

-   Spontaneous fission (JSH),

-   Multiple-source particle types (JSH/MTS),

-   Repeated structures source-path improvement (LLC/JSH),

-   Positron sources (HGH),

-   Sources on cylindrical surfaces (GWM/JSH),

-   Source particles that may be specified by character descriptors
    (JSH), and

-   translated sources with dependence (GWM).

Tallies

-   Pulse-height tallies (PHTs) with variance reduction (JSH/GWM),

-   Default dose functions (LSW/JSH),

-   Pulse-height light tally with anticoincidence: FT8 PHL (GWM),

-   Coincidence capture tally: FT8 CAP (MTS/SJT/DRM/JSH),

-   Residual nuclei tally: FT8 RES (JSH),

-   Lattice tally speedup by orders of magnitude (GWM),

-   Proton reaction multipliers with FM cards (JSH),

-   Photonuclear reaction multipliers with FM cards (JSH/GWM),

-   Expanded radiography tally specification (JSH), and

-   Cosine bins that may be specified in degrees and for F2 tallies
    (JSH).

Other Input File Extensions

-   Logarithmic interpolation on input cards (JSH),

-   Auxiliary input files (JSH),

-   HISTP file size control (LSW),

-   DXTRAN improvements (GWM), and

-   PTRAC file for all particles and for coincidence counting
    (MTS/JSH/GWM).

Graphics

-   Mesh tallies (tally grid superimposed over geometry) plotted within
    MCNPX (JSH/GWM);

-   Two-dimensional (2D) color tally contour plots of lattice and
    radiography tallies (GWM);

-   Enhanced color geometry plots (GWM/JSH);

-   Geometry plot of WWG superimposed mesh (JSH);

-   i,j,k lattice indexing in geometry plots (JSH);

-   Proton and photonuclear cross-section plots (JSH); and

-   Pause command for tally and cross-section plots (JSH).

Parallel Processing

-   Distributed memory multiprocessing for all particles and energies
    (GWM),

-   Message passing interface (MPI) multiprocessing (JL/GWM), and

-   Significant speedup of criticality problems run with MPI (GWM/NAC).

Compiling, Configuration, and Installation

-   FORTRAN90 modularity and dynamic memory allocation (GWM);

-   F90 autoconfiguration (TLR/HWE);

-   64-bit integer support (GWM); and

-   NAG/IBM/INTEL compiler extensions for Unix, Linux, and Windows
    (GWM).

## Guarantee

MCNPX is guaranteed. We are so confident of the quality of MCNPX that we
will pay \$20 or €20 to the first person finding anything that does not
meet or exceed the capabilities of MCNPX 2.3.0 and MCNP4C3. We also will
pay a brand new \$2 bill for any error in MCNPX that has been inherited
from its constituent codes.[^2]

MCNPX is a better quality code than MCNP4C3.[^3] First, it corrects many
MCNP4C3 problems. Second, cash awards have been earned less frequently
with MCNPX than with MCNP4C3 and its predecessors, and most of those
awards are for problems carrying over from older code versions; very few
have been put into the new MCNPX versions. A list of winners is
available at <http://mcnpx.lanl.gov/>. MCNPX bugs are described in the
newsletter for each MCNPX version.

## Availability

MCNPX 2.5.0 is available from the Radiation Safety Information and
Computational Center (RSICC) in Oak Ridge, Tennessee, USA, at
<http://www-rsicc.ornl.gov/>. MCNPX 2.5.0 also is expected to be
available from the Organization for Economic Cooperation and
Development**/**Nuclear Energy Agency (OECD/NEA) Data Bank in Paris,
France, at <http://www.nea.fr/>.

An essential part of the MCNPX software quality assurance plan is the
beta test program. Before a code version goes to RSICC or OECD/NEA, it
is made available to more than 1600 MCNPX beta testers worldwide. MCNPX
beta test versions are available to authorized beta testers on the MCNPX
website at <http://mcnpx.lanl.gov/>. Participants in MCNPX workshops
automatically become MCNPX beta testers (<http://mcnpxworkshops.com/>).
In special cases, others may also obtain beta test access by contacting
mcnpx@lanl.gov.

All beta test, RSICC, and OECD/NEA versions of MCNPX are guaranteed with
cash awards.

# PHYSICS extensions

The following physics capability extensions are available and are
controlled by the input cards listed in parentheses:

-   Seamless transport of 34-particle types at nearly all energies by
    mixing and matching of nuclear data table and model physics (PHYS:N,
    PHYS:H, MX);

-   Extension of neutron model physics below 20 MeV (PHYS:N, PHYS:H,
    PHYS:x);

-   Fission multiplicity (PHYS:N, FMULT);

-   Light-ion recoil (PHYS:N, PHYS:H);

-   Photonuclear model physics (PHYS:P);

-   Photon Doppler broadening from MCNP5 (PHYS:P);

-   Inline generation of double differential cross sections and residual
    nuclei (LCA);

-   CEM2k physics (LCA);

-   INCL4/ABLA physics models (LCA, LEA, LCC);

-   Improved S(*α,β*) physics (MT); and

-   User specification of fission multiplicity constants (FMULT).

In the following descriptions of input cards, **bold face type** is used
to highlight extensions from earlier MCNP or MCNPX input.

## Neutrons: PHYS:N

PHYS:N *EMAX EAN IUNR DNB **TABL FISM RECL***

> *EMAX =* upper energy limit (default = 100 MeV).
>
> **Note:** EMAX must be higher than the highest energy in the problem
> or the physics *is wrong.* However, setting EMAX too high causes
> excessively coarse energy binning of model physics, which is
> inaccurate. For problems with energies above 100 MeV, EMAX should be
> chosen carefully; the default is appropriate for problems with
> energies below 100 MeV.
>
> *EAN =* analog capture below *EAN;* implicit capture above EAN
> (default = 0 MeV).
>
> *IUNR =* unresolved resonance range probability table treatment.
>
> = 0/1 = on/off (default = 0) when unresolved data are available.
>
> *DNB =* delayed neutrons from fission.
>
> -1 = analog production of delayed neutrons from fission (default).
>
> 0 = treat prompt and delayed neutrons as prompt.

n = biased production: produce up to *n* delayed neutrons per fission

(*n* \> 0 disallowed in KCODE).

> ***TABL*** = use data tables below *TABL*, physics models above
> *TABL.*
>
> = -1 (default) (mix and match). When tables are available, use them up
> to\
> their upper limit and use the physics models above.
>
> ***FISM** =* fission multiplicity.

= 0 (default) MCNP treatment. The number of neutrons per fission is the

> integer above or below . If = 2.7, then the number of neutrons will
> be\
> two 30% of the time and three 70% of the time.
>
> = FWHM: sample Gaussian distribution with full-width half-maximum of\
> FWHM about *.*
>
> = -1: sample Gaussian distribution with FWHM appropriate for
> fissioning nuclide (recommended).
>
> ***RECL** =* 0 (default) no light-ion recoil.
>
> ≥ 0: light ion recoil. Produce 0 ≤ *RECL* ≤ 1 light ions (*h*, *d*,
> *t*, *s*, *a*) at each elastic scatter with light nuclei (H, D, T,
> ^3^He, ^4^He). The ionization potential is accounted for, and the
> proper two-body kinematics is used (with neutron free-gas thermal
> treatment if appropriate) to bank the created particles with the
> proper energy and angle. *MODE n h d t s a* . . . is required to
> produce light ions *h*, *d*, *t*, *s*, *a*. *CUT:x 2J 0* for *x* =
> *h*, *d*, *t*, *s*, *a* is recommended so that the low-energy recoil
> ions produced are not killed by energy cutoff.

### Mix and Match (TABL)

The *TABL* entry on the PHYS:N and PHYS:H cards controls mix and match.
If the energy, *E*, is greater than *TABL,* physics models will be used
above *E* = *TABL*. Physics models can be used at any energy, even below
20 MeV, although physics models are often poor below 100 to 200 MeV. If
*tabl* = -1, then mix and match is used and the data tables are used for
their full energy range, when available, for each nuclide, even if other
nuclides in the problem do not have data tables available in that energy
range. The MCNPX mix-and match-capability is illustrated in Fig. 1.

Two nuclides are present in this example: oxygen and uranium. The oxygen
data table ranges from 0 to 150 MeV; the uranium data table ranges from
0 to 20 MeV. If *TABL* = 20, then nuclear data tables are used below 20
MeV and physics models above; the superior table data for oxygen between
20 to 150 MeV are ignored. If *TABL* = 150, then nuclear data tables are
used below 150 MeV and physics models above; the values for uranium at
20 MeV are used (incorrectly) in the range of 20 to 150 MeV. However, if
*TABL* = -1 (default), mix and match is used: both oxygen and uranium
use the table data where they exist and physics models above the maximum
energy.

See the discussion of mix and match for the MX card in Section 2.5.

![](media/image1.png){width="5.1819444444444445in" height="3.95in"}

Fig. 1. Mix and match.

### Fission Multiplicity (FISM)

Fission multiplicity is the number of neutrons produced in a fission
reaction. The nuclear data tables provide the average number fission
neutrons, , per fission. The default MCNPX treatment is the MCNP
treatment: the number of neutrons per fission is the integer above or
below . If = 2.7, then the number of neutrons will be two 30% of the
time and three 70% of the time. If the sixth entry, *FISM,* on the
PHYS;N card is positive, then is sampled from a Gaussian distribution.
The full-width at half-maximum (FWHM) of the Gaussian distribution is
obtained from the table in Section 2.10. For induced fission, the
average value of generally comes from the nuclear data library and is a
function of the incident neutron energy. For spontaneous fission
sources, the fission multiplicity is sampled from a cumulative
distribution when available or sampled from a Gaussian distribution when
unavailable. The values of the FWHM and spontaneous fission may be
overridden using an **FMULT** card (Section 2.10).

When a Gaussian distribution is used, the average number of fission
neutrons becomes biased above its true value. The high value of occurs
because can have sampled values between , and when the negative values
of are sampled they are set to zero, causing a slightly high bias. The
negative Gaussian bias for fission multiplicity has a small correction
to preserve the average number of neutrons per fission. The following
options are available for the sampling of fission multiplicity[^4]:

> **PHYS**:N 5J *FISM ,*

where

*FISM* = 0: no fission multiplicity: MCNP method---sample *ν* from
nearest integer,

> which preserves the first moment but gives incorrect multiplicity;
>
> = 1: correct sampled *ν* to preserve average multiplicity
> (recommended);
>
> = 2 : preserve multiplicity by increasing *ν* threshold.

= 3 : sample Gaussian distribution without correction;

= 4 : MCNP method in presence of spontaneous fission or FMULT card.

*Defaults:* If an **FMULT** card is present or if a spontaneous fission
source is used (par = sf on SDEF card), then the default is *FISM* = 1;
otherwise, the default is *FISM* = 0.

If *FISM* = -1 (MCNPX 2.5.E specification), it is changed to +1 for
backward compatibility. To specify a width for fission multiplicity
(*FISM*\> 0 in MCNPX 2.5.E), the **FMULT** card must be used.

The fission multiplicity---both the number of fission neutrons and the
moments of the distributions, are printed in Print Table 117 (Section
3.3.3).

### Light-Ion Recoil (RECL)

MCNPX can produce and track ions created by elastic recoil from neutrons
or protons. Neutrons and protons undergoing elastic scatter with light
nuclei (H, D, T, ^3^He, and ^4^He) can create ions (protons, deuterons,
tritons, ^3^He, and *α*) that are banked for subsequent transport.

Figure 2 shows the energy-angle production of alphas created from 15-MeV
neutrons striking ^4^He. Note that in the forward bin, cosine 0.8 \< *μ*
\< 1, the *α* energy goes up to the theoretical maximum of 9.6 MeV. The
theoretical maxima in the other cosine bins (0.8, 0.6, 0.4, and 0.2) are
6.144, 3.456, 1.536, and 0.384.

The input file for this example is as follows.

Test of light ion recoil

1 1 1e-5 -1

2 0 1

1 so 1.e-5

mode n a

imp:n,a 1 0

phys:n 6j 1

sdef erg=15

print -161 -162

tmp1 1e-20 0

fcl:n 1 0

m1 2004 .2

cut:a j 0

nps 1000000

f51:a 1

e51 .1 100log 20

c51 -.8 8i 1 t

fq51 e c

The plot commands to produce Fig. 2 are presented in the following plot
command file.

rmct lir.m tal 51 xlim .1 15 loglog &

title 1 \"Light-Ion Recoil: 15 MeV Neutrons on 4He\" &

title 2 \"Alpha Energy vs Cosine\" &

fix c 11 label 1 \"cos total\" cop fix c 6 label 2 \"cos -1-.2\" &

cop fix c 7 label 3 \"cos .2-.4\" cop fix c 8 label 4 \"cos .4-.6\" &

cop fix c 9 label 5 \"cos .6-.8\" cop fix c 10 label 6 \"cos .8-1.\"

## Photons: PHYS:P

> PHYS:P *EMCPF IDES NOCOH ISPN **NODOP** ,*

where

> *EMCPF* = upper energy for detailed physics (default = 1 MeV);
>
> *IDES* = generate electrons by mode E or thick-target-bremsstrahlung
>
> (default = 0 = on);
>
> *NOCOH* = coherent (Thomson) scattering (default = 0 = on);
>
> ***ISPN*** = -1/0/1 = analog/off/biased. The bias causes a
> photonuclear event at each\
> photoatomic event. (default = 0 = off); and
>
> ***NODOP*** = 0/1 = on/off Doppler energy broadening. (default = 1 =
> off).

### Photonuclear Model Physics (ISPN)

Photonuclear physics models enable (*γ*,*n*) and other photonuclear
reactions when photonuclear data tables are unavailable. When some
photonuclear data tables are available, MCNPX will mix and match, using
tables when available and physics models when no tables are available.
The user interface is not changed; the user simply gets the latest
gdr.dat file from the MCNPX source file, src/DATA/CEM/gdr.dat. The
threshold for photonuclear reactions is 1 MeV.

The MCNPX photonuclear physics model modifications adjust the
photofission cross sections in CEM by the ratio of fission and neutron
evaporation-level density parameters (*wam*). For selected isotopes
(thorium, uranium, and plutonium), a list of energy-dependent *wam*
parameters was generated by matching the fissility of the BOFOD
photonuclear cross-section data. The lists are entered into the code as
data statements in subroutine set\_*wam*. This subroutine provides a
*wam* parameter for a requested nuclide "IZA" at an incident photon
energy of "t0mev" interpolating between two tabulated values. If data
for a requested nuclide are not available, *wam* is returned as 0.0. We
hope to provide *wam* parameters for more isotopes in the future. Also,
an improved fit of photoabsorption cross sections in the updated CEM2k
pnxs.F routine occurs based on experimental data from Muccifora,
Bianchi, Brookes, Michalowski, and Chollet.[^5]

![](media/image2.png){width="6.0in" height="4.5in"}

Fig. 2. Light-ion recoil.

Photonuclear models are designed to work with the CEM2k physics package
and are fragile or may not work with other packages.

### Photon Doppler Broadening (NODOP)

The MCNP5 photon Doppler broadening capability[^6] developed by Avneet
Sood is available in MCNPX. Unlike MCNP5, the default is "off" (PHYS:P
4j 1) because the effect is small but time consuming. When photon
Doppler broadening is turned on (NODOP = 0), it has no effect unless
photon Doppler broadening momentum profile data are available in the
photon library. These data are available in the MCPLIB03 and MCPLIB04
photon libraries with ZAID identifiers .03p and .04p.

The electron binding effect on a scattered photon's energy distribution
appears as a broadening of the energy spectrum due to the precollision
momentum of the electron. The effect applies to incoherent scatter
(Compton scatter). The scattered energy of a Doppler-broadened photon is
calculated by selecting an orbital shell, sampling the projected
momentum from the Compton profile, and then calculating the scattered
photon energy.

Two research notes on photon Doppler broadening are available on the
MCNPX web site, <http://mcnpx.lanl.gov/documents/>. The following is
quoted from one of these research notes (X‑5:AS-02-16).

Incoherent scattering of an incident photon can occur with a bound
electron in a shell of a material and will generate a Compton electron
and a scattered photon. The electron binding effects become important
when the incident photon energy is near a few hundred keV. The result of
the binding effects on the angle and energy of the scattered photon must
be taken into account for accurate simulation of low-energy photon
transport. The effect of the bound electron on the scattered photon's
angular distribution appears as a reduction in the total scattering
cross section in the forward direction. MCNP currently accounts for the
electron binding effects on the angular distribution of the scattered
photon by modifying the Klein-Nishina differential cross section with a
form factor. The electron binding effect on the scattered photon's
energy distribution appears as a broadening of the energy spectrum due
to the precollision momentum of the electron. This effect on the energy
distribution of the incoherently scattered photon is called Doppler
broadening.

## Protons: PHYS:H

PHYS:H *EMAX EAN **TABL** J **ISTRG** J **RECL ,***

where

> *EMAX =* upper energy limit (default = 100 MeV).

**Note:** EMAX must be higher than the highest energy in the problem or
the physics *is wrong.* However, setting EMAX too high causes
excessively coarse energy binning of model physics, which is inaccurate.
For problems with energies above 100 MeV, EMAX should be chosen
carefully; the default is satisfactory for problems with energies below
100 MeV.

> *EAN* = analog capture below *EAN;* implicit capture above EAN
> (default =
>
> 0 MeV)
>
> ***TABL*** = use data tables below *TABL*; physics models above *TABL*
>
> = -1 (default) (mix and match). If tables are available, use them up
> to
>
> the upper energy limit and use physics models above
>
> *J* = jump (unused)
>
> ***ISTRG** =* charged particle straggling control
>
> = 0 Vavilov (default, best)
>
> = 1 continuous slowing down approximation
>
> = -1 old MCNPX 2.2.4 method
>
> *J* = jump (unused)
>
> ***RECL*** = light-ion recoil. Produce 0 \< *RECL \<* 1 light ions
> (*h*, *d*, *t*, *s*, *a*) at each\
> elastic scatter with light nuclei (H, D, T, ^3^He, ^4^He). The
> ionization\
> potential is accounted for, and the proper two-body kinematics is
> used\
> to bank the created particles with the proper energy and angle. MODE
> h\
> d t s a ... is required to produce light ions (h, d, t, s, a). *CUT:x
> 2J 0* for x\
> = h, d, t, s, a is recommended so that the low-energy recoil ions\
> produced are not killed by energy cutoff. Note that protons colliding\
> with hydrogen to produce more protons can produce an overwhelming\
> number of protons; caution is required, and *RECL \< 1* may be needed.

See the definitions of *TABL* and *RECL* in the description of the
neutron physics card, PHYS:N, above.

## Charged Particles: PHYS:x

PHYS:x *EMAX 3J **ISTRG *****,**

where

*EMAX =* upper energy limit (default = 100 MeV).

**Note:** EMAX must be higher than the highest energy in the problem or
the physics *is wrong.* However, setting EMAX too high causes
excessively coarse energy binning of model physics, which is inaccurate.
For problems with energies above 100 MeV, EMAX should be chosen
carefully; the default is satisfactory for problems with energies below
100 MeV.

*J* = jump (unused)

***ISTRG** =* charged particle straggling control

> = 0 Vavilov or Prael's straggling model, which is an energy
>
> correction addressing stopping powers. (default, best)

= 1 continuous slowing down ionization model

= -1 old MCNPX 2.2.4 method

## Mix-and-Match Nuclide Replacement: MX

The **MX** card enables materials substitution for different particle
types. This card is an extension of, and replacement for, the MPN card
for photonuclear data

**MXn:p *zaid1 zaid2* \... ,**

where

n = material number of an Mn card that MUST precede the MXn card,

p = particle type (n, p, h), and

*zaidn* = ZA of replacement nuclide for the n^th^ nuclide on the Mn
card.

Only particle types n (neutron), p (photonuclear), and h (proton) are
allowed on the MX card. No substitutions are allowed for photoatomic (p)
and electron (e) data because those data sets are complete. The MXn:P
card is an exact replacement of the MPNn card and specified photonuclear
nuclide substitutions (library type u). zaidn = 0 is allowed on MXn:P
(photonuclear substitution) to specify no photonuclear data for a
specific photoatomic reaction. zaidn = model is allowed on the MXn:N and
MXn:H (neutron and proton substitution) to allow models to be mixed with
tabular data. As an example, consider the following input file.

