DYN3D Code Research Articles

Verification of the code DYN3D for calculations of neutron flux fluctuations

Insufficiently explained magnitudes and patterns of flux fluctuation observed mainly in KWU PWRs are recently investigated by various European institutions. Among the numerical tools used to investigate the neutron flux fluctuations is the time-domain reactor dynamics code DYN3D. As DYN3D and comparable codes have not been developed with the primary intention to simulate low-amplitude neutron flux fluctuations, their applicability in this field has to be verified.In order to contribute to the verification of DYN3D for the simulation of neutron flux fluctuations, two special cases of perturbations of the neutron flux (a localized absorber of variable/oscillatory strength and a travelling oscillatory perturbation) are considered with DYN3D on the one hand and with the frequency-domain neutron noise tool CORE SIM as well as analytical frequency-domain approaches, respectively, on the other hand. The obtained results are compared with respect to the distributions of the amplitude and the phase of the induced neutron flux fluctuations. The comparisons are repeated with varied amplitudes and frequencies of the perturbation.The results agree well both qualitatively and quantitatively for each of the conducted calculations. The remaining deviations between the DYN3D results and the reference results exhibit a dependence on the perturbation magnitude, which is attributed to the neglect of higher-order terms (linear theory) of the perturbed quantities in the calculation of the reference solutions.

Optimization of multi-group energy structures for diffusion analyses of sodium-cooled fast reactors assisted by simulated annealing – Part I: Methodology demonstration

This study presents an approach to the selection of optimal energy group structures for multi-group nodal diffusion analyses of Sodium-cooled Fast Reactor cores. The goal is to speed up calculations, particularly in transient calculations, while maintaining an acceptable accuracy of the results.In Part I of the paper, possible time-savings due to collapsing of energy groups are evaluated using 24-group energy structure as a reference. Afterwards, focusing on energy structures with a number of groups leading to significant calculation speedups, optimal grid configurations are identified. Depending on a number of possible energy grid configurations to explore, the optimization is conducted by either a direct search or applying the simulated annealing method. Speedup and optimization studies are performed on a selected case of the Superphénix static neutronic benchmark by using the nodal diffusion DYN3D code. The results demonstrate noticeable improvements in DYN3D performance with a marginal deterioration of the accuracy.

Adaptation of the gas gap simplified model in DYN3D code to new types of fuel

Abstract Currently, new types of fuel are being considered to be introduced or already in the introduction process at Ukrainian NPPs with WWER. By means of a new version of the TRANSURANUS code, new functions of the gas gap thickness in dependence on the burnup have been created and implemented into the gas gap model of the reactor dynamics code DYN3D. These new functions cover all actual and perspective fuel types for the Ukrainian NPPs with WWER.

Modeling of the FFTF isothermal physics tests with the Serpent and DYN3D codes

In this study, the isothermal physics tests performed on the fully loaded core of the Fast Flux Test Facility are analyzed with the Monte Carlo code Serpent and with the deterministic core simulator DYN3D. The selected tests comprise two neutron spectra measurements and a multitude of reactivity measurements (32 cases with control rod movements and one case for isothermal temperature coefficient). While the flux spectra are calculated only with Serpent, the reactivity effects are evaluated with the both codes. The homogenized few-group cross sections for DYN3D calculations are generated with Serpent. The obtained numerical results are in a very good agreement as compared to the experimental data. Additionally, a comparison of radial power distributions is presented between DYN3D and Serpent calculations, demonstrating an adequate performance of the nodal code DYN3D. This paper provides an additional contribution to the validation of both codes for neutronic analyses of Sodium cooled Fast Reactor cores.

