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AZNHEX is a novel 3D neutron diffusion code for nuclear core analysis with hexagonal-z geometry, and it is part of the Mexican project on the development of domestic nuclear analysis software “AZTLAN Platform”. Currently, the code is under development but some important steps have been made in the verification procedure of the code. A composite nodal method for prismatic hexagons has been developed and applied to solve the neutron diffusion equations for cores built by the union, side by side, of hexagonal prisms. A standard finite element method is used starting by the strong form to get the corresponding weak form. The verification and validation process of the AZNHEX code started with some academic exercises. In the case of this paper, in order to move to much more realistic scenarios, the OECD/NEA UAM Sodium Fast Reactor Benchmark has been used together with the SERPENT code for further verification. One of the exercises of the OECD/NEA Sodium Fast Reactor Benchmark is focused on the neutronic characterization of global parameters ( k eff , sodium void worth, Doppler, etc.) and feedback coefficient evaluation. This Benchmark is intended to confirm the ability of participants and their neutronic codes to provide generally consistent results when analyzing SFR core characteristics and thus, it represents a very good exercise for the AZNHEX development team. The differences in k eff of AZNHEX versus the well validated deterministic codes DYN3D and PARCS are 70 pcm and 113 pcm respectively when using the same set of XS previously generated with the stochastic code SERPENT. Also, the radial power distribution compared between the deterministic codes over the main diagonal of the core presented very good agreement among them. In other exercises, differences in the order of 200 pcm, in the worst cases, were found when comparing integral parameters with SERPENT. The AZNHEX results give the confidence to keep developing the code in order to convert it into the standard domestic tool for hexagonal-z geometry core analysis. Future developments on the code will be focused on the implementation of discontinuity factors and the thermal expansion effects in an operating core as well as the time dependent implementation.

Venting timing and duration are key issues for the development and assessment of severe accident guidelines and mitigation alternatives. In BWRs, venting from wetwell has the advantage of gaining fission product scrubbing. In this study, two strategies are investigated to avoid hydrogen deflagration in venting pipelines. The starting point of the vent pipe is a penetration on the wall of wetwell’s suppression chamber of a BWR Mark II containment. A three-dimensional pipeline model was developed for the CFD type code GASFLOW, to better determine conditions leading to risk of flame acceleration and hydrogen deflagration. The analysis starts with a base case, in which venting occurs when pressure reaches 4.5 kg f /cm 2 and the vent pipe is full of air. Then, the first strategy to reduce hydrogen deflagration risk consists of inertization with nitrogen at specific locations along the vent pipe through rupture disks with three opening pressure setpoints (2.0, 3.0. 4.0 kg f /cm 2 ). Three different locations are considered in this study. The second strategy is the volume enlargement of the last section of the vent pipeline. Two different expansions additional to the base case were considered for analysis. The results show that the inertization with nitrogen at the lower pressure setpoints (2.0 and 3.0 kg f /cm 2 ) does effectively, for practical applications of safety analysis, highly reduces the risk of flame acceleration anywhere in the vent pipeline. However, lowering the opening pressure value implies earlier venting. If it is preferable to keep the disk opening pressure at the higher pressures (4.0 and 4.5 kg f /cm 2 ), the results show that it is necessary to choose an appropriate location to set the rupture disk, to effectively diminish flame acceleration risk. Regarding the second venting strategy, the results show that increasing the volume of the last section of the vent pipe is also an effective way to reduce hydrogen deflagration risk. Thus, although flame acceleration still could occur, those conditions for that to happen will be restricted to a shorter period. For actual practical applications, this second strategy seems more plausible to be carried out, because all relevant changes to the vent pipeline would be focused on the parts already outside reactor building.

AZTRAN is a 3D parallel SN neutron transport code developed for the AZTLAN Platform project, an initiative devoted to developing a Mexican platform for analysis and design of nuclear reactors. AZTRAN solves the multi-group discrete-ordinates form of the neutron transport equation in 3D Cartesian geometry applying the nodal RTN-0 method and the domain decomposition method for its parallelization. To verify and start the validation process, the C5G7 MOX Benchmarks were used. The accuracy of the serial solver was demonstrated by comparing the numerical results of the keff and normalized pin powers with the reference solutions. Excellent agreement and good consistency were obtained, showing errors reduce as the spatial and angular discretizations refine. The parallel implementation is also demonstrated. The speed-up is significant as the number of processors increases without losing quality in the numerical results and reaching efficiencies over 80% in the 2D cases and 85% in the 3D cases.