> mode n h p
>
> m1 1002 1 1003.6 1 6012 1 20040 1 nlib .24c
>
> mx1:n j model 6000 20000
>
> mx1:h model 1001 j j
>
> mpn1 6012 0 j j

MCNPX will issue the following warnings.

warning. MPNn will soon be obsolete. Use MXn:p instead.

warning. photonuclear za = 6012 different from nuclear za = 1002

warning. photonuclear za = 0 different from nuclear za = 1003

Models will be used for neutron tritium and proton deuterium. The MPN
card still works but has a warning. The mixing and matching is
summarized in Print Table 101:

> 1particles and energy limits print table 101

+---+---+------+--------+--------+--------+--------+--------+--------+
| * |   |      | **Part | **M    | **Sm   | **Lar  | **     | **     |
| * |   |      | icle** | aximum | allest | gest** | Always | Always |
| P |   |      |        | Pa     | Table  |        | Use    | Use    |
| a |   |      | **Cu   | rticle | Max    | *      | T      | M      |
| r |   |      | toff** | En     | imum** | *Table | able** | odel** |
| t |   |      |        | ergy** |        | Max    |        |        |
| i |   |      | **En   |        |        | imum** | **B    | **A    |
| c |   |      | ergy** |        |        |        | elow** | bove** |
| l |   |      |        |        |        |        |        |        |
| e |   |      |        |        |        |        |        |        |
| T |   |      |        |        |        |        |        |        |
| y |   |      |        |        |        |        |        |        |
| p |   |      |        |        |        |        |        |        |
| e |   |      |        |        |        |        |        |        |
| * |   |      |        |        |        |        |        |        |
| * |   |      |        |        |        |        |        |        |
+---+---+------+--------+--------+--------+--------+--------+--------+
|   |   |      |        |        |        |        |        |        |
+---+---+------+--------+--------+--------+--------+--------+--------+
| 1 | n | neu  | 0.00   | 1.00   | 1.50   | 1.50   | 0.00   | 1.50   |
|   |   | tron | 00E+00 | 00E+37 | 00E+02 | 00E+02 | 00E+00 | 00E+02 |
+---+---+------+--------+--------+--------+--------+--------+--------+
| 2 | p | ph   | 1.00   | 1.00   | 1.00   | 1.00   | 1.00   | 1.00   |
|   |   | oton | 00E-03 | 00E+02 | 00E+05 | 00E+05 | 00E+05 | 00E+05 |
+---+---+------+--------+--------+--------+--------+--------+--------+
| 9 | h | pr   | 1.00   | 1.00   | 1.50   | 1.50   | 0.00   | 1.50   |
|   |   | oton | 00E+00 | 00E+02 | 00E+02 | 00E+02 | 00E+00 | 00E+02 |
+---+---+------+--------+--------+--------+--------+--------+--------+
|   |   |      |        |        |        |        |        |        |
+---+---+------+--------+--------+--------+--------+--------+--------+

The MCNPX mix-and-match capability[^7] enables mixing and matching of
physics models and data tables. It is possible to specify some nuclides
with models and other nuclides with data tables (isotope "mixing"). It
is also possible to use data tables up to their maximum energy value and
then use models above that energy, even when the maximum table energy
differs from nuclide to nuclide ("energy matching").

As an example, consider protons interacting with a bismuth-germanium
particle detector (BGO). Because the present LA-150 nuclear data
libraries[^8] do not include germanium, it was previously possible to
use physics models alone for all nuclides in the entire problem. With
the mix-and-match capability, bismuth and oxygen can be modeled with the
more accurate nuclear data tables, whereas germanium uses a physics
model.

As another example, consider a neutron problem with deuterium and
tritium. The available deuterium library goes up to 150 MeV, but the
tritium library goes up to only 20 MeV. Previously, either neutron
physics models above 20 MeV (neglecting the deuterium table data up to
150 MeV) or nuclear data tables below 150 MeV (using the 20-MeV tritium
data throughout the entire 20- to 150-MeV range) had to be used. With
the mix-and-match capability, deuterium uses tables up to 150 MeV and
uses physics models above 150 MeV; tritium uses data tables up to 20 MeV
and uses physics models above 20 MeV.

Figure 3 shows an example of the energy-matching capability. The 100-MeV
neutrons are incident on an 8.433-cm-long, 3.932-cm-radius BGO crystal.
The crystal contains 21% bismuth, 16% germanium, and 63% oxygen.
Currently, no germanium libraries are available. The solid line
represents flux in the crystal with the full mix-and-match capability,
which uses all libraries up to their energy limits and physics models
above those limits and for germanium. The dashed-line calculation uses
the old method of substituting arsenic for the missing germanium
library, using the libraries up to 20 MeV and using physics models
above. The dotted line uses bismuth and oxygen libraries up to their
limits of 150 MeV; the arsenic library is used up to its limit of
20 MeV, and then the 20-MeV data are used from 20 to 150 MeV; above 150
MeV, physics models are used for all three nuclides. This last option is
least desirable but often was used in past code versions to take
advantage of the 150‑MeV libraries, even though many data libraries go
only to 20 MeV.

![](media/image3.png){width="5.302083333333333in" height="3.975in"}

Fig. 3. Comparison of different germanium library and model options.

The mix-and-match capability is particularly useful for photonuclear
calculations because few photonuclear data tables are available
currently. Libraries are used when available, and models are used
otherwise. Note that photonuclear physics is modeled with the CEM2k
model, regardless of whether CEM is used for other particles.

Different nuclides can be substituted for different particle types. For
example, natural carbon and calcium can be used for neutrons, whereas
^12^C and ^40^Ca can be used for protons and photonuclear reactions.

## Physics Models: LCA

LCA *IELAS IPREQ IEXISA ICHOIC JCOUL NEXITE NPIDK **NOACT** **ICEM***
**,**

where

> *IELAS =* 0 / 1 / 2 (default) = no/neutron only/neutron and proton
> elastic\
> scattering
>
> *IPREQ =* 0 = no preequilibrium model will be used
>
> = 1 (default) = use preequilibrium model after intranuclear cascade
>
> = 2 = select IPREQ = 1 and IPREQ = 3 randomly if IEXISA = 0
>
> = 3 = use preequilibrium model instead of the intranuclear cascade if\
> IEXISA = 0
>
> *IEXISA =* 0 = do not use ISABEL intranuclear cascade (default if ICEM
> = 2 =\
> INCL4)
>
> = 1 (default) = use Bertini for nucleons and pions and ISABEL for
> other\
> particles
>
> = 2 = use ISABEL for all particles
>
> *ICHOIC =* ISABEL model control (0023 = default)
>
> *JCOUL =* 0 / 1 (default) = no/yes Coulomb barrier for incident
> charged particles
>
> *NEXITE =* 1 (default) / 2 = subtract nuclear recoil energy to get
> excitation energy
>
> *NPIDK =* 0 (default) / 1 = force pi- nuclear capture below pi- energy
> cutoff
>
> ***NOACT** =* -2 = source particles immediately collide; all progeny
> escape. Used to\
> compute double differential cross sections and residual nuclei.
>
> = -1 = nuclear interactions of source particles only---transport and
> slowing\
> down are off. Used to compute double-differential cross sections with\
> XSEX code.
>
> = 0 = turn off all nonelastic reactions
>
> = 1(default) = normal transport
>
> = 2 = attenuation mode---transport primary source particles without\
> nonelastic
>
> ***ICEM** =* 0 (default) = Use Bertini or ISABEL model determined by
> IEXISA
>
> = 1 = use CEM2k model
>
> = 2 = use INCL4 model (default ABLA fission-evaporation)

### Inline Generation of Double Differential Cross Sections and Residual Nuclei

The *NOACT* option on the LCA card enables MCNPX to calculate a single
interaction in the physics model region. If the eighth entry on the LCA
card is 2,

LCA 7J **2 ,**

then the source particle immediately will have a collision and all
subsequent particles will escape. That is, LCA(8) = 2 causes the source
particle to have a distance to collision of zero, and all subsequent
tracks have a distance to collision of infinity.

The double differential cross sections and distributions of residual
nuclei for a single nuclear interaction thus may be calculated directly
in MCNPX. Tallying of the residual nuclei is discussed later in the FT8
RES tally description (see Section 4.9). Tallying of the differential
cross section can be done with standard F1 surface tallies, as shown in
the following example. The input file models a 1.2-GeV proton source
having a single collision with ^208^Pb.

Test of p(1.2GeV)+Pb(208)

1 1 -11. -1 imp:h 1

2 0 1 imp:h 0

1 so .01

mode h n

sdef par h erg=1200 vec 0 0 1 dir 1

m1 82208 1

phys:h 1300 j 0

phys:n 1300 3j 0

nps 10000

fc1 \*\*\* neutron angle spectra tally \*\*\*

f1:n 1

ft1 frv 0 0 1

fq1 e c

\*c1 167.5 9i 17.5 0 T

e1 1 50log 1300 T

lca 2 1 1 23 1 1 0 -2 0

The differential cross section for neutron production is tallied in the
F1 current tally with energy and time bins. This tally is simply the
neutrons that are created from the single proton collision with lead and
then escape. These data may be plotted with MCNPX using the tally
plotter and then following execute line command

MCNPX Z ,

where the command file, COM91, is

rmctal=mctl91

file all loglog xlim 1 1300 ylim 1e-6 1 &

fix c 13 cop fix c 1 cop fix c 6 cop fix c 12

In Fig. 4, the first line (solid black) is the energy spectrum over all
angles, the second (blue dashed) is the 180 output, the third (red
dotted) is the 90 output, and the fourth (green broken) is the 0 output.

### CEM2k Physics Model (ICEM)

The MCNPX CEM physics model has been upgraded from CEM95 to CEM2k.[^9]
It still is controlled solely by the ninth entry, ICEM, on the LCA card.
The model assumes that reactions occur in three stages. The first stage
is the IntraNuclear Cascade (INC), in which primary particles can be
rescattered, thus producing secondary particles several times before
absorption by or escape from the nucleus. The excited residual nucleus
that remains after the cascade determines the particle-hole
configuration, which is the starting point for the preequilibrium stage
of the reaction. The subsequent relaxation of the nuclear excitation is
treated in terms of an improved modified exciton model of preequilibrium
decay, followed by the equilibrium evaporative final stage of the
reaction, which is competing with the fission and Fermi-breakup
channels.

CEM2k incorporates the latest approximations for the elementary cross
sections used in the cascade, using more precise values for nuclear
masses and pairing energies, corrected system atics for the
level-density parameters, and several other refinements. Improved
algorithms decrease the computing time by up to a factor of 6 for heavy
targets. Other improvements were

![](media/image4.png){width="5.860416666666667in"
height="4.009722222222222in"}

Fig. 4. Differential cross sections at all angles, 180, 90, and 0, for
1.3-GeV protons on $_{\text{82}}^{\text{208}}$Pb.

motivated by recent measured data on isotope production from
Gesellschaft fűr Schwerionenforschung (GSI) (mbH, Darmstadt, Germany)
experiments. Compared with earlier versions, CEM2k has a longer cascade
state, less preequilibrium emission, and less evaporation from more
highly excited compound nuclei. CEM2k also has better models of neutron,
radionuclide, and gas production in accelerator-transmutation-of-waste
spallation targets.

A photonuclear physics package has been added to CEM2k by Franz
Gallmeier \[Oak Ridge National Laboratory (ORNL)\] for photon energies
from 8 MeV to about 2 GeV and includes the Giant Dipole and
Quasi-Deuteron resonances. These photonuclear physics models compliment
existing photonuclear data evaluations in that they may be used for
nuclides where no tabular data exist. Also, they may be used for
tabular-data nuclides above the 150-MeV upper-energy boundary of the
tables.

### INCL4/ABLA (Cugnon/Schmidt) Physics Models (ICEM)

The IntraNuclear Cascade, Liege (INCL4) and ABLA fission-evaporation
models are available in MCNPX. They are controlled by the LCA, LCC, and
LEA input cards:

LCA(9) = **2** INCL4 model (sets LCA(3) =*IEXISA = 0)*

LEA(7) = **2** ABLA fission evaporation

**LCC** = control parameters for INCL4

INCL4 and ABLA may be used in various combinations with other physics
models.

Bertini/Dresner: LCA 2 1 1 23 1 1 0 1 0 \$ MCNPX Default: Bertini

LEA 1 4 1 0 1 0 0 1 \$ MCNPX Default: Dresner

CEM2k LCA 8J 1 \$ CEM2k works alone

ISABEL/Dresner LCA 2J 2 \$ ISABEL

Bertini/ABLA LEA 6J 2 \$ ABLA

ISABEL/ABLA LCA 2J 2 \$ ISABEL

LEA 6J 2 \$ ABLA

INCL4/Dresner LCA 8J 2 \$ INCL4

INCL4/ABLA LCA 8J 2 \$ INCL4

LEA 6J 2 \$ ABLA

These combinations also may be shown in the following table.

LCA(3) LCA(9) LEA(7)

*IEXISA ICEM IEVAP*

Bertini/Dresner 1 0 0

ISABEL/Dresner 2 0 0

Bertini/ABLA 1 0 2

ISABEL/ABLA 2 0 2

CEM2k - 1 -

INCL4/Dresner 0 2 0

INCL4/ABLA 0 2 2

The INCL4 model is based largely on the work of Joseph Cugnon at the
University of Liege in Liege, Belgium. This model generally is coupled
with the ABLA fission-evaporation model that was developed principally
by Karl-Heinz Schmidt at GSI. Its integration into MCNPX has been done
principally by Jean-Christophe David at Commissariat à l'Energie
Atomique-Saclay, France.

INCL4 and ABLA are intended for use in the 200-MeV to 2-GeV energy
range. Only two free parameters exist in INCL4: the potential depth V~0~
(default = 45 MeV) and the overall factor f~stop~ (default = 1.0). These
parameters are controlled by the LCC card.

The INCL4 model, in its present implementation, is much slower than the
Bertini and CEM2k models.

## INCL4 Options: LCC

**LCC *STINCL V0INCL ***

The LCC card specifies options for the INCL4 model and the ABLA
fission-evaporation model. INCL is invoked by setting the ninth LCA card
entry to 2, and ABLA is invoked by setting the seventh LEA card entry,
IEVAP, to 2. The LCC card entries are

(1) *STINCL* rescaling factor of the cascade duration (default = 1)

(2) *V0INCL* potential depth (default = 45 MeV)

## ABLA Fission-Evaporation Model: LEA

LEA *IPHT ICC NOBALC NOBALE IFBRK ILVDEN **IEVAP** NOFIS* ,

where

> *IPHT* = 0 / 1 (default) = no/yes generate de-excitation photons
>
> *ICC =* 0 / 1 / 2 / 3 / 4 (default) =
> continuum/Troubetzkoy/intermediate/
>
> spin-dependent/full PHT physics model.
>
> *NOBALC =* 0 / 1 (default) no/yes---turn off mass-energy balancing in
> cascade
>
> *NOBALE =* 0 (default) / 1 = mass-energy balancing in evaporation
>
> *IFBRK =* 0 / 1 (default) = Fermi breakup model nuclide range
>
> *ILVDEN =* -1 / 0 (default) / 1 = HETC/Gilbert/Julich---level density
> model
>
> ***IEVAP =*** 0 (default) / 1 / 2 = RAL/ORNL/ABLA fission-evaporation
> model
>
> *NOFIS =* 0 / 1 (default) = suppress/allow - fission

## Improved S(α,β) Treatment: MT

The *S(α,β)* thermal neutron treatment user interface is unchanged, but
the underlying secondary energy treatment is improved. In many cases,
the *S(α,β)* thermal neutron treatment uses discrete energies instead of
32-equiprobable bin histogram energies. The result is nonphysical spikes
in the thermal neutron spectrum that are smoothed out by an algorithm
provided by Bob Little and Eric Pitcher of Los Alamos National
Laboratory. MCNPX still can track the old algorithm by setting DBCN(20)
= 1.

## User Specification of Multiplicity Constants: FMULT

The FMULT card enables users to override or add additional fission
multiplicity data as

**FMULT** *ZAID \[ keyword = values \] ,*

where the keywords are

> *ZAID* = nuclide for which data are entered.
>
> *SFnu* = average ν for sampling spontaneous fission multiplicity from
> a Gaussian
>
> distribution with *Width*; or
>
> = cumulative probability of spontaneous fission multiplicity.
>
> *SFnu* is used for spontaneous fission only; for induced fission, ν is
> taken
>
> from the nuclear data library at the energy of the incident neutron.
>
> *Width* = Gaussian width for sampling ν for both spontaneous and
> induced fission,
>
> ignored for spontaneous fission when *SFnu* is specified as a
> cumulative
>
> probability distribution.
>
> *SFYield* = spontaneous fission yield ( n/s-g )---used for selecting
> the spontaneous
>
> fission nuclide when more than one is present in a material.
>
> *Watt* = watt energy spectrum parameters a and b for spontaneous
> fission neutron
>
> energy sampling. Induced fission energies are sampled from the nuclear
>
> data library.

*ZAID* must be specified. Defaults exist only for the most common
fission nuclei; these defaults are provided in Print Table 38 of the
MCNPX output. *Width* is used for both spontaneous and induced fission;
*SFYield* is used only for spontaneous fission. *SFnu* and *Watt* are
used only for spontaneous fission; for induced fission, the values are
sampled from the nuclear data library.

*Examples:*

> **FMULT** 98252 *SFYield = 2.34e12 SFnu = .002 .028 .155 .428 .732
> .917 .983 .998 1. Width = 1.207 Watt* = 1.18 1.03419
>
> **FMULT** 94239 Watt = .885247 3.8026 Width = 1.14 SFYield = .0218
> *SFnu* = 2.16

All values are printed in the MCNPX output file in Print Table 38. The
data in Table 1 are the default values in MCNPX, not the values modified
with the FMULT cards above. The default values[^10] have been obtained
from Brookhaven National Laboratory and other sources modified by Peter
Santi and referenced by him.[^11] The spontaneous fission multiplicity
table values are printed in Print Table 38 to only three-digit accuracy
but are accurate to seven digits in MCNPX.

**Table 1. MCNPX Output File Print Table 38 Displaying Fission
Multiplicity Data**

1fission multiplicity data. print table 38

zaid width watt1 watt2 yield sfnu

90232 1.079 .800000 4.00000 6.00E-08 2.140

92232 1.079 .892204 3.72278 1.30E+00 1.710

92233 1.041 .854803 4.03210 8.60E-04 1.760

92234 1.079 .771241 4.92449 5.02E-03 1.810

92235 1.072 .774713 4.85231 2.99E-04 1.860

92236 1.079 .735166 5.35746 5.49E-03 1.910

92238 1.230 .648318 6.81057 1.36E-02 0.048 .297 .722 .950 .993 1.00 1.00
1.00 1.00 1.

93237 1.079 .833438 4.24147 1.14E-04 2.050

94236 0.000 .000000 0.00000 0.00E+00 0.080 .293 .670 .905 .980 1.00 1.00
1.00 1.00 1.

94238 1.115 .847833 4.16933 2.59E+03 0.056 .267 .647 .869 .974 1.00 1.00
1.00 1.00 1.

94239 1.140 .885247 3.80269 2.18E-02 2.160

94240 1.109 .794930 4.68927 1.02E+03 0.063 .295 .628 .881 .980 .998 1.00
1.00 1.00 1.

94241 1.079 .842472 4.15150 5.00E-02 2.250

94242 1.069 .819150 4.36668 1.72E+03 0.068 .297 .631 .879 .979 .997 1.00
1.00 1.00 1.

95241 1.079 .933020 3.46195 1.18E+00 3.220

\* 96242 1.053 .887353 3.89176 2.10E+07 0.021 .168 .495 .822 .959 .996
.999 1.00 1.00 1.

96244 1.036 .902523 3.72033 1.08E+07 0.015 .131 .431 .764 .948 .991 1.00
1.00 1.00 1.

96246 0.000 .000000 0.00000 0.00E+00 0.015 .091 .354 .699 .917 .993 1.00
1.00 1.00 1.

96248 0.000 .000000 0.00000 0.00E+00 0.007 .066 .287 .638 .892 .982 .998
1.00 1.00 1.

97249 1.079 .891281 3.79405 1.00E+05 3.400

98246 0.000 .000000 0.00000 0.00E+00 0.001 .114 .349 .623 .844 .970 1.00
1.00 1.00 1.

98250 0.000 .000000 0.00000 0.00E+00 0.004 .040 .208 .502 .801 .946 .993
.997 1.00 1.

98252 1.207 1.180000 1.03419 2.34E+12 0.002 .028 .153 .427 .733 .918
.984 .998 1.00 1.

98254 0.000 .000000 0.00000 0.00E+00 0.000 .019 .132 .396 .714 .908 .983
.998 1.00 1.

100257 0.000 .000000 0.00000 0.00E+00 0.021 .073 .190 .390 .652 .853
.959 .993 1.00 1.

102252 0.000 .000000 0.00000 0.00E+00 0.057 .115 .207 .351 .534 .717
.863 .959 .997 1.

\* = used in problem.

All of the available data are printed when Print Table 38 is turned on
by the **PRINT** card. Data actually used are denoted by an \*. If any
data are overridden by FMULT user input, the user data replaces the
default data shown in Print Table 38.

[]{#__RefHeading___Toc100926598 .anchor}**2.10.1. Nuclides without
Transport Cross Sections**

MCNP(X) does not provide nuclear cross sections for transporting
^246^Cf, ^254^Cf, ^257^Fm, and ^252^No.

To model spontaneous fission from these nuclides, it is necessary to do
the transport either with a physics model or by substituting cross
sections. Physics models are not recommended at fission energies. To
make a nuclide substitution, the AWTAB and MX cards must be used. For
example,

M123 100257 1.

AWTAB 100257 257.

MX123:N 98252.

The AWTAB card provides the atomic weight ratio for ^257^Fm, which is
not provided in the standard MCNP(X) data libraries. The MX123:N card
substitutes ^252^Cf, for which there is neutron cross section data, for
the corresponding nuclide (100257) on the M123 materials card.

### 2.10.2. Nuclides without Multiplicity Data {#nuclides-without-multiplicity-data .unnumbered}

Fission widths, watt fission spectra parameters, and fission yields are
not available for the above nuclides, which do not have transport cross
sections: ^246^Cf, ^254^Cf, ^257^Fm, and ^252^No. They are also not
available for ^246^Pu, ^246^Cm, ^248^Cm, and ^250^Cf. Note that the
values for width, watt1, watt2, and yield in print table 38 (above) are
all zero for these eight nuclides. To have a spontaneous fission source
for these nuclides, an FMULT input file data card is required. For
example,

FMULT 96246 width = 1.1 watt = .2 4 sfyield 1 .

