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+\section{Contents of a 
+\dir{burnup} directory}\label{sect:burnupdir}
+
+This directory contains the main burnup information, namely  the multigroup flux and the
+isotopic concentration at each time or burnup step.
+
+\subsection{State vector content for the \dir{burnup} data structure}\label{sect:burnupstate}
+
+The dimensioning parameters for the \dir{burnup} data structure, which are stored in
+the state vector $\mathcal{S}^{b}$, represent:
+
+\begin{itemize}
+\item The type of solution considered $I_{s}=\mathcal{S}^{b}_{1}$ where
+\vskip -0.8cm
+\begin{displaymath}
+I_{s} = \left\{
+\begin{array}{rl}
+ 1 & \textrm{Fifth-order Cash-Karp method}\\
+ 2 & \textrm{Forth-order Kaps-Rentrop method}  
+\end{array} \right.
+\end{displaymath}
+
+\item The type of burnup considered $I_{t}=\mathcal{S}^{b}_{2}$ where
+\vskip -0.8cm
+\begin{displaymath}
+I_{t} = \left\{
+\begin{array}{rl}
+ 0 & \textrm{Out of core or zero flux/power depletion} \\
+ 1 & \textrm{Constant flux depletion} \\
+ 2 & \textrm{Constant fuel power depletion} \\
+ 3 & \textrm{Constant assembly power depletion} 
+\end{array} \right.
+\end{displaymath}
+
+\item Number of time steps for which burnup properties are present in this directory
+$N_{t}=\mathcal{S}^{b}_{3}$
+ 
+\item Total number of isotopes $N_{I}=\mathcal{S}^{b}_{4}$ 
+
+\item Number of depleting mixtures $N^{\rm depl}_{M}=\mathcal{S}^{b}_{5}$
+
+\item Number of depleting reactions $N^{\rm depl}_{R}=\mathcal{S}^{b}_{6}$
+
+\item Number of depleting isotopes $N^{\rm depl}_{I}=\mathcal{S}^{b}_{7}$
+
+\item Number of mixtures $N_m=\mathcal{S}^{b}_{8}$
+
+\item Microscopic reaction rate extrapolation option in solving the burnup equations
+$I_{e}=\mathcal{S}^{b}_{9}$ where
+\vskip -0.8cm
+\begin{displaymath}
+I_{e} = \left\{
+\begin{array}{rl}
+ 0 & \textrm{Do not extrapolate} \\
+ 1 & \textrm{Perform linear extrapolation} \\
+ 2 & \textrm{Perform parabolic extrapolation} \\
+\end{array} \right.
+\end{displaymath}
+
+\item Constant power normalization option for the burnup calculation
+$I_{g}=\mathcal{S}^{b}_{10}$ where
+\vskip -0.8cm
+\begin{displaymath}
+I_{g} = \left\{
+\begin{array}{rl}
+ 0 & \textrm{Compute the burnup using the power released in fuel} \\
+ 1 & \textrm{Compute the burnup using the power released in the global geometry} \\
+\end{array} \right.
+\end{displaymath}
+This option have an effect only in cases
+where some non-depleting mixtures are producing energy.
+
+\item Saturation of initial number densities $I_{s}=\mathcal{S}^{b}_{11}$ where
+\vskip -0.8cm
+\begin{displaymath}
+I_{s} = \left\{
+\begin{array}{rl}
+ 0 & \textrm{Do not store saturated initial number densities in the {\sc burnup}
+ object} \\
+ 1 & \textrm{Store saturated initial number densities} \\
+\end{array} \right.
+\end{displaymath}
+This option have an effect only in cases where some depleting isotopes are
+at saturation.
+
+\item Type of saturation model $I_{d}=\mathcal{S}^{b}_{12}$ where
+\vskip -0.8cm
+\begin{displaymath}
+I_{d} = \left\{
+\begin{array}{rl}
+ 0 & \textrm{Do not use Dirac functions in saturated number densities} \\
+ 1 & \textrm{Use Dirac functions in saturated number densities} \\
+\end{array} \right.
+\end{displaymath}
+This option have an effect only in cases where some depleting isotopes are
+at saturation.
