• Nuclear Engineering and Technology
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• 제55권5호
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• pp.1687-1707
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• 2023

This paper presents a molten salt reactor (MSR) design parameter sensitivity study using a nodal dynamic modelling methodology with explicitly modified point kinetics equation and Mann's model for heat transfer. Six parameters that can impact MSR safety are evaluated. A MATLAB-Simulink model inspired by Thorcon's 550MWth MSR is used for parameter evaluations. A safety envelope was formed to encapsulate power, maximum and minimum temperature, and temperature-induced reactivity feedback. The parameters are perturbed by ±30%. The parameters were then ranked by their subsequent impact on the considered safety envelope, which ranks acceptable parameter uncertainty. The model is openly available on GitHub.

• Nuclear Engineering and Technology
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• 제55권10호
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• pp.3549-3558
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• 2023

The jet pump can be used in a test device of a nuclear reactor for high flow amplification as it reduces inlet flow requirement and thereby size of the process components. In the present work, a miniature jet pump was designed to meet high flow amplification greater than 3. Subsequently, experiments were carried out using a test setup for design validation and performance evaluation of the jet pump for different parameters. It was observed that a minimum pressure of 0.6 bar (g) was required for the secondary fluid inside the jet pump to ensure cavitation free performance at high amplification. Spacing between the nozzle tip and the mixing chamber entry point had significant effect on the performance of the jet pump. Variation in primary flow, temperature and area ratio also affected the performance. It was observed that at high flow amplification, the analytical solution differed significantly from experimental results due to very large velocities encountered in the miniature size jet pump.

• Nuclear Engineering and Technology
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• 제48권5호
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• pp.1109-1119
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• 2016

Although, the direct use of spent pressurized water reactor (PWR) fuel in CANda Deuterium Uranium (CANDU) reactors (DUPIC) cycle is still under investigation, DUPIC cycle is a promising method for uranium utilization improvement, for reduction of high level nuclear waste, and for high degree of proliferation resistance. This paper focuses on the effect of DUPIC cycle on CANDU reactor safety parameters. MCNP6 was used for lattice cell simulation of a typical 3,411 MWth PWR fueled by $UO_2$ enriched to 4.5w/o U-235 to calculate the spent fuel inventories after a burnup of 51.7 MWd/kgU. The code was also used to simulate the lattice cell of CANDU-6 reactor fueled with spent fuel after its fabrication into the standard 37-element fuel bundle. It is assumed a 5-year cooling time between the spent fuel discharges from the PWR to the loading into the CANDU-6. The simulation was carried out to calculate the burnup and the effect of DUPIC fuel on: (1) the power distribution amongst the fuel elements of the bundle; (2) the coolant void reactivity; and (3) the reactor point-kinetics parameters.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.4003-4013
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• 2021

During the long-term cooling (LTC) phase of a loss-of-coolant accident (LOCA) in a pressurized water reactor (PWR), water is supplied from the containment sump to the reactor coolant system (RCS) by the flooded sump water to the Reactor Vessel (RV) through the broken pipes. As part of the technical efforts for resolving GSI-191 [( Reid and Crytzer, May. 2007) 1, consideration is needed for the consequences of debris penetrating the sump screen and propagating downstream into the RV. Injection of debris (fiberglass) into the RV during the LTC recirculation phase needs special attention to assure that reactor core cooling is maintained. The point of concern is the potential for debris to adversely affect the reactor core flow paths or heat transfer [2]. However, all the experiments for proving the coolability of RV have been done with the assumption of the most of debris would be transferred to the RV and the bottom nozzle of the FAs. The purpose of the tests is to quantify the amount of the debris that would be accumulated at the lower plenum and the debris that passes through the FAs since non-conservatism of other researches assumptions that have been used in the past experimental or analytical programs.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2803-2815
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• 2021

This paper addresses the uncertainty quantification and sensitivity analysis of a depleted light-water fuel assembly of the Turkey Point-3 benchmark. The uncertainty of the fuel assembly decay heat and isotopic densities is quantified with respect to three different groups of diverse parameters: nuclear data, assembly design, and reactor core operation. The uncertainty propagation is conducted using a two-step analysis code system comprising the lattice code STREAM, nodal code RAST-K, and spent nuclear fuel module SNF through the random sampling of microscopic cross-sections, fuel rod sizes, number densities, reactor core total power, and temperature distributions. Overall, the statistical analysis of the calculated samples demonstrates that the decay heat uncertainty decreases with the cooling time. The nuclear data and assembly design parameters are proven to be the largest contributors to the decay heat uncertainty, whereas the reactor core power and inlet coolant temperature have a minor effect. The majority of the decay heat uncertainties are delivered by a small number of isotopes such as ²⁴¹Am, ¹³⁷Ba, ²⁴⁴Cm, ²³⁸Pu, and ⁹⁰Y.

