GENERATING ALBEDO BOUNDARY CONDITIONS FOR NODAL DIFFUSION CODES USING SERPENT 2

We use the Serpent Monte Carlo code to produce total and partial albedo boundary conditions that can be used to model the Loviisa NPP VVER-440 core with the nodal neutronics tools of Fortum. The albedo generation process is described in detail. The dependence of the generated albedos on boron content and water density is investigated and a clear distinction is noted in water density dependence between regions containing mostly water and those containing mostly structural materials. The Serpent generated albedos are currently used in production calculations for modeling the Loviisa reactors at Fortum.

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Validating the Serpent-Ants Calculation Chain Using BEAVRS Fresh Core HZP Data

Journal of Nuclear Engineering and Radiation Science ◽

10.1115/1.4052731 ◽

2021 ◽

Author(s):

Ville Valtavirta ◽

Antti Rintala ◽

Unna Lauranto

Keyword(s):

Monte Carlo ◽

Best Practices ◽

Power Distribution ◽

Measured Data ◽

Monte Carlo Code ◽

Generation Process ◽

Isothermal Temperature ◽

Control Rod ◽

Temperature Coefficients ◽

Group Constant

Abstract The Serpent Monte Carlo code and the Serpent-Ants two step calculation chain are used to model the hot zero power physics tests described in the BEAVRS benchmark. The predicted critical boron concentrations, control rod group worths and isothermal temperature coefficients are compared between Serpent and Serpent-Ants as well as against the experimental measurements. Furthermore, radial power distributions in the unrodded and rodded core configurations are compared between Serpent and Serpent-Ants. In addition to providing results using a best practices calculation chain, the effects of several simplifications or omissions in the group constant generation process on the results are estimated. Both the direct and two-step neutronics solutions provide results close to the measured values. Comparison between the measured data and the direct Serpent Monte Carlo solution yields RMS differences of 12.1 mg/kg, 25.1 × 10-5 and 0.67 × 10-5 K-1 for boron, control rod worths and temperature coefficients respectively. The two-step Serpent-Ants solution reaches a similar level of accuracy with RMS differences of 17.4 mg/kg, 23.6 × 10-5 and 0.29 × 10-5 K-1. The match in the radial power distribution between Serpent and Serpent-Ants was very good with the RMS and maximum for pin power errors being 1.31 % and 4.99 % respectively in the unrodded core and 1.67 %(RMS) and 8.39 % (MAX) in the rodded core.

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ANALYSIS OF THE OECD/NEA SFR BENCHMARK WITH ANTS REDUCED-ORDER NODAL DIFFUSION SOLVER AND THE SERPENT MONTE CARLO CODE

EPJ Web of Conferences ◽

10.1051/epjconf/202124704021 ◽

2021 ◽

Vol 247 ◽

pp. 04021

Author(s):

Marton Szogradi

Keyword(s):

Monte Carlo ◽

Lattice Models ◽

Working Party ◽

Research Centre ◽

Monte Carlo Code ◽

Control Rod ◽

Reduced Order ◽

State Present ◽

Nodal Diffusion ◽

Full Core

In order to meet modern industrial and scientific demands the Kraken multi-physics platform’s development was recently launched at VTT Technical Research Centre of Finland. The neutronic solver of the framework consists of two calculation chains, providing full core solutions by the Serpent high fidelity code (1) and the AFEN/FENM-based reduced-order diffusion solver called Ants (2) capable of handling square and hexagonal geometries in steady-state. Present work introduces the simulation of a large 3600 MWth Sodium-cooled Fast Reactor (SFR) described within the activities of the Working Party on Scientific Issues of Reactor Systems (WPRS) of OECD. Full-core 3D results were obtained by Serpent for carbide- and oxide-fuel cores, moreover group constants were generated for Ants utilizing 2D super-cell and single assembly infinite lattice models of Serpent. The continuous-energy Monte Carlo method provided the reference results for the verification of the reduced-order method. Implementing the spatially homogenized properties, 3D solutions were obtained by Ants as well for both core configurations. Comparison was made between the various core designs and codes based on reactivity feedbacks (Doppler constant, sodium voiding, control rod worth) considering power distributions. Regarding reactivity sensitivity on geometry, axial fuel- and radial core expansion coefficients were evaluated as well.

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MACROSCOPIC CROSS SECTIONS GENERATION BY MONTE CARLO CODE MCS FOR FAST REACTOR ANALYSIS

EPJ Web of Conferences ◽

10.1051/epjconf/202124702007 ◽

2021 ◽

Vol 247 ◽

pp. 02007

Author(s):

Tung Dong Cao Nguyen ◽

Hyunsuk Lee ◽

Xianan Du ◽

Vutheam Dos ◽

Tuan Quoc Tran ◽

...

Keyword(s):

Monte Carlo ◽

Cross Sections ◽

Fast Reactor ◽

Three Dimensional ◽

Monte Carlo Code ◽

Metallic Fuel ◽

Sodium Fast Reactor ◽

The Difference ◽

Nodal Diffusion ◽

Core Problem

Recent researches have become more interested in the feasibility of using Monte Carlo (MC) code to generate multi-group (MG) cross sections (XSs) for fast reactor analysis using nodal diffusion codes. The current study, therefore, presents a brief methodology for MG XSs generation by the in-house UNIST MC code MCS, which can be compatibly utilized in nodal diffusion codes, PARCS and RAST-K. The applicability of the methodology is quantified on the sodium fast reactor (SFR) ABR-1000 design with a metallic fuel from the OECD/NEA SRF benchmark. The few-group XSs generated by MCS with a two-dimensional (2D) fuel assembly geometry are well consistent with those of SERPENT 2. Furthermore, the simulation of beginning-of-cycle (BOC) steady-state three-dimensional (3D) whole-core problem with PARCS and RAST-K is conducted using the generated 24-group XSs by MCS. The nodal diffusion solutions, including the core keff, power profiles and various of reactivity parameters, are compared to reference whole-core results obtained by MC code MCS. Overall, the code-to-code comparison indicates a reasonable agreement between deterministic and stochastic codes, with the difference in keff less than 100 pcm and the root-mean-square (RMS) error in assembly power less than 1.15%. Therefore, it is successfully demonstrated that the employment of the MG XSs generation by MCS for nodal diffusion codes is feasible to accurately perform analyses for fast reactors.

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