SMURE: Serpent - MCNP Utility for Reactor Evolution (see home page)
The main aim of the MURE/SMURE package is to perform nuclear reactor time-evolution using the widely-used particle transport Monte-Carlo codes MCNP or Serpent2. This tool is mainly developped at LPSC, in close collaboration with PACS group of IPN d’Orsay and Subatech Nantes.
Many depletion codes exist for determining time-dependent fuel composition and reaction rates. These codes are either based on solving Boltzman equation using deterministic methods or based on Monte-Carlo method for neutron transport. Among them, one has to cite MCNPX/CINDER 90, MONTEBURN, KENO/ORIGEN, MOCUP, MCB, VESTA/MORET, TRIPOLI-4D, Serpent, ... which provide neutron transport and depletion capabilities. However, the way to control (or interact with) the evolution are either limited to specific procedure and/or difficult to implement.
In (S)MURE, due to the Object-oriented programming, any user can define his own way to interact with evolution. From an academic point of view, it is also good to have lots of M-C evolution codes to compare and benchmark them to understand physics approximations of each one. Moreover, SMURE provides a simple graphical interface to visualize the results. It also provides a way to couple the neutronics (with or without fuel burn-up) and thermohydraulics using either an open source simple code developed in SMURE (BATH, Basic Approach of Thermal Hydraulics) or a sub-channel 3D code, COBRA-EN. But SMURE can also be used just as an interface to MCNP or Serpent to build geometries (e.g. for neutronics experiments simulation).
SMURE is based on C++ objects allowing a great flexibility in the use. There are 4 main parts in this library:
- Definition of the geometry, materials, neutron source, tallies, ...
- Construction of the nuclear tree, the network of links between neighbouring nuclei via radioactive decays and nuclear reactions.
- Evolution of some materials, by solving the corresponding Bateman's equations.
- Thermal-hydraulics: it couples neutronics, thermal-hydraulics and, if needed, fuel evolution.
- A robust diffusion method (Nodal Drift Method, NDM) for both spatial kinetics and burnup calculations (full core evolution with a precise refueling scheme).
- Part 1 can be used independently of the 3 others; it allows easy generation of Serpent/MCNP input files by providing a set of classes for describing complex geometries. The ability to make quick global changes to reactor component dimensions and the ability to create large lattices of similar components are two important features that can be implemented by the C++ interface. It should be noted that some knowledge of MCNP or Serpent is very useful in understanding the geometry generation philosophy.
- Part 2 builds the specific nuclear tree from an initial material composition (list of nuclei). The tree of each evolving nucleus is created by following the links between neighbours via radioactive decay and/or reactions until a self-consistent set of linked nuclei is extracted. Nuclei with half-lives very much shorter than the evolution time steps, could be removed from the tree; mothers and daughters of these removed nuclei are re-linked in the correct way. Part 2 can also be used independently of the other two parts to process cross-sections for MCNP/Serpent at the desired temperature.
- Part 3 simulates the evolution of the fuel within a given reactor over a time period of up to several years, by successive steps of Serpent/MCNP calculation and numerical integration of Bateman's equations. Each time the MC code is called, the reactor fuel composition will change due to the fission/capture/decay process occurring inside. Changes in geometry, temperature, external feeding or extraction during the evolution can also be taken into account. Obviously this part is not independent of the 2 others (see figure).
- Part 4 consists of coupling the Oak Ridge National Laboratory code COBRA-EN (COolant Boiling in Rod Arrays) with MURE. COBRA is a sub-channel code that allows steady-state and transient analysis of the coolant in rod arrays. The simulation of flow is based on a three or four partial differential equations : conservation of mass, energy and momentum vector for the water liquid/vapor mixture (optionally a fourth equation can be added which tracks the vapor mass separately). The heat transfer model is featured by a full boiling curve, comprising the basic heat transfer regimes : single phase forced convection, sub-cooled nucleate boiling, saturated nucleate boiling, transition and film boiling. Heat conduction in the fuel and the cladding is calculated using the balance equation.
