ORILL¶

ORILL is a code suite for nuclear research reactors design and operational safety.

ORILLCODE: 1D transmutation, fuel depletion, and radiological protection
MULTILLABY: multi-compartment model dispersion and transmutation
HARMATTILL: atmospheric plume dispersion and radiological consequences

ORILL has been cooked at the Nuclear Safety Unit of the Institut Laue-Langevin (ILL), Grenoble, France. Actually, there is a need for such pocket tools that not addressed by available heavy codes.

The name “ORILL” comes from “ILL” and the prefix “OR” is a tribute to the excellent ORIP_XXI freeware from E.G.Romanov, V.A.Tarasov, F.Z.Vahetov, Research Institute of Atomic Reactors, Dimitrovgrad, Russia.

ORILLCODE¶

Nuclear Data¶

Libraries¶

ORILL Code data are parsed from evaluated libraries in ENDF-6 or ACE format by using the PyNE toolkit. Parsing features are not provided with ORILL public version. However, data provided with public ORILL have been selected from several libraries in order to cover a wide range of situations:

Nuclide set consists of 3851 nuclides from the European Nuclear Data JEFF-3.3 decay sub-library.

Incident neutron data cover 816 nuclides cross sections and multiplicities from the European Activation File (EAF 2010) at 293.6K: (only (n,2n), (n,3n), (n,fission), (n,gamma), (n,p), (n,n+p), (n,d), (n,t), (n,3He), (n,alpha), (n,n+alpha) reactions are selected).

Direct (independent) induced fission yields of 83 nuclides are from TENDL 2011 TALYS nuclear model code system (only products with yields > 0.01% are selected).

Effective dose coefficients for inhalation (adult, public) includes 1975 nuclides from EAF 2010 and ICRP 72.

Fluence to effective dose coefficients (H*(10)) are from ICRP 74 (we consider ICRP 116 values questionable).

The choice of EAF 2010 and TENDL 2011 was made to enlarge the data set in terms of available cross sections and fission products (independent fission yields).

Radiological protection¶

The effective dose coefficients for inhalation (Sv/Bq) (adult, public) from EAF 2010 comprises ICRP 72 data and calculated coefficients by the EAF project. Be careful that EAF 2010 coefficients for inter gases (Ar, Kr, Xe) have been converted from dose rate unit per air concentration to Sv/Bq, based on the metabolic behavior of Yttrium (EAF 2010 biological, clearance and transport libraries, L. W. Packer and J-Ch. Sublet, EURATOM/CCFE Fusion Association). These coefficient are therefore much more conservative than in the ICRP 72.

ORILL Data Viewer¶

ORILLData.exe is a graphical user interface to visualize ORILL nuclear data. A demo program for Windows can be found here: ORILLDataDemo.exe.

Warning¶

ORILL Code public version is provided for educational purpose, with limited nuclear data. Additional data is available for individuals or institutions that would like to collaborate.

ORILLCODE ¶

Purpose¶

ORILLCODE is a 1D transmutation, fuel depletion (burn-up), and radiological protection code useful for nuclear research reactors design and operational safety. ORILL has been cooked at the Nuclear Safety Unit of the Institut Laue Langevin (ILL), Grenoble, France.

General Features¶

Small:	Code and nuclear data are small (only 4 Mo).
Fast:	Vectorization and parallelization associated with Python/Scipy is transparently available.
User friendly:	ORILL is designed to address practical questions about isotopes inventories, decay heat, dose rate, photons spectrum. It is usable by everyone for 1D point-flux calculation at the speed of light.

Prerequisites¶

ORILL needs only Python 2.7 or later, with Scipy and standard modules installed (re, io, yaml, time). Python is available on all platforms (WinPhyton, MiniConda, Debian/Ubuntu, Intel Distribution,…).

Installation¶

Uncompress ORILL files in a directory where a Python 2.7+ shell is available (with Scipy). In the Python shell, type:

>>> from ORILL import ORILL_CODE
>>> ORILL_CODE('ORILL_TEST.yml')

The command input file ‘ORILL_TEST.yml’ located in (/input) subdirectory is processed and the corresponding output files are created in (/output) subdirectory. Have a look at ORILL_TEST.yml file to understand the file syntax. It is recommended to edit this file with Notepad++ with the YAML markup language.