Because the multiplicities are provided as a table with 10 bins, the
width is ignored for the spontaneous fission source; however, a value is
still required for induced fission. For spontaneous fission, the energy
distribution is sampled from the two watt-fission spectra parameters;
for induced fission, the energy spectra is chosen from parameters in the
nuclear data tables of the transport cross sections. Finally, the
spontaneous fission yield must be specified if more than one spontaneous
fission source nuclide occurs. The yield is used to determine the\
relative sampling between spontaneous fission source nuclides. These
parameters have no default values; if the FMULT card is missing, a fatal
error message is issued:

fmult card needed for za = 96246 .

# SOURCE features

Most MCNPX source extensions are controlled by the SDEF card:

-   source particles that may be specified by descriptors,

-   positron sources,

-   spontaneous fission,

-   multiple-source-particle types,

-   sources on cylindrical surfaces,

-   repeated structures source-paths, and

-   translated sources with dependence.

## Particle-Type Specification: SDEF

The source particle type may be specified on the SDEF card by either its
symbol

SDEF PAR = **h**

or by its number

SDEF PAR = 9 .

## Positron Sources: SDEF

Positron sources may be specified as

SDEF PAR = **-e**

or

SDEF PAR = **-3** .

Note that positron physics in MCNPX, just as with MCNP and the
Integrated Tiger Series (ITS), is identical to electron physics, except
for positron annihilation. Electrons below the energy cutoff are
terminated, whereas positrons below the energy cutoff produce
annihilation photons. Also, the positrons have a positive charge and may
be tallied using the FT card ELC option.

## Spontaneous Fission Sources: SDEF

The spontaneous fission source can be specified in either of two ways:

SDEF par = **SF** normalize summary and tally information by the number
of spontaneous fission neutrons, or

SDEF par = **SF** normalize summary and tally information by the number
of histories (generally the number of spontaneous fissions).

### Spontaneous Fission Physics

Eighteen nuclides are available: ^232^Th, ^232^U, ^233^U, ^234^U,
^235^U, ^236^U, ^238^U, ^237^Np, ^238^Pu, ^239^Pu, ^240^Pu, ^241^Pu,
^242^Pu, ^241^Am, ^242^Cm, ^244^Cm, ^249^Bk, and ^252^Cf.

Cells are sampled according to the usual SI and SP distributions. If
more than one spontaneous fission nuclide is present in a source cell,
the fissioning nuclide will be chosen proportionately to the product of
its atom fraction and the spontaneous fission yield for each nuclide. If
no spontaneous fission nuclide is found in a specified source cell, the
code exits with a *bad trouble* error, "spontaneous fission impossible."

The number of spontaneous fission neutrons then is sampled. The
spontaneous fission multiplicity data from Santi and references cited by
him (Section 2.10) are used. The energies are sampled from a Watt
spectrum with appropriate spontaneous fission parameters for the
selected nuclide. Currently, only the first spontaneous fission neutron
from each history is printed. If the spontaneous fission samples a
multiplicity of zero---that is, no neutrons for a given spontaneous
fission---then the history is omitted from the first 50 history lists of
Print Table 110. The number of source particles is equal to the number
of spontaneous fission neutrons, which will be times the requested
number of source histories on the NPS card.

Fission multiplicity for induced fissions (6h entry, PHYS:N card)
automatically is turned on with the default width (FISM = -1 = nuclide
dependent). If FISM \> 0 on the PHYS:N card, then that value will be
used.

### Normalization of Summary and Tally Information

The MCNPX spontaneous fission source is different from most other SDEF
fixed sources. Let

N = NPS from the NPS card = the number of source particle histories,

W = the average source particle weight, and

= the average number of spontaneous fission neutrons per fission.

For most other MCNPX fixed source (SDEF) problems,

1.  summary table source tracks = N,

    summary table source weight = W, and

    summary tables and tallies are normalized by N.

For the spontaneous fission source, SDEF par = SF ,

4.  summary table source tracks = ·N,

    summary table source weight = W, and

summary tables and tallies are normalized by ·N.

For the spontaneous fission source, SDEF par = SF,

6.  summary table source tracks = ·N,

    summary table source weight = ·W, and

summary tables and tallies are normalized by N.

### Fission Multiplicity Output

Multiplicity and moments are printed in Print Table 117 for both
spontaneous and spontaneous plus induced fissions.

spontaneous fission source multiplicity and moments. print table 117

\-\-\-\-\-- by number \-\-\-\-\-\-\-\-\-\--
\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-- by weight
\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\--

fission multiplicity fission multiplicity

fissions neutrons fraction fissions neutrons fraction error

nu = 0 697 0 6.97000E-02 3.24005E-02 0.00000E+00 6.97000E-02 0.0373

nu = 1 2295 2295 2.29500E-01 1.06685E-01 1.06685E-01 2.29500E-01 0.0197

nu = 2 3261 6522 3.26100E-01 1.51590E-01 3.03180E-01 3.26100E-01 0.0161

nu = 3 2518 7554 2.51800E-01 1.17051E-01 3.51153E-01 2.51800E-01 0.0187

nu = 4 1025 4100 1.02500E-01 4.76478E-02 1.90591E-01 1.02500E-01 0.0305

nu = 5 183 915 1.83000E-02 8.50688E-03 4.25344E-02 1.83000E-02 0.0736

nu = 6 21 126 2.10000E-03 9.76199E-04 5.85720E-03 2.10000E-03 0.2181

total 10000 21512 1.00000E+00 4.64857E-01 1.00000E+00 1.00000E+00 0.0073

factorial moments by number by weight

nu 2.15120E+00 0.0054 2.15120E+00 0.0054

nu(nu-1)/2! 1.91100E+00 0.0116 1.91100E+00 0.0116

nu(nu-1)(nu-2)/3! 8.86800E-01 0.0219 8.86800E-01 0.0219

nu(nu-1) \.... (nu-3)/4! 2.25500E-01 0.0439 2.25500E-01 0.0439

nu(nu-1) \.... (nu-4)/5! 3.09000E-02 0.0987 3.09000E-02 0.0987

nu(nu-1) \.... (nu-5)/6! 2.10000E-03 0.2180 2.10000E-03 0.2180

spontaneous and induced fission multiplicity and moments. print table
117

\-\-\-\-\-- by number \-\-\-\-\-\-\-\-\-\--
\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-- by weight
\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\-\--

fission multiplicity fission multiplicity

fissions neutrons fraction fissions neutrons fraction error

nu = 0 700 0 6.90949E-02 3.25400E-02 0.00000E+00 6.90949E-02 0.0372

nu = 1 2301 2301 2.27125E-01 1.06964E-01 1.06964E-01 2.27125E-01 0.0197

nu = 2 3286 6572 3.24351E-01 1.52752E-01 3.05504E-01 3.24351E-01 0.0161

nu = 3 2558 7674 2.52492E-01 1.18910E-01 3.56731E-01 2.52492E-01 0.0186

nu = 4 1059 4236 1.04531E-01 4.92283E-02 1.96913E-01 1.04531E-01 0.0300

nu = 5 205 1025 2.02349E-02 9.52956E-03 4.76478E-02 2.02349E-02 0.0695

nu = 6 22 132 2.17155E-03 1.02269E-03 6.13611E-03 2.17155E-03 0.2131

total 10131 21940 1.00000E+00 4.70946E-01 1.01990E+00 1.00000E+00 0.0072

factorial moments by number by weight

nu 2.16563E+00 0.0054 2.16563E+00 0.0054

nu(nu-1)/2! 1.94393E+00 0.0115 1.94393E+00 0.0115

nu(nu-1)(nu-2)/3! 9.16395E-01 0.0216 9.16395E-01 0.0216

nu(nu-1) \.... (nu-3)/4! 2.38279E-01 0.0426 2.38279E-01 0.0426

nu(nu-1) \.... (nu-4)/5! 3.32642E-02 0.0932 3.32642E-02 0.0932

nu(nu-1) \.... (nu-5)/6! 2.17155E-03 0.2130 2.17155E-03 0.2130

In the above problem, there were 10000 source histories; all were
spontaneous fissions. There were 21512 spontaneous fission neutrons
produced. The number of source particles and the source weight listed in
the problem summary table for neutrons is also 21512 and 2.1512E+00.
Also, from the problem summary table for neutrons, there were 131
induced fissions, producing 428 fission neutrons.

In the second Print Table 117, there are 10000 + 131 = 10131 fissions.
There are 21512 + 428 = 21940 fission neutrons. For nu = 0, the
multiplicity fraction is 700 / 10131 = 0.0690949. The total multiplicity
fraction always sums to 1. By-weight fissions sum to 10131 / 21512 =
0.470946. By-weight fission neutrons sum to 21940 / 21512 = 1.01990.

The by-weight multiplicity fractions are the same as the by-number
multiplicity fractions because analog capture is used in this problem
and the entire source is spontaneous fission.

If SDEF PAR = -SF is used, the tallies, summary information, and some
entries in Print Table 117 are normalized by source history. By-weight
fissions is 10131 / 10000 = 1.01310 instead of 0.470946. By-weight
fission neutrons is 2.19400 instead of 1.01990. The by-weight
multiplicity fractions are divided by the total fission weight, 10131.0,
to sum to unity.

The first moment is nu = 21940 / 10131 = 2.16563. The second moment is (
2 1 / 2! 3286 + 3 2 / 2! 2558 + 4 3 / 2! 1059 + 5 4 / 2! 205 + 6 5 / 2!
22 ) / 10131 = 1.94393.The third moment is ( 3 2 1/3! 2558 + 4 3 2 / 3!
1059 + 5 4 3 / 3! 205 + 6 5 4 / 3! 22 ) / 10131 = 0.916395.The fourth
moment is ( 4 3 2 1 / 4! 1059 + 5 4 3 2 / 4! 205 + 6 5 4 3 / 4! 22 ) /
10131 = 0.238279.The fifth moment is ( 5 4 3 2 1 / 5! 205 + 6 5 4 3 2 /
5! 22 ) / 10131 = 0.0332642.The sixth moment is 6 5 4 3 2 1 / 6! 22 /
10131 = 0.00217155.

## Multiple Particles on SDEF Card

Multiple-source-particle types may be specified in fixed-source
problems. PAR on the SDEF card may be either a distribution or a
dependent distribution.

### Source Particle Type as the Independent Variable

sdef par=d1 pos fpar d2 erg fpar d3

si1 L h n

sp1 2 1

sb1 1 2

ds2 L 0 0 0 15 0 0

ds3 L 2 3

The particle types are protons (at 0,0,0, energy 2 MeV) and neutrons (at
15,0,0, energy 3 MeV).

### Source Particle Type as a Dependent Variable

sdef par=fpos d2 pos=d1 erg fpos d3

si1 L 0 0 0 15 0 0

sp1 2 1

sb1 1 2

ds2 L h n

ds3 L 2 3

The source particles are protons and neutrons, depending on the problem
source positions. This source tracks the source described in Section
3.4.1 exactly.

### Additional Capabilities and Cautions

-   The characters L, S, Q, F, and T are reserved as SI card options. L
    = discrete source variables, S = distribution numbers, etc. If the
    first entry on the SI card is L, S, Q, F, or T, it will be
    interpreted as a distribution option. To use source particle types
    L, S, Q, F, or T, either the corresponding particle numbers (10, 33,
    30, 27, or 32) must be used or L, S, Q, F, or T must be used as the
    second or later particle type. Generally, it is best to specify the
    discrete source variable option; therefore, L will be the first
    entry, followed by the particle types. A second L will be
    interpreted correctly as particle type L. For example,

SI99 L H N L Q F T S .

-   Antiparticles may be designated. For example,

SI77 L E N --H .

-   Either characters (n, p, e, h, d, s, t, a, etc.) or numbers (1, 2,
    3, 9, 31, 32, 33, 34, etc.) may be used. For example,

SI98 L --H 3 32 N .

-   Spontaneous fission may be used as a particle type. For example,

SI87 L SF N .

-   Particle types may be listed multiple times to give them different
    energy distributions, angular distributions, etc., in different
    parts of the problem. For example,

SI23 L N n 1 n N .

### Tally and Summary Table Normalizations

Tallies are normalized by dividing the total source weight by the number
of source histories. Weight (WGT on the SDEF card) cannot be a source
distribution (either independent or dependent). The weight of particles
in the summary tables is controlled by the SI, SP, SB, and DS cards for
the particle distribution. In the example shown in Section 3.4.1,

> sdef par=d1 pos fpar d2 erg fpar d3
>
> si1 L h n
>
> sp1 2 1
>
> sb1 1 2
>
> ds2 L 0 0 0 15 0 0
>
> ds3 L 2 3 .

The total source weight is WGT = 1.0 by default. From the SP1 card, the
weight of the neutrons that are produced is 0.3333 and the weight of the
protons that are produced is 0.6667. From the SB1 card, the total number
of neutron tracks is 0.6667 NPS for neutrons and 0.3333 NPS for protons
(where NPS is the number of source histories on the NPS card). The
energy per source particle is normalized to the source particle weight
for each source particle type. If the particle type is not a source
particle (e.g., photons in the above problem), then the energy per
source particle is normalized to the source particle weight of the
lowest particle type. In this example, photon source energy would be
normalized in the photon creation-and-loss summary table by 0.3333,
which is the weight of the source neutrons produced.

Fission multiplicity and FT8 CAP capture tallies (Print Tables 117 and
118) "by number" quantities are always normalized by NPS, the number of
histories run (NPS card). NPS generally equals the number of fissions,
not fission neutrons, plus other source particles. The "by-weight"
quantities are normalized by NPS for PAR = SF and the "by-fission"
neutrons (plus other source particles) for PAR = SF.

Spontaneous fission has two normalizations: PAR = SF and PAR = SF. For
PAR = SF, the normalization is total source weight, WGT, divided by the
total neutron source weight: spontaneous fission neutrons plus other
source neutrons.

## Sources on Cylindrical Surfaces: SDEF

Cylindrical surface sources may be specified on the SDEF card.
Furthermore, particle directions relative to the cylindrical source
normal may be specified. The cylindrical surface can be, but does not
have to be, a cell-bounding problem surface. Likewise, a spherical
surface source no longer has to be on a cell-bounding problem surface.

In MCNP and earlier MCNPX versions, the only way to specify a
cylindrical surface source was to have a degenerate cylindrical volume
source (radius = constant) that is not also a problem surface. Particle
direction had to be isotropic from the source point.

### Examples

The following examples now work, whereas all failed in previous MCNP and
MCNPX versions.

Cylindrical problem surface with default cosine distribution relative to
surface normal:

SDEF pos = 0 0 0 rad = 1 ext = d1 axs = 1 0 0 sur = 5 .

Cylindrical problem surface with specified angle distribution relative
to surface normal:

SDEF pos = 0 0 0 rad = 1 ext = d1 axs = 1 0 0 sur = 5 dir = d2 .

Cylindrical surface (degenerate volume source) with specified angle
distribution relative to surface normal:

SDEF pos = 0 0 0 rad = 1 ext = d1 axs = 1 0 0 dir = d2 .

Spherical surface (degenerate volume source) with specified angle
distribution relative to surface normal:

SDEF pos = 0 0 0 rad = 1 dir = d2 .

### Additional Information and Cautions

If the cylindrical surface is a problem surface, then the surface number
must be specified on the SDEF card with the SUR parameter.

If both DIR and VEC are specified, then particle directions are relative
to VEC rather than to the cylindrical surface normal.

All former capabilities remain the same:

-   volume sources have a default isotropic direction;

-   particle directions are relative to VEC if VEC is specified;

-   only capabilities that formerly were disallowed are different:

```{=html}
<!-- -->
```
-   SUR may be specified on a cylindrical surface, and the default VEC
    is the surface normal;

-   DIR may be specified without VEC (so that VEC defaults to the
    surface normal) for spheres and cylinders; and

-   the outward normal to cylindrical surfaces is generated whenever DIR
    is specified and VEC is not.

## Repeated Structures Source Specifications: SI

The CEL source specification for repeated structures geometries now is
consistent with the tally specification. The old MCNP4C specification
still works, but the MCNPX is

> SDEF CEL=d3 POS=0 6 0 EXT=d1 RAD=d2 AXS= 0 1 0
>
> SI3 L (1\<10\[0 0 0\]\<11) (1\<10\[1 0 0\]\<11) (1\<10\[2 0 0\]\<11)
>
> (1\<10\[0 1 0\]\<11) (1\<10\[1 1 0\]\<11) (1\<10\[2 1 0\]\<11) .

## Translated Sources with Dependence

MCNPX allows translating sources to different locations with the TR
option on the SDEF card. This capability is most useful for setting up
the source as an accelerator beam and then using the translation as a
distribution to repeat the accelerator source at different locations and
orientations.

The TR option on the SDEF card can be dependent on other source
variables. For example, the particle type can depend on the translated
source location:

SDEF CEL=FTR=D3 PAR=FTR=D1 TR=D2 .

Or the translated source location can be a dependent distribution
function of cell:

SDEF CEL=D2 TR=FCEL=D5 .

Thus, the TR parameter on the SDEF card is consistent with the other
source parameters.

# TALLY features

The MCNPX tally extensions provide a wide range of significant
capabilities. Note that logarithmic interpolation (Section 5.1) is
particularly useful for many tally cards. The applicable tally input
cards are found in parentheses after the features they control.

-   PHTs with variance reduction (F8);

-   Default dose functions (DF);

-   Lattice tally speedup by orders of magnitude (F);

-   Radiography tallies: translate new MCNP5 FIP, FIR, and FIC
    radiography input cards as MCNPX radiography tallies PI, TIR, and
    TIC;

-   Proton and photonuclear reaction multipliers (FM);

-   Cosine bins that may be specified in degrees and for F2 tallies (C);

-   Pulse-height light tally with anticoincidence: FT8 PHL;

-   Coincidence capture tally: FT8 CAP;

-   Residual nuclei tally: FT8 RES.

## 4.1. Pulse-Height Tallies with Variance Reduction {#pulse-height-tallies-with-variance-reduction .unnumbered}

MCNPX allows variance reduction techniques (VRTs) with F8 PHTs. Unlike
standard Monte Carlo tallies, which score events as they occur, PHTs
combine the energy deposition of many events in a single Monte Carlo
"history." The entire history tree must score, not just a single event
in the tree. Further, VRT events such as Russian roulette can kill the
entire tree if they occur after a physical event such as photon pair
production.

MCNPX enables VRTs with PHTs.[^12] Results are very problem dependent,
and variance reduction is more challenging than with other MCNPX
tallies. On a space satellite problem, calculations were sped up by two
orders of magnitude.[^13] Although the MCNPX capability greatly enhances
most F8 PHT calculations, it has not been extended to all VRTs or
physics combinations. See Section 4.1.1, "Future Work."

No user interface changes are required to use VRTs with PHTs. If VRTs
are used with PHTs, a fatal error no longer occurs unless the VRTs are
incompatible with PHTs. Several warning messages occur if it appears
that the VRTs will be inefficient with PHTs. If no VRTs are specified,
MCNPX uses the MCNP4C3 analog algorithm without constructing PHT trees.

### 4.1.1. Future Work {#future-work .unnumbered}

PHTs work with VRTs in the most widely used F8 PHT applications:
electrons and photons.

They do not work with other particle types and have other extensions
that must await further funding. Among these future work items are

-   enable neutrons and other particles to use VRTs with PHTs;

-   enable more than a single F8 tally per problem;

-   enable DXTRAN and SPABI VRTs; and

-   extend to FT special treatments, particularly the light tally with
    anticoincidence, residual nuclei, and capture options (FT PHL, RES,
    and CAP options).

## 4.2. Default Dose Functions (DF) {#default-dose-functions-df .unnumbered}

The DE/DF dose function cards are unchanged but now have extensions. As
before, dose conversions may be input as a table. Note that the
interpolation int = log or int = lin may be placed anywhere, and n =
tally number, which implies particle type.

> DEn E1 E2 int E3 \...
>
> DFn F1 int F2 F3 \...

The dose conversion capability is extended to provide standard default
dose functions. These are invoked by omitting the DE card and using
keywords on the DF card

**DFn iu=j fac=F int ic=I** ,

where all of the following entries are optional.

> iu = 1 = US units (rem/h)
>
> iu = 2 = international units (sieverts/h)
>
> Default: iu = 2 international units (sieverts/h)

fac = normalization factor for dose (acr is also accepted instead of
fac).

> fac = -1 = normalize results to Q = 20 by dividing the parametric form
> of Q \[5.0+17.0 exp(-(ln(2E)) 2/6)\] from ICRP60 (1990), paragraph
> A12.
>
> fac = -2 = apply LANSCE albatross response function.

Default: fac = 1.0.

int = "log" or "lin" results in "log" or "lin" interpolation of energy;
the dose function is always linear. That is, "lin" results in "linlin"
interpolation, and "log" results in "loglin" interpolation.

Default: for ic = 10, 40: log

for ic = 20,31-39: recommended analytic parameterization.

ic = i = standard dose function.

