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Best Estimate Plus Uncertainty (BEPU) Technology: NINE is a world-wide leading organization about BEPU both for the development of the methodologies and for the application in licensing framework. This is also confirmed by the BEPU conferences that are organized by NINE every 2 years. In BEPU-2018 more than 300 people attended the conference in Lucca and in 2024 a new event is going to take place in Sicily, BEPU-2024, with an expected attendance larger than in 2018.

  

         

Multi-Physics and Multi-Scale Technology: NINE is one of the organizations which is very active in developing and deploying new technologies for carrying-out safety analysis. NINE is actively involved in the OECD/NEA/NSC “Expert Group on Multi-physics Experimental Data, Benchmarks and Validation (EGMPEBV)” and it leads the first of the kind international benchmark devoted to Multiphysics: the MPCMIV (Multi-physics, Pellet Clad Mechanical Interaction Validation)                                                                                            
       

Management and Preservation of Experimental and Simulation Data: NINE is supporting the International Experimental Thermal-Hydraulic Safety Experiments (TIETHYS) through the application of its own methodology for collecting and organizing the experimental database (SCCRED, Standardized and Consolidated Calculated and Reference/Experimental Database)

 

    Multi-Level Training Courses (MLTC): NINE is world-wide recognized as one of the leading organizations for training. Every year NINE offers three platforms for training activities devoted to code users (NRSHOT), to safety analysists (MMARS) and to managers and licensing specialists (SUNBEAM)
       
Development and Conduction of Competence Building Program for Embarking Countries (NUCOMBIP): NINE experience is based upon the NINE’s personnel support to IAEA SAET (Safety Assessment and Education Training) since 2009 for Poland, Vietnam, Malaysia, Jordan, Bangladesh, Bulgaria, Romania and UAE. In addition, since 2014 NINE’s personnel organizes twice a year “Inspector and Plant Walkdown” trainings under the guidance of US NRC inspectors at abandoned NPP at Zwentendorf, Austria.      


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BEST ESTIMATE PLUS UNCERTAINTY ANALYSIS OF METAL-WATER REACTION TRANSIENT EXPERIMENT

Alan Matias Avelar a, Camila Diniz a, Fábio de Camargo b, Claudia Giovedi c, Alfredo Abe c, Marco Cherubini d, Alessandro Petruzzi d, Marcelo Breda Mourão a

 

a Department of Metallurgical and Materials Engineering, University of São Paulo, Professor Mello Moraes, 2463 São Paulo, SP, Brazil
b Amazônia Azul Tecnologias de Defesa S.A., Corifeu de Azevedo Marques, 1847 São Paulo, SP, Brazil
c Nuclear and Energy Research Institute, University of São Paulo, Professor Lineu Prestes, 2242 São Paulo, SP, Brazil
d Nuclear and Industrial Engineering, Via della Chiesa XXXII 759, 55100 Lucca, Italy

Nuclear Engineering and Design, Volume 411, September 2023, 112414

Abstract — Uncertainty analysis is applied in the licensing process for nuclear installations to complement best estimate analysis and to verify that the upper bound value is less than the threshold corresponding to the safety parameter of interest. Metal-water reaction is a critical safety phenomenon of water-cooled nuclear reactors at accident conditions, e.g. Loss-Of-Coolant Accidents (LOCA). AISI 348 cladding is able to increase the accident tolerance comparing to Zr-based alloys and differently from other accident tolerant fuel cladding options, there is operational experience of nuclear power plants with stainless steel. In this study, a transient oxidation experiment of AISI 348 by steam was conducted and the major sources of uncertainty were addressed. An evaluation model was developed to calculate the evolution of mass gain during the experiment. Meanwhile, uncertainty propagation of experimental data was performed. The results show that the mass gain predicted by the transient metal-water reaction model lays within the experimental data uncertainty band. Furthermore, the selection of the oxidation kinetics model seems to be important whether the analysis wills to provide conservative results.


 
 
 

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TOWARDS MODELLING DEFECTIVE FUEL RODS IN TRANSURANUS: BENCHMARK AND ASSESSMENT OF GASEOUS AND VOLATILE RADIOACTIVE FISSION PRODUCT RELEASE

L. Giaccardia, M. Cherubinia, G. Zullob, D. Pizzocrib, A. Magnib, L. Luzzib

a Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII 759, 55100, Lucca, Italy

b Politecnico di Milano, Department of Energy, Nuclear Engineering Division, Via La Masa 34, 20156, Milan, Italy

ANNALS OF NUCLEAR ENERGY, volume 197, March 2024, Article number 110249

Abstract —This work presents the results of a collaborative benchmark activity between different organizations towards the use of the TRANSURANUS code to estimate the release of gaseous and volatile radioactive fission products from defective fuel rods, into the primary coolant of pressurized water reactors. First, the radioactive release from the fuel to the gap is evaluated according to three approaches: the coupling between TRANSURANUS and SCIANTIX, the development of TRANSURANUS devoted subroutines, and the use of the ANS 5.4-2010 methodology. Fuel-to-gap release calculations are benchmarked and assessed against measured data from the CONTACT1 irradiation experiment. Then, TRANSURANUS has been used to estimate the radioactive release into the primary coolant by applying a first-order phenomenological rate theory and tested against measured data of fission product coolant concentrations from irradiation experiments of the CRUSIFON program.