The AZTLAN Platform project is a Mexican national initiative which aims to have a platform for analysis and design of nuclear reactors. In order to enhance the AZTLAN Platform with Uncertainty and Sensitivity (U&S) capabilities, the AZTUSIA (AZtlan Tool for Uncertainty and SensItivity Analysis) code has been developed to perform U&S analysis. The main characteristics and capabilities of AZTUSIA are briefly described together with some verification exercises. The versatility of the AZTUSIA code has been used to analyze a transient of a BWR-5 with the SIMULATE 3-K code. An artificial perturbation has been caused for creating a more challenging case and to demonstrate the capabilities of an extended interface in which AZTUSIA can be easily adapted to other codes outside AZTLAN Platform. The U&S analysis conducted using the AZTUSIA code reflect, as expected, the impact of uncertain parameters and all capabilities of the tool are demonstrated. It is possible to assess the influence of uncertainty in the calculations and to distinguish trends and readily group and/or separate results by means of scatter and cobweb color plots. As a main conclusion, a new tool has been developed allowing U&S analysis. It is foreseen to continue increasing the tool capabilities.

After the Fukushima Daiichi Nuclear Power Plant (NPP) accident, the analysis of severe accidents (SA) has become an important field of study in order to define appropriate severe accident management and emergency planning. The Station Blackout (SBO) scenario, in which all the external power is lost, is a reference to the simulation of a severe accident initiating event. All nuclear power plants are designed to face partially a SBO. However, a long SBO scenario can result in a beyond design basis accident and evaluations in order to define severe accident management guidelines (SAMGS) must be performed. The hydrogen generation due to the high temperatures of the melting core in contact with steam is an issue in the assessments of SA. Usually, in boiling water reactors (BWR), the primary containment has an inert atmosphere (with Nitrogen) in order to prevent a detonable hydrogen–oxygen mixture. However in a SA, the primary containment is highly jeopardized and can fail with the subsequent release of the inert atmosphere to the secondary containment where the hydrogen can be in contact with the oxygen of the air and a detonation may occur (as in Fukushima Daiichi NPP). An option in order to avoid a detonation inside the reactor building in the case of a primary containment failure is by means of the installation of a Hard Venting Containment System (HVCS). A methodology in order to assess the transport of gases (especially hydrogen, steam and nitrogen) inside the hard venting pipe has been developed. Taking the results of the MELCOR code, mainly mass fluxes as well as temperature and pressure fields inside the primary containment, a source term for the computational fluid dynamics (CFD) code GASFLOW is calculated in order to predict the hydrogen behaviour inside the venting system. The calculations performed with GASFLOW showed the risk of the occurrence of a detonation inside the venting pipe due to the hydrogen volumetric concentrations entering in contact with the air inside the venting pipe (if it is initially filled with air).

A model of the Triga Mark III reactor of the National Institute for Nuclear Research (ININ) of Mexico was developed with the Monte Carlo codes Serpent and MCNP. The models were verified and validated by means of the experiments carried out as part of the starting tests to change the mixed fuel to Low Enrichment Fuels (LEU) of the TRIGA reactor core. The reactor data used in the V&V process consisted of fuel loading measurements, simulating the different stages of loading of fuel elements to the core to reach the reactor core criticality and the additional loading to achieve the reactivity excess to operate the reactor, as well as the evaluation of the shutdown margin reactivity and the control rods worth. The validated models constitute a trustworthy computational tool to analyse the most important neutronic core parameters as well as to have the numerical capabilities for fuel utilisation studies and analysis for extension of experimental facilities.

This paper presents the results of the accelerated solution of the linear algebraic system Av = b arising from a nodal finite element method implemented in the neutron diffusion code AZKIND to solve 3D problems. The numerical solution of full nuclear reactor cores with AZKIND implies the generation of large sparse algebraic systems that produce bottle-necks in the iterative solution. Aiming to alleviate the overload of the algorithm, an acceleration technique has to be implemented. Consequently, a Fortran plug-in of the open source linear algebra PARALUTION library (C ++) was integrated into the AZKIND source code (Fortran 95). This implementation allows AZKIND to use GPUs as well as CPUs, threading into the GPU thousands of arithmetic operations for parallel processing. Selected examples of preliminary investigations performed for a cluster of nuclear fuel assemblies are presented and the obtained results are discussed in this paper.

DYNSUB is a novel two-way pin-based coupling of the simplified transport ( SP 3 ) version of DYN3D with the sub-channel code SUBCHANFLOW. The new coupled code system allows for a more realistic description of the core behaviour under steady state and transient conditions. The details of the developed internal coupling approach of both codes together with its implementation are presented and discussed. Comparisons of the results predicted by DYNSUB with the ones of coarser coupled solutions have shown very good agreement in the global parameters ( k eff and power distribution at steady state and position and magnitude of the power peak in the transient cases) validating the correctness of the coupling strategy. At local level however, important (and expected) deviations in the local safety parameters (maximal clad, fuel and moderator temperatures) have arisen. Differences up to 150 K in the centreline fuel rod temperature were found. It demonstrates the novel capabilities of the developed coupled system DYNSUB. The more detailed coupling solution had also an important impact on the convergence process, mainly in the neutronics internal convergence due to a smoother gradient on the thermal-hydraulics feedback parameters between neighbour sub-channels, compared to the gradient between assembly level channels. DYNSUB has successfully been applied to analyze the behaviour of one eight of a PWR core during a REA transient by a pin-by-pin simulation consisting of a huge amount of nodes.