- Part5, essential to core design studies, has been recently added. NDM has been originally conceived for the calculation of transients by spatial kinetics from representative diffusion data (previously computed and tabulated). After validation on a CANDU LOCA (cf. Nuttin et al., 2016), this method has been generalized for the calculation of a Rod Ejection Accident in a PWR (cf. Prévot et al., 2017 ). Used on the much longer times of fuel cycle studies, NDM allows also to evaluate the final burnup which depends on the way the core is refueled.
Comparative analysis of high conversion achievable in thorium-fueled slightly modified CANDU and PWR reactors
We have studied the conversion performance of thorium-fueled standard or only slightly modified CANDU and PWR reactors with unchanged core envelope and equipments, to be eventually used as the third and last tier of symbiotic scenarios. For instance, plutonium extracted from the spent fuel of UOX PWRs could be converted in Th/Pu CANDUs to uranium (mainly 233U), finally used to feed a thorium-fueled water-cooled high converting third component. This could be a convenient way to replace likely delayed Generation IV in the case of an important increase of uranium-based energy demand. In order to assess the competitiveness of such symbiotic scenarios, detailed burnup and conversion data are obtained by means of a core-equivalent simulation methodology developed for CANDU-6 and adapted to N4-type PWR.
Once-through cycles in CANDU are firstly evaluated for various Th/Pu and Th/233U fuels as regards detailed conversion and basic safety performance. Breeding in Th/233U CANDU is achieved for a 1.30 wt% homogeneous fissile enrichment and a relatively short burnup of 7 GWd/t. Small increase of enrichment (to 1.35 wt%) considerably extends cycle length (to 14 GWd/t) at the cost of slight sub-breeding. Heterogeneity of fissile load can bring another 70 % gain on burnup with no significant impact on conversion. Multirecycling gives even shorter burnup (about 5 GWd/t) for the breeding case, while performance close to the once-through 1.35 wt% case is obtained for a slightly sub-breeding regime sustained by a small add of uranium from Th/Pu CANDU. Th/U cycle neutronic analysis explains the convenient feature of almost constant burnup as 233U load is unchanged at each recycle. Two symbiotic scenarios based on UOX PWRs, Th/Pu CANDUs and Th/233U CANDUs in a first open version or optimized Th/U CANDUs in a second closed version are compared.
At standard power and moderation levels, Th/233U PWR conversion performance is much lower than CANDU with only a bit more than half of initial fissile load remaining after 50 GWd/t. Contrary to CANDU, fuel heterogeneity does not increase burnup. Conversion is mainly improved by enhanced sub-moderation down to minimal acceptable water over fuel volume ratio of 0.8 at standard power. In this limit case, a 3.00 wt% enrichment ensures a burnup of 33 GWd/t with 80 % of initial fissile load remaining. By comparing a few Th/233U CANDU and PWR high converting cases, we understand that main part of the CANDU-PWR conversion gap results from neutron-economical CANDU operation conditions based on frequent online refueling and therefore why sub-moderation improves PWR conversion. From this better understanding, we deduce and preliminarily evaluate two possible ways to really higher conversion with thorium fuel in PWR envelope based on faster spectra either with light water and power derating or with heavy water and Spectral Shift Control.
Development of new academic simulation tools within MURE for safety studies with coupled thermalhydraulics and spatial kinetics
Safety analysis of innovative reactor designs requires three dimensional modeling to ensure a sufficiently realistic description, starting from steady state. Actual Monte Carlo (MC) neutron transport codes are suitable candidates to simulate large complex geometries, with eventual innovative fuel. But if local values such as power densities over small regions are needed, reliable results get more difficult to obtain within an acceptable computation time. In this scope, NEA has proposed a performance test of full PWR core calculations based on Monte Carlo neutron transport, which we have used to define an optimal detail level for convergence of steady state coupled neutronics. Coupling between MCNP for neutronics and the subchannel code COBRA for thermal-hydraulics has been performed using the C++ tool MURE, developed for about ten years at LPSC and IPNO. In parallel with this study and within the same MURE framework, a simplified code of nodal kinetics based on two-group and few-point diffusion equations has been developed and validated on a typical CANDU LOCA. Methods for the computation of necessary diffusion data have been defined and applied to NU (Nat. U) and Th fuel CANDU after assembly evolutions by MURE. Simplicity of CANDU LOCA model has made possible a comparison of these two fuel behaviours during such a transient.