Extension of the reactor dynamics code DYN3D to SFR applications – Part I: Thermal expansion models

The reactor dynamics code DYN3D, initially developed for LWR applications, is being extended for steady-state and transient analyses of Sodium cooled Fast Reactor (SFR) cores. In contrast to LWRs, thermal expansions of SFR core and reactor components such as fuel, cladding, diagrid, control rod (CR) drivelines, vessel, etc. provide essential reactivity feedbacks under normal and transient conditions.Since DYN3D was originally oriented to the LWRs analyses, the modeling of thermal expansion mechanisms was not considered in the code. Therefore, the development of a new thermal-mechanical module accounting for thermal expansions has been initiated as a part of the SFR related activities. At first step, the DYN3D code was extended with the capability of treating two important thermal expansion effects occurring within the core, namely axial expansion of fuel rod and radial expansion of diagrid.Part I of the paper provides a detailed description of the newly implemented models and presents the results of the initial verification study performed on a full core level using a large oxide SFR core from the OECD/NEA benchmark and a Phenix end-of-life core from the static neutronic IAEA benchmark.Two IAEA benchmarks on the Phenix end-of-life experiments were used for a more detailed validation of the extended version of DYN3D. While PartII presents the validation study performed against the static neutronic benchmark of the control rod withdrawal tests, the results of the initial stage of the natural circulation test are covered in PartIII of the paper.

Validation of the DYN3D-Serpent code system for SFR cores using selected BFS experiments. Part II: DYN3D calculations

The capability of the DYN3D-Serpent codes system to simulate highly heterogeneous sodium-cooled fast reactor cores has been studied. The BFS-73-1 and the BFS-62-3A critical assemblies were chosen for the investigation. The study was performed in two parts. In the first part of the study, a 3D full model of each of the assemblies was simulated using the Serpent Monte-Carlo (MC) code, and the basic neutronic characteristics were evaluated and compared against experimental values. In the second part of the study, which is the subject of this paper, the assemblies were modeled using the DYN3D nodal diffusion code. The few-group cross-sections for the DYN3D analysis were generated using the Serpent MC code. The generation of effective few-group cross-sections of such assemblies is quite a challenge due to the substantial heterogeneity of the assemblies configuration. Therefore, the use of homogenization techniques was considered and evaluated. Initially, the GET and SPH techniques were applied for the analysis of the BFS-73-1 assembly core fuel rods, and of selected fuel rods from the BFS-62-3A assembly. Then, the SPH method was implemented and demonstrated for a pin-by-pin calculation of the BFS-73-1 assembly. It was shown that the GET and the SPH method noticeably improve the prediction accuracy of the DYN3D code. The results of the DYN3D pin-by-pin calculation with the SPH correction agree very well with that of the full assembly Serpent results, which in turn agree very well with the experimental data.

Validation of the DYN3D-Serpent code system for SFR cores using selected BFS experiments. Part I: Serpent calculations

A comparative study has been performed to evaluate the prediction capability of the DYN3D-Serpent code system for sodium fast reactor (SFR) cores. In this study, the calculation system was tested against the BFS-73-1 and BFS-62-3A experiments conducted at the Russian Institute of Physics and Power Engineering (IPPE). These experiments were designed for full-scale modeling of SFR cores, and for validation of codes and nuclear data. The study was performed in two parts. The first part is aimed at developing and validating a 3D full-core heterogeneous model of each of the experiments using the Serpent Monte-Carlo code (MC) code. This part meant as a first step towards the use of the Serpent MC code as a tool for preparation of homogenized group constants, and as a reference solution for code-to-code comparison with the DYN3D code. The second part is devoted to the homogenized group constants generation procedure and the DYN3D steady-state calculations. This paper covers the first part of the study. The experiments were simulated using the Serpent MC code, and the basic neutronic characteristics were evaluated and compared against experimental values. The calculated results agreed well with the measured values on most of the neutronic characteristics. It suggests that the Serpent MC code can be used for the preparation of homogenized group constants and as a reference solution for code-to-code verification with the DYN3D code.

The reactor dynamics code DYN3D

Abstract The article provides an overview on the code DYN3D which is a three-dimensional core model for steady-state, dynamic and depletion calculations in reactor cores with quadratic or hexagonal fuel assembly geometry being developed by the Helmholtz-Zentrum Dresden-Rossendorf for more than 20 years. The current paper gives an overview on the basic DYN3D models and the available code couplings. The verification and validation status is shortly outlined. The paper concludes with the current developments of the DYN3D code. For more detailed information the reader is referred to the publications cited in the corresponding chapters.

