• Nuclear Engineering and Technology
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• 제37권1호
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• pp.25-78
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• 2005

As a key and core knowledge for the design of various types of nuclear reactors, the discipline of reactor physics has been advanced continually in the past six decades and has led to a very sophisticated fabric of analysis methods and computer codes in use today. Notwithstanding, the discipline faces interesting challenges from next-generation nuclear reactors and innovative new fuel designs in the coming. After presenting a brief overview of important tasks and steps involved in the nuclear design and analysis of a reactor, this article focuses on the currently-used design and analysis methods, issues and limitations, and current activities to resolve them as follows: (1) Derivation of the multi group transport equations and the multi group diffusion equations, with representative solution methods thereof. (2) Elements of modem (now almost three decades old) diffusion nodal methods. (3) Limitations of nodal methods such as transverse integration, flux reconstruction, and analysis of UO2-MOX mixed cores. Homogenization and related issues. (4) Description of the analytic function expansion nodal (AFEN) method. (5) Ongoing efforts for three-dimensional whole-core heterogeneous transport calculations and acceleration methods. (6) Elements of spatial kinetics calculation methods and coupled neutronics and thermal-hydraulics transient analysis. (7) Identification of future research and development areas in advanced reactors and Generation-IV reactors, in particular, in very high temperature gas reactor (VHTR) cores.

• 한국전산유체공학회지
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• 제20권4호
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• pp.14-20
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• 2015

Without electirc power to cool down the hot reactor core, passive systems utilizing natural circulation are becoming a big specialty of recent neculear systems after the severe accident in Fukusima. When we consider the natural circulation in a pool, thermal mixing phenomena may start from single phase circulation and can continue to two phase condition. Since the CUPID code, which has been developed for two-phase flow analysis, can deal with the phase transition phenomena, the CUPID would be pertinent to natural convection problems in single- and two-phase conditions. Thus, the CUPID should be validated against single- and two-phase natural circulation phenomena. For the first step of the validation process, this study is focused on the validation of single-phase natural circulation. Moreover, the CUPID code solves the fluid properties by the relationship to pressure and temperature from the steam table considering non-condensable gas effects, so that the effects from variable properties are included. Simple square thermal cavity problems are tested for laminar and turbulent conditions against numerical and experimental data. Throughout the investigation, it is found that the variable properties can affect the flow field in laminar condition, but the effect becomes weak in turbulence condition, and the CUPID code implementing steam table is capable of analyzing single phase natural circualtion phenomena.

• Nuclear Engineering and Technology
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• 제53권2호
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• pp.455-465
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• 2021

The Molten Salt Reactor (MSR), one of the six advanced reactor types of the 4th generation nuclear energy systems, has many impressive features including economic advantages, inherent safety and nuclear non-proliferation. This paper introduces a system analysis code named TREND, which is developed and used for the steady and transient simulation of MSRs. The TREND code calculates the distributions of pressure, velocity and temperature of single-phase flows by solving the conservation equations of mass, momentum and energy, along with a fluid state equation. Heat structures coupled with the fluid dynamics model is sufficient to meet the demands of modeling MSR system-level thermal-hydraulics. The core power is based on the point reactor neutron kinetics model calculated by the typical Runge-Kutta method. An incremental PID controller is inserted to adjust the operation behaviors. The verification and validation of the TREND code have been carried out in two aspects: detailed code-to-code comparison with established thermal-hydraulic system codes such as RELAP5, and validation with the experimental data from MSRE and the CIET facility (the University of California, Berkeley's Compact Integral Effects Test facility).The results indicate that TREND can be used in analyzing the transient behaviors of MSRs and will be improved by validating with more experimental results with the support of SINAP.

• Nuclear Engineering and Technology
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• 제53권9호
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• pp.2775-2787
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• 2021

At present, most of the widely used system codes for nuclear safety analysis are one-dimensional, which cannot effectively simulate the flow field of the reactor core or other structures. This is true even for the system codes containing three-dimensional modules with limited three-dimensional simulation function such as RELAP-3D. In contrast, the computational fluid dynamics (CFD) codes excel at providing a detailed three-dimensional flow field of the reactor core or other components; however, the computational domain is relatively small and results in the very high computing resource consuming. Therefore, the development of coupling codes, which can make comprehensive use of the advantages of system and CFD codes, has become a research focus. In this paper, a review focus on the researches of coupled CFD and thermal-hydraulic system codes was carried out, which summarized the method of coupling, the data transfer processing between CFD and system codes, and the verification and validation (V&V) of coupled codes. Furthermore, a series of problems associated with the coupling procedure have been identified, which provide the general direction for the development and V&V efforts of coupled codes.