+
+\item Perturbation flag for cross sections $I_{p}=\mathcal{S}^{b}_{13}$ where
+\vskip -0.8cm
+\begin{displaymath}
+I_{p} = \left\{
+\begin{array}{rl}
+ 0 & \textrm{Time-dependent cross sections will be used if available} \\
+ 1 & \textrm{Time-independent cross sections will be used} \\
+\end{array} \right.
+\end{displaymath}
+
+\item Neutron flux recovery flag $I_{f}=\mathcal{S}^{b}_{14}$ where
+\vskip -0.8cm
+\begin{displaymath}
+I_{f} = \left\{
+\begin{array}{rl}
+ 0 & \textrm{Neutron flux is recovered from a L\_FLUX object} \\
+ 1 & \textrm{Neutron flux is recovered from the embedded macrolib present in a} \\
+ & \textrm{L\_LIBRARY object} \\
+\end{array} \right.
+\end{displaymath}
+
+\item Fission yield data recovery flag $I_{y}=\mathcal{S}^{b}_{15}$ where
+\vskip -0.8cm
+\begin{displaymath}
+I_{y} = \left\{
+\begin{array}{rl}
+ 0 & \textrm{Fission yield data is recovered from {\tt DEPL-CHAIN} directory (see \Sect{microlibdirdepletion})} \\
+ 1 & \textrm{Fission yield data is recovered from {\tt PIFI} and {\tt PYIELD} records in /isotope/} \\
+ & \textrm{directory (see Table~\ref{tabl:tabiso3})} \\
+\end{array} \right.
+\end{displaymath}
+\end{itemize}
+
+\subsection{The main \dir{burnup} directory}\label{sect:burnupdirmain}
+
+On its first level, the
+following records and sub-directories will be found in the \dir{burnup} directory:
+
+\begin{DescriptionEnregistrement}{Main records and sub-directories in \dir{burnup}}{8.0cm}
+\CharEnr
+  {SIGNATURE\blank{3}}{$*12$}
+  {Signature of the \dir{burnup} data structure ($\mathsf{SIGNA}=${\tt L\_BURNUP\blank{4}}).}
+\IntEnr
+  {STATE-VECTOR}{$40$}
+  {Vector describing the various parameters associated with this data structure
+  $\mathcal{S}^{b}_{i}$, as defined in \Sect{burnupstate}.}
+\RealEnr
+  {EVOLUTION-R\blank{1}}{$5$}{}
+  {Vector describing the various parameters associated with the burnup calculation options
+$R_{i}$}
+\CharEnr
+  {LINK.LIB\blank{4}}{$*12$}
+  {Name of the {\sc microlib} on which the last depletion step was based.}
+\RealEnr
+  {DEPL-TIMES\blank{2}}{$N_{t}$}{$10^{8}$ s}
+  {Vector describing the various time steps at which burnup information has been saved
+$T_{i}$}
+\RealEnr
+  {FUELDEN-INIT}{$3$}{}
+  {Vector giving the initial density of heavy element in the fuel $\rho_{f}$ (g
+  cm$^{-3}$), the initial mass of heavy element in the fuel $m_{f}$ (g) and the
+  initial mass of heavy element in the fuel divided by the global geometry
+  volume (g cm$^{-3}$)}
+\RealEnr
+  {VOLUME-MIX\blank{2}}{$N_m$}{cm$^3$}
+  {Vector giving the mixture volumes}
+\RealEnr
+  {FUELDEN-MIX\blank{1}}{$N_m$}{g}
+  {Initial mass of heavy element contained in each mixture}
+\RealEnr
+  {WEIGHT-MIX\blank{2}}{$N_m$}{g}
+  {Initial mass of all the isotopes contained in each mixture}
+\IntEnr
+  {DEPLETE-MIX\blank{1}}{$N_m \times N^{\rm depl}_{I}$}
+  {Matrix giving the index in the {\tt ISOTOPESDENS} record of each depleting
+   isotope in each mixture.}
+\CharEnr
+  {ISOTOPESUSED}{$(N_{I})*12$}
+  {Alias name of the isotopes}
+\IntEnr
+  {ISOTOPESMIX\blank{1}}{$N_{I}$}
+  {Mixture number associated with each isotope}
+\IntEnr
+  {MIXTURESBurn}{$N_m$}
+  {Depletion flag array. A component is set to 1 to indicate that a mixture is depleting.}
+\IntEnr
+  {MIXTURESPowr}{$N_m$}
+  {Power flag array. A component is set to 1 to indicate that a mixture is producing power.}
+\DirVar