> [i neutron dose function]{.underline}
>
> 10 = ICRP-21 1971
>
> 20 = NCRP-38 1971, ANSI/ANS-6.1.1-1977
>
> 31 = ANSI/ANS-6.1.1-1991 (AP anterior-posterior)
>
> 32 = (PA posterior-anterior)
>
> 33 = (LAT side exposure)
>
> 34 = (ROT normal to length and rotationally symmetric)
>
> 40 = ICRP-74 1996 ambient dose equivalent

[i photon dose function]{.underline}

10 = ICRP-21 1971

> 20 = Claiborne & Trubey, ANSI/ANS 6.1.1-1977
>
> 31 = ANSI/ANS-6.1.1-1991 \[AP (anterior-posterior)\]
>
> 32 = PA (posterior-anterior)
>
> 33 = (LAT side exposure)
>
> 34 = (ROT normal to length and rotationally symmetric)
>
> 35 = (ISO isotropic)

Default: ic = 10

> *Examples:*
>
> DF4
>
> DF0 ic 40 iu 1 lin fac 123.4
>
> DF1 iu=2 acr=-2 log ic=34

## 4.3. Lattice Tally Speedup (F) {#lattice-tally-speedup-f .unnumbered}

With no change in the user interface, input files with large lattice
tallies run 10 to 1000 times faster than in MCNP4C3 if the following
apply:

-   The lattice is specified fully on the cell fill card, e.g., FILL
    50:50 --50:50\
    --50:50.

-   The tally chain refers to no more than one cell at each level,
    except for the lattice cell, which must include the entire range of
    all indices specified on the corresponding FILL card:

> F4:p ( 1 \< 2 \< 3\[50:50 --50:50 --50:50\] \< 4 ) .

The following,

F4:p ( 1 \< 2 \< 3\[50:50 --50:50 --50:50\] \< 4 5 ) ,

will not work because two cells, 4 and 5, were specified at the top
level.

*Example:*

The following example of a simple input file runs 70 times faster with
MCNPX than with MCNPX 2.4.0 and MCNP4C versions.

21x21x21 void lattice of balls

11 0 -31 u=1 imp:p=1

12 0 31 u=1 imp:p=1

16 0 -32 u=2 imp:p=1

lat=1 fill= -10:10 -10:10 -10:10 1 9260R

17 0 -33 fill=2 imp:p=1

18 0 33 imp:p=0

31 sph 0 0 0 .5

32 rpp -1 1 -1 1 -1 1

33 rpp -21 21 -21 21 -21 21

mode p

print

prdmp 2j -3

sdef

nps 10000

f4:p (11\<16\[-10:10 -10:10 -10:10\]\<17)

Larger lattices and nested lattices offer even more dramatic speedups
but take longer to demonstrate than the simple example shown.

## 4.4. Radiography Tally (FIP, FIR, FIC) {#radiography-tally-fip-fir-fic .unnumbered}

Radiography tallies may be input with either the standard MCNPX card
names (PI, TIR, TIC) or the MCNP5 card names (**FIP, FIR, and FIC**). In
all cases, the radiographical plane is specified with C and FS cards for
the s and t axes. The tally number is n and the particle type is p,
which may be either neutron, N, or photon, P.

### 4.4.1. Pinhole Radiography {#pinhole-radiography .unnumbered}

> **FIPn:p** *X1 Y1 Z1 R0 X2 Y2 Z2 F1 F2 F3*
>
> **PIn:p** *X1 Y1 Z1 R0 X2 Y2 Z2 F1 F2 F3* ,

where

> *X1 Y1 Z1 =* coordinates of the pinhole,
>
> *R0 =* 0,
>
> *X2 Y2 Z2 =* reference coordinates (center of object),
>
> *F1 =* collimator radius,
>
> *F2 =* pinhole size, and
>
> *F3 =* distance from pinhole to detector grid defined by C and FS
> cards.

### 4.4.2. Transmitted Image Projection {#transmitted-image-projection .unnumbered}

*Rectangular grid:*

**FIRn:p** *X1 Y1 Z1 R0 X2 Y2 Z2 F1 F2 F3*

**TIRn:p** *X1 Y1 Z1 R0 X2 Y2 Z2 F1 F2 F3*

*Cylindrical grid:*

**FICn:p** *X1 Y1 Z1 R0 X2 Y2 Z2 F1 F2 F3*

**TICn:p** *X1 Y1 Z1 R0 X2 Y2 Z2 F1 F2 F3* ,

where

> *X1 Y1 Z1 =* coordinates of center of grid defined by C and FS cards,
>
> *R0 =* 0,
>
> *X2 Y2 Z2 =* reference coordinates (center of object),
>
> *F1 =* -1 / 0 = scattered only/direct + scattered---contributions to
> detector,
>
> *F2 =* radial field of view, and
>
> *F3 =* 0 / -1 contributions scored at center/random---on grid.

## 4.5. Photonuclear and Proton Reaction Multipliers (FM) {#photonuclear-and-proton-reaction-multipliers-fm .unnumbered}

Photonuclear and proton cross sections may be used in tally multipliers
on the FM card. For example,

M102 92235 1 pnlib = 27u

F2:P 1

FM2 (-1 102 18 1018) .

Photonuclear cross-section reaction numbers all are positive, unlike the
photoatomic reaction numbers, which are negative. The principal
photonuclear cross sections are the following: 1 = total, 2 =
nonelastic, 3 = elastic, 4 = heating, and \>4 = various reactions such
as 18 = *(,f)*. The photonuclear yields (multiplicities) for various
secondary particles are specified by adding 1000 times the secondary
particle number to the reaction number. For example, 31001 is the total
yield of deuterons (particle type d = 31), 34001 is the total yield of
alphas (particle type = 34), and 1018 is the total number of neutrons
(particle type n = 1) from fission.

Proton reaction numbers are similar to the neutron reaction
numbers---all positive. The principal proton cross sections are 1 =
total, 2 = nonelastic, 3 = elastic, 4 = heating, \>4 = various
reactions. On the LA150H proton library, the only available reaction is
mt = 5 and its multiplicities, 1005, 9005, 31005, etc. The multiplicity
reaction numbers are specified by adding 1000 times the secondary
particle number to the reaction number. For interaction reaction mt = 5,
the multiplicities are 1005 for neutrons, 9005 for protons, 31005 for
deuterons, etc. The proton multiplicity, mt = 9001, 9004, 9005, etc., is
generally available, along with the total cross-section and heating
number, mt = 1, mt = 4.

It is always wise to plot the desired cross sections first to see if
they are available with the expected reaction numbers in the data
library. The tally multipliers treat the data the same as the data are
treated in transport: the cross section at the lowest energy is extended
down to E = 0 for protons with mt \< 0; the cross section at the highest
energy of the table is extended to E = for proton interaction cross
sections with mt \< 0, and for photonuclear interaction cross sections,
mt \< 1000. These extrapolations can be seen in the cross-section plots.

## 4.6. Expanded Cosine Specification (C) {#expanded-cosine-specification-c .unnumbered}

Cosines may be specified in degrees. They also may be specified with
surface flux (F2) tallies as

\*C2 150 120 90 60 30 0 .

The \* on the C2 card interprets cosines as in degrees. Entries must be
such that the cosine is monotonically increasing.

## 4.7. Pulse-Height Light Tally with Anticoincidence: FT8 PHL {#pulse-height-light-tally-with-anticoincidence-ft8-phl .unnumbered}

Pulse-height (F8) tallies may be based on energy/light deposition in one
or two other regions as specified via one or two F6 tallies. Thus, this
tally is dependent on results from another tally, which works because
the F8 tally is posted at the end of the particle history where the F6
tallies are accumulated along each track of the particle history. The
format of the FT PHL card is

> FT8 PHL N T~A1~ B~A1~ T~A2~ B~A2~ \... M T~B1~ B~B1~ T~B2~ B~B2~ \...
> ,

where

> N = number of F6 tallies for the first detector region,
>
> T~A1~ B~A1~ \... = pairings of tally number and f-bin number for the N
> F6 tallies of the first detector region,
>
> M = number of F6 tallies for the second detector region, and
>
> T~B1~ B~B1~ \... = pairings of tally number and f-bin number for the M
> F6 tally of the second detector region.

When M is nonzero, indicating the use of two detector regions, an FU
card is required for the F8 tally. The entries on the FU card are
presented in units of electron-equivalent light (MeV~ee~) and must
increase monotonically. The particle type indicated on the F8 tally does
not matter because this tally allows a combination of light output from
various particle types. If B~An~ is zero, then the number of cell bins
on the F8 card must match that on the corresponding T~An~ tally card.
Setting B~An~ to zero allows for a lattice pulse-height PHL tally.

*Examples:*

Case 1

F8:n 5

FT8 PHL 1 6 1 0

E8 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0

F6:e 5

DE6 LIN 1.0 1.5 2.0 2.5 3.0 3.5 10.0

DF6 LIN 1.0 1.1 1.2 1.3 1.4 1.5 1.6

FT6 GEB A B C

Case 2

F8:n 5

FT8 PHL 1 6 1 1 16 1

E8 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0

FU8 1.5 2.5 3.5 4.5 5.5 6.5 7.5 8.5

F6:e 5

DE6 LIN 1.0 1.5 2.0 2.5 3.0 3.5 10.0

DF6 LIN 1.0 1.1 1.2 1.3 1.4 1.5 1.6

FT6 GEB A B C

F16:e 6

DE16 LIN 1.0 1.5 2.0 2.5 3.0 3.5 10.0

DF16 LIN 1.0 1.1 1.2 1.3 1.4 1.5 1.6

FT16 GEB A B C

In both cases, the F6 tallies convert energy deposition to equivalent
light (units in millielectron volts). SD cards are not required with the
F6 tallies because these divisors renormalize only the printed output
for the F6 tallies and not the values stored in the tally arrays (thus,
the F8 tally will result in the same value, regardless of whether the F6
tally has an SD card). The DE/DF conversion is based on the incident
particle energy, and the values on the DF card should be the dL/dE for
that incident particle energy. Thus, the F6 tally will multiply the
dL/dE values by the energy deposition to give the light output (ΔL)
summed over each track. Also, no energy bins exist for the F6 tallies.
The F8 tally uses the total light output. Energy bins (E6 card) can be
added, but the F8 tally will use the value from the total bin.
Similarly, for other bins associated with the F6 tally, in each case,
the TFC bin is used to extract the value for the F8 tally (see the TF
card to alter this). The FT GEB cards are used to perform Gaussian
broadening on these tally values; however, this process is done only at
the end of the particle history to determine the light output value used
in the PHT.

In Case 1, the electron light output from only one region (cell 5) is
used to subdivide the PHT. In this case, a pulse of 1 (input source
weight) is put into the first E8 bin when the light output in cell 5 is
\<1 MeV. It is placed in the second E8 bin when the light output is
between 1 and 2 MeV, etc. A zero F6 tally will result in no F8 tally.

In Case 2, the light output from two regions (cells 5 and 6) is used to
subdivide the PHT. This case is useful for coincidence/anticoincidence
applications. A pulse of 1 (input source weight) is put into the first
E8 bin and into the first FU8 bin when the light output in cell 5 is
\<1.0 MeV *and* the light output in cell 6 is \<1.5 MeV. This pulse is
put into the first E8 bin and into the second FU8 bin when the light
output in cell 5 is \<1.0 MeV *and* the light output in cell 6 is
between 1.5 and 2.5 MeV. A zero light output in both cells will result
in no F8 tally. A zero light output in cell 5 (tally 6) with a nonzero
light output in cell 6 (tally 16) will result in a pulse in the
corresponding FU8 bin. Similarly, for a zero light output in cell 6 and
a nonzero light output in cell 5, a pulse will be put into the
corresponding E8 bin. The E8 and FU8 bins do not have to be the same and
typically would not be unless the detector regions were of similar
material and size. Separate F6 tallies (as in Case 2, F6 and F16) are
needed only when the two regions have different light conversion
functions. If the two regions are of the same material, then a single F6
tally can be used as follows.

F8:n 5

FT8 PHL 1 6 1 1 6 2

E8 1.0 2.0 3.0 4.0 5.0 6.0 7.0 8.0

FU8 1.5 2.5 3.5 4.5 5.5 6.5 7.5 8.5

F6:e 5 6

DE6 LIN 1.0 1.5 2.0 2.5 3.0 3.5 10.0

DF6 LIN 1.0 1.1 1.2 1.3 1.4 1.5 1.6

Currently, it is not important to determine what cell is listed on the
F8 card because this tally is made only at the end of a particle history
and depends only on the tally results of the listed F6 tallies. Having
multiple cells listed on the F8 card is meaningful only when the f-bin
parameter (i.e., B~An~ or B~Bn~) of the FT PHL option is zero,
indicating a lattice grid of detector regions. Otherwise, each
additional F8 cell bin simply will be a duplicate of the first cell bin.

## 4.8. Coincidence Capture Tally (FT8 CAP) {#coincidence-capture-tally-ft8-cap .unnumbered}

The FT8 capture tally scores the number of captures in specified
combinations of nuclides at the end of each history. The tally is
particularly useful for neutron coincidence detectors. In addition,
captures may be written to an auxiliary output file, PTRAC. Section 5.5
describes the PTRAC capture file:

FT8 \[-Mc\] \[-Mo\] I1 I2 ... \[Gate td tw \] ,

where

> Mc = optional maximum number of captures (default = 21),
>
> Mo = optional maximum number of moments (default = 12),
>
> In = capture nuclide(s) such as 3006 or 5010 for ^6^Li or ^10^B, and
>
> Gate td tw = optional time gate with pre-delay time, td, and time gate
> width, tw.

*Example:*

F8:n 2 (5 6) 7 T

FT8 cap 3006 5010

T8 1 7LOG 1E8 .

In this example, captures and moments are tallied in cells 2, 7, in the
combination of 5 and 6 and in the total over cells 2, 5, 6, 7. The
captures by either ^6^Li or ^10^B are tallied. Results are tabulated in
time bins at 1, 10, 100, 1000, 1e4, 1e5, 1e6, 1e7, and 1e8 shakes---that
is, in the range of 10 ns to 1 s.

### 4.8.1. Restrictions {#restrictions .unnumbered}

Coincidence counting of capture multiplicities and moments requires
analog capture: CUT:N 2J 0 0. Calculations must be totally analog, with
no variance reduction. Fission multiplicity also is required: PHYS:N J
100 3J --1. An FT8 CAP tally in an input file automatically will set
analog capture, fission multiplicity, and exit with error messages if
variance reduction is used. The capture multiplicities and moments are
stored in 80 cosine bins, which are printed out with the F8 tally. A
much more readable table of capture multiplicities and moments is given
in Print Table 118. The captures and moments can be compared with Print
Table 117, which has the spontaneous-fission source and induced-fission
summaries of fission neutrons and moments (Section 3.3.3).

The capture tallies may be written to a PTRAC file for further analysis
by auxiliary codes. See Section 5.5 on the PTRAC card extensions.

### 4.8.2. Interpreting Capture Tally Output {#interpreting-capture-tally-output .unnumbered}

The FT8 CAP coincidence capture tally option produces both a standard
tally, which is generally unreadable, and a coincidence capture table,
Print Table 118. An example is provided to help in the interpretation of
this table.

neutron captures, moments & multiplicity distributions. tally 8 print
table 118

cell: 999

neutron captures on 3he

captures captures multiplicity fractions

histories by number by weight by number by weight error

captures = 0 700 0 0.00000E+00 7.00000E-02 3.25400E-02 0.0364

captures = 1 2285 2285 1.06220E-01 2.28500E-01 1.06220E-01 0.0184

captures = 2 3223 6446 2.99647E-01 3.22300E-01 1.49823E-01 0.0145

captures = 3 2489 7467 3.47109E-01 2.48900E-01 1.15703E-01 0.0174

captures = 4 1022 4088 1.90033E-01 1.02200E-01 4.75084E-02 0.0296

captures = 5 209 1045 4.85775E-02 2.09000E-02 9.71551E-03 0.0684

captures = 6 51 306 1.42246E-02 5.10000E-03 2.37077E-03 0.1397

captures = 7 12 84 3.90480E-03 1.20000E-03 5.57828E-04 0.2885

captures \> 7 9 73 3.39345E-03 9.00000E-04 4.18371E-04 0.3332

total 10000 21794 1.01311E+00 1.00000E+00 4.64857E-01 0.0056

factorial moments by number by weight

3he 2.17940E+00 0.0056 1.01311E+00 0.0056

3he(3he-1)/2! 2.01890E+00 0.0128 9.38499E-01 0.0128

3he(3he-1)(3he-2)/3! 1.06390E+00 0.0291 4.94561E-01 0.0291

3he(3he-1) \.... (3he-3)/4! 3.93800E-01 0.0744 1.83061E-01 0.0744

3he(3he-1) \.... (3he-4)/5! 1.34100E-01 0.1636 6.23373E-02 0.1636

3he(3he-1) \.... (3he-5)/6! 4.43000E-02 0.2666 2.05932E-02 0.2666

3he(3he-1) \.... (3he-6)/7! 1.12000E-02 0.3808 5.20640E-03 0.3808

3he(3he-1) \.... (3he-7)/8! 1.70000E-03 0.5548 7.90257E-04 0.5548

The capture tally input for this problem was

F8:n 999 input F8 card

FT8 CAP -8 -8 2003 input FT8 CAP card

Note that the line captures \> 7 9 73 indicate that nine histories had
eight or more neutrons captured. This implies that 8 histories had 8 8 =
64 neutrons captured and 1 history had 1 9 neutrons captured, for a
total of 73 neutrons captured. The table of captures evidently was too
short, and the problem should have been run with FT8 CAP -9 -9 or even
more captures and moments. Not specifying enough capture rows affects
only the captures \> 7 lines and the error estimate on the totals
capture line; all other information is correct as if more captures and
moments were listed.

As an interpretation of the neutron captures on the 3he table, Column 1
is the number of histories according to the number of captures by the
designated material (2003 = 3he) in the designated cell (999). This
number sums to the total number of source histories for the problem, nps
= 10000.

Column 2 is the number of captures by 2003 in cell 999 = 21794. Because
analog capture is the default for F8 tallies, the total weight captured
is also 21794.

Column 3 is the total weight captured divided by the tally
normalization. For SDEF PAR = -SF, the tally normalization is the number
of source histories = number of spontaneous fissions = 10000. For SDEF
PAR = -SF, column 3 would be 21794.0 / 10000 = 2.17940. In this problem,
SDEF PAR = SF, and the tally normalization is the source particles =
spontaneous fission neutrons = 21512. Thus, captures by weight are
21794.0 / 21512 = 1.01311.

Column 4 is the multiplicity fraction by number, which is Column 1
divided by the number of source histories. The total is always 1.00000.

Column 5 is the multiplicity fraction by weight, which is the weight of
histories undergoing capture divided by the tally normalization. For
SDEF PAR = -SF, this fraction would be 10000.0 / 10000 = 1.00000. In
this problem, SDEF PAR = SF and the multiplicity fraction by weight is
10000.0 / 21512 = 0.464857.

The interpretation of the factorial moments table now follows.

The first moment by number is the number of captures divided by the
number of source histories = 21794 / 10000 = 2.17940.

The first moment by weight is the total weight of capture divided by the
tally normalization. For SDEF PAR = -SF, this moment would be 21794.0 /
10000 = 21794.0. In this problem, SDEF PAR = SF and the first moment by
weight is 21794.0 / 21512 = 1.01311.

The second moment is N (N-1)/2, where N is the number of captures. In
this problem,

N N (N-1)/2 histories product

1 0 2285 = 0

2 1 3223 = 3223

3 3 2489 = 7467

4 6 1022 = 6132

5 10 209 = 2090

6 15 51 = 765

7 21 12 = 252

8 28 8 = 224

9 36 1 = 36

> sum = 20189 ,

and the second moment by number is divided by the number of histories,

> 20189 / 10000 = 2.01890 .

Because of analog capture, the second moment weight is 20189.0. The
second moment by weight is divided by the tally normalization. For SDEF
PAR = -SF, this moment would be 20189.0 / 10000 = 2.01890, which is the
same as the second moment by number. In this problem, SDEF PAR = SF, and
the second moment by weight is

20189.0 / 21512 = 0.938499 .

The seventh moment is

7 6 5 4 3 2 1/7! = 1 12 = 12

8 7 6 5 4 3 2/7! = 8 8 = 64

9 8 7 6 5 4 3/7! = 36 1 = 36

sum = 112 ;

thus, 112 / 10000 = 0.0112 .

The eighth moment is

8 7 6 5 4 3 2 1/8! = 1 8 = 8

9 8 7 6 5 4 3 2/8! = 9 1 = 9

sum = 17 ;

thus, 17 / 10000 = 0.0017 .

And the ninth moment is

9 8 7 6 5 4 3 2 1/9! = 1 1 = 1 ;

thus, 1 / 10000 = 0.0001 .

### 4.8.3. FT8 Capture Tallies with Time Gating {#ft8-capture-tallies-with-time-gating .unnumbered}

The coincidence capture tally optionally allows specification of
predelay and gate width[^14] with the "gate" keyword on the FT8 card.
The "gate" keyword may appear anywhere after the "cap" keyword and is
part of the "cap" command. Immediately following, the "gate" keyword
must be the predelay time and the total gate width, both in units of
shakes (1.0e-8 s).

The addition of the predelay and time gate width changes the capture
tally scoring. When a neutron is captured at time *t0* in the specified
cell by the specified nuclide (22 and ^3^He = 2003 in all three tallies
below), the gate is "turned on." If the predelay is *t1* and the gate
width is *t2*, then all captures between *t0 + t1* and *t0 + t1 + t2*
are counted. For a history with no captures, no events are scored. With
one capture, 0 events are scored. With two captures, the first turns on
the time gate are at time *t0* and scores 0; the second will score one
event if it is captured between\
*t0 + t1* and *t0 + t1 + t2* or score another 0 if outside the gate.

Other entries after the "cap" keyword may be placed in any order, as
shown in the following examples. The negative entries change the allowed
number of captures and moments (defaults 21 and 12 are changed to 40 and
40 in F78 below). The list of capture nuclides (^3^He = 2003 in all
three tallies below) also may be placed anywhere after "cap."