• 시스템엔지니어링학술지
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• 제18권2호
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• pp.58-74
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• 2022

Deterministic accident analysis plays a central role in the nuclear power plant (NPP) safety evaluation and licensing process. Traditionally the conservative approach opted for the point kinetics model, expressing the reactor core parameters in the form of reactivity and power tables. However, with the current advances in computational power, high fidelity multi-physics simulations using real-time code coupling, can provide more detailed core behavior and hence more realistic plant's response. This is particularly relevant for transients where the core is undergoing reactivity anomalies and uneven power distributions with strong feedback mechanisms, such as reactivity initiated accidents (RIAs). This work addresses a RIA, specifically a control element assembly (CEA) withdrawal at power, using the multi-physics analysis tool RELAP5/MOD 3.4/3DKIN. The thermal-hydraulics (TH) code, RELAP5, is internally coupled with the nodal kinetics (NK) code, 3DKIN, and both codes exchange relevant data to model the nuclear power plant (NPP) response as the CEA is withdrawn from the core. The coupled model is more representative of the complex interactions between the thermal-hydraulics and neutronics; therefore the results obtained using a multi-physics simulation provide a larger safety margin and hence more operational flexibility compared to those of the point kinetics model reported in the safety analysis report for APR1400. The systems engineering approach is used to guide the development of the work ensuring a systematic and more efficient execution.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2001년도 춘계학술대회논문집D
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• pp.821-827
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• 2001

As one of the advanced design features of the Korea next generation reactor, direct vessel injection (DVI) system is being considered instead of conventional cold leg injection (CLl) system. It is known that the DVI system greatly enhances the reliability of the emergency core cooling (ECC) system. However, there is still a dispute on its performance in terms of water delivery to the reactor core during the reflood period of a large-break loss-of-coolant accident (LOCA). Thus, experimental validation is under progress. In this paper, a new scaling method, using time and velocity reduced linear scaling law, is suggested for the design of a scaled-down experimental facility to investigate the direct ECC bypass phenomena in PWR downcomer.

• Nuclear Engineering and Technology
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• 제38권1호
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• pp.33-44
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• 2006

This paper presents a review of small-to-medium-sized, pressurized-water-cooled nuclear power reactors whose major primary coolant systems are integrated into a reactor pressure vessel, the concepts categorized as Integrated Primary System Nuclear Reactors (IPSRs). Typical examples of these proposals of interest in this review are CAREM, SMART, IRIS and IMR, all of which are being aimed at the near term deployment. Emphasis is placed on thermal hydraulic aspects. A brief characterization of the IPSR concepts is made and comparisons of plant key parameters are shown. Discussions will follow for the core cooling under rated power conditions and natural circulation heat removal on the basis of the design data available in the public domain.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.873-879
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• 2024

A sodium-cooled fast reactor (SFR) core has a potential of prompt criticality due to a change of core material distribution during a severe accident, and the resultant energy release has been one of the safety issues of SFRs. In this study, the safety assessment of an unprotected loss-of-flow (ULOF) in a small SFR (SSFR) has been performed using the SIMMER-IV computer code, which couples the models of space- and time-dependent neutronics and multi-component, multi-field thermal hydraulics in three dimensions. The code, therefore, is applicable to the simulations of transient behaviors of extended disrupted core material motion and its reactivity effects during the transition phase (TP) of ULOF, including a potential of prompt-criticality power excursions driven by fuel compaction. Several conservative assumptions are used in the TP analysis by SIMMER-IV. It was found out that one of the important mechanisms that drives the reactivity-inserting fuel motion was sodium vapor pressure resulted from a fuel-coolant interaction (FCI), which itself was non-energetic local phenomenon. The uncertainties relating to FCI is also evaluated in much conservative way in the sensitivity analysis. From this study, the ULOF characteristics in an SSFR have been understood. Occurrence of recriticality events under conservative assumptions are plausible, but their energy releases are limited.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.116-124
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• 2019

$M^2LFR-1000$ is a medium-power modular lead-cooled fast reactor, developed by University of Science and Technology of China (USTC), aiming at achieving a reactor design fulfilling the Gen IV nuclear system requirements and meanwhile emphasizing the optimum safety and economics. In order to evaluate the safety performance of $M^2LFR-1000$ reactor core, three typical transients are selected from initiating events, which are unprotected transient overpower (UTOP), unprotected loss of offsite power (ULOHS+ULOF) and increase of feedwater flowrate with primary pumps trip (IFW+PLOF). These three transients presented and discussed in this paper are performed with the code Analysis of THermal-hydraulics of LEaks and Transients (ATHLET), which is developed by Gesellschaft $f{\ddot{u}}r$ Anlagen-und Reaktorsicherheit gGmbH (GRS). The results indicate that the $M^2LFR$ is safe enough under these three transients due to the good inherent safety features of the reactor, without human intervention, the reactor will reach a new steady state under UTOP condition.

• Nuclear Engineering and Technology
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• 제56권6호
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• pp.2375-2387
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• 2024

The passive safety systems (PSSs) within water-cooled reactors are meticulously engineered to function autonomously, requiring no external power source or manual intervention. They depend exclusively on inherent natural forces and the fundamental principles of reactor physics, such as gravity, natural convection, and phase changes, to manage, alleviate, and avert the release of radioactive materials into the environment during accident scenarios like a loss-of-coolant accident (LOCA). PSSs are already integrated into such operating commercial reactors as the Advanced Pressurized Reactor-1000 MWe (AP1000) and the Water-Water Energetic Reactor-1200 MWe (WWER-1200) are adopted in most of the upcoming small modular reactor (SMR) designs. Examples of water-cooled SMR PSSs are the passive emergency core-cooling system (ECCS), passive containment cooling system (PCCS), and passive decay-heat removal system, the designs of which vary based on reactor system-design requirements. However, understanding the accident-event progression and phases of a LOCA is pivotal for adopting a specific PSS for a new SMR design. This study covers the accident-event progression for direct vessel injection (DVI) small-break loss-of-coolant accident (SB-LOCA), associated physics phenomena, knowledge gaps, and important figures of merit (FOMs) that may need to be evaluated and assessed to validate thermal-hydraulics models with an available experimental dataset to support new SMR design and development.