DYNSUB is a novel two-way pin-based coupling of the simplified transport (SP3) version of DYN3D with the subchannel code SUBCHANFLOW. The new coupled code system allows for a more realistic description of the core behaviour under steady state and transients conditions, and has been widely described in Part I of this paper. Additionally to the explicit coupling developed and described in Part I, a nested loop iteration or fixed point iteration (FPI) is implemented in DYNSUB. A FPI is not an implicit scheme but approximates it by adding an iteration loop to the current explicit scheme. The advantage of the method is that it allows the use of larger time steps; however the nested loop iteration could take much more time in getting a converged solution that could be less efficient than the explicit scheme with small time steps. A comparison of the two temporal schemes is performed. The results using FPI are very promising and represent a very good option in order to optimize computational times without losing accuracy. However it is also shown that a FPI scheme can produce inaccurate results if the time step is not chosen in agreement with the analyzed transient.

Materials surveillance programs are required to detect and prevent degradation of safety-related structures and components of a nuclear power reactor. In this work, following the directions in the Regulatory Guide 1.190, a calculational methodology is implemented as additional support for a reactor pressure vessel and internals surveillance program for a BWR. The choice of the neutronic methods employed was based on the premise of being able of performing all the expected future survey calculations in relatively short times, but without compromising accuracy. First, a geometrical model of a typical BWR was developed, from the core to the primary containment, including jet pumps and all other structures. The methodology uses the Synthesis Method to compute the three-dimensional neutron flux distribution. In the methodology, the code CORE-MASTER-PRESTO is used as the three-dimensional core simulator; SCALE is used to generate the fine-group flux spectra of the components of the model and also used to generate a 47 energy-groups job cross section library, collapsed from the 199-fine-group master library VITAMIN-B6; ORIGEN2 was used to compute the isotopic densities of uranium and plutonium; and, finally, DORT was used to calculate the two-dimensional and one-dimensional neutron flux distributions required to compute the synthesized three-dimensional neutron flux. Then, the calculation of fast neutron fluence was performed using the effective full power time periods through six operational fuel cycles of two BWR Units and until the 13th cycle for Unit 1. The results showed a maximum relative difference between the calculated-by-synthesis fast neutron fluxes and fluences and those measured by Fe, Cu and Ni dosimeters less than 7%. The dosimeters were originally located adjacent to the pressure vessel wall, as part of the surveillance program. Results from the computations of peak fast fluence on pressure vessel wall and specific weld locations on the core shroud are also presented.

This paper presents a substantial extension of the pin power reconstruction (PPR) method used in the reactor dynamics code DYN3D with the aim to better describe the heterogeneity within the fuel assembly during reactor simulations. The flexibility of the new implemented PPR permits the local spatial refinement of one fuel assembly, of a cluster of fuel assemblies, of a quarter or eight of a core or even of a whole core. The application of PPR in core regions of interest will pave the way for the coupling with sub-channel codes enabling the prediction of local safety parameters. One of the main advantages of considering regions and not only a hot fuel assembly (FA) is the fact that the cross flow within this region can be taken into account by the subchannel code. The implementation of the new PPR method has been tested analysing a rod ejection accident (REA) in a PWR minicore consisting of 3 × 3 FA. Finally, the new capabilities of DNY3D are demonstrated by the analysing a boron dilution transient in a PWR MOX core and the pin power of a VVER-1000 reactor at stationary conditions.

Numerical modeling simulations and the use of high-performance computing are fundamental for detailed safety analysis, control and operation of a nuclear reactor, allowing the study and analysis of problems related with thermal-hydraulics, neutronic and the dynamic of fluids which are involved in these systems. In this work we introduce the bases for the implementation of the smoothed particle hydrodynamics (SPH) approach to analyze heat transfer in a nuclear reactor core. Heat transfer by means of convection is of great importance in many engineering applications and especially in the analysis of heat transfer in nuclear reactors. As a first approach, the natural convection in the gap (space that exists between the fuel rod and the cladding) can be analyzed helping to reduce uncertainty in such calculations that usually relies on empirical correlations while using other numerical tools. The numerical method developed in this work was validated while comparing the results obtained in previous numerical simulations and experimental data reported in the literature showing that our implementation is suitable for the study of heat transfer in nuclear reactors. Numerical simulations were done with the DualSPHysics open source code that allows to perform parallel calculations using different number of cores. The current implementation is a version written in CUDA (Compute Unified Device Architecture) that allows also the use of GPU processors (Graphics Processor Unit) to accelerate the calculations in parallel using a large number of cores contained in the GPU. This makes possible to analyze large systems using a reasonable computer time. The obtained results verified and validated our method and allowed us to have a strong solver for future applications of heat transfer in nuclear reactors fuel inside the reactor cores.