Papers by Dumitru Serghiuta
Nuclear science and engineering, Apr 12, 2024
In this paper, results related to the reconstruction of intra-bundle fission density profile for ... more In this paper, results related to the reconstruction of intra-bundle fission density profile for a 37-pin CANDU-6 bundle with the highest enthalpy deposition during a postulated large LOCA stagnation break in a Bruce B core are presented. Bruce B is a nuclear power plant in Kincardine, Ontario (Canada)), on the shores of Lake Huron with 4 CANDU reactors that are rated at about 750 MWe. The reconstruction of the fuel pin fission densities is based on steady-state, three-dimensional simulations with the Monte Carlo code MCNP for a subset of 27 out of 69 time steps during the first two seconds of the power pulse predicted for the fuel bundle at core location V13/8. Two-group cross section data libraries are generated for MCNP at each time step by the lattice depletion neutron transport code HELIOS-1.7. To include the effect of the surrounding core environment, the calculations are performed with time-dependent albedo boundary conditions inferred from a full core simulation of the transient by the nodal diffusion code NESTLE with HELIOS homogenized cross-sections. It is found that the local peaking factor (LPF) in the outer ring varies during the transient, but never exceeds its value before the transient. Inclusion of the coremore » environment increases the LPF in the outer ring. For the analyzed case, the increase is 0.72% with a relative error of 0.01% for the LPF before the transient and 0.55% (with a relative error of 0.01%) for the maximum average LPF during the transient. The latter is based on only four selected transient time points. Note that the immediate environment of the 'hot bundle' does not contain any reactivity devices or other perturbing factors. As a result, the increases observed in the LPF in the outer ring may not be representative of the situations in which 'other' core environment perturbing factors are present. To determine the effect of these factors on the LPF, further analyses of a bundle in the proximity of control devices should be carried out. (authors)« less
Title: MCNP Simulation of Void Reactivity in a Simplified CANDU Core Sub-region. Authors: Rahnema... more Title: MCNP Simulation of Void Reactivity in a Simplified CANDU Core Sub-region. Authors: Rahnema, F.; Mosher, S.; Pitts, M.; Akhtar, P.; Serghiuta, D. Publication: Advanced Monte Carlo for Radiation Physics, Particle Transport Simulation and Applications. ...
Annals of Nuclear Energy, 2019
On the path to deployment of any reactor, modeling tool verification and validation is a key step... more On the path to deployment of any reactor, modeling tool verification and validation is a key step. Fluoride-Salt-Cooled High Temperature Reactors (FHR) pose challenges to neutronics modeling and simulation tools due to several design features of the reactor type. This paper presents a categorized list of phenomena that pose challenges to FHR modeling using current neutronics tools. These phenomena are presented in four categories: Fundamental Cross Section Data, Material Composition, Computational Methodology, and General Depletion. A short path forward for these phenomena is also presented. In addition, a discussion of the resulting gaps in current codes is presented.
Epj Web of Conferences, 2021
This paper reports on the development and testing of a comprehensive few-group cross section inpu... more This paper reports on the development and testing of a comprehensive few-group cross section input uncertainty library for the NESTLE-C nodal diffusion-based nuclear reactor core simulator. This library represents the first milestone of a first-of-a-kind framework for the integrated characterization of uncertainties in steady-state and transient CANDU reactor simulations. The objective of this framework is to propagate, prioritize and devise a mapping capability for uncertainties in support of model validation of best-estimate calculations. A complete framework would factor both input and modeling uncertainty contributions. The scope of the present work is limited to the propagation of multi-group cross-section uncertainties through lattice physics calculations down to the few-group format, representing the input to the NESTLE-C core simulator, and finally to core responses of interest.
Journal of Nuclear Engineering and Radiation Science, Jun 17, 2016
EPJ Web of Conferences, 2021
This paper reports on the development and testing of a comprehensive few-group cross section inpu... more This paper reports on the development and testing of a comprehensive few-group cross section input uncertainty library for the NESTLE-C nodal diffusion-based nuclear reactor core simulator. This library represents the first milestone of a first-of-a-kind framework for the integrated characterization of uncertainties in steady-state and transient CANDU reactor simulations. The objective of this framework is to propagate, prioritize and devise a mapping capability for uncertainties in support of model validation of best-estimate calculations. A complete framework would factor both input and modeling uncertainty contributions. The scope of the present work is limited to the propagation of multi-group cross-section uncertainties through lattice physics calculations down to the few-group format, representing the input to the NESTLE-C core simulator, and finally to core responses of interest.