Examples for three capture tallies now follow. The capture tally without
gating (f18) is shown for reference. An infinite gate (F38) results in a
very different Print Table 118: the number of captures is the same, but
the moments are offset by one. A finite gate (F78) has fewer captures,
as expected.

*Example 1: Capture Tally without Gate*

*Input:*

f18:n 22

ft18 cap 2003

*Output:*

1 neutron captures, moments and multiplicity distributions. tally 18
print table 118

weight normalization by source histories = 20000

cell: 22

neutron captures on 3he

captures captures multiplicity fractions

histories by number by weight by number by weight error

captures = 0 13448 0 0.00000E+00 6.72400E-01 6.72400E-01 0.0049

captures = 1 5550 5550 2.77500E-01 2.77500E-01 2.77500E-01 0.0114

captures = 2 588 1176 5.88000E-02 2.94000E-02 2.94000E-02 0.0406

captures = 3 238 714 3.57000E-02 1.19000E-02 1.19000E-02 0.0644

captures = 4 94 376 1.88000E-02 4.70000E-03 4.70000E-03 0.1029

captures = 5 40 200 1.00000E-02 2.00000E-03 2.00000E-03 0.1580

captures = 6 26 156 7.80000E-03 1.30000E-03 1.30000E-03 0.1960

captures = 7 8 56 2.80000E-03 4.00000E-04 4.00000E-04 0.3535

captures = 8 5 40 2.00000E-03 2.50000E-04 2.50000E-04 0.4472

captures = 9 1 9 4.50000E-04 5.00000E-05 5.00000E-05 1.0000

captures = 12 1 12 6.00000E-04 5.00000E-05 5.00000E-05 1.0000

captures = 16 1 16 8.00000E-04 5.00000E-05 5.00000E-05 1.0000

total 20000 8305 4.15250E-01 1.00000E+00 1.00000E+00 0.0128

factorial moments by number by weight

3he 4.15250E-01 0.0128 4.15250E-01 0.0128

3he(3he-1)/2! 1.59300E-01 0.0651 1.59300E-01 0.0651

3he(3he-1)(3he-2)/3! 1.47900E-01 0.2165 1.47900E-01 0.2165

3he(3he-1) \.... (3he-3)/4! 1.87750E-01 0.5063 1.87750E-01 0.5063

3he(3he-1) \.... (3he-4)/5! 2.96500E-01 0.7493 2.96500E-01 0.7493

3he(3he-1) \.... (3he-5)/6! 4.61900E-01 0.8727 4.61900E-01 0.8727

3he(3he-1) \.... (3he-6)/7! 6.15800E-01 0.9311 6.15800E-01 0.9311

3he(3he-1) \.... (3he-7)/8! 6.68950E-01 0.9626 6.68950E-01 0.9626

3he(3he-1) \.... (3he-8)/9! 5.83050E-01 0.9812 5.83050E-01 0.9812

3he(3he-1) \.... (3he-9)/10! 4.03700E-01 0.9918 4.03700E-01 0.9918

3he(3he-1) \.... (3he-10)/11! 2.19000E-01 0.9972 2.19000E-01 0.9972

3he(3he-1) \.... (3he-11)/12! 9.10500E-02 0.9994 9.10500E-02 0.9994

*Example 2: Infinite Gate*

*Input:*

f38:n 22

ft38 cap 2003 gate 0 1e11

*Output:*

1 neutron captures, moments and multiplicity distributions. tally 38
print table 118

weight normalization by source histories = 20000

cell: 22

neutron captures on 3he

time gate: predelay = 0.0000E+00 gate width = 1.0000E+11

pulses occurrences occurrences pulse fraction

in gate histogram by number by weight by number by weight error

captures = 0 6552 0 0.00000E+00 3.27600E-01 3.27600E-01 0.0101

captures = 1 1002 1002 5.01000E-02 5.01000E-02 5.01000E-02 0.0308

captures = 2 414 828 4.14000E-02 2.07000E-02 2.07000E-02 0.0486

captures = 3 176 528 2.64000E-02 8.80000E-03 8.80000E-03 0.0750

captures = 4 82 328 1.64000E-02 4.10000E-03 4.10000E-03 0.1102

captures = 5 42 210 1.05000E-02 2.10000E-03 2.10000E-03 0.1541

captures = 6 16 96 4.80000E-03 8.00000E-04 8.00000E-04 0.2499

captures = 7 8 56 2.80000E-03 4.00000E-04 4.00000E-04 0.3535

captures = 8 3 24 1.20000E-03 1.50000E-04 1.50000E-04 0.5773

captures = 9 2 18 9.00000E-04 1.00000E-04 1.00000E-04 0.7071

captures = 10 2 20 1.00000E-03 1.00000E-04 1.00000E-04 0.7071

captures = 11 2 22 1.10000E-03 1.00000E-04 1.00000E-04 0.7071

captures = 12 1 12 6.00000E-04 5.00000E-05 5.00000E-05 1.0000

captures = 13 1 13 6.50000E-04 5.00000E-05 5.00000E-05 1.0000

captures = 14 1 14 7.00000E-04 5.00000E-05 5.00000E-05 1.0000

captures = 15 1 15 7.50000E-04 5.00000E-05 5.00000E-05 1.0000

total 8305 3186 1.59300E-01 4.15250E-01 4.15250E-01 0.0291

factorial moments by number by weight

n 1.59300E-01 0.0651 1.59300E-01 0.0648

n(n-1)/2! 1.47900E-01 0.2165 1.47900E-01 0.2165

n(n-1)(n-2)/3! 1.87750E-01 0.5063 1.87750E-01 0.5062

n(n-1)(n-2) \... (n-3)/4! 2.96500E-01 0.7493 2.96500E-01 0.7492

n(n-1)(n-2) \... (n-4)/5! 4.61900E-01 0.8727 4.61900E-01 0.8726

n(n-1)(n-2) \... (n-5)/6! 6.15800E-01 0.9311 6.15800E-01 0.9311

n(n-1)(n-2) \... (n-6)/7! 6.68950E-01 0.9626 6.68950E-01 0.9626

n(n-1)(n-2) \... (n-7)/8! 5.83050E-01 0.9812 5.83050E-01 0.9812

n(n-1)(n-2) \... (n-8)/9! 4.03700E-01 0.9918 4.03700E-01 0.9918

n(n-1)(n-2) \... (n-9)/10! 2.19000E-01 0.9972 2.19000E-01 0.9972

n(n-1)(n-2) \... (n-10)/11! 9.10500E-02 0.9994 9.10500E-02 0.9994

n(n-1)(n-2) \... (n-11)/12! 2.80000E-02 1.0000 2.80000E-02 1.0000

*Example 3: Finite Gate*

*Input:*

f78:n 22

ft78 cap gate .5 .4 -40 -40 2003

*Output:*

1 neutron captures, moments and multiplicity distributions. tally 78
print table 118

weight normalization by source histories = 20000

cell: 22

neutron captures on 3he

time gate: predelay = 5.0000E-01 gate width = 4.0000E-01

pulses occurrences occurrences pulse fraction

in gate histogram by number by weight by number by weight error

captures = 0 7837 0 0.00000E+00 3.91850E-01 3.91850E-01 0.0118

captures = 1 394 394 1.97000E-02 1.97000E-02 1.97000E-02 0.0666

captures = 2 67 134 6.70000E-03 3.35000E-03 3.35000E-03 0.1542

captures = 3 6 18 9.00000E-04 3.00000E-04 3.00000E-04 0.4082

captures = 4 1 4 2.00000E-04 5.00000E-05 5.00000E-05 1.0000

total 8305 550 2.75000E-02 4.15250E-01 4.15250E-01 0.0624

factorial moments by number by weight

n 2.75000E-02 0.0717 2.75000E-02 0.0716

n(n-1)/2! 4.55000E-03 0.1654 4.55000E-03 0.1654

n(n-1)(n-2)/3! 5.00000E-04 0.4690 5.00000E-04 0.4690

n(n-1)(n-2) \... (n-3)/4! 5.00000E-05 1.0000 5.00000E-05 1.0000

Scratch space is needed to save capture times during the course of a
history. The times are stored temporarily in the capture and moment bins
of the tally. If sufficient bins are unavailable, then the number of
allowed captures and moments must be increased using the negative
entries after the "cap" keyword. The message " \*\*\* warning \*\*\*
dimension overflow. Some pulses not counted." is put in Print Table 118
if the space needs to be increased.

### 4.8.4. Possible Future Extensions of FT8 CAP Option {#possible-future-extensions-of-ft8-cap-option .unnumbered}

The coincidence counter capture tally model lacks many standard tally
features to make the algorithm faster. Adding the following tally
capabilities would be a straightforward extension:

-   extension to particle types other than neutrons,

-   tally cell and surface flagging,

-   extension to other special tally treatment options on the FT card,

-   tally segmenting,

-   tally multipliers, and

-   energy bins.

## 4.9. Residual Nuclei Tally: FT8 RES {#residual-nuclei-tally-ft8-res .unnumbered}

Residual nuclei from nuclear interactions in the model physics energy
range may be tallied with the FT tally special treatments RES option in
conjunction with PHTs. The form of the FT card is

FT8 **RES** ,

or

FT8 **RES** Z~1~ Z~2~ .

The interaction of high-energy particles with target nuclei causes the
production of many residual nuclei. The generated residual nuclei can be
recorded to an F8 tally if used with an FT8 RES special-treatment
option. The residuals are recorded at each interaction in the model
physics; residual nuclei are not tabulated at collisions using table
physics.

By default, the FT8 RES card will cause the corresponding F8 tally to
have 2000+ user bins for each possible residual nucleus ZAID. The range
of ZAID bins may be reduced by specifying Z~1~ and Z~2~, which
correspond to a range of possible Z values. If Z~1~ and Z~2~ are
specified and a residual is generated with a higher or lower Z, the
residual will not be scored in the F8 tally.

The FT8 RES capability is particularly useful with the eighth LCA card
entry, NOACT. When NOACT = 2 on the LCA card, the source particle
immediately collides in the source material. All subsequent daughter
particles then are transported without further collision, as if in a
vacuum. The F8 tally with an FT8 RES special tally treatment is then
simply the distribution of nuclides resulting from a single collision.

The following input files model a 1.2-GeV proton source having a single
collision with ^208^Pb.

Test of p(1.2GeV)+Pb(208)

1 1 -11. -1 imp:h 1

2 0 1 imp:h 0

1 so .01

mode h n

sdef par h erg=1200 vec 0 0 1 dir 1

m1 82208 1

phys:h 1300 j 0

phys:n 1300 3j 0

nps 10000

f8:h 1

ft8 res 1 99

fq8 u e

lca 2 1 1 23 1 1 0 -2 0

These data are plotted in Fig. 5, with MCNPX using the tally plotter and
the execute line command

MCNPX Z ,

![](media/image5.png){width="6.491666666666666in"
height="4.425694444444445in"}

Fig. 5. Residuals for ~81~T1 isotopes 189 to 201 from 1.3-GeV protons on
$_{\text{82}}^{\text{208}}$Pb.

where the command file, COM91, is

> rmctal=mctl91
>
> tally 8 free u xlim 81189 8120 ylim .0001 .01 .

# 5.0. Other input file extensions {#other-input-file-extensions .unnumbered}

-   Logarithmic interpolation on input cards,

-   Auxiliary input files (READ),

-   HISTP file size control (HISTP),

-   DXTRAN improvements (DBCN), and

-   PTRAC file for all particles and for coincidence counting (PTRAC).

## 5.1. Logarithmic Interpolation {#logarithmic-interpolation .unnumbered}

Logarithmic interpolation is allowed on all input cards where lists of
numbers are given. It is similar to the IJMR interpolation. For example,

E0 1.e-3 6log 1.e4

is interpreted as

E0 .001 .01 .1 1 10 100 1000 10000 .

## 5.2. Auxiliary Input File and Encryption {#auxiliary-input-file-and-encryption .unnumbered}

The MCNPX READ card enables

-   the reading of parts of the input file from other (auxiliary) files,

-   the suppression of the printing of the auxiliary input files to
    shorten output files and protect proprietary information, and

-   the encryption of auxiliary input files to protect proprietary
    information.

The READ card may appear anywhere after the title card of an MCNPX input
file but not in the middle of a card continuation.

### 5.2.1. Reading from an Auxiliary File {#reading-from-an-auxiliary-file .unnumbered}

READ FILE = *filename*

will cause input from the file "*filename*" to be inserted after the
READ command in the MCNPX input deck. Unlike most MCNPX input cards,
there may be as many READ cards and auxiliary input files as desired.

### 5.2.2. Suppressing Input Printing {#suppressing-input-printing .unnumbered}

READ NOECHO

will suppress printing of the input cards following the READ card. The
echoing of input cards is resumed with

READ ECHO .

Because the default of the READ card is ECHO, the echoing also will
resume when the next READ card is encountered without the NOECHO
command.

The echoing of the input cards also is resumed when an "end of file" is
encountered.

READ FILE = *filename* NOECHO

causes the input from the auxiliary file, *filename,* to be suppressed.
After the file *filename* is read, input transfers back to the input
file with the READ card and printing is no longer suppressed.

### 5.2.3. Encryption of Input {#encryption-of-input .unnumbered}

A simple encryption scheme is available in MCNPX. The encryption
capability can be used to protect proprietary designs of tools and other
systems modeled with MCNPX. To read an encrypted file,

READ DECODE *password* FILE = *filename .*

The encrypted input file will not be echoed, and many default print
tables are turned off (and cannot be turned back on) to protect the data
in the encrypted file.

To write an encrypted file,

READ ENCODE *password* FILE = *filename .*

The encryption capability is localized in subroutine ENCRYPT. The MCNPX
scheme is very simple; therefore, it protects nothing. To protect input,
the subroutine should be modified to a more sophisticated scheme known
only to those producing the data and only executable MCNPX versions
should be provided to users of the encrypted files.

## 5.3. HISTP File Size Control {#histp-file-size-control .unnumbered}

The HISTP card controls the writing of information to an external file
for analysis by the HTAPE3X program. HISTP can take two different
argument types:

HISTP --lhist icl~1~ icl~2~ icl~3~ .

lhist is the parameter that controls the number of words written to a
*histp* file. Once this limit is exceeded, a new file will be written
with the name *histpa* and the incrementing of the name will continue
until all particles are run. lhist must be entered as a negative number
and may appear anywhere on the card. The default value of lhist is
500,000,000 words.

icl~1~, icl~2~, ... are cell numbers. If no icl value is present, all
events will be written to *histp*. The user may enter as many icl values
as needed, and only events within these cells will be written to
*histp*.

*Example:*

HISTP --100000 5 6 3 10 .

Each *histp* file will contain a maximum of 100,000 words. Only events
within cells 3, 5, 6, and 10 will be written to the *histp* file.

HISTP

By default, each *histp* file will contain a maximum of 500,000,000
words (which virtually ensures that only one file will be written). All
events in all cells will be written to the file.

Note: writing *histp* files during multiprocessing is still under
development.

## 5.4. DXTRAN Improvements {#dxtran-improvements .unnumbered}

### 5.4.1. Detectors with Model Physics (Isotropic Approximation) {#detectors-with-model-physics-isotropic-approximation .unnumbered}

All collisions producing neutrons and photons now contribute to DXTRAN
and point detectors, including model physics interactions. When the
secondary neutron/photon angular scattering distribution function is
unknown (generally the case for model physics), isotropic scattering is
assumed (poor approximation). Although the extension to higher energies
often is approximate, energy bins for the point detector tally can
identify what portion of the tally is coming from high energies.
Further, this approximation is superior to neglecting charged-particle
and high-energy neutron collisions altogether.

### 5.4.2. Detector/DXTRAN Underflow Control (DBCN) {#detectordxtran-underflow-control-dbcn .unnumbered}

DXTRAN and point detector contributions are based on the next-event
estimator that makes contributions of

W p(μ) exp(-λ) / 2πR^2^ ,

where

> W = particle weight,
>
> p(μ) = the density function for scattering to the detector or DXTRAN
> sphere,
>
> R = the distance from collision to the detector or DXTRAN sphere, and
>
> λ = the attenuation factor, namely the sum of total macroscopic cross
> sections times the track length for each material region crossed
> between the collision and detector or DXTRAN sphere.

In MCNP4C, if λ \> 80, then exp(λ) = 0 and the score is terminated as
"underflow in transmission." These small contributions are
truncated---that is, they are neglected---which biases answers by
omitting them. Generally, they are insignificant to the final answer,
but in some cases, the underflow contribution is significant and needed.
When DXTRAN spheres or point detectors are used to get tally
contributions for generating weight windows, sometimes these underflow
contributions cannot be neglected.

It is possible to specify the underflow limit with the sixth entry on
the DBCN card. The default is 80, but values between 50 and 200 are
accepted.

If DXTRAN/detector underflow is significant in the calculation, serious
problems generally exist, such as not sampling enough collisions near
the detector. The underflow limit should be changed only with extreme
caution.

## 5.5. PTRAC for All Particles and for Coincidence Counting {#ptrac-for-all-particles-and-for-coincidence-counting .unnumbered}

### 5.5.1. PTRAC for All Particles (PTRAC) {#ptrac-for-all-particles-ptrac .unnumbered}

The PTRAC capability of writing particle tracks to a file for
post-processing has been extended to include all particle types and
provides appropriate information for all model collision/bank events.

### 5.5.2. PTRAC Capture File for Coincidence Counting (PTRAC) {#ptrac-capture-file-for-coincidence-counting-ptrac .unnumbered}

The FT8 CAP coincidence capture tallies may be written to a PTRAC file
for further analysis by auxiliary codes. A PTRAC file is created with
the following MCNPX input file (INP) input card:

PTRAC *EVENT = cap FILE = asc* .

Either an ascii (text) file (*FILE = asc*) or a binary file (*FILE =
bin*) may be created. The default is a binary file.

For *EVENT = cap,* most of the standard PTRAC capabilities are bypassed
(for speed) and the data written to each line (or binary file record)
are very different from the usual PTRAC data.

#### **5.5.2.1. History Information on PTRAC File** {#history-information-on-ptrac-file .unnumbered}

For binary files, the entries on each line are

NPS Time Cell .

For ascii files, the entries on each line are (format: i10,1p1e15.5,2i8)

NPS Time Cell Source ,

where

NPS = the particle history number,

Time = the time from source event to analog capture in any FT8 cap
tally,

Cell = the cell in which analog capture occurred, and

Source = the source particle number of a given history.

For example, after the usual PTRAC header, the capture events would be
recorded as

NPS Time Cell Source

1 4.5e01 22 4

1 0.0e00 0 3

1 2.6e02 22 2

1 3.5e02 22 2

1 1.5e00 23 2

1 0.0e00 0 1

3 0.0e00 0 2

3 0.0e00 0 1

In the previous example, source history 1 (NPS = 1) had four spontaneous
fission neutrons (4, 3, 2, 1 in the Source column). The fourth (Source
4---numbering is done as in the MCNPX bank: last in, first out) was
captured in Cell 22 at Time = 45 shakes after the source spontaneous
fission. The third (Source 3) was not captured (Time = 0.0, Cell = 0).
The second (Source 2) caused an induced fission---or possibly several
induced fissions---as flagged by the negative cell number. Captures in
Cell 22 were at 260 and 350 shakes for these neutrons from induced
fission (from the second neutron from a spontaneous fission). The second
spontaneous fission neutron was captured in cell 23; this neutron must
have been a branch \[after an (n,2n) split, perhaps\] that did not
undergo fission. The final (first, Source 1) spontaneous fission neutron
was not captured.

The second history (NPS = 2) is a spontaneous fission of multiplicity
zero; thus, no history 2 exists.

The third source particle history is a fission that results in two
fission neutrons (Source = 2, Source = 1), neither of which is captured.

For binary files (*FILE = bin*), only three words per record are
written. In the fourth column, "Source" is omitted.

#### 5.5.2.2. Changing the Name of the PTRAC File {#changing-the-name-of-the-ptrac-file .unnumbered}

The PTRAC file name may be changed on the MCNPX execution line,

mcnpx *ptrac = filename ,*

and the file name will be altered if an old file by that name already
exists. For example, if file *ptraca* was the last PTRAC file, then the
created PTRAC file would be *ptracb.*

#### 5.5.2.3. Possible Future Extensions of the PTRAC Capture Option {#possible-future-extensions-of-the-ptrac-capture-option .unnumbered}

Many standard PTRAC capabilities are missing from the PTRAC capture
tally output file capability. If desired, these capabilities could be
added and could include nearly all of the remaining PTRAC options:
BUFFER, MAX, MEPH, WRITE, other EVENTs, FILTER, particle type, NPS,
cell, surface, tally, and VALUE.

# 6.0. graphics {#graphics .unnumbered}

MCNPX has the following geometry, tally, and cross-section plotting
extensions:

-   mesh tallies (tally grid superimposed over geometry) plotted within
    MCNPX;

-   two-dimensional (2D) color contour plots for lattice and radiography
    tallies;

-   enhanced color geometry plots;

-   i,j,k lattice indexing in geometry plots;

-   geometry plot of WWG superimposed mesh;

-   proton and photonuclear cross-section plots; and

-   pause command for tally and cross-section plots.