Monte Carlo and nodal neutron physics calculations of the IAEA MTR benchmark using Serpent/DYN3D code system

As part of recent efforts to utilize NPPs computational methodologies to safety analysis of research reactors, the Serpent and DYN3D codes were extensively compared with a variety of static and burnup calculations as defined in the IAEA benchmark for 10 MW MTR pool-type reactor. These calculations include unit cell calculations and few group constants generation, unit cell and full core k-eigenvalue and burnup calculations, and full core 3D flux and power distributions. The Serpent code capabilities as a lattice code for MTR plate-type fuel assemblies were evaluated and compared with EPRI-CELL and WIMS-D4 results and reference solutions for full 3D core models were compared with MCNP5 and OpenMC results. The DYN3D nodal diffusion code capabilities in modeling full 3D MTR cores were also evaluated using few group cross sections and assembly discontinuity factors obtained by Serpent unit cell calculations. The DYN3D results were compared with Serpent, MCNP5 and OpenMC.

On the use of the SPH method in nodal diffusion analyses of SFR cores

A number of recent studies successfully demonstrated the feasibility of using Monte Carlo code Serpent to generate few-group cross sections (XS) for full core nodal diffusion analyses of SFR cores. The current study investigated the potential of the SPH method, applied to correct the few-group XS produced by Serpent, to further improve the accuracy of the nodal diffusion solutions. The procedure for the generation of SPH-corrected few-group XS is presented in the paper. The performance of the SPH method was tested on a large oxide SFR core from the OECD/NEA SFR benchmark. The reference SFR core was modeled with the DYN3D and PARCS nodal diffusion codes using the SPH-corrected few-group XS generated by Serpent. The nodal diffusion results obtained with and without SPH correction were compared to the reference full-core Serpent MC solution. It was demonstrated that the application of the SPH method improves the accuracy of the nodal diffusion solutions, particularly for the rodded core state.

Development of Cross-Section Library for DYN3D Code

At present, SSTC NRS uses the HELIOS code for generation of few-group cross-section libraries for WWER core calculations. There is an urgent issue of selecting the appropriate approach to implement the cross-section library into the DYN3D code. The paper overviews the application of approaches used by SSTC NRS, such as a multidimensional table and polynomial dependences. The capabilities and possible extension of each approach are described with inherent advantages and disadvantages. In addition, the model development and cross-section preparation for the WWER-1000 radial reflector taking into account discontinuity factors are discussed. Brief results of calculations with the use of different approaches are presented.

Effect of Fuel Burnup History on Neutronic Characteristics of WWER-1000 Core

In preparation of few-group cross-section libraries to be used in WWER macro-calculations, change in the nuclide composition during fuel burnup is commonly defined under invariable characteristics, averaged over the entire core (power, fuel and moderator temperature, moderator density etc.). In reality, conditions of fuel burnup are changing and this factor affects the fuel nuclide composition (the so-called spectral history effect). To account for real burnup history, it is necessary to take into consideration the dependence of cross-sections not only on burnup but also on history of neutron spectrum in the fuel burnup process.
 This paper analyzes fuel burnup history effect on neutronic characteristics of WWER-1000 core with use of the DYN3D code. The DYN3D code employs the local Pu-239 concentration as an indicator of burnup spectral history. The calculations have been performed for the first four fuel loadings of Khmelnitsky NPP unit 2 and stationary fuel loading with TVSA. The effect of fuel burnup history is shown both on macro-characteristics of the reactor core (boric acid concentration, fuel cycle duration, reactivity coefficients) and on local values of burnup and power.

Overview of major HZDR developments for fast reactor analysis

The upgrading of the DYN3D code for the application for fast reactors is described. After extension and validation, a diverse code with the possibility for steady state and transient core analysis on the basis of coupled thermal hydraulics/neutronics calculations is available. The work on the use of fine distributed moderating material in SFR cores is discussed with the target on enhancing the feedback effects in SFR cores without influencing the operational. Newly developed analytical solutions without separation of space and time for the analysis of ADS experiments are shown with good agreement for the YALINA experiment. The analytical solutions are a very promising tool for the development of a new method for the analysis of ADS experiments.