Nuclear Technology, 2019
Abstract This paper reviews the attributes and challenges of applying the functional failure conc... more Abstract This paper reviews the attributes and challenges of applying the functional failure concept and the use of Best-Estimate Plus Uncertainty methods in evaluating protective systems in the risk space. As an illustrative example, the paper uses the case of the effectiveness of CANada Deuterium Uranium (CANDU) reactor shutdown systems. A risk-informed formulation is first introduced for estimation of a reasonable limit for functional failure probability using the Swiss Cheese model. In the real application, there are several challenges in realistically estimating probabilities of exceeding a prescribed design or regulatory limit. Key challenges discussed in this critical review include the use of complex, computationally intensive predictive models; modeling completeness; assumptions about input distributions; validation; separation of uncertainties; and selection of statistical model and algorithms. The use of hybrid deterministic-probabilistic methods may address these challenges to a certain extent.
Annals of Nuclear Energy, 2019
On the path to deployment of any reactor, modeling tool verification and validation is a key step... more On the path to deployment of any reactor, modeling tool verification and validation is a key step. Fluoride-Salt-Cooled High Temperature Reactors (FHR) pose challenges to neutronics modeling and simulation tools due to several design features of the reactor type. This paper presents a categorized list of phenomena that pose challenges to FHR modeling using current neutronics tools. These phenomena are presented in four categories: Fundamental Cross Section Data, Material Composition, Computational Methodology, and General Depletion. A short path forward for these phenomena is also presented. In addition, a discussion of the resulting gaps in current codes is presented.
Advanced Monte Carlo for Radiation Physics, Particle Transport Simulation and Applications, 2001
Title: MCNP Simulation of Void Reactivity in a Simplified CANDU Core Sub-region. Authors: Rahnema... more Title: MCNP Simulation of Void Reactivity in a Simplified CANDU Core Sub-region. Authors: Rahnema, F.; Mosher, S.; Pitts, M.; Akhtar, P.; Serghiuta, D. Publication: Advanced Monte Carlo for Radiation Physics, Particle Transport Simulation and Applications. ...
Journal of Nuclear Engineering and Radiation Science, 2016
Transactions of the American Nuclear Society, 2008
In this paper, results related to a limited scope assessment of the geometry-distortion-induced e... more In this paper, results related to a limited scope assessment of the geometry-distortion-induced effects on key reactor physics parameters of a CANDU reactor are discussed. These results were generated by simulations using refined analytical methods and detailed modeling of CANDU reactor core with aged lattice cell geometry. (authors)
Volume 1: Plant Operations, Maintenance, Engineering, Modifications, Life Cycle and Balance of Plant; Nuclear Fuel and Materials; Plant Systems, Structures and Components; Codes, Standards, Licensing and Regulatory Issues, 2014
ABSTRACT
Nuclear Engineering and Design, 2015
h i g h l i g h t s • CANDU-type-lattice kinetics parameters are calculated using different adjoi... more h i g h l i g h t s • CANDU-type-lattice kinetics parameters are calculated using different adjoint-weighting approximations at different burnups. • Fine-group space-dependent adjoint weighting is the most accurate method of calculating the kinetics parameters. • Two-group lattice-homogenized adjoint weighting overestimates the effective delayed-neutron fraction by approximately 5%. • Fine-group lattice-homogenized adjoint weighting overestimates the effective delayed neutron fraction only by approximately 2%.
Annals of Nuclear Energy, 2011
A 3D stylized half-core Canadian deuterium uranium (CANDU) reactor benchmark problem is presented... more A 3D stylized half-core Canadian deuterium uranium (CANDU) reactor benchmark problem is presented. The benchmark problem is comprised of a heterogeneous lattice of 37-element natural uranium fuel bundles, heavy water moderated, heavy water cooled, with adjuster rods included as reactivity control devices. Furthermore, a 2-group macroscopic cross section library has been developed for the problem to increase the utility of this benchmark for full-core deterministic transport methods development. Monte Carlo results are presented for the benchmark problem in cooled, checkerboard void, and full coolant void configurations.
Nuclear Engineering and Design, May 1, 2015
h i g h l i g h t s • CANDU-type-lattice kinetics parameters are calculated using different adjoi... more h i g h l i g h t s • CANDU-type-lattice kinetics parameters are calculated using different adjoint-weighting approximations at different burnups. • Fine-group space-dependent adjoint weighting is the most accurate method of calculating the kinetics parameters. • Two-group lattice-homogenized adjoint weighting overestimates the effective delayed-neutron fraction by approximately 5%. • Fine-group lattice-homogenized adjoint weighting overestimates the effective delayed neutron fraction only by approximately 2%.
Annals of Nuclear Energy, 2011
ABSTRACT
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Papers by Dumitru Serghiuta