where
is the name of the input file and is the burnup step.
The results for each detector are written in a 13-column table, one bin
value per row. The variable is named DET.m, where is the
detector name. The values in each column are: 1. Value index (to

flux estimator yields zero values in void regions, resulting in a
systematic over-prediction of the homogenized values. The problem
was fixed in code update 1.1.3. NOTES: 1. The first two entries are the
total (one-group) value and the associated relative

the ENDF format fission yield library The spontaneous fission yield
library is optional. If the file path is not set, the code uses neutroninduced yields for spontaneous fission. The present code version does
not model spontaneous fission. A default direc

determine the isotopic composition explicitly. 3. The methodology is
available from code version 1.0.2 on. SEE ALSO: 1. Definition of
irradiation history (Sec. 8.3 on page 110) 5.15 Iteration 70 5.15
Iteration keff can be iterated to a desired value by al

for the Evaluated Nuclear Data File ENDF-6. BNL-NCS-44945-01/04Rev. Brookhaven National Laboratory, 2001. [7] N. Messaoudi and B.C. Na. VENUS-2 MOX-fuelled Reactor Dosimetry Calculations, Final
Report. NEA/NSC/DOC(2005)22. OECD/NEA, 2004. [8] J. Leppnen.

and options. Option Description Section Page pop (3-4) population
size and number of cycles 5.2 53 nbuf (1) source buffer 5.2 53 egrid
(1-3) energy grid reconstruction 5.3 55 dix (1) double indexing of
energy grids 5.3 56 acelib (1) file path for xs libra

emission energy is sampled. If the value is not set, the minimum value
allowed by the distribution is used. The third option is to define discrete
energy bins as: src sb . where is the number of source energy bins are
the energy bin boundaries are the bin

defined. The reaction rates are calculated by summing over all
nuclides in the material. MCNP also 7.1 Detector Input 99 uses some
code-specific negative reaction MTs, but the interpretations are
slightly different. 6. The fission energy deposition functi

simultaneous post-processing of several files. Each parameter is read
to a variable (scalar or vector) and a run index idx is assigned to
each file. Each time a new file is read, the index is first increased by, 1
so that the new data is placed on the nex

i is given by: n = 2(i 1)G + 2j 1 2. The production matrixes
include neutron multiplication in (n,xn) reactions. 6.1.27 Diffusion
parameters Parameter Values Description DIFFAREA 2G + 2
Diffusion area DIFFCOEF 2G + 2 Diffusion coefficient TRANSPXS
2G + 2

independently for each burnup step. 2. Soluble absorber must be
defined in the absorber iteration mode. 3. The -eigenvalue
calculation, albedo iteration and B1 mode are available from update
1.1.5 on. 4. The albedo- and B1-iteration modes are experimental

Defines the x-mesh where the reactions are scored dy Detector mesh
Defines the y-mesh where the reactions are scored dz Detector mesh
Defines the z-mesh where the reactions are scored dt Detector type
Special detector types ds Surface current detector Def

generation rate TOT_FISSRATE 2 Total fission rate TOT_ABSRATE
2 Total absorption rate TOT_LEAKRATE 2 Total leakage rate
TOT_LOSSRATE 2 Total loss rate TOT_SRCRATE 2 Total source
rate TOT_FLUX 2 Total flux TOT_RR 2 Total reaction rate
TOT_SOLU_ABSRATE 2 To

number of collisions SLOW_TIME 2 Average slowing-down time
THERM_TIME 2 Average thermal life time SLOW_DIST 2 Average
slowing-down distance THERM_DIST 2 Average thermal migration
distance THERM_FRAC 2 Average fraction of neutrons reaching
thermalization 6

the independent calculation mode is written in Matlab m- file format in
file
_dep.m, where
is the name of the input file. The
variables are summarized in Table 8.4. The number of burnup steps is N
and the number of inventory nuclides I. The material-wise

Sections There are two options for calculating the isotopic one-group
transmutation cross sections: set xscalc where is the method used for
cross section calculation In the default method ( = 2), the code
calculates these parameters using a highresolution

majorant. 2. Sample cross sections when the neutron enters a new
material. Use a pre-calculated majorant cross section corresponding to
the maximum probability table values. 3. Sample cross sections when
the neutron enters a new material. Switch to surfac

uniformly all over the geometry. The sampling volume can limited by
setting the boundaries in x-, y- and z-directions using: 9.2 Source
definition 127 src sx sy sz where is the source name is the minimum
boundary in x-direction is the maximum boundary in

burnup calculation is under development. 2. The group boundaries in
the few-group structure must match the boundaries in the microgroup structure. 3. Relative statistical errors are not included in the
results. 4. The fundamental mode calculation must not

upper and lower limits and the number must be consistent with the
number of given values ( - 1 values for groups). When it comes to
multiplying scattering reactions, such as (n,2n), (n,3n) or (n, 2n),
there is some ambiguity in the way group-to-group scat

URES_MODE 1 Probability table sampling mode URES_DILU_CUT
1 Infinite dilution cut-off URES_EMIN 1 Minimum energy for
unresolved resonance probability table data (MeV) URES_EMAX 1
Maximum energy for unresolved resonance probability table data
(MeV) URES_AV

SIX_FF_LF 2 Fast non-leakage probability SIX_FF_LT 2 Thermal
non-leakage probability SIX_FF_KINF 2 Six-factor k (four-factor
keff) SIX_FF_KEFF 2 Six-factor keff NOTES: 1. The parameters are
calculated using simple analog estimates and inteded mainly for t

Absorption includes all reactions in which the incident neutron is lost,
i.e. all capture reactions and fission. The default normalization is
absorption rate set to unity. Normalization to total loss rate is set
using: 5.8 Source rate normalization 62 set

entropy 6.1 Main output file 84 6.1.13 Fission source center
Parameter Values Description SOURCE_X0 2 X-coordinate of fission
source center SOURCE_Y0 2 Y-coordinate of fission source center
SOURCE_Z0 2 Z-coordinate of fission source center 6.1.14 Soluble

the detector material is the detector universe Detector cells can be
either physical or super-imposed on the geometry. Super-imposed cells
are not used for defining material regions. They must contain void
material and the universe number must be set to a

into account using the universe symmetry option: set usym where is
the universe number is symmetry type is the x-coordinate of symmetry
origin is the y-coordinate of symmetry origin Present version of
Serpent allows only quadrant symmetries ( = 4) in univ

uniform with respect to the lethargy variable. The plotter produces a
file
_xs.png, where
is the name of the input file and is
the burnup step. The file contains the energy grid vector, isotopic
reaction cross sections, material total cross sections and

created for the results. The response functions are listed in Table 7.2.
Negative entries define total reaction rates related to materials. The
total cross section (mt = -1), for example, is calculated from: R = 1 V Z
V Z Ei Ei+1 X j tot,j (r, E)(r, E) d

without running the depletion calculation. The code treats depleted
materials in fuel pins different from materials in ordinary cells. Each pin
type is treated separately and further divided into annular depletion
zones of equal volume. The division is im