## 6.1. Color Contour Plots for Mesh Tallies, Radiography Tallies, and Lattice Tallies {#color-contour-plots-for-mesh-tallies-radiography-tallies-and-lattice-tallies .unnumbered}

Tally output, including mesh tallies, radiography tallies, and lattice
tallies, may be plotted as 2D color contour plots. Mesh tallies also can
be plotted as superimposed over problem geometries. All of these plots
are done in MCNPX without the need of auxiliary post-processing codes.
The plots can be made either at the end of a calculation or while a
calculation proceeds using the MPLOT card.

The following combinations are available:

Post-processing, MCNPX Z option, RMCTAL = \<mctal filename\>

Mesh tallies, radiography tallies, lattice tallies, and all other
tallies

Post-processing, MCNPX Z option, RUNTPE = \<runtpe filename\>

Radiography tallies, lattice tallies, and all other tallies

In the PLOT mode mesh tally contours superimposed over problem
geometries

Run-time plotting: all of the above combinations are available during a
run with the MPLOT input file card.

### 6.1.1. User Interface for Contour Tally Plots {#user-interface-for-contour-tally-plots .unnumbered}

Mesh tally, radiography, and lattice tally contour plots may use all of
the usual MCNPX tally plot (MCPLOT) commands. The FREE and CONTOUR
commands have the following extensions.

#### **6.1.1.1. Free Command** {#free-command .unnumbered}

The MCPLOT free command is

FREE X\[Y\] \[nxm\] \[all\] \[noall\] .

For mesh tallies,

FREE *ij* ,

FREE *jk* , and

FREE *ik*

are used, where *i,j,k* refer to the CORA, CORB, and CORC mesh tally
dimensions.

For radiography tallies,

FREE *SC*

is used to make a contour plot of the *s-* and *t-* radiography axes.

For lattice tallies*, i,j,k,* refer to the *i,j,k* lattice indices.

If the lattice is fully specified, e.g.,

> fill=-2:2 -2:2 0:0 2 2 2 2 2
>
> 2 2 1 1 2
>
> 2 1 1 1 2
>
> 2 1 1 2 2
>
> 2 2 2 2 2 ,

then the \[nxm\] entries are not needed. If the lattice is not fully
specified, e.g.,

fill=2 ,

then the *i* and *j* dimensions must be given with the \[nxm\] entries,
e.g.,

FREE *ik 20x10* FIX *j=15 .*

The FIX command has its usual meaning, namely that the *j=15* slice of
the full *i,j,k* lattice will be used.

The "all" and "noall" keywords specify if the minimum and maximum
contour range should be taken from all of the tally bins ("all") or just
from those of the FIXED command slice ("noall"---default).

One-dimensional mesh, radiography, and lattice tallies are also possible
simply by giving the free dimension on the FREE command:

FREE *i* FIXED *j=10* FIXED *k=12 .*

Again, for lattice tallies that are not specified fully, the \[mxn\]
dimensions must be provided. Mesh and radiography tallies are always
specified fully, so \[mxn\] is never required for them.

#### 6.1.1.2. Contour Command {#contour-command .unnumbered}

#### The MCPLOT contour command is {#the-mcplot-contour-command-is .unnumbered}

####  {#section .unnumbered}

CONTOUR \[cmin cmax cstep\] \[command\] \[command\] .

All entries are optional. The \[cmin cmax cstep\] entries are the
minimum, maximum, and step values for contours. The \[cmin cmax cstep\]
entries are numbers and must appear together. The other commands may
appear anywhere after the CONTOUR command and are

\% interpret step values as percentages. The default is 5 95 10 %;

pct interpret step values as percentages. The default is 5 95 10 %;

lin interpret step values as absolute values of contour levels;

log contour levels logarithmically spaced between cmin and cmax, with
cstep values in between;

all contours normalized to min and max values of entire tally;

noall contours normalized to min and max values of (FIXED command)
contour slice;

line draw lines around contours (default);

noline do not draw lines around contours;

color make color contour plot (default); and

nocolor no color; draw line contours only.

*Examples:*

CONTOUR 5 95 10 % line color .

Ten contour lines will be at 5%, 15%, ..., 95% of the maximum value.
Lines will be drawn around the colored contours, as shown in Fig. 6. The
default setting is

CONTOUR 1e-4 1e-2 12 log .

Twelve contour lines will be logarithmically spaced between tally values
of 1e-4 and 1e-5.

### 6.1.2. Mesh Tally Contour Plot Example {#mesh-tally-contour-plot-example .unnumbered}

Mesh tallies may be plotted either in the MCNPX tally plotter (MCPLOT)
from MCTAL files or superimposed over geometry plots in the geometry
plotter (PLOT) from RUNTPE files.

#### 6.1.2.1. MCPLOT Mesh Tally {#mcplot-mesh-tally .unnumbered}

Figure 6 shows a mesh tally of a critical configuration of seven
identical barrels of fissionable material. The mesh tally is generated
from an MCTAL file in the MCPLOT tally plotter.

The input file for this problem is

> cylinders containing critical fluid in macrobody hex lattice
>
> 1 1 -8.4 -1 u=1 imp:n=1
>
> 2 0 -2 u=1 imp:n=1
>
> 3 2 -2.7 -3 1 2 u=1 imp:n=1
>
> 4 3 -.001 3 u=1 imp:n=1
>
> 10 3 -.001 -6 lat=2 u=2 imp:n=1 fill=-2:2 -2:2 0:0
>
> 2 2 2 2 2
>
> 2 2 1 1 2
>
> 2 1 1 1 2
>
> 2 1 1 2 2
>
> 2 2 2 2 2
>
> 11 0 -8 imp:n=1 fill=2
>
> 50 0 8 imp:n=0
>
> 1 rcc 0 0 0 0 12 0 5
>
> 2 rcc 0 12 0 0 8 0 5
>
> 3 rcc 0 -1 0 0 22 0 6
>
> 6 rhp 0 -1 0 0 22 0 9 0 0
>
> 8 rcc 0 -1 0 0 22 0 30
>
> m1 1001 5.7058e-2 8016 3.2929e-2
>
> 92238 2.0909e-3 92235 1.0889e-4
>
> m2 13027 1
>
> m3 7014 .8 8016 .2
>
> c
>
> fc14 total keff in each element
>
> f4:n (1\<10\[-2:2 -2:2 0:0\]\<11)
>
> fq4 f m
>
> sd4 1 24r
>
> f14:n (1\<10\[-1 1 0\]) (1\<10\[0 1 0\])
>
> (1\<10\[-1 0 0\]) (1\<10\[0 0 0\]) (1\<10\[1 0 0\])
>
> (1\<10\[0 -1 0\]) (1\<10\[1 -1 0\]) t
>
> fq14 f m
>
> sd14 1 7r
>
> tf14 4
>
> fm14 (-1 1 -6 -7)

![](media/image6.png){width="5.979861111111111in"
height="4.866666666666666in"}

Fig. 6. Mesh tally of barrel geometry.

> print -160
>
> prdmp 2j 1
>
> kcode 1000 1 10 50
>
> ksrc 0 6 0 18 6 0 -18 6 0 9 6 15 -9 6 15 9 6 -15 -9 6 -15
>
> tmesh
>
> rmesh12
>
> cora12 -30. 99i 30.
>
> corb12 0. 12.
>
> corc12 -30. 99i 30.
>
> endmd

The plot command is

> mcplot\> tal 12 free ik .

The geometry is shown in Fig. 7.

***6.1.2.2. Superimposed Geometry Plot Mesh Tally***

Figure 8 shows the mesh plot superimposed over the geometry plot. The
MCNPX Z option is used, and the commands are

![](media/image7.png){width="6.491666666666666in"
height="4.866666666666666in"}

Fig. 7. Geometry of the seven-barrel problem.

> MCPLOT\> RUNTPE=\<runtpe filename\>
>
> MCPLOT\> PLOT
>
> PLOT\> py 4 ex 40 or 0 4 0 la 0 1 tal12 color on la 0 0 con 0 100 % .

After the PLOT command, the MCNPX interactive geometry plotter pops up.
If the PLOT\> button (bottom center) is clicked, then the above command
after the PLOT\> prompt can be entered. Alternatively, the mesh tally
superimposed on the geometry can be viewed by clicking buttons of the
interactive tally plot. These options are described in Section
6.1.2.3**.**

Each outer barrel is expected to have the same source and flux
distributions and that the center barrel should have a higher value.
However, the mesh tally shows that the barrels do not have the expected
distribution. The asymmetric distribution is a known Monte Carlo
deficiency and arises in all Monte Carlo codes, including MCNP4C, MCNP5,
and MCNPX: eigenvector fluxes generally are converged falsely in
eigenvalue problems of critical systems. It is evident that the mesh
tally is useful in assessing such deficiencies.

![](media/image8.png){width="6.495833333333334in"
height="4.872222222222222in"}

Fig. 8. Mesh plot superimposed on geometry plot.

During the course of a calculation, an MPLOT card in the INP file can be
used to plot the mesh tallies. The plot of Fig. 8 would be achieved with
the following MPLOT card:

> mplot freq 3000 plot ex 40 py 4 la 0 1 tal12.1 color on la 0 0
>
> cont 0 100 pct

The "freq 3000" command causes a plot to be made every 3000 histories.
The "plot" command then transfers plotting from the tally plotter to the
geometry plotter. The remaining commands are described in Section
6.1.2.3**.**

When a calculation is completed, the mesh tallies may be plotted as
superimposed over geometries using the MCNPX Z option. Only RUNTPE files
can be used; MCPLOT files cannot be used because they do not contain the
geometry information. The RUNTPE file may be specified either in the
execution line,

MCNPX Z RUN=\<runtpe filename\> ,

or the RUNTPE file may be specified in the usual way anytime when doing
tally plots:

MCNPX Z

mcplot\> run=\<runtpe filename\> .

Then to get the mesh tallies, the geometry plot mode must be requested:

mcplot\> PLOT .

At this time, the interactive geometry plotter screen will pop up and
commands may be entered either interactively or in command mode by
striking the PLOT\> button in the bottom center of the screen.

The mesh tally boundaries also may be plotted by cycling through the
CellLine options in the interactive geometry plot or by using the
command mesh = 6 or mesh = 7 plot commands (Section 6.1.2.3).

***6.1.2.3. Commands for Superimposed Geometry Plot Mesh Tally***

To plot a mesh tally superimposed over a geometry plot, the geometry
first must be specified in the usual manner. For the geometry of Fig. 2,
the plot commands are

PLOT\> or 0 4 0 py 4 ex 40 la 0 0 .

The same geometry can be viewed by using the Zoom, Origin, XZ, and LABEL
interactive buttons.

Next, the tally must be selected as the "Edit" quantity. In the command
mode, the only method is to select the tally as the label quantity,
transfer the label to the color quantity, and then turn the labels back
off:

PLOT\> la 0 1 tal12.1 color on la 0 0 .

Note that only mesh tallies will be recognized. If multiple mesh tally
bins exist, e.g.,

rmesh11:h flux popul ,

then the number after the decimal indicates which bin. In this case,
tal11.1 refers to the "flux" mesh tally and tal11.2 refers to the
"popul" mesh tally.

Interactive plot buttons can be used to achieve the same result. First,
the mesh tallies must be made the "Edit" quantity by clicking the last
of the buttons in the far-right column of the screen.

> Click tal (make mesh tallies the "Edit" quantity),
>
> Click N (cycle through available mesh tally numbers), and
>
> Click IP (cycle through mesh tally bins).

The "Edit" quantity, e.g., tal12.1, now has been specified. Next, change
the color parameter (default = "mat"), by clicking COLOR twice.

> Click COLOR (will change "mat" to "off"),
>
> Click COLOR (will change "off" to the "Edit" quantity, "tal12.1"), and
>
> Click Redraw (bottom center button---to make new picture).

Two other commands are also useful: mesh and contour.

The actual mesh tally grid can be displayed by clicking "CellLine" and
cycling through the options to get either "MeshTaly" (which draws mesh
tally grid lines over the plot) or "MT+Cell" (which draws mesh tally
grid lines and cell surface lines over the plot). In the command prompt
mode, this is done with

PLOT\> mesh = 6 (mesh tally grid lines)

or

PLOT\> mesh = 7 (mesh tally grid lines plus cell surface lines) .

The contour levels can be adjusted using the contour command. No
interactive button is available for this; thus, to get from interactive
to command mode, either the "Click here or picture or menu" (bottom left
of interactive screen) or "PLOT\>" (bottom center) must be struck. The
contour command is

CONTOUR cmin cmax \[cstep\] command .

The "cmin cmax cstep" entries are the minimum, maximum, and step values
for contours. Superimposed geometry mesh tally plots do not use steps:
the values are shaded by 64 colors. Thus, the cstep entry is ignored and
can be omitted; it is allowed only for consistency with the CONTOUR
command in the tally plotter (see Section 6.1.1.2). The "cmin cmax
cstep" entries are numbers and must appear together. Once a CONTOUR
command is entered, subsequent CONTOUR commands use the previous "cmin
cmax cstep" values; thus, only the "command" entry is required. The
"command" entry may appear before cmin or after cstep or by itself.
"CONTOUR" may be abbreviated simply to "CON" or "CONT". The allowable
entries for "command" are

> \% interpret step values as percentages. The default is 5 95 10 %;
>
> pct interpret step values as percentages. The default is 5 95 10 %;
>
> lin interpret step values as absolute values of contour levels;
>
> log contour levels logarithmically spaced between cmin and cmax; and
>
> off use default: 0 100 % .

The remaining command options for tally plots ("all", "noall", "line",
"noline", "color", and "nocolor" (see Section 6.1.1.2) make no sense for
geometry mesh tally plots and are disallowed. Some examples are

cont 0 100 pct

con 5 95 10 %

contour off

cont 1E-4 2 log

### 6.1.3. Radiography Tally Coutour Plot Example {#radiography-tally-coutour-plot-example .unnumbered}

The following example is a radiograph of a 4-cm-radius, 1-cm-thick
^238^U disc with an embedded 4-mm-void sphere and skew-oriented 1-cm
1-cm 8-mm box. The input file is

Radiography Tally

1 5 -25.0 -1 4 5 imp:p=1

2 0 1 -2 imp:p=1

3 0 2 imp:p=0

4 0 -4 imp:p=1

5 0 -5 imp:p=1

1 RCC 0 0 0 0 0 1 4

2 RPP -100 100 -100 100 -100 100

4 SPH 3 0 .5 .4

5 BOX -1 1 .1 .6 .8 0 -.8 .6 0 0 0 .8

mode p

nps 100 5

sdef pos=0 0 -20 axs=0 0 1 rad=d1 ext=0 vec=0 0 1 dir=d2 erg=6

si1 0 .1

sp1 -21 1

si2 -1 1

sp2 -31 1

m5 92238 1

print

prdmp 2j 1

tir5:p 0 0 10 0 0 0 -100 0 100 0

fs5 -10. 99i 10.

c5 -10. 99i 10.

The x-y geometry plot of this geometry is given in Fig. 9.

To get the contour plot, run

MCNPX Z RUNTPE=*filename* .

The contour plots also may be read from a MCTAL file instead of the
RUNTPE. When the code gives you the MCPLOT prompt, enter two dimensions
with the free command; for example, S and C:

> MCPLOT\> free SC .

![](media/image9.png){width="5.766666666666667in"
height="4.322916666666667in"}

Fig. 9. Geometry plot of radiograph example.

Recall that the tally dimensions are

F surface/cell/detector F card bin,

D total/direct or flagged bin,

U user bin,

S segment or radiography s-axis bin,

M multiplier bin,

C cosine or radiography t-axis bin,

E energy bin, and

T time bin.

The results are plotted in Fig. 10. The embedded sphere and box are seen
plainly in the disc.

![](media/image10.png){width="5.288888888888889in"
height="4.079861111111111in"}

Fig. 10. Scattered photon radiographic image of ^238^U disc.

### 6.1.4. Lattice Tally Plot Examples {#lattice-tally-plot-examples .unnumbered}

Lattice tallies also may be plotted as either 1D or 2D contour plots. An
example of a lattice tally 1D plot is

free i 64x64 fix j = 38 fix k = 30 .

This command specifies a 1D lattice tally plot of the cell bins, which
should correspond to a lattice tally with 64 "i" index bins, 64 "j"
index bins, and at least 30 "k" index bins. With the "k" index set to 30
and the "j" index to 38, the offset into the f bins will be 29 64 64 +
37 64 = 121,152. The minimum and maximum values will be determined from
the 64 "i" bin values included in the plot. If the "j" and "k" indices
are not specified, their default value of 1 is assumed, which results in
an offset of 0.

An example of a lattice tally 2D contour plot is

free ij 10x30 all fix k=60 .

This example specifies a 10 30 2D contour plot, which should correspond
to a lattice tally with 10 "i" bins, 30 "j" bins, and at least 60 "k"
bins. Note that the "k" index is specified using the "fix" command,
which sets the offset into the f bins as 60 10 30 = 18,000. In this
case, the contour range is taken from all of the f-bin tally values.

### 6.1.5. Other Notes and Future Work {#other-notes-and-future-work .unnumbered}

A consequence of the mesh tally contour plotting capability is that
MDATA files are ignored in continue runs. MDATA files and the auxiliary
GRIDCONV code still are supported for the plotting of mesh tallies with
other postprocessors, such as MORITZ and TECPLOT. A new MDATA file is
written at the end of each initial or continue run.

When using COMMAND files (which use a preselected sequence of MCPLOT
tally plot commands), the "RETURN" command must follow geometry plot
commands to revert from the geometry plotter back to the tally plotter.

The future work is as follows:

1.  Treat the mesh tallies as regular tallies. Among the many
    consequences are the following.

-   RMESH, CORA, CORB, CORC, etc., could be anywhere in the input deck
    (TMESH and ENDMD would be ignored), and all other tally features,
    such as FM, DE, DF, FC, CF, SF, FU, FT, E, T, EM, and TM, would be
    enabled and consistent with other tallies; and

-   tally plot mesh tallies (Section 6.1.2.1) could use RUNTPE files and
    not be limited to MCTAL files.

2.  Plot theta and phi values from cylindrical and spherical meshes
    correctly in tally plots and plot spherical mesh tally grid lines in
    geometry plots.

## 6.2. Enhanced Color Geometry Plots: 64-Color Plotting and Shading by Cell Parameters {#enhanced-color-geometry-plots-64-color-plotting-and-shading-by-cell-parameters .unnumbered}

MCNP4C3 had 10-color plotting; MCNPX has 64-color plotting. Shading of
geometry plots may be used for any cell parameter. MCNP4C colored its
geometry plots by material only, giving a different color to each
material number. MCNPX can color geometry plots by any cell quantity.
Each cell can have a different color, or each repeated structure level
or universe can have a different color. Logarithmic shading of
importances, weight windows, and summary information are automatic. If a
superimposed weight-window mesh is used, coloring also may be done by
the value of the mesh weight windows.

In the interactive capability, the "SCALES n" button has been moved up
two lines (after the cursor) to make room for a larger "COLOR name"
button. The default is "COLOR mat", which colors problem cells by the
program material number. This button must be clicked to get "COLOR off"
(black and white) and then clicked again to color by whatever parameter
is listed after the "Edit" button. For example, in the right margin,
click "cel", which will make the "Edit" quantity "cel". Next, click
"COLOR" so that "COLOR cel" is displayed; on the next plot, the color
shades will be determined by program cell number.

For command/prompt plotting, enter

PLOT\> label 0 1 rho ;

the color command then must be set such as

PLOT\> color on ,

and the coloring will now be by rho, the atom density.

## 6.3. i,j,k Lattice Index Labeling {#ijk-lattice-index-labeling .unnumbered}

The i,j,k lattice indices of repeated structures/lattice geometries may
be used as plot labels in geometry plots, as illustrated in Fig. 11.

If the level (LEVEL command or button) is not a lattice cell level, then
the indices will be those of the next lattice in a higher level. To get
the lattice index labels, choose "ijk" as the "Edit" quantity by
clicking "ijk" in the right margin. Then click the send entry after
LABEL so that it reads "LABEL off ijk". For command/prompt plotting,
enter "Label 0 1 ijk".

![](media/image11.png){width="5.997222222222222in"
height="4.495833333333334in"}

Fig. 11. Lattice indexing in geometry plot.

## 6.4. Weight-Window-Generator Superimposed Mesh Plots {#weight-window-generator-superimposed-mesh-plots .unnumbered}

MCNPX can plot the WWG superimposed mesh specified on the MESH card in
an input file. MCNP4C3 and previous MCNPX versions could plot only the
weight-window superimposed mesh used in a problem, and a WWINP file had
to be provided. In the MCNPX geometry plotter, toggle CellLine for the
following options:

No Lines Plot cells not outlined in black

CellLine Plot geometric cells, outlined in black

WW MESH Plot the weight-window superimposed mesh (WWINP required)

WW+Cell Plot superimposed mesh and cells, outlined in black

WWG MESH Plot MESH card WWG mesh

WWG+Cell Plot WWG mesh and cells, outlined in black

MeshTaly Plot mesh tally boundaries

MT+Cell Plot mesh tally boundaries + CellLine

The CellLine and No Lines options are always available. WW Mesh and
WW+Cell are available only when the WWP card calls for using a
superimposed weight-window mesh (fifth entry negative) and a WWINP file
is provided. WWG MESH and WWG+Cell are available only when a MESH card
is in the input file and when the WWG requests superimposed mesh
generation (WWG card second entry = 0). In all cases, the cells may be
outlined in black (CellLine, WW+Cell, WWG+Cell) or the cells simply may
be colored without outlining (WW MESH, WWG MESH, No Lines).

