High fidelity simulation of conventional and innovative LWR with the coupled Monte-Carlo thermal-hydraulic system MCNP-SUBCHANFLOW

Ivanov, A.; Sánchez, V.; Ivanov, Kostadin

doi:10.1016/j.nucengdes.2013.05.008

Cited by 26 publications

(5 citation statements)

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“…Although we mostly avoid discussing stochastic transport in this thesis, we mention that several Monte Carlo transport codes have been loosely coupled to thermal hydraulics solvers. For example, the OpenMC [78], Serpent [22], MCNP [37], and MC21 [45] codes each have proof-of-concept implementations.…”

Section: Multiphysics Iterative Methodsmentioning

confidence: 99%

The NILO-CMFD Method for Iteratively Solving Coupled Neutron Transport–Thermal Hydraulics Problems

Adamowicz

Manera

Larsen

2022

Nuclear Science and Engineering

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The interactions between neutron radiation and matter are nonlinear. Specifically, certain neutron nuclear reactions (e.g., fission events) generate heat, which modifies the temperature and density of the surrounding material, which in turn modify the neutron interaction probabilities (through the macroscopic cross sections). This nonlinear process is referred to as neutron transport with thermal hydraulics feedback.Linear transport problems (without feedback) are accurately described by the Neutron Transport Equation (NTE). Discretized forms of the NTE are routinely solved using iterative transport acceleration schemes, for example, the standard Coarse Mesh Finite Difference (CMFD) method. CMFD and related acceleration schemes for linear transport problems are reliable and well-understood. However, when these same iterative methods are applied to nonlinear (i.e. multiphysics) problems, significant performance and stability issues often occur.In this thesis, we propose modifications to the CMFD procedure, so that it can more robustly and efficiently solve loosely-coupled neutron transport -thermal hydraulics multiphysics problems. We refer to the new method as Nonlinearly Implicit Low-Order CMFD (NILO-CMFD). In this method, the transport-corrected diffusion equationa foundational component of the CMFD method -is modifed to include approximate thermal hydraulics and nuclear data update physics.To begin, we provide a general derivation of the 3-D NILO-CMFD method. Then, to initially analyze and test the method, we develop an approximate 1-D multiphysics model of a 3-D nuclear reactor fuel pin. The 1-D model consists of: (i) a 1-D neutron transport equation describing neutron transport, (ii) a 1-D advection-diffusion equation describing fluid temperature variation, (iii) a fuel temperature that depends on the 1-D fission heat source, and (iv) model equations that define the neutron cross sections and related nuclear data in terms of the fluid and fuel temperatures. The implementation of this model in a 1-D test code shows that the NILO-CMFD method overcomes the instabilities that occur in the standard Relaxed-CMFD (R-CMFD) method.x To test the new NILO-CMFD method on realistic 3-D problems, we implemented an incomplete version of the method, NILO-A, in the 3-D neutron transport code MPACT. For multiphysics problems, MPACT is loosely coupled to the 3-D COBRA-TF (CTF) code, which performs subchannel thermal hydraulics calculations. In every 3-D problem that we ran, NILO-CMFD successfully reduced the number of outer iterations to the low number required by CMFD for single-physics calculations. Unfortunately, due to lack of time, we could not implement the full NILO-CMFD method, which would have significantly reduced the wall clock time of the simulations. Consequently, some of our 3-D simulations required more wall clock time to converge than R-CMFD.Nonetheless, our results show that the NILO-CMFD method is a legitimate generalization of single-physics CMFD to multiphysics problems. The number of outer iterations requi...

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Section: Multiphysics Iterative Methodsmentioning

confidence: 99%

The NILO-CMFD Method for Iteratively Solving Coupled Neutron Transport–Thermal Hydraulics Problems

Adamowicz

Manera

Larsen

2022

Nuclear Science and Engineering

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“…Besides sub-channel thermal-hydraulic tools, computational fluid dynamics codes have been coupled to deterministic neutron transport codes to analyze up to a pressurized water reactor (PWR) core with quarter symmetry under steady-state conditions (Kochuas et al, 2012;Weber et al, 2007). Numerous groups work on coupling Monte Carlo (MC) neutron transport methods with thermal-hydraulic tools (Bernnat et al, 2012;Espel et al, 2013;Ivanov et al, 2013Ivanov et al, , 2014Li and Wang, 2012;Vazquez et al, 2012).…”

Section: Introductionmentioning

confidence: 99%

Static and transient pin-by-pin simulations of a full PWR core with the extended coupled code system DYNSUB

Daeubler

Trost

Jiménez

et al. 2015

Annals of Nuclear Energy

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“…Besides sub-channel thermalhydraulic tools, computational fluid dynamics (CFD) codes have been used together with deterministic neutron transport codes to analyze a pressurized water reactor (PWR) core with quarter symmetry under steady-state conditions (Weber et al, 2007;Kochuas et al, 2012). Many groups couple Monte Carlo (MC) neutron transport methods to thermal-hydraulic tools (Espel et al, 2013;Li and Wang, 2012;Bernnat et al, 2012;Vazquez et al, 2012;Ivanov et al, 2013Ivanov et al, , 2014Gill et al, 2014). Most of these coupling efforts utilize the MC code MCNP.…”

Section: Introductionmentioning

confidence: 99%

High-fidelity coupled Monte Carlo neutron transport and thermal-hydraulic simulations using Serpent 2/SUBCHANFLOW

Daeubler

Ivanov

Sjenitzer

et al. 2015

Annals of Nuclear Energy

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High fidelity simulation of conventional and innovative LWR with the coupled Monte-Carlo thermal-hydraulic system MCNP-SUBCHANFLOW

Cited by 26 publications

References 1 publication

The NILO-CMFD Method for Iteratively Solving Coupled Neutron Transport–Thermal Hydraulics Problems

The NILO-CMFD Method for Iteratively Solving Coupled Neutron Transport–Thermal Hydraulics Problems

Static and transient pin-by-pin simulations of a full PWR core with the extended coupled code system DYNSUB

High-fidelity coupled Monte Carlo neutron transport and thermal-hydraulic simulations using Serpent 2/SUBCHANFLOW

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