Contribute¶

If you want to contribute, let us know.We have a mailing list located at: orill_code@googlegroups.com

Support¶

If you are having issues, please let us know. We have a mailing list located at: orill_code@googlegroups.com

License¶

Redistribution and use in source and binary forms, without modification, are permitted provided that the following conditions are met:

Redistribution of source code must retain the above copyright notice, this list of conditions and the following disclaimer.
Redistribution in binary form must reproduce the above copyright notice, this list of conditions and the following disclaimer in the documentation and/or other materials provided with the distribution.

THE SOFTWARE IS PROVIDED “AS IS”, WITHOUT WARRANTY OF ANY KIND, EXPRESS OR IMPLIED, INCLUDING BUT NOT LIMITED TO THE WARRANTIES OF MERCHANTABILITY, FITNESS FOR A PARTICULAR PURPOSE AND NON INFRINGEMENT. IN NO EVENT SHALL THE AUTHORS OR COPYRIGHT HOLDERS BE LIABLE FOR ANY CLAIM, DAMAGES OR OTHER LIABILITY, WHETHER IN AN ACTION OF CONTRACT, TORT OR OTHERWISE, ARISING FROM, OUT OF OR IN CONNECTION WITH THE SOFTWARE OR THE USE OR OTHER DEALINGS IN THE SOFTWARE.

Results¶

Based on neutron flux at each time period, ORILL Code compute:

nuclides inventory after irradiation (Atom, Becquerel, Gram)
fission rate and power (Fission/s, Watt)
decay power (Nucleon, Electron, Photon) (Watt)
effective dose per inhalation (Sv)
decay effective gamma dose rate (mSv/h.cm2)
gamma spectrum (decay and PGAA) with MicroShield®V5 formatted input file for further shielding studies

ORILL Code compute sets of nuclides and/or elements. Elements are automatically converted into nuclides according IUPAC natural abundance.

Technical features¶

Neutron flux spectrum¶

The neutron flux is specially formatted in:

1-point thermal flux at 0.0253 eV

2-points fast flux at neutron fission spectrum average (U-235) and 14 MeV

For thermal and cold neutrons transmutation, only the thermal value is useful. For fuel depletion, only the relative values of thermal and fast neutrons are useful (because of threshold reactions). A multiplication factor can be used to adjust the fission power to the right value, without changing the flux shape.

Evolution algorithm¶

The choice of the algorithm (Bateman/TTA, Matrix Exponential, Runge-Kutta, Adams, BDF…) is a tricky problem in case of loops and stiffness in nuclides chains. After some trials, only three methods were selected.

BDF¶

ORILL can use the F77 VODE code (Brown, Byrne, Hindmarsh, 1989) with Backward Differentiation Formula (BDF) as provided with Python/Scipy to solve the stiff differential equations. VODE/BDF is faster than MMPA on old CPU and the difference with MMPA is generally < 2%. This method is generally stable, but instabilities can occur sometimes.

MMPA¶

Mini-Max Polynomial Approximation MMPA of matrix exponential can be used in ORILL. It is extremely simple to code on sparse matrix, and has several advantages over Chebyshev Rational Approximation Method (CRAM). MMPA method is very stable. It is described by Yosuke Kawamoto and al., Numerical solution of matrix exponential in burn-up equation using mini-max polynomial approximation, Annals of Nuclear Energy, Volume 80, 2015, Pages 219-224.

DIAG¶

The DIAG method uses diagonalization of the evolution matrix in C field, which is a more general analytical method than the classic Bateman / TTA. If diagonalization is possible, each eigenvalue is the pseudo decay constant of the corresponding eigenvector. A decay matrix is always diagonalizable because it hasn’t any loop and it can be rearranged into a lower triangular matrix with its real eigenvalues on the diagonal. Unfortunately, large burn-up matrix with loops in transmutation chains due to neutron absorbtion are not always diagonalizable in C field with enough accuracy. If diagonalization is impossible, the DIAG method will fail.