Neutronic analysis of SFR core with HELIOS-2, Serpent, and DYN3D codes

In this study, HELIOS-2 deterministic transport code and Serpent Monte–Carlo (MC) reactor physics code were considered as tools for preparation of few-group constants for sodium cooled fast reactor (SFR) analysis. Initially, applicability of the mainly LWR-oriented HELIOS-2 code to the modeling of SFR lattices was investigated and recommendations for methodological modifications were given. At the next stage the methodology for few-group cross section generation for fuel and non-multiplying regions of SFR core was proposed. Afterward, few-group constants produced by HELIOS-2 and Serpent employing the proposed methodology were used by nodal diffusion code DYN3D for the analysis of a reference SFR core. Finally, the DYN3D results were verified against the full core Serpent MC solution. The full core DYN3D results obtained using few-group constants produced by Serpent agreed very well with that of the reference full core MC simulations. The use of HELIOS-2 based few-group constants led to somewhat reduced agreement between reference MC and DYN3D results. The implementation of the suggested modifications to the HELIOS-2 methodology can potentially improve its modeling accuracy for SFR lattices.

Modeling of SFR cores with Serpent–DYN3D codes sequence

DYN3D reactor dynamics nodal diffusion code was originally developed for the analysis of Light Water Reactors. In this paper, we demonstrate the feasibility of using DYN3D for modeling of fast spectrum reactors. A homogenized cross sections data library was generated using continuous energy Monte-Carlo code Serpent which provides significant modeling flexibility compared with traditional deterministic lattice transport codes and tolerable execution time. A representative sodium cooled fast reactor core was modeled with the Serpent–DYN3D code sequence and the results were compared with those produced by ERANOS code and with a 3D full core Monte-Carlo solution. Very good agreement between the codes was observed for the core integral parameters and power distribution suggesting that the DYN3D code with cross section library generated using Serpent can be reliably used for the analysis of fast reactors.

Studies on boiling water reactor design with reduced moderation and analysis of reactivity accidents using the code DYN3D-MG

Abstract The multi-group version of the DYN3D reactor dynamics code was used for calculations for a new concept of a boiling water reactor with tight lattice of fuel rods and reduced neutron moderation. For that purpose, a 5-group cross section library was prepared and connected to the DYN3D code. Comparison calculations with the steady-state finite-difference code ACADEM showed a very good agreement. The capability of the DYN3D multi-group code in modeling transients in boiling water reactors with tight fuel element lattices was demonstrated by the analysis of two reactivity accidents initiated by the ejection of one control rod and unauthorized withdrawal of a control rod bank from the reactor core. The corresponding analyses were performed for begin of cycle conditions, when the considered control rods are at their maximum insertion depth.

The simplified P3 approach on a trigonal geometry of the nodal reactor code DYN3D

Abstract DYN3D is a three-dimensional nodal code for steady-state and transient analyses of Light Water Reactors applicable for square and hexagonal fuel assembly geometries. Several versions of the DYN3D code are available including a multigroup diffusion and a simplified P3 (SP3) neutron transport option. The multi-group SP3 method based on a trigonal geometry was developed recently. This method is applicable to the analysis of reactor cores with hexagonal fuel assemblies and allows flexible mesh refinement. In this paper, the theoretical background for the SP3 method is briefly described. The consistency of the implementation of the trigonal SP3 methodology in DYN3D is demonstrated by means of a simplified VVER-440 core test example. The corresponding few-group homogenized cross sections and reference solutions were produced by the Monte Carlo code Serpent. The DYN3D results are in good agreement with the reference solutions.

Development and verification of the coupled 3D neutron kinetics/thermal-hydraulics code DYN3D-HTR for the simulation of transients in block-type HTGR

