•The paper presents a CAD based geometry type in Serpent 2.•The geometry model is constructed from triangulated STL solids.•The same data format is widely used in computer graphics and 3D ...printing.•The methodology and typical user examples are presented.•A workflow example is provided to demonstrate the practical use and performance of the geometry type.
The Serpent 2 Monte Carlo code features a CAD-based geometry option for the modeling of complicated and irregular systems. The methodology is based on the STL data format, which is commonly used for computer graphics and 3D printing, and supported by a wide range of software tools. The methodology has been available for several years, but not described in complete detail. This paper presents the geometry routine with its advantages, limitations and known flaws. A brief overview on practical applications and a workflow example involving neutron and photon transport calculations for a spent nuclear fuel storage rack are provided for discussion. It is concluded that the CAD-based geometry type is a convenient option for various neutron and radiation transport problems, and does not suffer from significant deterioration in computational efficiency compared to conventional CSG models. It is also noted, however, that taking full advantage of the methodology requires some level of understanding on the software tools, the STL data format and the Serpent geometry routine to avoid the most common pitfalls.
A deterministic importance solver has been implemented as an internal subroutine in the Serpent 2 Monte Carlo code for the purpose of producing weight-window meshes for variance reduction. The ...routine solves the adjoint transport problem using the response matrix method with coupling coefficients obtained from a conventional forward Monte Carlo simulation. The methodology can be applied to photon and neutron external source problems, and the solver supports multiple energy groups and several mesh types. Importances can be generated with respect to multiple responses, and an iterative global variance reduction sequence enables distributing the transported particle population evenly throughout the geometry. This paper describes the methodology applied in the response matrix solver and presents a verification for the generated importance functions through simple demonstrations. A practical example involving a photon shielding problem is included for performance evaluation.
•A new method was implemented in Serpent 2 to accelerate fission source convergence.•The method is based on an improved source guess for the first fission cycle.•The solution is obtained using the ...response matrix method.•The results show significant speed-up in source convergence for LWR calculations.
This paper presents a new response matrix based solver implemented in the Serpent 2 Monte Carlo code for the purpose of producing an improved initial guess to accelerate source convergence in criticality source simulations. The solver obtains coupling coefficients required for the response matrix solution from Monte Carlo simulations, and provides a spatial distribution that approximates the converged fission source. The implemented methodology is demonstrated by single-assembly and full-core PWR calculations. The results show that the improved initial guess leads to faster source convergence in terms of both inactive cycles and overall running time.
•The Serpent Monte Carlo reactor physics code has been developed at VTT since 2004.•The code is used in more than 100 universities and research organizations around the world.•This paper is a review ...on the status and user applications of Serpent at the end of year 2013.
The Serpent Monte Carlo reactor physics burnup calculation code has been developed at VTT Technical Research Centre of Finland since 2004, and is currently used in over 100 universities and research organizations around the world. This paper presents the brief history of the project, together with the currently available methods and capabilities and plans for future work. Typical user applications are introduced in the form of a summary review on Serpent-related publications over the past few years.
The topic of this paper is the computation of the matrix exponential in the context of burnup equations. The established matrix exponential methods are introduced briefly. The eigenvalues of the ...burnup matrix are important in choosing the matrix exponential method, and their characterization is considered. Based on the characteristics of the burnup matrix, the Chebyshev rational approximation method (CRAM) and its interpretation as a numeric contour integral are discussed in detail. The introduced matrix exponential methods are applied to two test cases representing an infinite pressurized water reactor pin-cell lattice, and the numerical results are presented. The results suggest that CRAM is capable of providing a robust and accurate solution to the burnup equations with a very short computation time.
•The basic tracking routines and result estimators in Serpent Monte Carlo code are briefly introduced.•The efficiency of the collision flux estimator used in Serpent depends on a user-defined virtual ...collision frequency.•The effect of adjusting the collision frequency on computational performance is demonstrated in a simple test case.
