• Nuclear Engineering and Technology
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• v.51 no.7
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• pp.1715-1720
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• 2019

Jozef Stefan Institute (JSI), TRIGA Mark II reactor employs the homogeneous mixture of uranium and zirconium hydride fuel type. Since its upgrade, a series of fresh fuel steady state experimental benchmarks have been conducted. The benchmark results have provided data for testing computational neutronics codes which are important for reactor design and safety analysis. In this work, we investigated the JSI TRIGA Mark II reactor neutronics characteristics: the effective multiplication factor and two safety parameters, namely the control rod worth and the fuel temperature reactivity coefficient using SuperMC. The modeling and real-time cross section generation methods of SuperMC were evaluated in the investigation. The calculation analysis indicated the following: the effective multiplication factor was influenced by the different cross section data libraries; the control rod worth evaluation was better with Monte Carlo codes; the experimental fuel temperature reactivity coefficient was smaller than calculated results due to change in water temperature. All the results were in good agreement with the experimental values. Hence, SuperMC could be used for the designing and benchmarking of other TRIGA Mark II reactors.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.103-110
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• 2021

In the development of a Risk Monitor probabilistic safety assessment (PSA) model from the basic PSA model of a nuclear power plant, the modeling of common-cause failure (CCF) is very important. At present, some approximate modeling methods are widely used, but there lacks criterion of modeling accuracy and error analysis. In this paper, aiming at ensuring the accuracy of risk assessment and minimizing the Risk Monitor PSA models size, we present three basic issues of CCF model resulted from the changes of a nuclear power plant configuration, put forward corresponding modeling methods, and derive accuracy criteria of CCF modeling based on minimum cut sets and risk indicators according to the requirements of risk monitoring. Finally, a nuclear power plant Risk Monitor PSA model is taken as an example to demonstrate the effectiveness of the proposed modeling method and accuracy criteria, and the application scope of the idea of this paper is also discussed.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3388-3400
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• 2023

Detailed analysis of the neutron pathway through matter inside the nuclear reactor core is exceedingly needed for safety and economic considerations. Due to the constant development of high-performance computing technologies, neutronics analysis using computer codes became more effective and efficient to perform sophisticated neutronics calculations. In this work, a commercial pressurized water reactor (PWR) presented by Virtual Environment for Reactor Applications (VERA) Core Physics Benchmark are modeled and simulated using a high-fidelity simulation of OpenMC code in terms of criticality and fuel pin power distribution. Various problems have been selected from VERA benchmark ranging from a simple two-dimension (2D) pin cell problem to a complex three dimension (3D) full core problem. The development of the code capabilities for reactor physics methods has been implemented to investigate the accuracy and performance of the OpenMC code against VERA SCALE codes. The results of OpenMC code exhibit excellent agreement with VERA results with maximum Root Mean Square Error (RMSE) values of less than 0.04% and 1.3% for the criticality eigenvalues and pin power distributions, respectively. This demonstrates the successful utilization of the OpenMC code as a simulation tool for a whole core analysis. Further works are undergoing on the accuracy of OpenMC simulations for the impact of different fuel types and burnup levels and the analysis of the transient behavior and coupled thermal hydraulic feedback.

• Nuclear Engineering and Technology
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• v.46 no.3
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• pp.313-342
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• 2014

A Nuclear Energy Agency (NEA), Organization for Economic Co-operation and Development (OECD) benchmark for Uncertainty Analysis in Modeling (UAM) is defined in order to facilitate the development and validation of available uncertainty analysis and sensitivity analysis methods for best-estimate Light water Reactor (LWR) design and safety calculations. The benchmark has been named the OECD/NEA UAM-LWR benchmark, and has been divided into three phases each of which focuses on a different portion of the uncertainty propagation in LWR multi-physics and multi-scale analysis. Several different reactor cases are modeled at various phases of a reactor calculation. This paper discusses Phase I, known as the "Neutronics Phase", which is devoted mostly to the propagation of nuclear data (cross-section) uncertainty throughout steady-state stand-alone neutronics core calculations. Three reactor systems (for which design, operation and measured data are available) are rigorously studied in this benchmark: Peach Bottom Unit 2 BWR, Three Mile Island Unit 1 PWR, and VVER-1000 Kozloduy-6/Kalinin-3. Additional measured data is analyzed such as the KRITZ LEU criticality experiments and the SNEAK-7A and 7B experiments of the Karlsruhe Fast Critical Facility. Analyzed results include the top five neutron-nuclide reactions, which contribute the most to the prediction uncertainty in keff, as well as the uncertainty in key parameters of neutronics analysis such as microscopic and macroscopic cross-sections, six-group decay constants, assembly discontinuity factors, and axial and radial core power distributions. Conclusions are drawn regarding where further studies should be done to reduce uncertainties in key nuclide reaction uncertainties (i.e.: $^{238}U$ radiative capture and inelastic scattering (n, n') as well as the average number of neutrons released per fission event of $^{239}Pu$).

