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Session Overview
7.01: Reactor Physics & Analysis-Transient Simulations
Monday, 16/Mar/2020:
1:30pm - 3:00pm

Session Chair: Muhammad Imron, Abu Dhabi Polytechnic, United Arab Emirates
Location: R-2013

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Verification of SARAX code for the transient analysis of sodium-cooled fast reactor

Xiaoqian Jia, Youqi Zheng, Xianan Du

Xi'an Jiaotong University

This paper describes the verification work of SARAX code for the transient analysis of a sodium-cooled fast reactor (SFR). The Advanced Burner Test Reactor (ABTR) benchmark created by Argonne National Laboratory (ANL) was modeled and calculated. The reference core is the 250 MWt sodium-cooled fast reactor, which includes neutronics calculation of the core at the beginning of equilibrium cycle, and also several transient analysis sequence such as ULOF (Unprotected Loss-of-Flow) accident and ULOHS (Unprotected Loss-of-Heat-Sink) accident. The SARAX code is a neutronics analysis package developed by the NECP team at Xi’an Jiaotong University and aiming for the advanced reactor R&D. It consists a cross-section generation code named TULIP, a steady state neutronics calculation code named LAVENDER and a transient analysis code named DAISY. In this paper, the 33-group homogenized cross sections of all materials were generated using TULIP. LAVENDER gave the results of steady state parameters like power distribution, critical control rod position, reactivity coefficients and kinetics parameters. Then, DAISY simulated the transient progress with a space-dependent point-kinetics model and a parallel multi-channel thermal-hydraulics model and gave the results of peaking fuel temperature, cladding temperature and coolant temperature. The simulation of ULOF and ULOHS transients showed that SARAX gave comparable results with the design code of ANL and the SAS4A code, which verified the complete code system for transient calculations of SFR.

Transient Capability of a Multigroup Pin Homogenized SP3 Code SPHINCS

Hyun Ho Cho1, Junsu Kang1, Joo Il Yoon2, Han Gyu Joo1

1Seoul National University; 2KEPCO Nuclear Fuel

The conventional two-step method is still a powerful and practical tool for nuclear design and analyses which involve repeated steady-state and transient calculations with reasonable accuracy. In this regard, the nTRACER/SPHINCS advanced two-step calculation system employing the Simplified P3 (SP3) method utilizing pin-homogenized group constants and pin-sized finite difference method (FDM) is developed at Seoul National University. For the complete pin-by-pin core analyses, a transient calculation module has been recently implemented in the SPHINCS code in addition to the pin-wise depletion and thermal hydraulic calculation modules. The transient capability of SPHINCS involves the solution of the time-dependent SP3 equation that is properly reformulated to be applicable to the FDM solver. In this two-step procedure, the nTRACER direct whole core calculation code provides the pin-homogenized group constants from the single assembly level calculations. SPHINCS then performs core calculation based on pin-homogenized group constants with proper super-homogenization (SPH) factors and leakage feedback methods.

In this work, the C5G7-TD benchmark problems are solved by both SPHINCS and nTRACER at the whole core level. The nTRACER whole core solution which employs faithful models of the core configuration and transient control parameters is used as the reference for the assessment of the SPHINCS results. The SPHINCS results obtained with pin-homogenized group constants updated with properly generated SPH factors proves the soundness of pin-homogenized SP3 transient calculations. This comparison involves obviously the comparison of the state-state results which confirms the soundness of nTRACER-SPHINCS two-step core analyses system for both steady-state and transient calculations.

Calculation of PWR MOX/UO2 Benchmark using Serpent2/ADPRES

Muhammad Imron1, Donny Hartanto2

1Abu Dhabi Polytechnic; 2University of Sharjah

This paper presents the solution of PWR MOX/UO2 transient benchmark by Serpent 2/ADPRES. The presences of MOX fuels and burn-up variation in the benchmark’s reactor core pose challenges for reactor simulators due to severe flux gradient across fuel assemblies. In this work, the two-step method was used, in which the assembly level two-group constants were generated from single assembly calculations using Monte Carlo Serpent 2 code, and later the core calculation was performed using ADPRES nodal reactor simulator. ADPRES is based on Nodal Expansion Method (NEM), it was recently developed in Abu Dhabi Polytechnic and has the capabilities to simulate reactor behaviors during static and transient conditions. Finally, the solutions of Serpent 2/ADPRESS, including multiplication factor, power distribution, control rod worth, and critical boron concentration, were compared against solutions from Serpent 2 full-core where the fuel and cladding are explicitly modeled. The results showed that Serpent 2/ADPRES were able to predict the full-core Monte Carlo solutions very well with reasonable differences.

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