Computer Programs

NAME, COMPUTER, PROBLEM, SOLUTION, RESTRICTIONS, CPU, STATUS, REFERENCES, REQUIREMENTS, LANGUAGE, OPERATING SYSTEM, OTHER RESTRICTIONS, AUTHOR, MATERIAL, CATEGORIES

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Program name | Package id | Status | Status date |
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MORET 5.D.1 | NEA-1905/01 | Tested | 15-MAY-2019 |

Machines used:

Package ID | Orig. computer | Test computer |
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NEA-1905/01 | Linux-based PC | Linux-based PC |

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3. DESCRIPTION OF PROGRAM OR FUNCTION

Initially designed to perform calculations to support criticality safety assessments, the MORET code is a Monte Carlo simulation tool that solves the transport equation for neutrons. From its inception, the code has been constructed with the aim of providing a great flexibility to implement new components or methods without any difficulties. The result is a code that allows users to model complex three-dimensional geometrical configurations, use various evaluations and treatments for nuclear data to describe the materials, select the best adapted simulation method (from all methods available) related to their problem, define their own tallies and analyse the results.

Geometry modelling

The description of the geometry relies on combinatorial geometry for which user has to create volumes that will contain the materials. Operators may be added to describe the combinatorial properties of the volumes. The code gives the possibility of defining the geometry as a composition of several regions. This modular geometry allows to model complex geometries by using basic building blocks, called "modules". These parts of the whole geometry can be embedded one in another using "holes". A geometry-plotting function is available to display the modeling.

Description of materials

The MORET code allows two calculation routes for the treatment of the cross sections energy dependence: multi-group approach (with direct use of macroscopic cross section sets, which result from preliminary cell code calculations), and continuous energy calculations using the nuclear data libraries included in this distribution.

Simulation and neutron tracking

The Monte Carlo simulation consists of simulating a number of individual neutrons by reproducing as accurately as possible their elementary behaviour. It is an iterative process, following neutrons from their birth until death (whatever the cause). For each cycle, the new neutron distribution is defined based on fissions generated in the previous one. The MORET code embeds several techniques for tracking (such as the Woodcock method) and sampling source neutrons among potential fission sites to address two concerns: force neutrons to visit all fissile volumes to reduce the risk of weak coupling, and accelerate the source convergence. The sampling methods are the analogue method, the stratified sampling, the fission matrix method, the importance method, the oversampling method, the superhistory powering and the Wielandt method.

Nuclear data

The nuclear data distribution consists in continuous-energy nuclear data libraries in ACE format. These nuclear data libraries were all generated using NJOY 99.259 using the same reconstruction tolerances (0.1 0 %).

These nuclear data libraries are based on JEFF 3.1.1 and ENDF/B-VII.0 nuclear data evaluations and are available at various temperatures: 293.6, 300, 600, 900, 1200, 1500 and 1800 K.

The JEFF 3.1.1 and ENDF/B-VII.0 data libraries also have S(α,β) ACE files associated to them for a number of materials: H in H20, H in CH2, Be metal, Be in BeO and graphite.

The following nuclides and elements are available (for the given evaluation source):

JEFF 3.1.1:

H1, H2, H3, HE3, HE4, LI6, LI7, BE9, B10, B11, C, N14, N15, O16, O17, F19, NA22, NA23, MG24, MG25, MG26, AL27, SI28, SI29, SI30, P31, S32, S33, S34, S36, CL35, CL37, AR36, AR38, AR40, K39, K40, K41, CA40, CA42, CA43, CA44, CA46, CA48, SC45, TI46, TI47, TI48, TI49, TI50, V, CR50, CR52, CR53, CR54, MN55, FE54, FE56, FE57, FE58, C058, C058M, C059, NI58, NI59, NI60, NI6I, NI62, NI64, CU63, CU65, ZN, GA, GE70, GE72, GE73, GE74, GE76, AS75, SE74, SE76, SE77, SE78, SE79, SE80, SE82, BR79, BR81, KR78, KR80, KR82, KR83, KR84, KR85, KR86, RB85, RB86, RB87, SR84, SR86, SR87, SR88, SR89, SR90, Y89, Y90, Y91, ZR90, ZR91, ZR92, ZR93, ZR94, ZR95, ZR96, NB93, NB94, NB95, MO92, MO94, MO95, MO96, MO97, MO98, MO99, MO100, TC99, RU96, RU98, RU99, RU100, RU101, RU102, RU103, RU104, RU105, RU106, RH103, RH105, PD102, PD104, PD105, PD106, PD107, PD108, PD110, AG107, AG109, AG110M, AG111, CD106, CD108, CD110, CD111, CD112, CD113, CD114, CD115M, CD116, IN113, IN115, SN112, SN114, SN115, SN116, SN117, SN118, SN119, SN120, SN122, SN123, SN124, SN125, SN126, SB121, SB123, SB124, SB125, SB126, TE120, TE122, TE123, TE124, TE125, TE126, TE127M, TE128, TE129M, TE130, TE132, I127, I129, I130, I131, I135, XE124, XE126, XE128, XE129, XE130, XE131, XE132, XE133, XE134, XE135, XE136, CS133, CS134, CS135, CS136, CS137, BA130, BA132, BA134, BA135, BA136, BA137, BA138, BA140, LA138, LA139, LA140, CE140, CE141, CE142, CE143, CE144, PR141, PR142, PR143, ND142, ND143, ND144, ND145, ND146, ND147, ND148, ND150, PM147, PM148, PM148M, PM149, PM151, SM144, SM147, SM148, SM149, SM150, SM151, SM152, SM153, SM154, EU151, EU152, EU153, EU154, EU155, EU156, EU157, GD152, GD154, GD155, GD156, GD157, GD158, GD160, TB159, TB160, DY160, DY161, DY162, DY163, DY164, HO165, ER162, ER164, ER166, ER167, ER168, ER170, LU175, LU176, HF174, HF176, HF177, HF178, HF179, HF180, TA181, TA182, W182, W183, W184, W186, RE185, RE187, OS, IR191, IR193, PT, AU197, HG196, HG198, HG199, HG200, HG201, HG202, HG204, TL, PB204, PB206, PB207, PB208, BI209, RA223, RA224, RA225, RA226, AC225, AC226, AC227, TH227, TH228, TH229, TH230, TH232, TH233, TH234, PA231, PA232, PA233, U232, U233, U234, U235, U236, U237, U238, NP235, NP236, NP237, NP238, NP239, PU236, PU237, PU238, PU239, PU240, PU241, PU242, PU243, PU244, PU246, AM241, AM242, AM242M, AM243, AM244, AM244M, CM240, CM241, CM242, CM243, CM244, CM245, CM246, CM247, CM248, CM249, CM250, BK247, BK249, BK250, CF249, CF250, CF251, CF252, CF254,ES253,ES254,ES255,FM255.

ENDF/B-VII.0:

H1, H2, H3, HE3, HE4, LI6, LI7, BE7, BE9, B10, B11, C, N14, N15, O16, O17, F19, NA22, NA23, MG24, MG25, MG26, AL27, S128, S129, S130, P31, S32, S33, S34, S36, CL35, CL37, AR36, AR38, AR40, K39, K40, K41, CA40, CA42, CA43, CA44, CA46, CA48, SC45, TI46, TI47, TI48, TI49, TI50, V, CR50, CR52, CR53, CR54, MN55, FE54, FE56, FE57, FE58, C058, C058M, C059, NI58, NI59, NI60, NI6I, NI62, NI64, CU63, CU65, ZN, GA69, GA71, GE70, GE72, GE73, GE74, GE76, AS74, AS75, SE74, SE76, SE77, SE78, SE79, SE80, SE82, BR79, BR81, KR78, KR80, KR82, KR83, KR84, KR85, KR86, RB85, RB86, RB87, SR84, SR86, SR87, SR88, SR89, SR90, Y89, Y90, Y91, ZR90, ZR91, ZR92, ZR93, ZR94, ZR95, ZR96, NB93, NB94, NB95, MO92, MO94, MO95, MO96, MO97, MO98, MO99, MO100, TC99, RU96, RU98, RU99, RU100, RU101, RU1022, RU103, RU104, RU105, RU106, RH103, RH105, PD102, PD104, PD105, PD106, PD107, PD108, PD110, AG107, AG109, AG110M, AG111, CD106, CD108, CD110, CD111, CD112, CD113, CD114, CD115M, CD116, IN113, IN115, SN112, SN113, SN114, SN115, SN116, SN117, SN118, SN119, SN120, SN122, SN123, SN124, SN125, SN126, SB121, SB123, SB124, SB125, SB126, TE120, TE122, TE123, TE124, TE125, TE126, TE127M, TE128, TE129M, TE130, TE132, I127, I129, I130, I131, I135, XE123, XE124, XE126, XE128, XE129, XE130, XE131, XE132, XE133, XE134, XE135, XE136, CS133, CS134, CS135, CS136, CS137, BA130, BA132, BA133, BA134, BA135, BA136, BA137, BA138, BA140, LA138, LA139, LA140, CE136, CE138, CE139, CE140, CE141, CE142, CE143, CE144, PR141, PR142, PR143, ND142, ND143, ND144, ND145, ND146, ND147, ND148, ND150, PM147, PM148, PM148M, PM149, PM151, SM144, SM147, SM148, SM149, SM150, SM151, SM152, SM153, SM154, EU151, EU152, EU153, EU154, EU155, EU156, EU157, GD152, GD153, GD154, GD155, GD156, GD157, GD158, GD160, TB159, TB160, DY156, DY158, DY160, DY161, DY162, DY163, DY164, HO165, HO166M, ER162, ER164, ER166, ER167, ER168, ER170, LU175, LU176, HF174, HF176, HF177, HF178, HF179, HF180, TA181, TA182, W182, W183, W184, W186, RE185, RE187, IR191, IR193, AU197, HG196, HG198, HG199, HG200, HG201, HG202, HG204, PB204, PB206, PB207, PB208, BI209, RA223, RA224, RA225, RA226, AC225, AC226, AC227, TH227, TH228, TH229, TH230, TH232, TH233, TH234, PA231, PA232, PA233, U232, U233, U234, U235, U236, U237, U238, U239, U240, U241, NP235, NP236, NP237, NP238, NP239, PU236, PU237, PU238, PU239, PU240, PU241, PU242, PU243, PU244, PU246, AM241, AM242, AM242M, AM243, AM244, AM244M, CM241, CM242, CM243, CM244, CM245, CM246, CM247, CM248, CM249, CM250, BK249, BK250, CF249, CF250, CF251, CF252, CF253, CF254, ES253, ES254, ES255, FM255

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6. TYPICAL RUNNING TIME

The calculation time depends on the complexity and required precision for a given case. It depends on the complexity of geometry modelling, the number of material compositions, and the number of particles to be simulated. For a very precise solution, a few hours of computation may be necessary, depending on the speed of the machine.

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10. REFERENCES

Background documentation:

Alexis Jinaphanh, Nicolas Leclaire, Bertrand Cochet: "Continuous-energy sensitivity coefficients in the MORET code", Nuclear Science and Engineering, 184, pp 53-68 (2016)

N. Leclaire, B. Cochet, F.X. Le Dauphin, W. Haeck, 0. Jacquet: "Use of the ETA-1 reactor for the validation of the multi-group APOLL02-MORET 5 code and the Monte Carlo continuous energy MORET 5 code", Annals of Nuclear Energy, 76, pp 530-539 (2015}

B. Cochet, A. Jinaphanh, L. Heulers, 0. Jacquet: "Capabilities overview of the MORET 5 Monte Carlo code", Annals of Nuclear Energy, 82, pp 74-84 (2015)

NEA-1905/01, included references:

- B. Cochet and A. Jinaphanh:"MORET User's Manual- Version 5.D.1 ", IRSN PSN-EXP/SNC/2017-282, Institut de

Radioprotection et de Surete Nucleaire, France (20 17)

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Package ID | Computer language |
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NEA-1905/01 | C-LANGUAGE, FORTRAN |

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NEA-1905/01

Executables for Linux systems (no sources are included) Nuclear data in ENDF and ACE format: 2 libraries based on JEFF 3.1.1 and

ENDF/B-VII.O evaluations

Test cases

Scripts

Installation procedure (README files)

Electronic documentation

Keywords: Monte Carlo simulation, continuous energy calculations, criticality safety, neutron 3D geometry modelling, particle transport, reactor physics.