Les études de systèmes complexes impliquant des réactions nucléaires dans un milieu fluide ou solide avec des transferts d’énergie et de masse, des changements de phases et des réactions chimiques, s’appuient sur deux volets :
- des développements numériques seuls capables de prendre en compte la totalité des phénomènes physiques
- des développements expérimentaux conçus pour valider certains aspects des modèles numériques (modèles de turbulences, effets radiatifs…)
Ces activités scientifiques mettent en jeu des connaissances utilisables dans plusieurs types d’applications (réacteurs nucléaires, cibles de production de neutrons ou d’isotopes…) et sont regroupées sous l’acronyme M.E.N (Multiphysique Expérimentale et Numérique). Elles s'appuient sur une plateforme de recherche expérimentale dont l’acronyme est F.E.S.T (Fluids Experiments and Simulations in Temperature) qui regroupe à la fois les installations expérimentales en température ainsi que les maquettes en eau et les simulations numériques nécessaires à la definition du design des expériences
Différents composants de cette platerforme sont décrits dans les liens suivants :
- Boîtes_à Gants
- Boucle FFFER (en cours de mise à jour)
- Installation SWATH (en cours de mise à jour)
Pour ceux qui veulent découvrir ce qu'est un sel fondu, voici un lien utile : les Sel_Fondus, qu'est-ce que c'est ?
Permanent staff
- Bidaud Adrien -- Senior lecturer INPG -- 04 76 28 40 45
- Billebaud Annick -- Directrice de Recherche CNRS -- 04 76 28 40 57
- Capellan Nicolas -- Senior lecturer INPG -- 04 76 28 41 90
- Chabod Sébastien -- Chargé de Recherche CNRS-- 04 76 28 40 93
- Ghetta Véronique -- Chargée de Recherche CNRS -- 04 76 28 41 85
- Méplan Olivier -- Senior lecturer UGA -- 04 76 28 40 57 (Head of the team)
- Nuttin Alexis -- Senior lecturer INPG -- 04 76 28 41 90
This email address is being protected from spambots. You need JavaScript enabled to view it. -- Professor UGA -- 04 76 28 40 27- Rubiolo Pablo -- Professor INPG -- 04 76 28 40 68
- Sage Christophe -- Senior lecturer INPG -- 04 76 28 41 34
PhD students and post-docs
- Lopez Pamela -- PhD student -- 04 76 28 40 45
- Reygadas-tello Daniela -- PhD student --ILL/LPSC
This email address is being protected from spambots. You need JavaScript enabled to view it. -- PhD student -- (ILL/LPSC)This email address is being protected from spambots. You need JavaScript enabled to view it. -- PhD student -- 04 76 28 43 38This email address is being protected from spambots. You need JavaScript enabled to view it. -- PhD student -- 04 76 28 41 34- Jonas Narvaez -- PhD student -- cotutelle with the Polytechnic Institute of Milan(POLIMI)
Visitors
This email address is being protected from spambots. You need JavaScript enabled to view it. -- Chargé de Recherche CNRS -- IJCLab -- 04 76 28 40 47
Engineer / technicians
- M. Baylac, P. Boge, T. Cabanel, E. Froidefond, E. Labussière, S. Rey (pôle accélérateurs et sources d’ions)
- J. Giraud, J. Menu, S. Roudier, Y. Odièvre (Service Études et Réalisations Mécaniques)
- O. Guillaudin, M. Rousseau, S. Marcatili, J. Marpaud, J.F. Muraz, O. Zimmermann (Service Détecteurs et Instrumentation)
- J. Bouvier, D. Tourrès (service électronique)
- G. Dargaud (service informatique)
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Former Members
This email address is being protected from spambots. You need JavaScript enabled to view it. -- Engineer-Researcher -- CEA Cadarache /SPRC/LEPh
Former PhD Students and post-docs
- Mantulet Gabin
- Blanco Juan --
- Stolyarova Elena --