• Nuclear Engineering and Technology
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• v.54 no.2
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• pp.532-545
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• 2022

In-depth convergence analyses for neutronics/thermal-hydraulics (T/H) coupled calculations are performed to investigate the performance of nonlinear methods based on the Fixed-Point Iteration (FPI). A simplified neutronics-T/H coupled system consisting of a single fuel pin is derived to provide a testbed. The xenon equilibrium model is considered to investigate its impact during the nonlinear iteration. A problem set is organized to have a thousand different fuel temperature coefficients (FTC) and moderator temperature coefficients (MTC). The problem set is solved by the Jacobi and Gauss-Seidel (G-S) type FPI. The relaxation scheme and the Anderson acceleration are applied to improve the convergence rate of FPI. The performances of solution schemes are evaluated by comparing the number of iterations and the error reduction behavior. From those numerical investigations, it is demonstrated that the number of FPIs is increased as the feedback is stronger regardless of its sign. In addition, the Jacobi type FPIs generally shows a slower convergence rate than the G-S type FPI. It also turns out that the xenon equilibrium model can cause numerical instability for certain conditions. Lastly, it is figured out that the Anderson acceleration can effectively improve the convergence behaviors of FPI, compared to the conventional relaxation scheme.

• Nuclear Engineering and Technology
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• v.42 no.5
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• pp.520-545
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• 2010

This paper presents the neutronics modeling capabilities of the fast reactor simulation system SHARP, which ANL is developing as part of the U.S. DOE's NEAMS program. We discuss the three transport solvers (PN2ND, SN2ND, and MOCFE) implemented in the UNIC code along with the multigroup cross section generation code $MC^2$-3. We describe the solution methods and modeling capabilities, and discuss the improvement needs for each solver, focusing on massively parallel computation. We present the performance test results against various benchmark problems and ZPR-6 and ZPPR critical experiments. We also discuss weak and strong scalability results for the SN2ND solver on the ZPR-6 critical assembly benchmarks.

• Nuclear Engineering and Technology
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• v.44 no.2
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• pp.177-198
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• 2012

This paper reviews the fast reactor physics and computational methods. The basic reactor physics specific to fast spectrum reactors are briefly reviewed, focused on fissile material breeding and actinide burning. Design implications and reactivity feedback characteristics are compared between breeder and burner reactors. Some discussions are given to the distinct nuclear characteristics of fast reactors that make the assumptions employed in traditional LWR analysis methods not applicable. Reactor physics analysis codes used for the modeling of fast reactor designs in the U.S. are reviewed. This review covers cross-section generation capabilities, whole-core deterministic (diffusion and transport) and Monte Carlo calculation tools, depletion and fuel cycle analysis codes, perturbation theory codes for reactivity coefficient calculation and cross section sensitivity analysis, and uncertainty analysis codes.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.933-940
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• 2024

A macroscopic cross section generation model was developed for the conceptual horizontal, compact high temperature gas reactor (HC-HTGR). Because there are many sources of spectral effects in the design and analysis of the core, conventional LWR methods have limitations for accurate simulation of the HC-HTGR using a neutron diffusion core neutronics simulator. Several super-cell model configurations were investigated to consider the spectral effect of neighboring cells. A new history variable was introduced for the existing library format to more accurately account for the history effect from neighboring nodes and reactivity control drums. The macroscopic cross section library was validated through comparison with cross sections generated using full core Monte Carlo models and single cell cross section for both 3D core steady-state problems and 2D and 3D depletion problems. Core calculations were then performed with the AGREE HTR neutronics and thermal-fluid core simulator using super-cell cross sections. With the new history variable, the super-cell cross sections were in good agreement with the full core cross sections even for problems with significant spectrum change during fuel shuffling and depletion.

• Nuclear Engineering and Technology
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• v.37 no.1
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• pp.25-78
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• 2005

As a key and core knowledge for the design of various types of nuclear reactors, the discipline of reactor physics has been advanced continually in the past six decades and has led to a very sophisticated fabric of analysis methods and computer codes in use today. Notwithstanding, the discipline faces interesting challenges from next-generation nuclear reactors and innovative new fuel designs in the coming. After presenting a brief overview of important tasks and steps involved in the nuclear design and analysis of a reactor, this article focuses on the currently-used design and analysis methods, issues and limitations, and current activities to resolve them as follows: (1) Derivation of the multi group transport equations and the multi group diffusion equations, with representative solution methods thereof. (2) Elements of modem (now almost three decades old) diffusion nodal methods. (3) Limitations of nodal methods such as transverse integration, flux reconstruction, and analysis of UO2-MOX mixed cores. Homogenization and related issues. (4) Description of the analytic function expansion nodal (AFEN) method. (5) Ongoing efforts for three-dimensional whole-core heterogeneous transport calculations and acceleration methods. (6) Elements of spatial kinetics calculation methods and coupled neutronics and thermal-hydraulics transient analysis. (7) Identification of future research and development areas in advanced reactors and Generation-IV reactors, in particular, in very high temperature gas reactor (VHTR) cores.

• Nuclear Engineering and Technology
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• v.55 no.1
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• pp.310-323
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• 2023

The k-eigenvalue neutronics/thermal-hydraulics coupling calculation is a key issue for reactor design and analysis. Jacobian-free Newton-Krylov (JFNK) method, featured with super-linear convergence rate and high efficiency, has been attracting more and more attention to solve the multi-physics coupling problem. However, it may converge to the high-order eigenmode because of the multiple solutions nature of the k-eigenvalue form of multi-physics coupling issue. Based on our previous work, a modified JFNK with a line search method is proposed in this work, which can find the fundamental eigenmode together with thermal-hydraulics feedback in a wide range of initial values. In detail, the existing modified JFNK method is combined with the line search strategy, so that the intermediate iterative solution can avoid a sudden divergence and be adjusted into a convergence basin smoothly. Two simplified 2-D homogeneous reactor models, a PWR model, and an HTR model, are utilized to evaluate the performance of the newly proposed JFNK method. The results show that the performance of this proposed JFNK is more robust than the existing JFNK-based methods.