The Serpent Monte Carlo code was originally developed for the purpose of spatial homogenization and other computational problems encountered in the field of reactor physics. However, during the past few years the implementation of new methodologies has allowed expanding the scope of applications to new fields, including radiation transport and fusion neutronics. These applications pose new challenges for the tracking routines and result estimators, originally developed for a very specific task. The purpose of this paper is to explain how the basic collision estimator based cell flux tally in Serpent 2 is implemented, and how it is applied for calculating integral reaction rates. The methodology and its limitations are demonstrated by an example, in which the tally is applied for calculating collision rates in a problem with very low physical collision density. It is concluded that Serpent has a lot of potential to expand its scope of applications beyond reactor physics, but in order to be applied for such problems it is important that the code users understand the underlying methods and their limitations.
•The Serpent Monte Carlo code has been developed for spatial homogenization for more than 10years.•The code has a large and active international user basis.•This paper describes the methodology used ...in Serpent for spatial homogenization, including its limitations and major flaws.
This paper describes the methods used in the Serpent 2 Monte Carlo code for producing homogenized group constants for nodal diffusion and other deterministic reactor simulator calculations. The methodology covers few-group reaction cross sections, scattering matrices, diffusion coefficients and poison cross sections condensed in infinite and B1 leakage-corrected critical spectra, as well as the calculation of discontinuity factors, pin-power form factors, delayed neutron parameters and total and partial albedos. Also included is a description of an automated burnup sequence, which was recently implemented for the handling of restart calculations with branch variations. This capability enables covering the full range of local operating conditions required for the parameterization of group constants within a single run. The purpose of this paper is to bring the methodological description provided in earlier publications up to date, and provide insight into the developed methods and capabilities, including their limitations and known flaws.
•A new energy deposition treatment was implemented for Serpent 2.•Both the spatial fidelity and the tallied total energy deposition are improved.•The treatment is demonstrated with simple test ...cases.•A high-level comparison with MCNP6 shows good agreement.•Accurate energy deposition estimates for multi-physics calculations are now possible.
In this work a new energy deposition treatment for the Serpent 2 Monte Carlo transport code is introduced. The treatment is implemented as a set of energy deposition modes, each of which offers a different combination of accuracy and required computational time. The most accurate mode uses the coupled neutron-photon transport available in Serpent 2. In addition to the new modes, a few other improvements such as a new energy deposition detector response was developed. The energy deposition treatment is demonstrated with simple LWR and LFR assembly level test cases. In the LWR cases the results are compared with MCNP6 on a high-level and a good agreement is achieved. In the LFR case differences in the energy deposition estimates of the different modes are highlighted. Based on the test cases, problems in the energy deposition data with ENDF/B-VII.1 and ENDF/B-VIII.0 are presented. Finally, future work concerning energy deposition capabilities of Serpent 2 are briefly discussed.
•We present a new Monte Carlo method to perform sensitivity/perturbation calculations.•Sensitivity of keff, reaction rates, point kinetics parameters to nuclear data.•Fully continuous implicitly ...constrained Monte Carlo sensitivities to scattering distributions.•Implementation of the method in the continuous energy Monte Carlo code SERPENT.•Verification against ERANOS and TSUNAMI generalized perturbation theory results.
In this work, the implementation of a collision history-based approach to sensitivity/perturbation calculations in the Monte Carlo code SERPENT is discussed. The proposed methods allow the calculation of the effects of nuclear data perturbation on several response functions: the effective multiplication factor, reaction rate ratios and bilinear ratios (e.g., effective kinetics parameters). SERPENT results are compared to ERANOS and TSUNAMI Generalized Perturbation Theory calculations for two fast metallic systems and for a PWR pin-cell benchmark. New methods for the calculation of sensitivities to angular scattering distributions are also presented, which adopts fully continuous (in energy and angle) Monte Carlo estimators.
This paper presents a methodology for applying continuously varying density distributions in Monte Carlo particle transport simulation. The capability is implemented in the Serpent 2 code, as part of ...an effort for developing a universal multiphysics interface for the coupling of Monte Carlo neutronics to thermal hydraulics and fuel performance codes. The method is based on rejection sampling of particle path lengths, but despite its close resemblance to the Woodcock delta-tracking method, the routine can be used with conventional surface tracking as well. The modified tracking routine is put to the test in a simple boiling water reactor pin-cell calculation with continuously changing void distribution in the coolant channel.