### 6.4.1. Geometry Plot of Weight-Window-Generator Superimposed Mesh {#geometry-plot-of-weight-window-generator-superimposed-mesh .unnumbered}

MCNPX can plot the WWG superimposed mesh specified on the MESH card in
an input file. The following options are available.

0\. No Lines Plot cells not outlined in black

1\. CellLine Plot geometric cells, outlined in black

2\. WW MESH Plot WW mesh (WWINP, WWP 4j 1 required)

3\. WW+Cell Plot WW mesh (WWINP, WWP 4j 1 required) + CellLine

4\. WWG MESH Plot WWG mesh (MESH, WWG J 0 required)

5\. WWG+Cell Plot WWG mesh (MESH, WWG J 0 required) + CellLine

6\. MeshTaly Plot mesh tally boundaries (RMESH, CORA, etc., required)

7\. MT+Cell Plot mesh tally boundaries + CellLine

In the interactive MCNPX geometry plotter, toggle CellLine for the above
options.

In the command-mode geometry plotter, the option is *mesh = n,* where
*n* is the number 0 to 7, as determined by the options list above.

The CellLine and No Lines options always are available. WW MESH and
WW+Cell are available only when the WWP card calls for using a
superimposed weight-window mesh (fifth entry negative) and when a WWINP
file is provided. WWG MESH and WWG+Cell are available only when a MESH
card is present in the input file and when the WWG requests superimposed
mesh generation (WWG card second entry = 0). In all cases, the cells may
be outlined in black (CellLine, WW+Cell, WWG+Cell) or the cells simply
may be colored without outlining (WW MESH, WWG MESH, No Lines).

### 6.4.2. Example {#example .unnumbered}

Input file: *inp10*

Demonstration of WWG Plot

1 1 1.0 -1 imp:p 1

2 0 1 imp:p 0

1 rcc 0 0 0 0 10 0 5

mode p

sdef sur 1.3 vec 0 1 0 dir 1 erg 100

m1 1001 2 8016 1

nps 1000

f1:p 1.2

wwg 1 0

mesh geom=cyl origin=0 -1 0 ref=0 .1 0 axs=0 1 0 vec=1 0 0

imesh 6 iints 7 jmesh 12 jints 7 kmesh 1 kints 3

Com file: *com10*

ex 10 lab 0 0 px 0 mesh 4

pause

py 5

pause

mcnpx i = *inp10* com = *com10* ip .

Or, instead of using the command file (with plot commands in command
mode), the interactive plotter can be used.

mcnpx i = *inp10* ip ,

click CellLine to get WWG + Cell

label sur to turn off surface labels

XY to get px = 0 view (axial view, Fig. 12)

Zoom 10 to get 10 magnification (click twice)

Origin click in the center of material to center picture

ZX to get py = 5 view (radial view, Fig. 13)

The above Com file or plot commands in the command mode generate the two
plots shown in Figs. 12 and 13.

![](media/image12.png){width="5.322916666666667in"
height="4.009722222222222in"}

## Fig.12. WWG mesh plot, axial view. {#fig.12.-wwg-mesh-plot-axial-view. .unnumbered}

## 6.5. Proton and Photonuclear Cross-Section Plots {#proton-and-photonuclear-cross-section-plots .unnumbered}

### 6.5.1. Photonuclear Cross-Section Plots {#photonuclear-cross-section-plots .unnumbered}

MCNPX can plot photonuclear data, in addition to the photoatomic data of
MCNP.

Photoatomic reaction numbers are all negative: -1 = incoherent, -2 =
coherent, -3 = photoelectric, -4 = pair production, -5 = total, and -6 =
heating. For the MCNPX photonuclear cross-section plotting, the reaction
numbers all are positive. The principal photonuclear cross sections are
as follows: 1 = total, 2 = nonelastic, 3 = elastic, 4 = heating, and \>4
= various reactions such as 18 = (,f). The photonuclear yields
(multiplicities) for various secondary particles are specified by adding
1000 times the secondary particle number to the reaction number. For
example, 31,001 is the total yield of deuterons (particle type d = 31),
34,001 is the total yield of alphas (particle type = 34), and 1018 is
the total number of neutrons (particle type n = 1) from fission. To find
out which reactions are available for a particular nuclide or material,
enter an invalid reaction number, such as mt = 99, and MCNPX will list
the available photonuclear reactions and the available yields, such as
1018; 31,018; and 34,018. Entering a bad nuclide, xs = 12345.67u, will
cause MCNPX to list the available nuclides.

Figure 14 illustrates a photonuclear cross-section plot of the total
photonuclear cross section, mt = 1, for material 11 and its
constituents, carbon and lead.

![](media/image13.png){width="5.331944444444445in"
height="4.195138888888889in"}

Fig. 13. WWG plot, radial view.

![](media/image14.png){width="5.502083333333333in"
height="4.0993055555555555in"}

Fig. 14. Photonuclear cross-section plot.

### 6.5.2. Proton Cross-Section Plots {#proton-cross-section-plots .unnumbered}

MCNPX can plot proton cross sections. The reaction numbers are similar
to the neutron reaction numbers: all positive. The principal proton
cross sections are the following: 1 = total, 2 = nonelastic, 3 =
elastic, 4 = heating, and \>4 = various reactions. On the LA150H proton
library, the only available reaction is mt = 5 with its multiplicities,
1005; 9005; 31,005; etc. The multiplicity reaction numbers for
interaction reaction mt = 5 are 1005 for neutrons, 9005 for protons,
31,005 for deuterons, etc. To find out which reactions are available for
a particular nuclide or material, enter an invalid reaction number, such
as mt = 99, and MCNPX will list the available proton reactions and the
available yields, such as 1005; 32,001; and 34,002. The proton
multiplicity, mt = 9001, 9004, 9005, etc., generally is available, along
with the total cross-section and heating number, mt = 1, mt = 4.
Entering a bad nuclide, xs = 12345.67h, will cause MCNPX to list the
available proton nuclides.

## 6.6. Pause Command {#pause-command .unnumbered}

The MCNPX geometry plot *PAUSE* command is extended to tally and
cross-section plots. When the word *PAUSE N* is put in a tally plotting
the COM input file, the picture will display for *N* seconds. If the
command *PAUSE* (without the *N*) is in the COM file, the display will
hold until a key is struck.

The reading of the COM file can be interrupted and control of plotting
can be returned to the terminal by entering "*quit*." After each picture
is displayed, the message to the command terminal is

*Strike any key to continue. Quit returns to interactive mode.*

# 7.0. Parallel processing {#parallel-processing .unnumbered}

-   Distributed memory multiprocessing for all particles and energies

-   MPI multiprocessing

-   Significant speedup of criticality problems run with MPI

##  {#section-1 .unnumbered}

## 7.1. Distributed Memory Multitasking for the Entire Energy Range of All Particles {#distributed-memory-multitasking-for-the-entire-energy-range-of-all-particles .unnumbered}

The entire MCNPX code may be run in parallel using message passing.
Table physics problems (i.e., standard MCNP4C3 problems) can be run with
threading and/or message passing. Fault tolerance and load balancing are
available, and multiprocessing can be done across a network of
heterogeneous platforms. Parallel Virtual Machine (PVM) or MPI software
can be used.

The parallel message passing capability applies to all particles for
their entire energy range and for all physics models.

## 7.2. MPI Multiprocessing {#mpi-multiprocessing .unnumbered}

The MPI parallel communication software is supported by MCNPX. To
compile MCNPX with MPI, it is necessary to use the "MPILB" configuration
option and provide a path to the MPI header file. This may not be
necessary on some systems if these header files are already in the
standard include path:

Configure --with-MPILIB\[="/path/to/MPI/libraries -lmpich"\]

> \--with --FFLAGS="-I/path/to/MPI/include/files" .

To run an MCNPX problem with MPI, simply start the MPI daemon (which
typically is running already on most systems) and then start MCNPX using
"MPIRUN". An example is

mpirun -np 4 mcnpx inp=gwm na=gwm1. \... .

MPI is quite different from the PVM, which required knowledge about
setting certain links, environment variables, and the PVM console
commands. An example of the PVM execution command is

mcnpx inp=gwm n=gwm1. tasks=-12 .

See "installation notes for building MCNPX with MPI" on the MCNPX
documents page at <http://mcnpx.lanl.gov/documents.html>.

## 7.3. MPI Speedup for Criticality Calculations {#mpi-speedup-for-criticality-calculations .unnumbered}

Note: The MPI speedup capability speeds up MPI KCODE calculations but
also has some collateral consequences for sequential KCODE calculations.

Criticality calculations (KCODE source) will run 10 to 1000 times faster
than MCNP4C3 in parallel with MPI multiprocessing. The speedup depends
on the number of CPUs and the number of histories per cycle. As the
number of CPUs increases above \~8 and the histories per cycle approach
\~1M, the speedup increases rapidly from a factor of \~10 to above 1000
with 64 CPUs.

The speedup has been achieved by having the next-generation fission
source points on each processor stay on that processor. No longer are
the fission source points from all processors combined at the end of
each cycle for rebroadcast in the next cycle. The considerable expense
of grouping the particles together at each processor rendezvous has been
eliminated.

### 7.3.1. DXTRAN and Detector Tracking Differences {#dxtran-and-detector-tracking-differences .unnumbered}

All calculations with DXTRAN and detectors---not just KCODE or parallel
calculations---generally no longer track. Answers are still correct, but
these next-event estimator problems do not track unless the twenty-first
entry on the DBCN card is set to nonzero or the default Russian roulette
detector game (DD card) is not played. The tracking difference occurs
because the default Russian roulette game no longer adjusts the roulette
criteria at the two-hundredth history. The Russian roulette criteria are
still set at all tally fluctuation chart intervals, NPD, the fifth entry
on the PRDMP card. The default for NPD is 1000 histories for
fixed-source problems but is NPD = 1 (at the end of each cycle) for
KCODE problems. Thus, a fixed-source problem that previously started
playing Russian roulette on next-event estimates at history 200 now
waits until history 1000.

### 7.3.2. User Interface Changes for Both Sequential and Parallel KCODE Problems {#user-interface-changes-for-both-sequential-and-parallel-kcode-problems .unnumbered}

The MPI speedup capability has consequences for both sequential and
parallel KCODE calculations.

-   Tally fluctuation charts are printed only at the ends of KCODE
    cycles rather than at specific particle intervals.

-   The units for the fifth entry of the PRDMP card are changed for
    KCODE calculations. The fifth entry is NPD, the tally fluctuation
    chart print interval. For KCODE calculations, NPD is the number of
    cycles (default = 1).

-   KCODE problems with DXTRAN and point detectors are still correct but
    may not track previous versions if the default Russian roulette game
    on small scores is played. The tracking difference occurs because
    the roulette criteria are updated in different places. DXTRAN and
    detectors using negative DD card entries (to have a constant Russian
    roulette game criteria) still track.

-   Setting the twenty-first entry on the DBCN card to 1 will cause
    MCNPX to track previous MCNPX versions without the MPI speedup
    capability:

    DBCN 20J 1 .

### 7.3.3. User Interface Changes for Parallel KCODE Problems {#user-interface-changes-for-parallel-kcode-problems .unnumbered}

-   The KCODE speedup is available only for negative JTASKS: *tasks = -n
    x m* on the MCNPX execution line. Microtasking/load-balancing
    (positive JTASKS) calculations use the slower former MCNP4C
    communication algorithm for KCODE source updating. A warning is
    issued if JTASKS \> 0 in a KCODE calculation.

-   In rare cases, a subtask may deplete its source particles. Only
    problem efficiency is affected, and a warning is issued.

### 7.3.4. Compiler and Operating System Problems for Parallel KCODE Problems {#compiler-and-operating-system-problems-for-parallel-kcode-problems .unnumbered}

-   CTRL-C and quit on a multiple processor MPI run in Linux do not
    finish writing the OUTP file before MCNPX exits. This failure
    appears to be an MPI error (possibly feature) in the MPI_FINALIZE
    call, where the last processor kills all subtasks and the master.

-   CTRL-C interrupt does not function properly in Windows 2000
    (probably general to all Windows). The user can attempt to quit, but
    MCNPX will hang and the user will be forced to kill the process
    through multiple CTRL-C commands.

-   The Portland Group pgcc compiler (Version 4.02) will not compile the
    ANL mpich-1.2.5 properly on a Linux system; however, GCC 3.2.3 will
    work to build the MPICH libraries.

# 8.0. Compiling, configuration, and installation {#compiling-configuration-and-installation .unnumbered}

-   F90 modularity and dynamic memory allocation

-   F90 autoconfiguration

-   64-bit-integer support

-   NAG/IBM/INTEL compiler extensions

## 8.1. F90 Modularity and Dynamic Memory Allocation {#f90-modularity-and-dynamic-memory-allocation .unnumbered}

The F90 conversion provides improvements in code modularity,
standardization of functions such as timing across platforms, and
compiler reliability. F90 will run more slowly on some systems.
Specifically, we have eliminated equivalences as a means of dynamic
storage allocation by using F90 pointers and allocable arrays. We have
replaced most common calls with F90 modules. The code will compile in
both free and fixed F90 formats.

MCNPX can be modified by patches, and as much of the MCNP4C coding as
possible has been preserved so that MCNP4C patches can be applied
directly to MCNPX.

Continuing improvements in the F90 structure are ongoing, especially
where they concern physics modules that have been brought into the code.

## 8.2. F90 Autoconfiguration {#f90-autoconfiguration .unnumbered}

MCNPX has a unique autoconfiguration build system that allows a variety
of compilation options to be executed easily on a large number of
platforms. MCNPX supports Linux, Windows PC, Sun Solaris UNIX, HPUX
UNIX, IBM AIX UNIX, and SGI UNIX.

Test problems and their solution templates are available for all
supported systems. The 65 test problems consist of the 42-problem MCNP5
test set supplemented by 23 problems testing MCNPX exclusive features.
The test problems on all systems can be run and compared with the
solution templates with a single, simple auto-configuration command.

## 8.3. 64-Bit-Integer Support {#bit-integer-support .unnumbered}

MCNPX has been restructured to enable 64-bit integers. MCNP and MCNPX
always have had 64-bit floating-point reals, either by use of compiler
directives on supercomputers or by double precision on "cheap" computers
(most systems). However, the integers have been 4 byte (32 bit) on all
but a few supercomputers; 4-byte integers limit the number of histories
that can be run to about 2 billion. They also limit the number of cross
sections and tallies in a problem because the cross-section data use
integer pointers.

With the 64-bit-integer capability, up to 1e18 histories can be run.
Memory access problems can be avoided, and nearly all the cross sections
and tallies needed can be used.

The print field for the number of histories run in a problem has been
increased to 12 digits to accommodate the long runs made possible with
parallel processing.

### 8.3.1. Recompilation Requirements for 64-Bit Integers {#recompilation-requirements-for-64-bit-integers .unnumbered}

The default for MCNPX is still 4-byte integers. The executable versions
provided on the MCNPX Beta Test page are all 4-byte integer versions. To
use the 64-bit version, it is necessary to

1.  recompile with special options and

2.  either use Type-1 ascii cross-section tables or rebuild the Type-2
    binary cross-section tables with MAKXS using 64-bit integer
    pointers.

Users who generally recompile their own MCNPX versions and use type-1
ascii cross sections can change to the 64-bit integers rather easily.
When configuring, simply use the options

\--with-FFLAGS=-i8 \--with-NOCHEAP .

The --with-NOCHEAP directive is not available on all systems. If it is
not recognized, then add the compile directive

\--with-FFLAGS=-i8

and remove --DCHEAP=1 from src/mcnpx/mcnpf/Makefile.h.

The 64-bit integer capability is unavailable on PCs. The --i8 compile
option is not available on the CVF F90 compiler and possibly other
compilers. The 64-bit integers will increase the size of the code.

### 8.3.2. Additional Notes {#additional-notes .unnumbered}

The 64-bit-integer capability was an extensive change to MCNPX because
it affected most F90 "kind" specifications. We also eliminated all
"REAL" declarations where floating point variables went to different
default lengths (usually 32 bit) on various systems. A consequence is
that real variables are all 64 bit unless some other length is required
for interfaces to plotting or interrupts or other special purposes. The
tracking between different compilers and platforms is now more
consistent. Finally, this capability made extension to new compilers and
platforms easier (see Section 8.4).

## 8.4. NAG/IBM/INTEL F90 and Compiler Extensions {#nagibmintel-f90-and-compiler-extensions .unnumbered}

MCNPX runs on a wide variety of platforms and operating systems by using
more standard F90 constructs and some fairly extensive code and
autoconfiguration changes. In particular, it runs on the G5 OS X system,
using the NAG 4.2 compiler for the Apple Power Mac G5. MCNPX also runs
on this platform with the IBM compiler.

The NAG compiler passes the MCNPX SGI-UNIX test suite on the Mac with
"-O1" optimization. The IBM compiler passes with "-O2" optimization and
is 17% faster. A trial version of the IBM compiler is available from
<http://www.ibm.com/>; see "support & downloads \-\-- trials & betas A-Z
\-\-- XL Fortran." No MCNPX parallel capability is available yet for the
NAG compiler.

MCNPX can use the Intel f90 and C compilers. Thus, on Linux, both the
Intel and Portland Group compilers can be used. On Windows, both the
Intel and Compaq Digital Fortran (CDF) or earlier Digital Visual Fortran
(DVF) f90 compilers can be used. The Windows Intel executable runs the
MCNPX test suite an average of 24% faster than the CVF executable.
Therefore, the Intel compiler is the default for compilation with
Windows PC.

The four configure options on Windows are

Configure (Build with the Intel compiler)

Configure CVF (Build with the CVF compiler)

Configure I8 (Build with 8-byte integers)

Configure MPI (Build the MPI version)

# 9.0. Summary of MCNPX Capabilities Since MCNP4B (by version) {#summary-of-mcnpx-capabilities-since-mcnp4b-by-version .unnumbered}

The extensions of MCNPX beyond MCNP4B are listed according to version
release; the initials of the principal developers are shown in
parentheses.[^15]

*MCNPX 2.5.f (February 2005)*

-   Mesh tally plots without auxiliary codes Mesh tallies (tally grid
    superimposed over geometry) plotted within MCNPX (JSH/GWM);

-   PHT with Variance Reduction (JSH/GWM);

-   Windows Intel Compiler Support (GWM);

-   Improved Photonuclear Model Data (FXG);

```{=html}
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```
-   FT8 Capture Tallies with Time Gating (JSH)

-   Improved or User Specified Fission Multiplicity Constants---FMULT
    (JSH)

-   Translated Sources with Dependence (GWM)

*MCNPX 2.5.e (February 2004)*

-   2D color tally contour plots, including lattices and radiography
    (GWM);

-   significant speedup of criticality problems run with MPI (GWM/NAC);

-   64-bit-integer support (GWM);

-   NAG/IBM/INTEL compiler extensions (GWM); and

-   photon Doppler broadening (from MCNP5) (AS).

*MCNPX 2.5.d (August 2003)*

-   INCL4/ABLA physics models (JCD/JSH);

-   lattice tally speedup (GWM);

-   auxiliary input files: READ (JSH);

-   geometry plot of WWG superimposed mesh (JSH);

-   pulse-height light tally with anticoincidence: FT8 PHL (GWM);

-   coincidence capture tally and PTRAC file: FT8 CAP (MTS/SJT/DRM/JSH);

-   residual nuclei tally: FT8 RES (JSH); and

-   inline generation of double differential cross sections and
    residuals (JSH).

*MCNPX 2.5.c (April 2003)*

-   message passing interface (MPI) multiprocessing (JL/GWM);

-   i,j,k lattice indexing in geometry plots (JSH);

-   enable of WWG in physics model region (FXG/ JSH);

-   enablement of exponential transform in physics model region
    (FXG/JSH);

-   extension of neutron model physics below 20 MeV (JSH);

-   ^3^He coincidence detector modeling (HGH/JSH); and

-   F90 autoconfiguration (TLR).

*MCNPX 2.5.b (November 2002)*

-   CEM2k physics (SGM/AJS/FXG);

-   mix and match (JSH);

-   positron sources (HGH); and

-   spontaneous fission (JSH).

*MCNPX 2.4.0 (August 2002)*

-   F90 modularity and dynamic memory allocation (GWM);

-   distributed memory multiprocessing for the entire energy range of
    all particles (GWM);

-   repeated structures source-path improvement (LLC/JSH);

-   default dose functions (LSW/JSH);

-   light-ion recoil (JSH);

-   enhanced color geometry plots (GWM/JSH);

-   photonuclear cross-section plots (JSH);

-   proton cross-section plots (JSH);

-   proton reaction multipliers with FM cards (JSH);

-   photonuclear reaction multipliers with FM cards (JSH/GWM);

-   some speedups (GWM/JSH);

-   logarithmic interpolation on input cards (JSH);

-   cosine bins that may be specified in degrees (JSH);

-   cosine bins that may be specified for F2 flux tallies (JSH);

-   source particles that may be specified by descriptors (JSH); and

-   pause command for tally and cross-section plots (JSH).

*MCNPX 2.3.0 and previous MCNPX versions (1995 to 2001)*

-   physics for 34 particle types (HGH);

-   high-energy physics above the tabular data range (REP);

-   photonuclear physics (MCW);

-   neutron, proton, and photonuclear 150-MeV libraries and use (RCL);

-   mesh tallies (tallies in a superimposed mesh) (LSW/ECS);

-   radiography tallies (JSH/ECS);

-   secondary-particle production biasing (ECS); and

-   autoconfiguration build system for compilation (TLR/HWE).

*MCNP4C3, MCNP4C2, and MCNP4C features added after MCNP4B (1997--2001)*

-   PC enhancements: Linux and Windows capable (LLC/GWM);

-   easier geometry specification with macrobodies (LLC);

-   interactive geometry plotting (JSH);

-   improved variance reduction with the superimposed mesh WWG
    (TME/JAF/JSH);

-   superimposed mesh plotting (JSH);

-   delayed neutrons (CJW);

-   unresolved resonance range probability tables (LLC/RCL);

-   perturbations for material-dependent tallies (GWM/LLC/JSH);

-   ENDF/B-VI extensions (MCW);

-   electron physics enhancements (upgrade to ITS3.0)[^16] (KJA/HGH);

-   weight-window enhancements (JSH/JAF); and

-   distributed memory multiprocessing (GWM).

# 10.0. Future work {#future-work-1 .unnumbered}

-   CINDER90 capabilities:

```{=html}
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```
-   delayed neutrons physics models;

-   delayed gamma physics models; and

-   transmutation.

```{=html}
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```
-   improved high-energy physics with the LAQGSM model;

-   heavy-ion tracking and interactions;

-   criticality:

```{=html}
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```
-   externally driven sources and

```{=html}
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```
-   improved stability of eigenfunctions.

