Neutronics benchmark analysis of the EBR-II SHRT-45R with SAC-3D

Lyu, Siyu; Lu, Daogang; Sui, Danting

doi:10.1016/j.nucengdes.2020.110679

Cited by 3 publications

(3 citation statements)

References 3 publications

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“…At first, we assumed that each assembly's boundary condition in the radial direction is a total reflection and kept the axial direction boundary conditions the same as in the core neutronics calculation. en, we get the neutron flux distribution in the axial direction by conducting the onedimensional S N (Discrete ordinates method) calculation and do the needed cross-section homogenization as follows [8]:…”

Section: Neutronics Calculation Modelmentioning

confidence: 99%

Benchmark Analysis on Loss-of-Flow-without-Scram Test of FFTF Using Refined SAC-3D Models

Lyu

Sui

2021

Science and Technology of Nuclear Installations

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The Fast Flux Test Facility (FFTF) is a liquid sodium-cooled nuclear reactor designed by the Westinghouse Electric Corporation for the U.S. Department of Energy. In July 1986, a series of unprotected transients were performed to demonstrate the passive safety of FFTF. Among these, a total of 13 loss-of-flow-without scram (LOFWOS) tests were conducted to confirm the liquid metal reactor safety margins, provide data for computer code validation, and demonstrate the inherent and passive safety benefits of specific design features. In our preliminary work, we have performed relatively coarse modeling of the FFTF. To better predict the transient behavior of FFTF LOFWOS test #13, we modeled it using a more refined thermal-hydraulics model. In this paper, we simulate FFTF LOFWOS test #13 with the system safety analysis code SAC-3D according to the benchmark specifications provided by Argonne National Laboratory (ANL). The simulation range includes the primary and secondary circuits. The reactor core was modeled by the built-in 3D neutronics calculation module and the parallel-channel thermal-hydraulics calculation module. To better predict the reactivity feedback introduced by coolant level variations within the GEMs, a real-time macro cross-section homogenization processing module was developed. The steady-state power distribution was calculated as the transient simulation initial boundary conditions. In general, both the steady-state calculation results and the whole-plant transient behavior predictions are in good agreement with the measured data. The relatively large deviations in transient simulation occur in the outlet temperature predictions of the PIOTA in row 6. It can be preliminarily explained by the reason for neglecting the heat transfer between channels in this model.

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Section: Neutronics Calculation Modelmentioning

confidence: 99%

Benchmark Analysis on Loss-of-Flow-without-Scram Test of FFTF Using Refined SAC-3D Models

Lyu

Sui

2021

Science and Technology of Nuclear Installations

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“…A point neutronic calculation model is useful for analyzing a specific case of a power reactor system. In a large-scale system, a 3D neutronic calculation model was developed to analyze the plant system [15] more accurately. Using a 3D model applied in an established VVER-1000 NESTLE code also enhances the calculation of the spent nuclear fuel of a power reactor [16].…”

Section: Introductionmentioning

confidence: 99%

Improvement of a Code-Based Kartini Reactor Simulator for Education and Training

Sutanto,

Nafisah,

R. Antariksawan

et al. 2023

Int. J. Adv. Sci. Eng. Inf. Technol.

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A code-based Kartini reactor simulator was improved as a facility for the human resource development of a nuclear reactor. The simulator simulates the plant dynamics regarding a change of a control rod position. Reactor operation parameter calculations of the reactor power, coolant flow, and fuel temperatures adopt a one-channel method with assumptions of homogeneous radial power distribution and a cosine function of the axial power distribution. Point reactor kinetics, radial conduction heat transfer, and mass and energy conservation are the calculation code's governing equations. Reactivity feedback due to the coolant density and fuel temperature changes are considered. Reactor pressure is fixed at 1 atm due to an open pool-type research reactor. A graphical user interface was developed to operate the simulator. The operation results of the simulator show that the power calculation agrees well with the experimental data. An accident of excess reactivity due to a control ejection is assumed to happen, causing a positive reactivity insertion of 1.11$. However, the safety criterion of the cladding temperature is satisfied due to the negative reactivity feedback. Besides, early application of 3D virtual reality was carried out to provide an immersive interaction between the users and the virtual Kartini reactor plant. The further development of integrating both the virtual reality and the simulator in the recent Kartini reactor-based internet reactor laboratory is interesting to provide a facility with features of remote as well as immersive education and training.

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“…Various computational analyses of the EBR-II have been performed for different areas of interest, but only limited studies focused on neutronics analysis (e.g., [2][3][4][5]). Furthermore, at the time of this writing, only the EBR-II IRPhEP benchmark authors have published computational analyses based on the new IRPhEP benchmark specifications [1,6,7].…”

Section: Introductionmentioning

confidence: 99%

Nuclear Data Sensitivity Study for the EBR-II Fast Reactor Benchmark Using SCALE with ENDF/B-VII.1 and ENDF/B-VIII.0

Bostelmann

Ilas

Wieselquist

2021

JNE

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The EBR-II benchmark, which was recently included in the International Handbook of Evaluated Reactor Physics Benchmark Experiments, served as a basis for assessing the performance of the SCALE code system for fast reactor analyses. A reference SCALE model was developed based on the benchmark specifications. Great agreement was observed between the eigenvalue calculated with this SCALE model and the benchmark eigenvalue. To identify potential gaps and uncertainties of nuclear data for the simulation of various quantities of interest in fast spectrum systems, sensitivity and uncertainty analyses were performed for the eigenvalue, reactivity effects, and the radial power profile of EBR-II using the two most recent ENDF/B nuclear data library releases. While the nominal results are consistent between the calculations with the different libraries, the uncertainties due to nuclear data vary significantly. The major driver of observed uncertainties is the uncertainty of the 235U (n,γ) reaction. Since the uncertainty of this reaction is significantly reduced in the ENDF/B-VIII.0 library compared to ENDF/B-VII.1, the obtained output uncertainties tend to be smaller in ENDF/B-VIII.0 calculations, although the decrease is partially compensated by increased uncertainties in 235U fission and ν¯.

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Neutronics benchmark analysis of the EBR-II SHRT-45R with SAC-3D

Cited by 3 publications

References 3 publications

Benchmark Analysis on Loss-of-Flow-without-Scram Test of FFTF Using Refined SAC-3D Models

Benchmark Analysis on Loss-of-Flow-without-Scram Test of FFTF Using Refined SAC-3D Models

Improvement of a Code-Based Kartini Reactor Simulator for Education and Training

Nuclear Data Sensitivity Study for the EBR-II Fast Reactor Benchmark Using SCALE with ENDF/B-VII.1 and ENDF/B-VIII.0

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