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One of the main objectives of the reactor physics analysis is to calculate the power generated and its axial and radial distribution within the whole core, taking the fuel depletion into account. These investigations are inherently three-dimensional and involve more than a single calculation methodology, hence more than one code, to model stationary and time-dependent neutron-physical phenomena, to perform criticality analyses and to carry out safety-related assessment of core design . The deterministic analyses require managing the cross-sections as a preliminary step: “continuous” energy spectrum data are processed (e.g. by NJOY) to obtain “discrete” libraries (hundreds of groups); the latter are then fed to lattice physics codes  to create two- or few-group cross-sections data sets through homogenization procedures. The so-obtained cross-sections are then used in nodal diffusion codes, possibly coupled to TH codes to provide time-dependent feedbacks. Statistical calculations, based on the Monte Carlo method, are also performed: they can either support deterministic evaluation or constitute standalone analyses. Core cycle evaluations, including the effects of chemical shim and burnable poisons, are also addressed by the reactor physics analysis. NINE’s experience includes the use of the following codes: SERPENT and MCNP for 2D and 3D Monte Carlo calculations; SCALE/TRITON as a lattice code to solve 2D transport equation for cross-sections generation; PARCS and NESTLE as nodal diffusion codes to perform 3D full-core simulations. All the mentioned tools are applied in industrial projects as well as within an International research context (e.g. OECD/NEA activities). The reactor physics calculations are then supported by the evaluation of the uncertainty by means of SCALE/TSUNAMI tool. NINE’s engineers are currently working also on the development of a novel approach for the uncertainty quantification of core physics simulations.


Computational Fluid Dynamics (CFD) IS a class of computer simulation tools that solve fluid mechanics balance equations (mass, momentum, energy and other transported quantities) to provide three-dimensional description of time and space distribution of the quantities that characterize fluid flows (local and instantaneous variables such as pressure, velocity, temperature, species concentration, turbulent parameters, etc., along with integral quantities such as pressure drops, heat transfer coefficients etc.) Most CFD codes are normally based on either finite-volume or hybrid finite-volume/finite-element methods and rely on parallel computing when complex, industrial-scale problems are involved.
NINE’s specialists have long-standing experience with general-purpose commercial CFD codes (such as ANSYS CFX, Fluent, Star-CCM+), as well as research or open-source CFD codes (such as Neptune_CFD, Trio_U, Code_Saturne, OpenFOAM).


Structural mechanics deals with the assessment of the behaviour of solid materials when subjected to mechanical and thermal loadings, and particularly aims at assessing stability, integrity, functionality and reliability of mechanical parts and components. The most advanced and accurate approach to structural mechanics relies on the Finite Element Analysis, i.e. on computer simulation tools, based on finite-element methods, which solve mechanical balance equations over complex 3D domains, accounting for realistic material properties, constitutive relations and closure models. The modelling capabilities range from simple static 1D linear elastic problems to complex transient 3D non-linear elastic-plastic problems and even modal methods for problems involving vibrations and seismic scenarios.
NINE’s specialists have experience with commercial finite-element codes such as ANSYS Mechanics/Multiphysics and open-source codes such as Code_Aster.


Nowadays best-estimate computer codes are routinely used for safety assessment purposes. Such codes are meant to provide realistic prediction of the system evolution during the investigated transients. However, a consistent use of the results of such powerful tools, especially for licensing purposes, requires the quantification of the related uncertainties, which in turn stem from the uncertainties on the input information (boundary and initial conditions, code models/correlations, nodalizations, etc.)
NINE’s engineers have a long-standing experience in the field of uncertainty quantification (UQ) of system thermal-hydraulic calculations and they can master the methods based on statistical propagation of input parameters as well as the method based on accuracy extrapolation.
The scope of NINE UQ-related activities extends to other classes of computer codes, such as CFD and fuel performance codes.
The Best-Estimate Plus Uncertainty (BEPU) approach has been adopted by NINE’s engineers to support the licensing of a PHWR.
NINE’s engineers participate in OECD/NEA expert groups dealing with benchmarking of UQ methodologies, both for system thermal-hydraulic and CFD codes, and is developing a new approach based on Bayesian techniques.


System thermal-hydraulic analysis addresses the simulation of complex thermal-hydraulics phenomena occurring during postulated accident scenarios in nuclear installations (power plants, research reactors, spent fuel pools and experimental facilities).
NINE’s engineers have long-standing experience and competence in applying best-estimate system thermal-hydraulic codes (such as RELAP5  and TRACE) to reproduce the complex nuclear systems with very accurate and detailed models, accounting for all the relevant phenomena and processes and even including the role of I&C systems. NINE is able to perform safety analyses by adopting either the “combined conservative” or the more advanced Best-Estimate Plus Uncertainty (BEPU) approach, the latter relying on proper uncertainty evaluation to complement the best-estimate code results. A robust and standardised methodology has been developed to support, document, review and qualify the results of a system thermal-hydraulic analysis.
Furthermore, coupling techniques are adopted to integrate system thermal-hydraulic, reactor physics, fuel behaviour and containment analyses within a multi-scale multi-physics framework.
The validation of the system thermal-hydraulic tools and of the associated coupling techniques is also addressed by activities that NINE’s engineers carry out in cooperation with OECD/NEA and IAEA expert groups.


