Search

• Fuel Assembly bowing is the lateral change in the FA shapes in the reactor core. • The change in the distance between FAs caused by the bowing leads that affects power distribution. • This perturbation is time-dependent because the boric acid concentration change during the reactor operation cycle. • Neutron noise is used in the analysis and investigation of neutron perturbation time-dependent. • In this research, by using an adiabatic approximation in neutron noise theory has been made to investigate the influence of the bowing phenomenon of FAs. • The noise amplitude caused by the bowing during the cycle depends on the changes in the boric acid concentration, burnup, size, and direction of the FA bowing. Thermomechanical and radiation loads in the reactor core cause lateral changes in the shape of Fuel Assemblies (FAs). This phenomenon, known as FA bowing, has significant implications for reactor safety and operation. Bowing results in alterations to the spacing between FAs, which in turn leads to neutron flux perturbations affecting power distribution. These perturbations are time-dependent due to variations in boric acid concentration during the cycle. Neutron noise theory is employed to analyze these time-dependent neutron perturbations. In this research, it is investigated the bowing effect during the cycle in a VVER-1000 reactor using the neutron noise adiabatic approximation. It was initially considered a VVER-1000 reactor core to verify the calculations algorithm without bowed FAs. Results had acceptable accuracy in power, initial boric acid, and cycle length. Subsequently, it was explored various modes of FA bowing as neutron noise sources the results revealed that the power noise caused by bowing varies throughout the cycle, as expected. The noise amplitude during the cycle depends on changes in boric acid concentration, burnup, size, and the direction of bowing. [ABSTRACT FROM AUTHOR]

Highlights: [•] A sensitivity analysis of unfolding methods to different parameters is done. [•] ISM is proposed to reconstruct the noise source of type vibrating absorber. [•] Four algorithms are used to solve the system of equations in the inversion method. [•] The neutron noise due to a new defined noise source (ILFAIP) is calculated. [•] We find that the scanning method is a reliable algorithm for source reconstruction. [Copyright &y& Elsevier]

Abstract: In pebble bed reactors, the core is filled with thousands of graphite and fuel pebbles. Fuel pebbles in these reactors consist of TRISO particles, which are embedded in a graphite matrix stochastically. The reactor core is also stochastically filled with pebbles. These two stochastic geometries comprise the so-called double heterogeneous nature of this type of reactor. In this paper, a pebble bed reactor, the HTR-10, is used to demonstrate a treatment of this double heterogeneity using the MCNP5 Monte Carlo code and MATLAB programming. In this technique, TRISO particles are modeled in a pebble using the expanded FILL and LATTICE features of MCNP5. MATLAB is used to generate random numbers which represent the location of pebbles in the core. Centers of pebbles are generated stochastically and uniformly and then transferred into the MCNP5 input file as the centers of spherical surfaces. In this model, there is no approximation to the actual geometry. In other words, the double heterogeneous nature is preserved while truncating neither the pebbles in the core nor the particles in the pebble matrix. Finally, to validate the model, benchmark problems of IAEA are used. Very good agreement with experimental results is observed. [Copyright &y& Elsevier]

Abstract: In the present study, the neutron noise, i.e. the stationary fluctuation of the neutron flux around its mean value, is calculated in 2-group forward and adjoint diffusion theory for both hexagonal and rectangular reactor cores. To this end, the static neutron calculation is performed at the first stage. The spatial discretization of equations is based on linear approximation of Galerkin Finite Element Method (GFEM) using unstructured triangle elements. Using power iteration method, forward and adjoint fluxes with the corresponding eigenvalues are obtained. The results are then benchmarked against the valid results for BIBLIS-2D and IAEA-2D benchmark problems and DONJON computer code. The dynamic calculations are performed in the frequency domain which leads to reducing the dimension of the variable space of the noise equations. The forward/adjoint noise in two energy group is obtained by assuming the neutron noise source as an absorber of variable strength type. The neutron noise induced by a vibrating absorber type of noise source is also obtained from the calculated adjoint Green''s function. Comparison of the calculated neutron noise at zero frequency with the results of static calculation, and utilizing the results of adjoint noise calculations are two different procedures to validate the neutron noise calculations. [Copyright &y& Elsevier]

Abstract: Effective delayed neutron fraction β eff and neutron generation time Λ are important factors in reactor physics calculation and transient analysis. In the first stage of this research, these kinetics parameters have been calculated for two states of Tehran Research Reactor (TRR), i.e. cold (fuel, clad and coolant temperature 20°C) and hot (fuel, clad and coolant temperature 65, 49 and 44°C, respectively) states using MTR_PC computer code. The ratio of (β eff ) i /(β eff )core plays an important role in reactivity accident analysis codes. This parameter and its contribution to effective delayed neutron fraction from each nucleus have been calculated in cold and hot reactor states. Uncertainty of effective delayed neutron fraction is evaluated in terms of following four quantities; basic delayed neutron constants, delayed neutron spectra, energy dependence of delayed neutron yield (ν d ) and fission cross-section of 235U and 238U. In the second stage, these parameters have been measured with an experimental method based on Inhour equation. The calculated and measured values are in good agreement. Relative Percent Errors (RPEs) are 2.8% for β eff and 5.7% for Λ in the cold state. [ABSTRACT FROM AUTHOR]

• The best estimate approaches for transients modeling in NPP needs the employment of multi-physics codes and methods. • The core geometry has been fully modeled by Neutronics/Thermal-hydraulics coupling, to enable one-to-one mapping of SBLOCA, aligning with the research objectives. • DRAGON solvers was utilized to generate the PARCS macroscopic cross-sections library in each core state after SBLOCA. • The accident sequence, initiation time of the safety systems and critical controlled parameters of the power plant have been predicted with high precision. • The results depict the impact of core parameters asymmetric distribution on natural circulation and heat transfer in SGs. With the development of computational tools, the accident analysis would be performed by best estimate approaches. The employment of multi-physics codes would be considered as new calculation procedure to obtain high valuable results by improving calculations with specialized codes. The small break loss of coolant accident (SBLOCA) is an accident that has typically been analyzed using system codes with a limited number of Thermal-Hydraulics (TH) channels in the core region, neglecting the use of multi-physics calculations. This approach results in insufficient accuracy and resolution in both neutronics and TH calculations within the core zone. To address this issue, this study proposes the implementation of a Neutronics/Thermal-Hydraulics coupling with one-to-one mapping in the core zone, achieved through the development of a comprehensive plant TH model. This approach aims to enhance the accuracy and resolution of calculations, enabling the identification of local phenomena, parameters, and feedback during accident simulations, and the incorporation of these effects on the macroscopic cross-sections at each time step. The results of SBLOCA simulation reveals the effects of conservative/best estimate approach on the sequence of events, the initiation time of the safety systems, the critical controlled parameters of the power plant, and the precise understanding of the accident. Also, it was found that the radial power distribution becomes disrupted and loses its symmetry immediately after the SCRAM leading to different characteristics of steam generators (SGs) and natural circulation flow rate in each loop of primary side. [ABSTRACT FROM AUTHOR]