 
  

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PRELIMINARY SAFETY ANALYSIS AT THE DECOMMISSIONING OF THE WWR-M RESEARCH REACTOR

Yu. M. Lobach1, S. Yu. Lobach2, V. M. Shevel1

 

1 Institute for Nuclear Research, National Academy of Sciences of Ukraine, Kyiv, Ukraine

2 Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy

АТОМНА ЕНЕРГЕТИКА ATOMIC ENERGY, Volume 110609, March 2020

Abstract — Following the demands established by the current Ukrainian legislation, the Decommissioning Concept for the WWR-M research reactor was recently approved. The Concept envisages a strategy of immediate dismantling; it identifies and justifies the main technical and organizational measures for the preparation and implementation of decommissioning, the sequence of planned works and activities, as well as the necessary conditions and infrastructure. Decommissioning requires proper planning and demonstration that all planned dismantling works will be carried out safely. Presented safety assessment is a mandatory component of the Concept and the most important element of the overarching technological scheme. The purpose of the safety analysis is to provide input for detailed planning on how to ensure safety during decommissioning. Based on the results of the safety analysis, the measures to ensure radiation protection are defined while justifying their necessity and sufficiency.


 
 

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ASSESSMENT OF THE DOSE LOADDURING THE DISMANTLING OF THE WWR-M REACTOR

Yu. M. Lobach1, S. Yu. Lobach2, E. D. Luferenko1, V. M. Shevel1

 

1 Institute for Nuclear Research, National Academy of Sciences of Ukraine, Kyiv, Ukraine

2 Nuclear and INdustrial Engineering (NINE), Via della Chiesa XXXIII, 759, Lucca, Italy

ЯДЕРНА ФІЗИКА ТА ЕНЕРГЕТИКА / NUCL. PHYS. AT. ENERGY 23 (2022) 234-244

Abstract — The WWR-M is a light-water-cooled and moderated heterogeneous research reactor with a thermal output of 10 MW.The final decommissioning planning is in progress now. The general decommissioning strategy consists of the dismantling and separate removal of the bulky elements as a whole (in one piece) without preliminary segmentation. The dismantling of the primary and secondary cooling loops is considered as one of the key tasks; a separate dismantling design has been developed. The baseline principles for the technical solution and safety are presented in the given paper. Results of the dose assessment showed that the work can be performed at a collective dose of less than 20 man-mSv.Keywords: WWR type research reactor, decommissioning, cooling loops, dismantling, exposure dose.

 

 

NINE ha implementato una serie di interventi di adeguamento della propria infrastruttura informatica per una più efficiente ed efficace gestione e protezione dei dati, sia con investimenti in hardware che in software. Allo stesso tempo sono stati sostenuti costi per far fronte alle nuove esigenze lavorative causate dall’emergenza epidemiologica COVID-19, inclusi arredi da ufficio.

 

Importo investimento € 50.841,00

Importo contributo concesso: € 25.670,50

N.IN.E. Srl


Operazione/progetto co-finanziato/finanziato dal
POR FESR Toscana 2014-2020”

 

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Studsvik R2 Materials Test Reactor Ad Hoc Depletion Strategy for the Derivation of the Fuel Isotopic Composition of the MPCMIV Benchmark

L. Giaccardi a, S. Di Pasquale a, S. Dulla b, M. Cherubini a and A. Petruzzi a

aNuclear and Industrial Engineering (NINE), Via della Chiesa XXXII 759, 55100 Lucca, Italy

bPolitecnico di Torino, Dipartimento Energia, NEMO group, Corso Duca degli Abruzzi 24, 10129 Torino, Italy

International Conference on Physics of Reactors 2022 (PHYSOR 2022)
Pittsburgh (PA), USA, May 15–20, 2022

Abstract — The Ad Hoc Depletion Strategy elaborated by the NINE company, developed in support of the organization of the MPCMIV (Multi-physics Pellet Cladding Mechanical Interaction Validation) benchmark input and output specifications, is presented. This work aims at illustrating the strategy itself and then showing the results obtained with its application over the Studsvik R2 Testing Reactor, which is analyzed in the benchmark. The objective of the application of the strategy is to compute the fuel elements isotopic compositions at the beginning of some core loadings of interest for the benchmark. To this objective, it is necessary to implement first the simulation model of the three single assembly types and perform the infinite lattice depletions, then, to build the full core model and to perform the simulation of the core cycle. All the models and simulations were carried out with the use of the Monte Carlo particle transport code Serpent 2. Finally, the simulations results are assessed against Studsvik isotopic compositions of the fuel elements discharged from the R2 Reactor at the end of the core loading. Several assumptions were necessary during all the steps of the strategy, to overcome the lack of information regarding the core management. For this reason, the solution found at the end of the current analysis may not be completely optimized and further improvements regarding the model assumptions will be tested in a future work.