DYN3D is a nodal diffusion code for 3D steady-state and transient analysis of Light Water Reactor (LWR) cores with hexagonal or square fuel element geometry. In addition to the neutron kinetics, it comprises of a thermal-hydraulics model for flow in parallel coolant channels. Macroscopic cross section data libraries generated with variation of burn-up, reactor poisons concentrations and thermal-hydraulic feedback parameters are linked to the code. Two-group and multi-groups versions of the code are available.Currently, at the Helmholtz-Zentrum Dresden-Rossendorf (HZDR), the DYN3D code is being extended and adopted for the application to block-type High Temperature Gas-Cooled Reactors (HTGRs). In this paper, we give an overview of the latest developments of DYN3D concerning block-type HTGR.The simplified P3 (SP3) transport approximation is implemented into the multi-group DYN3D code to take anisotropy of the neutron flux and heterogeneity of the core more precisely into account. The SP3 method previously implemented into DYN3D for square fuel element geometry of LWR is being extended for hexagonal geometry of the graphite blocks, where the hexagons are subdivided into triangular nodes to be able to perform a systematic mesh refinement.One of the main challenges in cross section generation for the HTGR core calculations is the treatment of the so-called “double heterogeneity”. The modified Reactivity-Equivalent Physical Transformation (RPT) approach is applied in order to eliminate the double-heterogeneity of HTGR fuel elements in the deterministic lattice calculations. The main steps of the RPT method are described. The use of the method for the cross section generation of a simplified HTGR core including its verification is presented.A 3D heat conduction module coupled with a channel-type coolant flow model is implemented to take the temperature reactivity feedback to neutronics physically correctly into account. It is shown that there is significant redistribution of the produced heat by heat conduction between the graphite blocks.

Development of the coupled 3D neutron kinetics/thermal-hydraulics code DYN3D-HTR for the simulation of transients in block-type HTGR

Abstract The Light Water Reactor (LWR) dynamics code DYN3D is extended and adopted for the application to block-type High Temperature Gas-Cooled Reactor (HTGR). A procedure for the cross section generation for the HTGR core calculations was developed. The modified Reactivity-Equivalent Physical Transformation (RPT) approach is applied in order to eliminate the double-heterogeneity of HTGR fuel elements in the deterministic lattice calculations. A full core analysis of the reference simplified HTGR core is performed with DYN3D using macroscopic nodal cross sections provided by HELIOS. The SP3 transport approximation is integrated into the multi-group DYN3D code to take anisotropy of the neutron flux and heterogeneity of the core more precisely into account. The SP3 method was developed for hexagonal geometry of the graphite blocks, where the hexagons are subdivided into triangular nodes. A 3D heat conduction module coupled with a channel-type coolant flow model is implemented into the code. It is shown that there is significant redistribution of the produced heat by heat conduction between the graphite blocks.

Development and verification of a nodal approach for solving the multigroup SP 3 equations

The core model DYN3D which has been developed for three-dimensional analyses of steady states and transients in thermal reactors with quadratic or hexagonal fuel assemblies is based on nodal methods for the solution of the two-group neutron diffusion equation. Loading cores with higher content of MOX fuel, the increase of the fuel cycle length, and the consideration of new reactor types are challenging for these standard methods. A nodal expansion method for solving the equations of the simplified P 3 (SP 3) approximation of the multigroup transport equation was developed to improve the accuracy of the DYN3D code. The method described in the paper is verified with pinwise calculations of a steady state of the OECD/NEA and US NRC PWR MOX/UO 2 Core Transient Benchmark. The used 16-group cross section library was generated for DORT calculations with homogenized pin cells. Two different approximations of the diffusion coefficient which occurs in the within-group form of the SP 3 equations are investigated. Using the transport cross section for the calculation of the diffusion coefficient gives much better results than those obtained with the removal cross section. The improvement of the results in comparison to a pinwise diffusion calculation is shown. The results are compared with the DORT and the heterogeneous reference solution of the code DeCART. Concerning the SP 3 calculation using the diffusion coefficient based on the transport cross section (DYN3D-SP3-TR) the deviations of the eigenvalue k eff and the assembly powers from the transport solutions of DORT and DeCART are in the same order as those between the two transport solutions themselves. The improvement of the DYN3D-SP3-TR results in comparison to the diffusion calculation is presented. As the DYN3D-SP3-TR and DORT calculations are performed with homogenized pin cells, the pin powers of the two calculations are closer to each other than to the pin powers of the DeCART solution. To estimate the contribution of higher flux-moments, the ratio of the second flux moment to the zeroth flux moment is investigated along a horizontal row of pins. It is shown that it is low in the fuel assemblies with small peaks at the water channels. Considering the baffle and the water reflector regions higher values are obtained in the fast neutron groups.