This email address is being protected from spambots. You need JavaScript enabled to view it. -- Milano Multiphysics, Founder, CEO -- Milano (Italie)This email address is being protected from spambots. You need JavaScript enabled to view it. -- IRSN Fontenay-aux-roses -- +33 (0)1 58 35 98 06This email address is being protected from spambots. You need JavaScript enabled to view it. -- Engineer-Researcher -- CEA Cadarache -- /SPRC/LEPhThis email address is being protected from spambots. You need JavaScript enabled to view it. -- IRSN Fontenay-aux-roses -- +33 (0)1 58 35 92 41This email address is being protected from spambots. You need JavaScript enabled to view it. -- Directeur de Recherche CNRS -- IJCLab -- +33 (0)1 69 15 69 52This email address is being protected from spambots. You need JavaScript enabled to view it. -- Chargé de Recherche CNRS -- IJCLab -- +33 (0)1 69 15 51 96This email address is being protected from spambots. You need JavaScript enabled to view it. -- EDF/UNIE/GECC Lyon --This email address is being protected from spambots. You need JavaScript enabled to view it. -- Ingénieur physique des réacteurs -- EDF - SEPTEN -- Villeurbanne -- +33 (0)4 72 82 78 51This email address is being protected from spambots. You need JavaScript enabled to view it. -- Chargée de Recherche CNRS -- IPHC Strasbourg -- +33 (0)3 88 10 62 81This email address is being protected from spambots. You need JavaScript enabled to view it. -- Post-doctorant -- EPFL --This email address is being protected from spambots. You need JavaScript enabled to view it. -- Chargé de Recherche CNRS -- CENBG -- +33 (0)5 57 12 08 87This email address is being protected from spambots. You need JavaScript enabled to view it. -- Engineer-Researcher -- CEA Grenoble -- +33 (0)4 38 78 93 59This email address is being protected from spambots. You need JavaScript enabled to view it. -- Engineer - IJCLab -- +33 (0)1 69 15 71 58This email address is being protected from spambots. You need JavaScript enabled to view it. -- Senior lecturer -- Ecole des Mines de Nantes/SUBATECH -- +33 (0)2 51 85 86 42
The reactor physics group works on the issue of innovative reactor development for future nuclear energy. Concerning nuclear data measurements, the group is at present involved in fission product yield measurements at the Lohengrin spectrometer at ILL. In the early 2000 years, the lab built a slowing-down time spectrometer coupled to a neutron source, creating the PEREN platform (Plateforme d'Etude et de Recherche sur l'Energie Nucléaire) which allowed in particular the access to first data on neutron elastic scattering on carbon and fluor, data of main interest for MSR (Molten Salt Reactor). This platform added a "molten salt" facility in 2005, in order to process fluoride materials, and is the place where bubbling experiments now take place (FFFER), a technique investigated for molten salt reactor as it allows the on-line cleaning of the salt by removing some fission products from the fuel. This new reactor core concept was the subject of many simulations which contributed to the definition of conditions for exploiting the thorium cycle with good safety and breeding abilities for different neutron energy spectra. The different ways to produce the fissile fuel required for starting breeder reactors, fed afterwards with fertile thorium or uranium 238 fuel (1 ton per GWe.year) were calculated and gave birth to deployment scenarios by the end of the century, taking into account the world situation of energy consumption. Other short term scenarios are also under study.
ADSs (Accelerators Driven Systems), studied for their possible ability to incinerate long-lived and high activity nuclear waste, are the subject of a demonstrator construction project (MYRRHA) led by the SCK-CEN (Belgium). Since the 90's our group works on the issue of on-line reactivity control of an ADS, a crucial issue for the system safety. It was one of the objectives of the MUSE program, completed in 2004, performed at the MASURCA reactor coupled to the pulsed neutron source GENEPI, designed and built by the LPSC. This project led to a new step in the validation of the methodolodgies we investigated, the GUINEVERE project, which consists in coupling the Belgian reactor VENUS (SCK-CEN), turned to a fast neutron reactor, with a new neutron source GENEPI-3C, allowing operation in both continuous and pulsed modes, designed and constructed by a CNRS/IN2P3 collaboration, coordinated by the LPSC. Started in the frame of the EUROTRANS-IP project, this program is still going on in the FREYA project (2011-2016), with additional objectives aiming at preparing the MYRRHA reactor construction.