```{=html}
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```
-   plotting of physics model total and absorption cross sections;

-   forced collisions for neutral particles extended to physics models;

-   secondary particle angle biasing for isotropic distributions;

-   neutral particle perturbation techniques extended to physics model
    region;

-   heavy-ion tracking and interactions;

-   interactive tally and cross-section plotting;

-   detectors and DXTRAN for all neutral particles at all energy ranges;

-   a capability to continue runs that write HTAPE files;

-   integration of HTAPE tallies directly into MCNPX;

-   CAD link; and

-   magnetic field tracking.

# 11.0. BIBLIOGRAPHY {#bibliography .unnumbered}

## 11.1. Code Manuals {#code-manuals .unnumbered}

Laurie S. Waters, ed., "MCNPX User's Manual, Version 2.4.0," Los Alamos
National Laboratory report LA-CP-02-408 (September 2002).

Laurie S. Waters, ed., "MCNPX User's Manual, Version 2.3.0," Los Alamos
National Laboratory report LA-UR-02-2607 (April 2002).

J. F. Briesmeister, ed., "MCNP---A General Monte Carlo N-Particle
Transport Code---Version 4C," Los Alamos National Laboratory report
LA-13709-M (April 2000).

Laurie S. Waters, ed., "MCNPX User's Manual---Version 2.1.5," Los Alamos
National Laboratory report LA-UR-99-6058 (November 14, 1999).

J. F. Briesmeister, ed., "MCNP---A General Monte Carlo N-Particle
Transport Code---Version 4B," Los Alamos National Laboratory report
LA-12625-M (March 1997).

J. A. Halbleib, R. P. Kensek, T. A. Mehlhorn, G. D. Valdez, S. M.
Seltzer, and M. J. Berger, "ITS Version 3.0: The Integrated TIGER Series
of Coupled Electron/Photon Monte Carlo Transport Codes," Sandia National
Laboratories report SAND91-1634 (March 1992).

## 11.2. Code Version Release Announcements {#code-version-release-announcements .unnumbered}

John S. Hendricks et al., "MCNPX, Version 2.5.F," Los Alamos National
Laboratory report, LA-UR-05-0891 (February 2005).

John S. Hendricks et al., "MCNPX, Version 2.5.E," Los Alamos National
Laboratory report, LA-UR-04-0569 (February 2004).

John S. Hendricks et al., "MCNPX, Version 2.5.D," Los Alamos National
Laboratory report, LA-UR-03-5916 (August 2003).

John S. Hendricks et al., "MCNPX, Version 2.5.C," Los Alamos National
Laboratory report LA‑UR-03-2202 (April 2002).

John S. Hendricks and the MCNPX Team, "MCNPX, Version 2.5.B," Los Alamos
National Laboratory report LA-UR-02-7086 (November 2002).

MCNPX Team, "MCNPX, Version 2.4.0," Los Alamos National Laboratory
report LA-UR-02-5253 (August 2002).

John S. Hendricks, "MCNP4C3," Los Alamos National Laboratory report
LA-UR-01-2244 (April 13, 2001).

John S. Hendricks, "MCNP4C2," Los Alamos National Laboratory report
LA-UR-01-858 (January 30, 2001).

John S. Hendricks, "MCNP4C," Los Alamos National Laboratory memorandum
X-5:JSH-2000-30 (February 29, 2000).

John S. Hendricks, "MCNP4B," Los Alamos National Laboratory memorandum
XTM:JSH-97-13 (January 31, 1997).

## 11.3. Reports {#reports .unnumbered}

John S. Hendricks, "MCNPX Model/Table Comparison," Los Alamos National
Laboratory report LA-14030 (March 2003).

## 11.4. Papers {#papers .unnumbered}

J. S. Hendricks, G. W. McKinney, J. W. Durkee, M. R. James, D. B.
Pelowitz, and H. R. Trellue, "The MCNPX Radiation Transport Code," Los
Alamos National Laboratory report LA-UR-05-0683, submitted to M&C 2005,
the International Topical Meeting on Mathematics and Computation,
Avignon, France (September 12--15, 2005).

John S. Hendricks and Jane M. Burward-Hoy, "Monte Carlo Radiation
Detector Modeling in Space Systems," Los Alamos National Laboratory
report LA-UR-05-1841, submitted to Space Nuclear Conference (SNC '05),
San Diego, California (June 6--9, 2005).

M. James, J. Durkee, J. Hendricks, G. McKinney, D. Pelowitz, and F.
Gallmeier, "Physics Improvements in MCNPX 2.5.0," International
Collaboration on Advanced Neutron Sources, ICANS XVII, Santa Fe, New
Mexico (April 25--29, 2005).

G. McKinney, J. Durkee, J. Hendricks, M. James, D. Pelowitz, and F.
Gallmeier, "New MCNPX 2.5.0 Features," International Collaboration on
Advanced Neutron Sources, ICANS XVII, Santa Fe, New Mexico (April
25--29, 2005).

John S. Hendricks, "Monte Carlo Sampling of Fission Multiplicity," Los
Alamos National Laboratory report LA-UR-04-6967, submitted to Monte
Carlo 2005, Chattanooga, Tennessee (April 17--21, 2005).

John S. Hendricks and Gregg W. McKinney, "Pulse-Height Tallies with
Variance Reduction," Los Alamos National Laboratory report
LA-UR-04-8431, submitted to Monte Carlo 2005, Chattanooga, Tennessee
(April 17--21, 2005).

Joshua P. Finch, John S. Hendricks, Chan K. Choi, "Vacation Matrix
Method for Correct Source Distribution in Monte Carlo Criticality
Calculations," Los Alamos National Laboratory report LA-UR-04-8595,
submitted to Monte Carlo 2005, Chattanooga, Tennessee (April 17--21,
2005).

G. W. McKinney, J. W. Durkee, J. S. Hendricks, M. R. James, D. B.
Pelowitz, and L. S. Waters, "MCNPX 2.5.0---New Features Demonstrated,"
Los Alamos National Laboratory report LA-UR-04-8695, submitted to Monte
Carlo 2005, Chattanooga, Tennessee (April 17--21, 2005).

Martyn T. Swinhoe, John S. Hendricks, and Douglas R. Mayo, "MCNPX for
Neutron Multiplicity Detector Simulation," Los Alamos National
Laboratory report LA-UR-04-8025, submitted to European SAfeguards
Research and Development Association (ESARDA) Symposium on Safeguards
and Nuclear Material Management, London, United Kingdom (June 2005).

J. S. Hendricks, G. W. McKinney, L. S. Waters, and F. X. Gallmeier, "New
MCNPX Capabilities," Los Alamos National Laboratory report
LA-UR-04-0193, *Transactions of the American Nuclear Society,*
Pittsburgh, PA **90,** 475 (June 13--17, 2004).

J. S. Hendricks, J. P. Finch, and C. Choi, "Calculation of Eigenfunction
Fluxes in Nuclear Systems," Los Alamos National Laboratory report
LA-UR-03-2917, American Nuclear Society Topical Meeting, Radiation
Protection and Shielding Division, ICRS-10/RPS 2004, Madiera, Portugal,
published by Instituto Tecnologico e Nuclear, Sacavem, Portugal (May
10--14, 2004).

G. W. McKinney, J. S. Hendricks, H. R. Trellue, and L. S. Waters,
"Eigenfunction Convergence and Transmutation Enhancements in MCNPX," Los
Alamos National Laboratory report LA-UR-03-7714, submitted to
PHYSOR-2004, American Nuclear Society Topical Meeting, Reactor Physics
Division, Chicago, Illinois (April 25--29, 2004).

J. S. Hendricks, M. T. Swinhoe, S. J. Tobin, W. H. Geist, and D. R.
Mayo, "Neutron Multiplicity Counting for Nuclear Safeguards with MCNPX,"
Los Alamos National Laboratory report LA-UR-03-0456 (2003).

J. S. Hendricks, G. W. McKinney, L. S. Waters, and Franz X. Gallmeier,
"MCNPX Advances for Accelerator Applications," *Sixth International
Meeting on the Nuclear Applications of Accelerator Technology*
(AccApp'03), American Nuclear Society Embedded Topical Meeting, San
Diego, California (June 1--5, 2003), Los Alamos National Laboratory
report LA-UR-03-0075.

John S. Hendricks, Gregg W. McKinney, and Franz G. Gallmeier, "Mix and
Match with MCNPX," *Nuclear Mathematical and Computational Sciences: A
Century in Review, A Century Anew*, Gatlinburg, Tennessee, on CD-ROM
(American Nuclear Society, LaGrange Park, Ilinois), ANS order no.
700300, ISBN: 0‑89448-674-8 (April 6--11, 2003), Los Alamos National
Laboratory report LA-UR-02-6415 (2002).

J. S. Hendricks, G. W. McKinney, L. S. Waters, H. G. Hughes, and E. C.
Snow, "New MCNPX Developments," Los Alamos National Laboratory report
LA-UR-02-2181 (April 2002).

Stepan G. Mashnik and Arnold J. Sierk, "Recent Developments of the
Cascade-Exciton Model of Nuclear Reactions," International Conference on
Nuclear Data for Science and Technology, October 7--12, 2001, Tsukuba,
Japan. *Nuclear Science and Technology, Supplement 2*, 720--725 (2002):
<http://lib-www.lanl.gov/la-pubs/00818526.pdf>.

E. C. Selcow, G. W. McKinney, T. E. Booth, J. F. Briesmeister, L. J.
Cox, R. A. Forster, J. S. Hendricks, R. D. Mosteller, R. E. Prael, A.
Sood, and S. W. White, "MCNP Capabilities at the Dawn of the 21st
Century: Neutron-Gamma Applications," Los Alamos National Laboratory
report LA-UR-00-5017 (2000).

J. S. Hendricks, K. J. Adams, T. E. Booth, J. F. Briesmeister, L. L.
Carter, L. J. Cox, J. A. Favorite, R. A. Forster, G. W. McKinney, and R.
E. Prael, "Present and Future Capabilities of MCNP," *Applied Radiation
and Isotopes* **53**:4-5, 857--861 (2000).

J. S. Hendricks, "Advances in MCNP4C," Los Alamos National Laboratory
report LA-UR-00-2643 (2000).

G. W. McKinney, T. E. Booth, J. F. Briesmeister, L. J. Cox, R. A.
Forster, J. S. Hendricks, R. D. Mosteller, R. E. Prael, and A. Sood,
"MCNP Applications for the 21st Century," in "Proceedings of the 4th
International Conference on Supercomputing in Nuclear Applications,"
Tokyo, Japan (September 4--7, 2000).

J. S. Hendricks, G. W. McKinney, and L. J. Cox, "Monte Carlo Advances
for the Eolus ASCI Project," PHYSOR2000 International Topical Meeting on
Advances in Reactor Physics and Mathematics and Computation, Pittsburgh,
Pennsylvania, May 7--11, 2000.

K. J. Adams, T. E. Booth, J. F. Briesmeister, L. L. Carter, J. D. Court,
L. J. Cox, J. A. Favorite, R. A. Forster, J. S. Hendricks, G. W.
McKinney, and R. E. Prael, "Present and Future Capabilities of MCNP," in
"Fourth Topical Meeting on Industrial Radiation and Radioisotope
Measurement Applications (IRRMA 99)," Raleigh, North Carolina (October
3--7, 1999), p. 38.

H. G. Hughes, K. J. Adams, M. B. Chadwick, J. C. Comly, L. J. Cox, H. W.
Egdorf, S. C. Frankle, J. S. Hendricks, R. C. Little, R. E. MacFarlane,
S. G. Mashnik, R. E. Prael, A. J. Sierk, L. S. Waters, M. C. White, P.
G. Young, Jr., F. Gallmeier, and E. C. Snow, "MCNPX for Neutron-Proton
Transport," in "Proceedings of Mathematics and Computation, Reactor
Physics, and Environmental Analysis in Nuclear Applications," Madrid,
Spain, Vol. 2, pp. 939--948 (September 1999).

Stepan G. Mashnik and Arnold J. Sierk, "Improved Cascade Exciton Model
of Nuclear Reactions," in "Proceedings of the Fourth Workshop on
Simulating Accelerator Radiation Environments (SARE4)," Knoxville,
Tennessee, September 14--16, 1998, T. A. Gabriel, Ed. (Oak Ridge
National Laboratory, Oak Ridge, Tennessee, 1999), pp. 29--51.

H. G. Hughes, K. J. Adams, M. B. Chadwick, J. C. Comly, L. J. Cox, H. W.
Egdorf, S. C. Frankle, F. X. Gallmeier, J. S. Hendricks, R. C. Little,
R. E. MacFarlane, R. E. Prael, E. C. Snow, L. S. Waters, M. C. White,
and P. G. Young, Jr., "Recent Developments in MCNPX," in "Proceedings of
the Second International Topical Meeting on Nuclear Applications of
Accelerator Technology," Gatlinburg, Tennessee (September 20--23, 1998),
p. 281.

#  {#section-2 .unnumbered}

# 12.0. AcknowledgmentS {#acknowledgments .unnumbered}

The authors wish to thank Lisa Rothrock and Tamara Hallman of group IM-1
for their editing of this document.

[^1]: Kenneth J. Adams (KJA), Nate A. Carstens (NAC), Leland L. Carter
    (LLC), Skip Egdorf (HWE), Thomas M. Evans (TME), Jeffrey A. Favorite
    (JAF), Franz X. Gallmeier (FXG), John S. Hendricks (JSH), H. Grady
    Hughes (HGH), Julian Lebenhaft (JL), Robert C. Little (RCL), Stepan
    G. Mashnik (SGM), Douglas R. Mayo (DRM), Gregg W. McKinney (GWM),
    Eric J. Pitcher (EJP), Richard E. Prael (REP), Teresa L. Roberts
    (TLR), Arnold J. Sierk (AJS), Edward C. Snow (ECS), Avneet Sood
    (AS), Martyn T. Swinhoe (MTS), Stephen J. Tobin (SJT), Laurie S.
    Waters (LSW), Christopher J. Werner (CJW), and Morgan C. White
    (MCW).

[^2]: Cash Award Fine Print: Offer is subject to cancellation or
    modification without notice. A bug is defined as an error in the
    source code that we choose to correct. We make awards for even the
    most trivial or insignificant problems but not for proposed code
    enhancements or proposed extended capabilities. Awards are given
    only to the first MCNPX user reporting a problem. Reported problems
    must be reproducible, and awards are paid when the correction is
    integrated into a forthcoming MCNPX version. We believe that MCNPX
    and its predecessor codes are the most error-free and robust Monte
    Carlo radiation transport capabilities, and we back them with a cash
    guarantee.

[^3]: J. F. Briesmeister, ed., "MCNP---A General Monte Carlo N-Particle
    Transport Code---Version 4C," Los Alamos National Laboratory report
    LA-13709-M (April 2000).

[^4]: John S. Hendricks, "Monte Carlo Sampling of Fission Multiplicity,"
    Los Alamos National Laboratory report LA-UR-04-6967, submitted to
    Monte Carlo 2005, Chattanooga, Tennessee (April 17--21, 2005).

[^5]: Franz Gallmeier, private communication, April 2004.

[^6]: X-5 Monte Carlo Team, "MCNP---A General Monte Carlo N-Particle
    Transport Code, Version 5, Volume II: User's Guide," Los Alamos
    National Laboratory report LA-CP-03-0245 (April 24, 2003), pp.
    2--59.

[^7]: John S. Hendricks, "MCNPX Model/Table Comparison," Los Alamos
    National Laboratory report LA-14030 (March 2003).

[^8]: M. B. Chadwick, P. G. Young, S. Chiba, S. C. Frankle, G. M. Hale,
    H. G. Hughes, A. J. Kongin, R. C. Little, R. E. MacFarlane, R. E.
    Prael, and L. S. Waters, "Cross Section Evaluations to 150 MeV for
    Accelerator-Driven Systems and Implementation in MCNPX," Nuclear
    Science and Engineering **131**: 3 (March 1999), 293.

[^9]: S. G. Mashnik and A. J. Sierk, "Recent Developments of the
    Cascade-Exciton Model of Nuclear Reactions," Los Alamos National
    Laboratory report LA-UR-01-5390, and International Conference on
    Nuclear Data for Science and Technology, Tsukuba, Japan (October
    7--12, 2001).

[^10]: John S. Hendricks, "New Spontaneous Fission Data," Los Alamos
    National Laboratory internal memorandum D-5:JSH-2005-064 (December
    14, 2004).

[^11]: P. Santi, D. Beddingford, and D. Mayo, "Revised Prompt Neutron
    Emission Multiplicity Distributions for\
    236, 238-Pu," Los Alamos National Laboratory report LA-UR-04-8040
    (December 2004).

    > N. E. Holden and M. S. Zucker, BNL-NCS-35513-R. IAEA Advisory
    > Group Meeting on Neutron Standard Reference Data, November 12--16,
    > 1984, Geel, Belgium.
    >
    > N. Ensslin, W. C. Harker, M. S. Krick, D. G. Langer, M. M.
    > Pickrell, and J. E. Stewart, "Application Guide to Neutron
    > Multiplicity Counting," Los Alamos National Laboratory report
    > LA-13422-M (1998).
    >
    > M. S. Zucker and N. E. Holden, "Californium-252 and ^238^U Nuclear
    > Parameters of Safeguards Interest," Brookhaven National Laboratory
    > report BNL-34804 (1983).
    >
    > N. E. Holden and M.S. Zucker, "Prompt Neutron Multiplicity for the
    > Transplutonium Nuclides," Brookhaven National Laboratory report
    > BNL-NCS-36379. *Proc. Int. Conf. Nucl. Data for Basic and Applied
    > Science,* May 13--17, 1985, Santa Fe, New Mexico (1985).
    >
    > D. A. Hicks, J. Ise, Jr., and R. V. Pyle, "Probabilities of
    > Prompt-Neutron Emission from Spontaneous Fission," *Phys. Rev.*
    > **101**, 1016 (1956).
    >
    > W. W. T. Crane, G. H. Higgins, and H. R. Bowman, "Average Number
    > of Neutrons per Fission for Several Heavy-Element Nuclides,"
    > *Phys. Rev.* **101**, 1804 (1956).
    >
    > J. W. Boldeman and M.G. Hines, *Nucl. Sci. Eng.* **91**, 114
    > (1985).
    >
    > B. C. Diven, H. C. Martin, R. F. Taschek, and J. Terrell,
    > "Multiplicities of Fission Neutrons," *Phys. Rev.* **101**, 1012
    > (1956).

[^12]: John S. Hendricks and Gregg W. McKinney, "Pulse-Height Tallies
    with Variance Reduction," Los Alamos\
    National Laboratory report LA-UR-04-8431 (2004).

[^13]: John S. Hendricks and Jane M. Burward-Hoy, "Monte Carlo Radiation
    Detector Modeling in Space Systems," Los\
    Alamos National Laboratory report LA-UR-05-0278 (2005).

[^14]: Martyn T. Swinhoe, John S. Hendricks, and Douglas R. Mayo, "MCNPX
    for Neutron Multiplicity Detector\
    Simulation," Los Alamos National Laboratory report LA-UR-04-8025
    (2004).

[^15]: Kenneth J. Adams (KJA), Nate A. Carstens (NAC), Leland L. Carter
    (LLC), Jean-Christophe David (JCD), Skip Egdorf (HWE), Thomas M.
    Evans (TME), Jeffrey A. Favorite (JAF), Franz X. Gallmeier (FXG),
    John S. Hendricks (JSH), H. Grady Hughes (HGH), Julian Lebenhaft
    (JL), Robert C. Little (RCL), Stepan G. Mashnik (SGM), Douglas R.
    Mayo (DRM), Gregg W. McKinney (GWM), Eric J. Pitcher (EJP), Richard
    E. Prael (REP), Teresa L. Roberts (TLR), Arnold J. Sierk (AJS),
    Edward C. Snow (ECS), Avneet Sood (AS), Martyn T. Swinhoe (MTS),
    Stephen J. Tobin (SJT), Laurie S. Waters (LSW), Christopher J.
    Werner (CJW), and Morgan C. White (MCW).

[^16]: J. A. Halbleib, R. P. Kensek, T. A. Mehlhorn, G. D. Valdez, S. M.
    Seltzer, M. J. Berger, "ITS Version 3.0: The\
    Integrated TIGER Series of Coupled Electron/Photon Monte Carlo
    Transport Codes," Sandia National\
    Laboratories report SAND91-1634 (March 1992).