Inside ORILL engine¶

Nuclide names are translated in ZAAAm numbers, like ‘Am-242m’ in 952421, or ‘He-4’ in 20040. ORILL uses Python dictionaries and Numpy arrays to process data before hard computation. A nuclide set is defined at the beginning: it depends on the problem and on the computation depth into chains (set by the user). In the set, nuclides are sorted on an index [0,1,…,N-1], from which a sparse evolution matrix is build. The sparse matrix is processed with previously described numerical methods. At the end of each time step, the desired values are computed and the nuclide list is build back from the index.

Usage and download¶

Download¶

ORILL is located at Github

If you want to contribute, let us know.We have a mailing list located at: orill_code@googlegroups.com

Run test¶

Uncompress ORILL files in a directory where a Python 2.7+ shell is available (with Scipy). In the Python shell, type:

>>> from ORILL import ORILL_CODE
>>> ORILL_CODE('ORILL_TEST.yml')

The command input file ‘ORILL_TEST.yml’ located in (/input) subdirectory is processed and the corresponding output files are created in (/output) subdirectory.

Input file syntax¶

Have a look at ‘ORILL_TEST.yml’ file to understand the file syntax. It is recommended to edit this file with Notepad++ with the YAML markup language.

Flux: [1.0e+14, 1.0e+14, 1.0e+14]

# 3-points flux at [0.0253 eV, fission spectrum average (U-235), 14.0 MeV] (neutron/cm2/s)
# Fuel depletion: 3-points flux values, adjust the flux multiplier in "Periods:" to adequate Fission Power
# Transmutation with cold neutrons only: adjust thermal flux with wavelength(cold)/wavelenght(T) factor [adjusted_thermal_flux, 0.0, 0.0]

Nuclides:
 Cl: 0.1
 Na: 0.1
 U-235: 0.1

# Valide nuclides names are: 'U','PU239','94Pu238','PU-240','Bi-194m','BI194M2','83-BI-194m2',...
# Names are not case sensitive, metastable states are m = m1, m2, m3
# Elements like 'U','Al','Ni' are converted into isotopes (according natural abundance fraction)

Mass: 10.0
Unit: 'w'

# Mass (total, grams) used only if Unit = 'w'
# Unit for nuclides set: mass fraction('w'), grams('g'), mol('mol'), atoms('atm'), becquerels('bq')

Periods: [['Period1', [3600], 1.0, 'MMPA'], ['Period2', [1.0e+3,1.0e+4,1.0e+5], 0.0, 'DIAG']]

# [['Identifier', [Duration(s),...], Flux multiplyer, 'Method'],...]
# 'Identifier' is used to identifiy output files on each period
# [Duration(s),...] are durations steps to compute in each period
# Flux multiplyer: 1.0 (full flux), 0.0 (zero flux, pure decay), 0.5 (50% flux)
# 'Method' are: 'MMPA', 'BDF' or 'DIAG'

Depth: 6

# Computation depth into chains: 6 to 12 recommanded (more depth = more nuclides = more computation time)

MULTILLABY¶

Multi-compartment model dispersion and transmutation.

Purpose¶

MULTILLABY is a multi-compartment model for nuclide dispersion with the following physical phenomenon:

mixing:	in a compartment volume
transfert:	from one compartment to another compartment
deposition:	on a compartment walls or inside a filter
filtration:	from one compartment to another
release:	from walls of a compartment or from a filter into another compartment
decay:	between isotopes individualized with one or several chemical form

Isotopes are individualized with one or several chemical form (example Iodine: I2, ICH3 or I aerosol). Parameters of each physical phenomenon can be set for each chemical form.

General Features¶

Small:	Code and nuclear data are small (only 200 Ko).
Fast:	Quantities and integral of quantities are computed directly from generalized H. Bateman equations.
User friendly:	A simple input instruction text file is needed.
Limits:	The choice of the algorithm enables maximal speed and precision but loops are not possible (example: recirculation of air in a loop from a compartment back to a previous compartment is not computed).

HARMATTILL¶

HARMATTIL¶

Atmospheric plume dispersion and radiological consequences