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The objective of this paper is to introduce a new direct method for neutronic calculations. This method, called direct discrete method, is simpler than the application of the neutron transport equation and more compatible with the physical meanings of the problem. The method, based on the physics of the problem, initially runs through meshing of the desired geometry. Next, the balance equation for each mesh interval is written. Considering the connection between the mesh intervals, the final discrete equation series are directly obtained without the need to pass through the set up of the neutron transport differential equation first. In this paper, one and multigroup neutron transport discrete equation has been produced for a cylindrical shape fuel element with and without the associated clad and the coolant regions each with two different external boundary conditions. The validity of the results from this new method is tested against the results obtained by the MCNP-4B and the ANISN codes.

In the present study, the development of a neutron noise simulator, DYN-ACNEM, using the Average Current Nodal Expansion Method (ACNEM) in 2-G, 2-D hexagonal geometries is reported. In first stage, the static neutron calculation is performed. The neutron/adjoint flux distribution and corresponding eigen-values are calculated using the algorithm developed based on power iteration method by considering the coarse meshes. The results of the static calculation are validated against the well-known IAEA-2D benchmark problem. In the second stage, the dynamic calculation is performed in the frequency domain in which the dimension of the variable space of the noise equations is lower than the time dependent equations. Induced neutron/adjoint noise distribution due to the neutron noise source of type absorber of variable strength is calculated as well. Two different methods are used to validate the neutron noise calculation. The Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position (ILOFAIP) as an important neutron noise source is investigated using ACNEM in this study. [ABSTRACT FROM AUTHOR]

In this paper, the development of the MCNPX-ESUT (MCNPX-Energy Engineering of Sharif University of Technology) computer code for simulation of neutron/gamma pulse height distribution is reported. Since liquid organic scintillators like NE-213 are well suited and routinely used for spectrometry in mixed neutron/gamma fields, this type of detectors is selected for simulation in the present study. The proposed algorithm for simulation includes four main steps. The first step is the modeling of the neutron/gamma particle transport and their interactions with the materials in the environment and detector volume. In the second step, the number of scintillation photons due to charged particles such as electrons, alphas, protons and carbon nuclei in the scintillator material is calculated. In the third step, the transport of scintillation photons in the scintillator and lightguide is simulated. Finally, the resolution corresponding to the experiment is considered in the last step of the simulation. Unlike the similar computer codes like SCINFUL, NRESP7 and PHRESP, the developed computer code is applicable to both neutron and gamma sources. Hence, the discrimination of neutron and gamma in the mixed fields may be performed using the MCNPX-ESUT computer code. The main feature of MCNPX-ESUT computer code is that the neutron/gamma pulse height simulation may be performed without needing any sort of post processing. In the present study, the pulse height distributions due to a monoenergetic neutron/gamma source in NE-213 detector using MCNPX-ESUT computer code is simulated. The simulated neutron pulse height distributions are validated through comparing with experimental data (Gohil et al. Nuclear Instruments and Methods in Physics Research Section A: Accelerators, Spectrometers, Detectors and Associated Equipment, 664 (2012) 304–309.) and the results obtained from similar computer codes like SCINFUL, NRESP7 and Geant4. The simulated gamma pulse height distribution for a 137 Cs source is also compared with the experimental data. [ABSTRACT FROM AUTHOR]

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: Noise analysis techniques including Feynman-α (variance-to-mean) and Rossi-α (correlation) have been simulated by MCNP computer code to calculate the prompt neutron decay constant (α 0), effective delayed neutron fraction (βeff ) and neutron generation time (Λ) in a subcritical condition for the first operating core configuration of Tehran Research Reactor (TRR). The reactor core is considered to be in zero power (reactor power is less than 1W) in the entire simulation process. The effect of some key parameters such as detector efficiency, detector position and its dead time on the results of simulation has been discussed as well. The results of proposed method in the current study are validated against both the experimental data and the results of MTR_PC computer code. [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]