• Nuclear Engineering and Technology
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• 제54권3호
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• pp.940-947
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• 2022

Meta-heuristic algorithms have found their place in optimization problems. Henry gas solubility optimization (HGSO) is one of the newest population-based algorithms. This algorithm is inspired by Henry's law of physics. To evaluate the performance of a new algorithm, it must be used in various problems. On the other hand, the optimization of the proportional-integral-derivative (PID) gains for load-following of a nuclear power plant (NPP) is a good challenge to assess the performance of HGSO. Accordingly, the power control of a pressurized water reactor (PWR) is targeted, based on the point kinetics model with six groups of delayed-neutron precursors. In any optimization problem based on meta-heuristic algorithms, an efficient objective function is required. Therefore, the integral of the time-weighted square error (ITSE) performance index is utilized as the objective (cost) function of HGSO, which is constrained by a stability criterion in steady-state operations. A Lyapunov approach guarantees this stability. The results show that this method provides superior results compared to an empirically tuned PID controller with the least error. It also achieves good accuracy compared to an established GA-tuned PID controller.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1350-1364
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• 2023

2D/1D coupling method is an important neutron transport calculation method due to its high accuracy and relatively low computation cost. However, 2D/1D coupling method may diverge especially in small axial mesh size. To analyze the convergence behavior of 2D/1D coupling method, a Fourier analysis for k-eigenvalue neutron transport problems is implemented. The analysis results present the divergence problem of 2D/1D coupling method in small axial mesh size. Several common attempts are made to solve the divergence problem, which are to increase the number of inner iterations of the 2D or 1D calculation, and two times 1D calculations per outer iteration. However, these attempts only could improve the convergence rate but cannot deal with the divergence problem of 2D/1D coupling method thoroughly. Moreover, the choice of axial solvers, such as DGFEM SN and traditional SN, and its effect on the convergence behavior are also discussed. The results show that the choice of axial solver is a key point for the convergence of 2D/1D method. The DGFEM SN based 2D/1D method could converge within a wide range of optical thickness region, which is superior to that of traditional SN method.

• Nuclear Engineering and Technology
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• 제52권7호
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• pp.1443-1451
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• 2020

The objective of this study is to design a robust power control system for a small pressurized water reactor (PWR) to achieve stable power operations under conditions of external disturbances and internal model uncertainties. For this purpose, the multiple-input multiple-output transfer function models of the reactor core at five power levels are derived from point reactor kinetics equations and the Mann's thermodynamic model. Using the transfer function models, five local reactor power controllers are designed using an H infinity (H_∞) mixed sensitivity method to minimize the core power disturbance under various uncertainties at the five power levels, respectively. Then a multimodel approach with triangular membership functions is employed to integrate the five local controllers into a multimodel robust control system that is applicable for the entire power range. The performance of the robust power system is assessed against 10% of full power (FP) step load increase transients with coolant inlet temperature disturbances at different power levels and large-scope, rapid ramp load change transient. The simulation results show that the robust control system could maintain satisfactory control performance and good robustness of the reactor under external disturbances and internal model uncertainties, demonstrating the effective of the robust power control design.

• Nuclear Engineering and Technology
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• 제50권7호
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• pp.1051-1059
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• 2018

Fast periodic pulsed reactor is a type of reactor in which the fission bursts are formed entirely with external reactivity modulation with a specified time periodicity. This type of reactors could generate much larger intensity of neutron beams for experimental use, compared with the steady state reactors. In the design of fast periodic pulsed reactors, the time dependent simulation of the power pulse is majorly based on a point kinetic model, which is known to have limitations. A more accurate calculation method is desired for the design analyses of fast periodic pulsed reactors. Monte Carlo computer code MCNP6 is used for this task due to its three dimensional transport capability with a continuous energy library. Some new routines were added to simulate the rotation of the movable reflector parts in the time dependent calculation. Fast periodic pulsed reactor IBR-2M was utilized to validate the new routines. This reactor is periodically in prompt supercritical state, which lasts for ${\sim}400{\mu}s$, during the equilibrium state. This generates long neutron fission chains, which requires tremendously large amount of computation time during Monte Carlo simulations. Russian Roulette was applied for these very long neutron chains in MCNP6 calculation, combined with other approaches to improve the efficiency of the simulations. In the power pulse of the IBR-2M at equilibrium state, there is some discrepancy between the experimental measurements and the calculated results using the point kinetics model. MCNP6 results matches better the experimental measurements, which shows the merit of using MCNP6 calculation relative to the point kinetics model.

• Nuclear Engineering and Technology
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• 제54권12호
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• pp.4393-4411
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• 2022

In recent years, beryllium oxide has been widely utilized in multiple compact nuclear reactors as the neutron moderator, the neutron reflector or the matrix material with dispersed nuclear fuels due to its prominent properties. In the past 70 years, beryllium oxide has been studied extensively, but rarely been systematically organized. This article provides a systematic review of the application history, thermal properties, mechanical properties, corrosion behavior and fabrication methods of beryllium oxide. Data from previous literature are extracted and sorted out, and all of these original data are attached as the supplementary material, so that subsequent researchers can utilize this paper as a database for beryllium oxide research in reactor design or simulation analysis, etc. In addition, this review article also attempts to point out the insufficiency of research on beryllium oxide, and the possible key research areas about beryllium oxide in the future.