The balance between sustainability, energy security and affordability are important trade-offs to consider in decarbonizing the current energy system. The two practical alternatives for moving forward seem to be either reducing the energy storage costs to enable deployment the intermittent renewables on a large scale or developing an affordable and more flexible nuclear power. A small modular boiling water reactor combined with external superheaters offers a significant improvement to the conventional nuclear system. The potential benefits include improvement in cycle thermal efficiency, reduction in the size of the vessel, and the capability to adjust load while maintaining the reactor operation at 100% of its full power. In this paper, the conceptual design of a Small Modular Boiling Water Reactor (SMBWR) combined with external superheaters is presented along with investigation into some of its core design performance characteristics. It is found that the 4-batch in-core fuel management scheme offers a more favorable performance compared to the 3-batch scheme as it has lower power peaking, less excess reactivity, and more negative coolant void coefficient (CVC). The combination of a multi-batch fuel arrangement, coolant temperature variation, and control rods are required to control the reactivity swing in the SMBWR while keeping the power peaking below the safety limit throughout the depletion cycle. (C) 2021 Elsevier Ltd. All rights reserved.

The hybrid Small Modular Boiling Water Reactor (SMBWR) is a new conceptual design of BWR-type SMR. The main features of SMBWR include a natural circulation loop in its coolant recirculation system and external superheaters system integrated into the steam cycle. A full core analysis of SMBWR is performed with the nodal diffusion code PANTHER using homogenised constant libraries generated by WIMS. The study compared a number of core geometry configurations and fuel management schemes to suppress excess reactivity throughout fuel depletion. Three options for SMBWR core aspect ratio using the same power density are investigated with the aim to assess the effect on the neutronic and thermal-hydraulic performance of the SMBWR. It is found that the thin and tall core configuration (192 fuel assemblies and 3.60 m) showed the least favourable performance out of the three options as it has the largest core pressure drop and thus requires taller chimney to develop natural circulation.

With the increased role of intermittent renewables in the energy mix, it is important for the Nuclear Power Plants (NPPs) to develop flexible load-follow capabilities in order to be economically competitive. By combining NPP with external superheaters, there is possibility to increase power conversion cycle efficiency and reduce load to some extent while maintaining the nuclear reactor operation at 100% of its rated power and thus maximizing its economic value. In this paper, parametric study of a Small Modular Boiling Water Reactor (SMBWR) conceptual design combined with external superheaters is presented quantifying its performance at different system pressure. The first part of the work is focusing on how operating pressure will affect the SMBWR system as a whole, which includes neutronics, thermal hydraulics, and thermodynamics. WIMS-PANTHER package and COBRA-EN were used as the neutronic and thermal hydraulic tools. In addition, MATLAB models of the SMBWR natural circulation loop and balance of plant were developed for the study. The second part of the work examines the maneuvering capability of SMBWR. It is found that increasing the pressure from 6.5 to 10.0 MPa has no significant neutronic effects, while thermal efficiency is slightly improved. It is also found that SMBWR is able to reduce the load to 65% while maintaining the reactor power at 100% of its rated value.

It is known that the Pressurized Water Reactors (PWRs), which are the most common type of nuclear reactor existing today, usually used to provide a base load electricity. In order to be able to compete with other generation types (fossil and renewables), it would be desirable to enable PWRs with a load following capabilities to cope with varying electricity demand, especially in deregulated markets. The thermal efficiency of PWRs can be increased by fitting the power plant with conventional fossil fuel superheaters. This hybrid system has been hypothesised to be able to adjust the power output and the cycle efficiency of PWRs. Such mode of operation would also improve the efficiency of converting the fossil fuel heat because it is applied only at the superheater stage. There are several ways to supply the heat to the superheaters, for example, by using the exhaust gas from the gas turbines and using the conventional gas burner. In this paper, the thermodynamic performance of the hybrid systems (PWR with superheater) is investigated for large reactor and Small Modular Reactor (SMR) application. The thermal efficiency of the AP1000 can be improved from 30.2% to 45.8% (with CCGT), 35.6% (with gas burner), and 36.6% (gas burner with reheating). The thermal efficiency of the SMR can be improved from 33.4% to nearly 45% (with CCGT), 35.5% (with gas burner), and 37.4% (gas burner with reheating). The analysis results show that it is possible for the hybrid system to operate between 65% and the full power load.

It is known that the Pressurized Water Reactors (PWRs), which are the most common type of nuclear reactor existing today, usually used to provide a base load electricity. In order to be able to compete with other generation types (fossil and renewables), it would be desirable to develop PWRs with flexible load following capabilities to cope with varying electricity demand, especially in deregulated markets.The thermal efficiency of PWRs can be increased by fitting the power plant with conventional fossil fuel superheaters. This hybrid system has been hypothesised to be able to adjust the power output and the cycle efficiency of PWRs. Such mode of operation would also improve the efficiency of converting the fossil fuel heat because it is applied only at the superheater stage. There are several ways to supply the heat to the superheaters, for example, by using the exhaust gas from the gas turbines and using the conventional gas burner.In this paper, the thermodynamic performance of the hybrid system (PWR with superheater) is investigated for large reactor and Small Modular Reactor (SMR) application. The thermal efficiency of the AP1000 can be improved from 30.2% to 45.8% (with CCGT), 35.6% (with gas burner), and 36.6% (gas burner with reheating). The thermal efficiency of the SMR can be improved from 33.4% to nearly 45% (with CCGT), 35.5% (with gas burner), and 37.4% (gas burner with reheating). The analysis results show that it is possible for the hybrid system to operate between 65% and the full power load. (C) 2016 Elsevier Ltd. All rights reserved.