The Fuel Behaviour analysis addresses the thermal-mechanical performance of the nuclear fuel, accounting for the strong modifications that its mechanical and physical properties undergo during the irradiation.
The numerous and complex phenomena involved can be predicted by means of suitable computational tools. Among the various computer codes adopted by the NINE’s specialists, TRANSURANUS was selected as the reference one to simulate the fuel performance under normal operation as well as accident conditions. Advanced and detailed studies are carried out by coupling the TRANSURANUS code with thermal-hydraulics and reactor physics analysis tools to allow realistic core behaviour simulation. Best Estimate analyses are then supported by uncertainty evaluation to quantify the safety margins.
Such coupling techniques have been exploited in the frame of both research programmes (e.g. RIA benchmarks conducted within the OECD/NEA/WGFS) and third-party contracts.


The containment constitutes the last barrier against the release of the Fission Products (FP). Knowing the containment behaviour during accidental scenarios is of outmost importance for the analysis for the source terms to the environment. Specific codes are used for this kind of analysis to reproduce the TH phenomena and the possible molten material concrete iteration occurring in the containment. Analyses of the structural integrity of the containment are performed to support the design of suitable safety measures and limit the radionuclides release. In addition, such analyses can involve the complex evaluation of FP distribution in the containment to derive the source term to the environment.


Severe Accident analysis consists in the simulation of the progression of an accident and of the relevant related  phenomena occurring in a nuclear facility causing the structural geometrical modification and the melting of the materials (fuel and structural) of the core, vessel and containment. NINE experts have experience in  the field of Severe Accident analyses, largely based on the WWER and PWR plant analysis and the use of the MELCOR code. NINE  performed SA analysis to evaluate the damage caused in the primary side (e.g. core melting, vessel failure),  the containment integrity evaluation (e.g. pressurization,  core-concrete interaction), the calculation of the source term (radionuclides release and distribution). NINE, to improve the quality and  extension of the Severe Accident analysis, has the capabilities to couple MELCOR with more detailed codes for the RCS simulation (e.g. RELAP), with structural analysis codes (ANSYS) and with codes for environment consequence evaluation (e.g. MACCS). NINE consolidated its capability to develop and to perform training activities related to the Severe Accident phenomenology and the use of MELCOR at every level of knowledge and for personnel and organization in the international framework (e.g. regulator bodies, IAEA activities). 


The Waste Management (WM) covers the activities from storage of spent fuel to the radwaste (RW) treatment and final disposal. The WM poses two main issues: the technological activities related to the best conditioning and treatment of the RW and the complex safety aspects related to the need of containing the radionuclides. The analysis covers the normal operation and the accident conditions because of the need to evaluate the radiological dose to the personnel and the population in all the conditions, accounting for the selected technological solutions for WM treatment and the adopted safety measures.

The activities performed in the presence of radioactive sources always pose the need of minimizing the radiological dose to the population and to the personnel. For the involved personnel this is obtained by the optimization of the work activities performed, the design of suitable shielding and the definition of specific monitoring and alarm systems. Concerning the population (surrounding the site where the radioactive sources is used), object of the analysis are the effectiveness of external shielding and, in case of radionuclides release, also the complex analysis of intake of the radionuclides, accounting for the characteristics of the area surrounding the sources, the atmospheric conditions and use of the ground.

Characterization of the radionuclides (RN) inventory in a nuclear installation is the first step of the analysis aimed at evaluating the radiological source term to the environment. The second step addresses distribution of the RN within the installation under consideration. This analysis requires the use of computer codes capable of simulating the dynamics of the RN, taking chemical, thermal-hydraulic and thermodynamic aspects into account for radioprotection purpose.


Excellent up-to-date in-house expertise in PSA methodology:

     ♦  NINE experts were principal developers of current IAEA Safety Standards in the area of PSA
        [Safety Guides on Level-1 (#SSG-3) and Level-2 PSA (#SSG-4) Development and Applications]

       NINE experts contributed to development of IAEA guidance on Level-3 PSA (currently suspended) and ANS/ASME PRA Standards

       NINE experts actively collaborate with IAEA in the area of new challenges related to PSA methodology,
         e.g. Coordinated Research Project (CRP) on Multi-Unit PSA (MUPSA).

Development of PSA models:

     In-house capabilities for Level-1 and Level-2 PSA

          ◊ All operational modes

           Internal IEs, internal and external hazards

           Practical experience in development of PSA models for WWER/PWR NPPs

     Possibilities for cooperation with reputable experts on Level-3 PSA.

 Review of PSA:
     NINE experts contributed to updating IAEA guidelines on PSA review  
     NINE experts have very good experience in conducting international  independent review of PSA studies,
        e.g. IAEA-led International Probabilistic Safety Assessment Review Team (IPSART) missions

           Different NPPs, including WWER plants (e.g. Bushehr, Armenian NPP)
          Various PSA scope

      NINE experts collaborate actively with IAEA , e.g. review of operational PSA applications in the framework
        of IAEA OSART Safety Service (May 2017, Krsko NPP, Slovenia)

R&D in the area of PSA:
     NINE experts contributed to post-Fukushima gap analysis related to  PSA methodology
     NINE experts contribute to development of PSA-based methods and tools for analysis of impact of extreme events
         Fault Sequence Analysis (FSA) method

Training on PSA:
     NINE experts have extensive experience with IAEA training approaches and training implementation
     NINE experts developed and delivered comprehensive training on PSA
         Theoretical (basic and advanced training)
         Practical (simulation-type) training  
         Case studies aimed at building practical skills on PSA (within IAEA COMPASS-M and COMPASS-J projects)