+  {\listedir{depldir}}
+  {Set of $N_{t}$ sub-directories containing the properties associated with each
+   burnup step $T_{i}$}
+\end{DescriptionEnregistrement}
+
+The set of directory \listedir{depldir} names $\mathsf{DEPLDIR}$ will be composed according to the
+following laws. The first eight character ($\mathsf{DEPLDIR}$\verb*|(1:8)|) will always be given by 
+\verb*|DEPL-DAT|. The last four characters  
+($\mathsf{DEPLDIR}$\verb*|(9:12)|) represent the time step saved. For the case where
+$N_{t}$ time steps were saved we would use the following FORTRAN instructions to create
+the last four characters of each of the directory names:
+$$
+\mathtt{WRITE(}\mathsf{DEPLDIR}\mathtt{(9:12),'(I4.4)')}\: J
+$$
+for $1\leq J \leq N_{t}$ with the time stamp associated with each directory being given by
+$T_{J}$.  For the case where ($N_{t}=2$), two such directory would be generated, namely
+
+\begin{DescriptionEnregistrement}{Example of depletion directories}{8.0cm}
+\DirEnr
+  {DEPL-DAT0001}{Sub-directories which contain the information associated with
+   time step 1}
+\DirEnr
+  {DEPL-DAT0002}{Sub-directories which contain the information associated with
+   time step 2}
+\end{DescriptionEnregistrement}
+
+\clearpage
+
+\subsection{The depletion sub-directory \dir{depldir} in
+\dir{burnup}}\label{sect:burnupdirdepletion}
+
+Inside each depletion directory the following records and sub-directories will be found:
+
+\begin{DescriptionEnregistrement}{Contents of a depletion sub-directory in \dir{burnup}}{7.0cm}
+\RealEnr
+  {ISOTOPESDENS}{$N_{I}$}{(cm b)$^{-1}$}
+  {Isotopic densities $\rho_{i}$ for each of the isotopes described in the \dir{microlib} directory
+   where the order of the isotopes is also specified}
+\RealEnr
+  {MICRO-RATES\blank{1}}{$N^{\rm dim}$}{$10^{-8}$ s$^{-1}\ $}
+  {Values of the microscopic reaction rate of the depleting reactions for each
+   depleting isotope and each mixture. The macroscopic reaction rate related to the
+   non-depleting isotopes is stored at location $N^{\rm depl}_{I}+1$. The
+   $N^{\rm depl}_{R}$ reaction types are stored in the order of the {\tt
+   'DEPLETE-IDEN'} array in Table~\ref{tabl:tabchain}, starting with the {\tt 'NFTOT'}
+   reaction. The flux-induced power factors are stored in location $N^{\rm depl}_{R}$.
+   The decay power (delayed) factors are stored in location $N^{\rm depl}_{R}+1$ Both
+   flux-induced and decay power are given in units of $10^{-8}$ MeV/s.
+   $N^{\rm dim}=(N^{\rm depl}_{I}+1)
+   \times (N^{\rm depl}_{R}+1) \times N_m$}
+\RealEnr
+  {INT-FLUX\blank{4}}{$N_m$}{cm s$^{-1}$}
+  {Integrated flux in each mixture.}
+\RealEnr
+  {FLUX-NORM\blank{3}}{$1$}{$1$}
+  {Flux normalization constant. It is zero for out of core depletion and
+   represents the
+   normalization of the flux $\phi_{r}^{g}$ that is used to ensure that the cell integrated flux or
+   power is that required when fixed flux or power burnup is requested}
+\RealEnr
+  {ENERG-MIX\blank{3}}{$N_m$}{$10^{-8}$ J}
+  {Energy realeased during the time step in each mixture}
+\OptRealEnr
+  {FORM-POWER\blank{2}}{1}{$I_{t}=3$}{1}
+  {Ratio of the global power released in the complete geometry divided by the
+   power released in fuel.}
+\RealEnr
+  {BURNUP-IRRAD}{$2$}{}
+  {Fuel burnup (MW d T$^{-1}$) and irradiation (Kb$^{-1}$) reached at this time step}
+\end{DescriptionEnregistrement}
+
+\eject