• Title/Summary/Keyword: Advanced reactors

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DEVELOPMENT OF THE SPACE CODE FOR NUCLEAR POWER PLANTS

  • Ha, Sang-Jun;Park, Chan-Eok;Kim, Kyung-Doo;Ban, Chang-Hwan
    • Nuclear Engineering and Technology
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    • v.43 no.1
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    • pp.45-62
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    • 2011
  • The Korean nuclear industry is developing a thermal-hydraulic analysis code for safety analysis of pressurized water reactors (PWRs). The new code is called the Safety and Performance Analysis Code for Nuclear Power Plants (SPACE). The SPACE code adopts advanced physical modeling of two-phase flows, mainly two-fluid three-field models which comprise gas, continuous liquid, and droplet fields and has the capability to simulate 3D effects by the use of structured and/or nonstructured meshes. The programming language for the SPACE code is C++ for object-oriented code architecture. The SPACE code will replace outdated vendor supplied codes and will be used for the safety analysis of operating PWRs and the design of advanced reactors. This paper describes the overall features of the SPACE code and shows the code assessment results for several conceptual and separate effect test problems.

Conceptual Design Based on Scale Laws and Algorithms Sub-critical Transmutation Reactors

  • Lee, Kwang-Gu;Chang, Soon-Heung
    • Proceedings of the Korean Nuclear Society Conference
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    • 1997.10a
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    • pp.475-480
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    • 1997
  • In order to conduct the effective integration of computer-aided conceptual design for integrated nuclear power reactor, not only is a smooth information flow required, but also decision making fur both conceptual design and construction process design must be synthesized. In addition to the aboves, the relations between the one step and another step and the methodologies to optimize the decision variables are verified, in this paper especially, that is, scaling laws and scaling criteria. In the respect with the running of the system, the integrated optimization process is proposed in which decisions concerning both conceptual design are simultaneously made. According to the proposed reactor types and power levels, an integrated optimization problems are formulated. This optimization is expressed as a multi-objective optimization problem. The algorithm for solving the problem is also presented. The proposed method is applied to designing a integrated sub-critical reactors.

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FAST (floating absorber for safety at transient) for the improved safety of sodium-cooled burner fast reactors

  • Kim, Chihyung;Jang, Seongdong;Kim, Yonghee
    • Nuclear Engineering and Technology
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    • v.53 no.6
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    • pp.1747-1755
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    • 2021
  • This paper presents floating absorber for safety at transient (FAST) which is a passive safety device for sodium-cooled fast reactors with a positive coolant temperature coefficient. Working principle of the FAST makes it possible to insert negative reactivity passively in case of temperature rise or voiding of coolant. Behaviors of the FAST in conventional oxide fuel-loaded and metallic fuel-loaded SFRs are investigated assuming anticipated transients without scram (ATWS) scenarios. Unprotected loss of flow (ULOF), unprotected loss of heat sink (ULOHS), unprotected transient overpower (UTOP) and unprotected chilled inlet temperature (UCIT) scenarios are simulated at end of life (EOL) conditions of the oxide and the metallic SFR cores, and performance of the FAST to improve the reactor safety is analyzed in terms of reactivity feedback components, reactor power and maximum temperatures of fuel and coolant. It is shown that FAST is able to improve the safety margin of conventional burner-type SFRs during ULOF, ULOHS, UTOP and UCIT.

MCCARD: MONTE CARLO CODE FOR ADVANCED REACTOR DESIGN AND ANALYSIS

  • Shim, Hyung-Jin;Han, Beom-Seok;Jung, Jong-Sung;Park, Ho-Jin;Kim, Chang-Hyo
    • Nuclear Engineering and Technology
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    • v.44 no.2
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    • pp.161-176
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    • 2012
  • McCARD is a Monte Carlo (MC) neutron-photon transport simulation code. It has been developed exclusively for the neutronics design of nuclear reactors and fuel systems. It is capable of performing the whole-core neutronics calculations, the reactor fuel burnup analysis, the few group diffusion theory constant generation, sensitivity and uncertainty (S/U) analysis, and uncertainty propagation analysis. It has some special features such as the anterior convergence diagnostics, real variance estimation, neutronics analysis with temperature feedback, $B_1$ theory-augmented few group constants generation, kinetics parameter generation and MC S/U analysis based on the use of adjoint flux. This paper describes the theoretical basis of these features and validation calculations for both neutronics benchmark problems and commercial PWR reactors in operation.

Techno-economic assessment of a very small modular reactor (vSMR): A case study for the LINE city in Saudi Arabia

  • Salah Ud-Din Khan;Rawaiz Khan
    • Nuclear Engineering and Technology
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    • v.55 no.4
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    • pp.1244-1249
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    • 2023
  • Recently, the Kingdom of Saudi Arabia (KSA) announced the development of first-of-a-kind(FOAK) and most advanced futuristic vertical city and named as 'The LINE'. The project will have zero carbon dioxide emissions and will be powered by clean energy sources. Therefore, a study was designed to understand which clean energy sources might be a better choice. Because of its nearly carbon-free footprint, nuclear energy may be a good choice. Nowadays, the development of very small modular reactors (vSMRs) is gaining attention due to many salient features such as cost efficiency and zero carbon emissions. These reactors are one step down to actual small modular reactors (SMRs) in terms of power and size. SMRs typically have a power range of 20 MWe to 300 MWe, while vSMRs have a power range of 1-20 MWe. Therefore, a study was conducted to discuss different vSMRs in terms of design, technology types, safety features, capabilities, potential, and economics. After conducting the comparative test and analysis, the fuel cycle modeling of optimal and suitable reactor was calculated. Furthermore, the levelized unit cost of electricity for each reactor was compared to determine the most suitable vSMR, which is then compared other generation SMRs to evaluate the cost variations per MWe in terms of size and operation. The main objective of the research was to identify the most cost effective and simple vSMR that can be easily installed and deployed.

Design of an Organic Simplified Nuclear Reactor

  • Shirvan, Koroush;Forrest, Eric
    • Nuclear Engineering and Technology
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    • v.48 no.4
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    • pp.893-905
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    • 2016
  • Numerous advanced reactor concepts have been proposed to replace light water reactors ever since their establishment as the dominant technology for nuclear energy production. While most designs seek to improve cost competitiveness and safety, the implausibility of doing so with affordable materials or existing nuclear fuel infrastructure reduces the possibility of near-term deployment, especially in developing countries. The organic nuclear concept, first explored in the 1950s, offers an attractive alternative to advanced reactor designs being considered. The advent of high temperature fluids, along with advances in hydrocracking and reforming technologies driven by the oil and gas industries, make the organic concept even more viable today. We present a simple, cost-effective, and safe small modular nuclear reactor for offshore underwater deployment. The core is moderated by graphite, zirconium hydride, and organic fluid while cooled by the organic fluid. The organic coolant enables operation near atmospheric pressure and use of plain carbon steel for the reactor tank and primary coolant piping system. The core is designed to mitigate the coolant degradation seen in early organic reactors. Overall, the design provides a power density of 40 kW/L, while reducing the reactor hull size by 40% compared with a pressurized water reactor while significantly reducing capital plant costs.