• Development of a Computer code for safety analysis of a nuclear reactor. • The Code works based on nodal discretization using MCNPX code. • The restrictions of previous methods in transient analysis have been removed. • The Code is capable to simulate the systems with time-varying geometry and cross sections. • Independency of the method of the code to the node size. This paper proposes a new method for solving the time-dependent neutron transport equation based on nodal discretization using the MCNPX code. Most valid nodal codes are based on the diffusion theory with differences in approximating the leakage term until now. However, the Monte Carlo (MC) method is able to estimate transport parameters without approximations usual in diffusion method. Therefore, improving the nodal approach via the MC techniques can substantially reduce the errors caused by diffusion approximations. In the proposed method, the reactor core is divided into nodes of arbitrary dimensions, and all terms of the transport equation e.g. interaction rates and leakage ratio are estimated using MCNPX. They are then employed within the time-dependent neutron transport equation for each node independently to compute the neutron population. Based on this approach, a time-dependent code namely MCNP-NOD (MCNPX code based on a NODal discretization) was developed for solving time-dependent transport equation in an arbitrary geometry considering feed backs. The MCNP-NOD is able to simulate multi-group processes using appropriate libraries. Several test problems are examined to evaluate the method. [ABSTRACT FROM AUTHOR]

Abstract: The time-dependent diffusion of neutrons in a spheroid as a function of the focal distance has been studied. The solution is based on an orthogonal basis and an extrapolation distanced related boundary condition for the spheroidal geometry. It has been shown that spheres and disks are two limiting cases for the spheroids, for which there is a smooth transition for the systems properties between these two limits. Furthermore, it is demonstrated that a slight deformation from a sphere does not affect the fundamental mode properties, to the first order. The calculations for both multiplying and non-multiplying media have been undertaken, showing good agreement with direct Monte Carlo simulations. [Copyright &y& Elsevier]

• The development of the SD-HACNEM neutron noise simulator. • Use of high-order flux expansion in order to simulate ILOFAIP neutron noise. • Use of nodes in the dimensions of a fuel assembly of Western nuclear reactors. • Use of fourth-degree polynomials in discretization of diffusion equations. In this study, the SD-HACNEM (Sharif Dynamic - High order Average Current Nodal Expansion Method) neutron noise simulator in two energy groups using a second-order flux expansion method for two-dimensional rectangular X Y-geometry has been developed. In the first step, the calculations were performed for the steady state and results of ACNEM (Average Current Nodal Expansion Method) and HACNEM (High order Average Current Nodal Expansion Method) were examined and compared. To solve the problem, the power iteration algorithm has been used to calculate the distribution of neutron flux and neutron multiplication factor by considering the coarse-mesh (each fuel assembly one node). To validate the steady state calculations, the results were compared with the western nuclear reactor BIBLIS-2D (reference model). In the next step, dynamic calculations were performed in the frequency domain. For this purpose, after discretizing the neutron noise equations by the high-order average current nodal expansion method (NEM), a neutron noise source of absorber with variable strength was simulated. The results were verified by performing simulation at zero frequency and adjoint calculations. Then, the distribution of neutron noise for Inadvertent Loading and Operation of a Fuel Assembly in an Improper Position (ILOFAIP) is obtained. The numerical results showed that the use of HACNEM compared with ACNEM, without reducing the size of the meshes in neutron noise simulation, presented more accurate. [ABSTRACT FROM AUTHOR]

• Development of a computer code for safety analysis of a nuclear reactor in the presence of xenon oscillation. • The Code works based on nodal discretization using a MC code named MCSP. • The restrictions of previous methods in transient analysis have been removed. • The Code is capable to simulate the systems with time-varying geometry and cross sections. • Independency of the method of the code to the node size. In this paper a probabilistic methodology based on core nodalization is proposed to estimate the core power in the presence of xenon oscillation. A time-dependent Monte Carlo neutron transport code named MCSP-NOD is developed for dynamic analysis in arbitrary 3D geometries to simulate xenon oscillations as well as sub-critical condition with feedbacks. The new code is based on the approach adopted in MCNP-NOD which was previously introduced as a tool for core transient analysis using the MCNPX platform. As before, the core is divided into nodes of arbitrary dimensions, and all terms of the transport equation e.g. interaction rates, leakage ratio are estimated using the MC techniques. However, as a new option the concentration of iodine and xenon are also estimated which enables us to predict the oscillatory behavior of reactor power following poison oscillation. These quantities are then employed within the time-dependent neutron transport equation for each node independently to compute the neutron population. Simulations prove the robustness of the method. [ABSTRACT FROM AUTHOR]