KEYWORDS: MPCMIV, Serpent 2, infinite lattice depletion, Core cycle, R2 Testing Reactor


Analysis of the Reactivity Effects Exercises of the Neutronics Benchmark of the CEFR Start-Up Tests

S. Di Pasquale a, M. Cherubini a, A. Petruzzi b and V. Giusti c

aNuclear and Industrial Engineering (NINE), Via della Chiesa XXXII 759, 55100 Lucca, Italy
bDipartimento di Ingegneria Civile ed Industriale (DICI), Università di Pisa, Italy, Largo Lucio Lazzarino (accanto all’edificio C), 56122 Pisa, Italy

International Conference on Physics of Reactors 2022 (PHYSOR 2022)
Pittsburgh (PA), USA, May 15–20, 2022

Abstract — The “Neutronics Benchmark of the CEFR Start-Up Tests” is an IAEA coordinated research project based on the simulation of the CEFR start-up tests. The main goals of the project are to improve the participant capabilities in SFR analysis and to perform an international validation of codes for Sodium Fast Reactor simulation. NINE-UNIPI work together on the creation of the Serpent 2 model and on the simulation of all the start-up tests proposed in the benchmark. In this work the three experiments related to the reactivity measurements are discussed. The geometry model is briefly described and the simulation set-up is presented. In particular, the geometry has been modeled considering the thermal expansion at the experimental temperatures. The nuclear data libraries used are the ENDF/B-VIII.0, pre-processed at the experimental temperatures and provided to the benchmark participants from SCK-CEN. The obtained results show a good agreement with the experimental data, except for the assembly-swap reactivity effect, which shows a small shift for all the considered cases. The results presented in this work could contribute to the validation of Serpent 2 for SFR criticality calculations.

KEYWORDS: CEFR. Serpent 2, SFR, Start-Up Tests, Validation


Thermal-Hydraulics Analysis of the IAEA CRP FFTF LOFWOS Test #13

Domenico De Luca, Kaiyue Zeng, Marco Cherubini, Alessandro Petruzzi

Nuclear and Industrial Engineering (NINE), Via della Chiesa XXXII 759, 55100 Lucca, Italy

HND2022 - 13th International Conference of the Croatian Nuclear Society
Zadar, Croatia, June 5 – 8, 2022

Abstract — Global interest in fast reactors has been growing since their inception in 1960 because they can provide efficient, safe and sustainable energy. Their closed fuel cycle can support long-term nuclear power development as part of the world’s future energy mix and decrease the burden of nuclear waste. Within this framework, the IAEA organized a Coordinated Research Projects (CRP) on FFTF Loss of Flow Without Scram (LOFWOS) Test #13, aimed at improving Member States’ fast reactor analytical simulation capabilities, international validation, and qualification of codes currently employed in the field of fast reactor. The Fast Flux Test Facility (FFTF) was a 400 MW thermal powered, oxide-fueled, liquid sodium cooled test reactor built to assist development and testing of advanced fuels and materials for fast breeder reactors. The present paper shows the work performed by NINE for the CRP focused on benchmark analysis of one of the unprotected passive safety demonstration tests performed at the FFTF. In particular, a detailed nodalization was developed following the NEMM (NINE Evaluation Model Methodology) already applied for LWR safety analysis. After achievement of acceptable steady-state results, transient analysis was performed. In addition, the NINE validation procedure was adopted in order to validate the Simulation Model (SM) against the experimental data. Two system thermal-hydraulic codes, namely RELAP5 and TRACE, were used to analyse the selected test and the comparison between the two SM results is also presented in this paper. The final goal of the activity is to present the main outcomes achieved through the use of codes currently employed in the field of fast reactor, and how the application of the NEMM procedures allows to develop and qualify the SM results and validate the computer codes against experimental data.