Mixed convection in vertical upward heated flow has been intensively studied in the last few years since it reduces the heat transfer capability of the system in the deteriorated turbulent heat transfer (DTHT) regime. Several mixed turbulent heat transfer correlations have been proposed, but the best correlation to be used in DTHT regime for water in vertical upward heated flow is yet to be determined. Furthermore, the relation between momentum flux and heat flux in mixed convection regime is still unclear. This paper presents comparison of several mixed turbulent heat transfer correlations performance of water up-flow in a circular tube by numerical approach using v2–f turbulence model. The existing correlation of Symolon et al. showed the best agreement with the numerical analysis results in predicting the heat transfer capability of a water-cooled system while Petukhov and Polyakov correlation showed a good agreement to correlate the momentum and heat transport behavior in the mixed convection turbulent regime.

In this paper, we suggest the conceptual design of a water-cooled reactor system for a low-pressure inherent heat sink nuclear desalination plant (LIND) that applies the safety-related design concepts of high temperature gas-cooled reactors to a water-cooled reactor for inherent and passive safety features. Through a scoping analysis, we found that the current LIND design satisfied several essential thermal–hydraulic and neutronic design requirements. In a thermal–hydraulic analysis using an analytical method based on the Wooton–Epstein correlation, we checked the possibility of safely removing decay heat through the steel containment even if all the active safety systems failed. In a neutronic analysis using the Monte Carlo N-particle transport code, we estimated a cycle length of approximately 6 years under 200 MWth and 4.5% enrichment. The very long cycle length and simple safety features minimize the burdens from the operation, maintenance, and spent-fuel management, with a positive impact on the economic feasibility. Finally, because a nuclear reactor should not be directly coupled to a desalination system to prevent the leakage of radioactive material into the desalinated water, three types of intermediate systems were studied: a steam producing system, a hot water system, and an organic Rankine cycle system.

Nanofluids application in nuclear systems have been intensively studied for the past few years since it has been found out that this type of fluids exhibit a substantially higher critical heat flux (CHF) compared to "clean" water. Most studies done in these field usually focus on heat transfer criteria of the nanofluid itself and its industrial application including nuclear systems. In vertical upward heating flow, the turbulent heat transfer can be deteriorated if it goes from forced to mixed convection regime or also known as deteriorated turbulent heat transfer (DTHT) regime. Hence, since nanofluids application is being considered in nuclear systems, it is quite interesting to study the mixed convection phenomenon in the nanofluids flow. To this end the current paper focuses on computational fluid dynamic (CFD) analysis of mixed convection flow of the nanofluids. The nanofluids selected for the study are the ones considered for applications related to nuclear systems. Three turbulent models; namely the low-Reynolds k-ε model, the v^2-f model, and the recently developed Elliptical Blending k-ε model are first examined to find out the best model to be used in nanofluids study. The impact of particles concentration in nanofluids on the mixed convection phenomenon is also investigated. This numerical work can give insights and a better understanding of the mixed convection phenomenon in the case of nanofluids-based systems and set a basis for future experimental verification.

The development of a small-sized nuclear heat-only plant with maximized safety features dedicated to seawater thermal desalination was proposed to address both a serious water crisis and nuclear safety issues, which continue to be perennial problems. In this study, the feasibility of a dedicated nuclear heat-only desalination system for a target country was evaluated in comparison with a target nuclear thermal desalination system. First, the target country was selected, and its current energy and desalination status was investigated. The suitable nuclear desalination options for the target country were then selected. Finally, using corresponding analysis tools, performance and economic analyses were conducted for a dedicated nuclear heat-only desalination system and the target nuclear thermal desalination system. The results of the analyses indicate that operating the small-sized nuclear heat-only plant at low pressures coupled with a seawater thermal desalination plant will considerably improve both the safety and economy without a significant loss in desalination performance. In conclusion, the proposed dedicated nuclear heat-only desalination system is expected to have high potential for solving both problems.

Passive safety is a primary motive behind the development of small and medium sized reactors of various coolants. After the Fukushima accident, there is an increased interest in a nuclear reactor's reliance on passive safety systems. Most of the existing passive systems, regardless of the reactor type, utilize buoyancy force to drive the cooling flow. Hence, it is essential to evaluate if the naturally developed cooling flow is sufficient to maintain the heated surface temperature of the fuel elements below the design limit. Evaluating passively driven flows can be quite a challenging task in both two phase natural circulation systems and also in single phase natural circulation systems. Previous research works have found that single phase heat transfer can be deteriorated and becomes uncertain when the driving force of a system is shifted from external force (forced convection) to self generated buoyancy force or a combination of both (natural or mixed convection). In this paper, single phase gas, water, and liquid metal reactors with passive systems are reviewed briefly. A simple theoretical analysis of each reactor type is performed to find the tendency of the shift in the operating heat transfer regime into the deteriorated region. The analysis results show that single phase water system can maintain operation within the forced convection regime but the operating regime gets closer to the deteriorating heat transfer regime as the system's physical size reduces from a large nuclear power plant to the small and medium reactor scale. The gas cooled system has a high tendency to operate in the deteriorated heat transfer regime when the driving force changes from forced to natural. Meanwhile the liquid metal system demonstrates more margins to operate outside from the deteriorated heat transfer region compared to the two other fluid systems. However further studies are needed to clearly identify the boundaries of the deteriorated heat transfer regime for each coolant since the deterioration greatly depends on the thermophysical properties variation of the coolant and the near-wall flow behavior of the coolant with respect to temperature change. (C) 2013 Elsevier B.V. All rights reserved.