MELCOR-To-MELCOR Coupling Method in Severe Accident Analysis Involving Core and Pent Fuel Pool

Hector Lopez, Alessandro Petruzzi, Walter Giannotti, Domenico De Luca

Nuclear and Industrial Engineering (NINE), Via della Chiesa XXXII 759, 55100 Lucca, Italy

HND2022 - 13th International Conference of the Croatian Nuclear Society
Zadar, Croatia, June 5 – 8, 2022

Abstract — A lot of effort has been spent to prevent the occurrence of SA in nuclear plant and to develop Severe Accidents (SA) Management to mitigate the consequences of a SA. Those consequences are mainly related to limit the release of fission product to the environment. The core in the vessel is not the only source of fission products as the Spent Fuel Pool (SFP) hosting the fuel removed by the core is, in some NPP, inside the containment and SA conditions can also occur. This is especially important in reactors having proximity between the RPV and SFP such as the VVER-1200. This close proximity implies that any SA occurring in the SFP potentially affects the RPV and vice-versa. This potential combination might cause unexpected evolution in the SA progression to whom the safety systems are not able to contain. MELCOR code is a widely used, flexible powerful SA code but it is incapable (due to the uniqueness of the COR package use inside the same input) to reproduce a situation in which both the fuel in vessel core and the fuel in the SFP, inside the same containment, are going to experience a severe accident scenario. The current study presents a MELCOR-to-MELCOR coupling method to simulate simultaneously scenarios with both, core and SFP, as sources capable of H2 generation, fuel damage and FP release in a VVER-1200 NPP. The coupling is performed by running two simulations in parallel and with the data exchange supervised and managed by a dedicated Python coupling supervising script developed at NINE.


 

Reactor Physics and Thermal Hydraulics Analyses for the OECD/NEA MPCMIV Benchmark

Luana Giaccardi, Domenico De Luca, Simone Di Pasquale, Marco Cherubini,Alessandro Petruzzi

Nuclear and Industrial Engineering (NINE), Via della Chiesa XXXII 759, 55100 Lucca, Italy

HND2022 - 13th International Conference of the Croatian Nuclear Society
Zadar, Croatia, June 5 – 8, 2022

Abstract — In order to complete the Multi-physics Pellet Cladding Mechanical Interaction Validation (MPCMIV) benchmark technical specifications, reactor physic and thermal hydraulic analyses have been performed. The work presented in this paper aims in particular to evaluate some of the missing Boundary and Initial Conditions necessary to complete the technical specifications, and also to perform some of the benchmark exercises connected with thermal hydraulic simulations. A far as the thermal hydraulic area is concerned, the analysis is carried out with the RELAP5 code. It is focused on the modelling of the in pile loop 1 located inside the R2 reactor core, in which a test fuel rodlet is inserted to perform some power ramp tests. The activity consists in the development of the simulation model of the in pile tube, the demonstration of the steady state achievement and the transient analysis of the first selected test, validating the simulation results against the benchmark experimental data. Considering the reactor physic area, the Monte Carlo code Serpent 2 is used to perform some single assemblies burn up calculations. The aim is to evaluate the initial composition of the fuel assemblies loaded in the core loadings of interest of the benchmark. Moreover, the temperature values to be used in the Serpent simulations are derived with thermal hydraulic simulations of the single assemblies. Further developments of the work will include the full core cycle analysis to validate the isotopic compositions and the complete model of the main circuit, using the gamma heating from the reactor physics calculations. Finally the TRANSURANUS fuel performance code will be adopted to compare the results against the available experimental data. A multi-physics effort is required to carry out the MPCMIV benchmark and appropriate coupling approach will be investigated and tested against the benchmark experimental results.


Simulation of the OECD/NEA Rod Bundle Heat Transfer (RBHT) Benchmark with RELAP5

Alessandro Del Ferraro, Domenico De Luca, Marco Cherubini, Alessandro Petruzzi

Nuclear and Industrial Engineering (NINE), Via della Chiesa XXXII 759, 55100 Lucca, Italy

HND2022 - 13th International Conference of the Croatian Nuclear Society
Zadar, Croatia, June 5 – 8, 2022

Abstract — The OECD/NEA RBHT (Rod Bundle Heat Transfer) Project is an International three-year NEA Joint Project whose objective is to conduct new experiments and evaluate system hydraulics and sub-channel codes in the simulation of reflood tests. Such tests are performed in a full height rod bundle facility equipped with advanced instrumentations capable to measure the real-time droplet field, cladding and steam/fluid temperatures, water carryover fraction and pressure drops. The test matrix encompasses both steady and oscillatory reflood inlet flow conditions. Within the RBHT project, a challenging benchmark exercise is conducted, including an open and a blind test phase providing a unique opportunity to project’s participants to validate codes and nodalization techniques. This paper presents a validation study of the RELAP5 code on the RBHT open test series. The simulations’ results generally well agree with the measured data, according to the accuracy metrics proposed by the benchmark team. A larger discrepancy is detected for experimental tests characterized by higher flooding rates with low subcooling degree. Several model’s parameters have been investigated including also different nodalization schemes to characterize the impact on the predicted results during the sensitivity analysis.


 

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