• Nuclear Engineering and Technology
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• v.49 no.8
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• pp.1669-1679
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• 2017

Background: Steam generator (SG) is one of the significant components in the nuclear steam supply system. A variety of SGs have been designed and used in nuclear reactor systems. Every SG has advantages and disadvantages. A brief account of some of the existing SG designs is presented in this study. A high surface to volume ratio of a SG is required in small modular reactors to occupy the least space. In this paper, performance improvement for SGs of integral small modular reactor is proposed. Aims/Methods: For this purpose, cross-grooved microfins have been incorporated on the inner surface of the helical tube to enhance heat transfer. The primary objective of this work is to investigate thermal-hydraulic behavior of the proposed improvements through modeling in RELAP5-3D. Results and Conclusions: The results are compared with helical-coiled SGs being used in IRIS (International Reactor Innovative and Secure). The results show that the tube length reduces up to 11.56% keeping thermal and hydraulic conditions fixed. In the case of fixed size, the steam outlet temperature increases from 590.1 K to 597.0 K and the capability of power transfer from primary to secondary also increases. However, these advantages are associated with some extra pressure drop, which has to be compensated.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.1974-1987
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• 2023

The load-following operation of small modular reactors (SMRs) requires accurate prediction of transient behaviors that can occur in the balance of plants (BOP) and the nuclear steam supply system (NSSS). However, 1-D thermal-hydraulics analysis codes developed for safety and performance analysis have conventionally excluded the BOP from the simulation by assuming ideal boundary conditions for the main steam and feed water (MS/FW) systems, i.e., an open loop. In this study, we introduced a lumped model of BOP fluid system and coupled it with NSSS without any ideal boundary conditions, i.e., in a closed loop. Various methods for coupling boundary conditions at MS/FW were tested to validate their combination in terms of minimizing numerical instability, which mainly arises from the coupled boundaries. The method exhibiting the best performance was selected and applied to a transient simulation of an integrated NSSS and BOP system of a SMART. For a transient event with core power change of 100-20-100%, the simulation exhibited numerical stability throughout the system without any significant perturbation of thermal-hydraulic parameters. Thus, the introduced boundary-condition coupling method and BOP fluid system model can expectedly be employed for the transient simulation and performance analysis of SMRs requiring daily load-following operations.

• Nuclear Engineering and Technology
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• v.52 no.5
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• pp.964-974
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• 2020

The small modular reactors (SMRs) of the integrated pressurized water reactor (IPWR) type have been widely developed owing to their enhanced safety features. The SMR-IPWR adopts passive residual heat removal system (PRHRS) to extract residual heat from the core. Because the PRHRS removes the residual heat using the latent heat of the water stored in the emergency cooldown tank, the PRHRS gradually loses its cooling capacity after the stored water is depleted. A quick restoration of the power supply is expected infeasible under station blackout accident condition, so an advanced PRHRS is needed to ensure an extended grace period. In this study, an advanced design is proposed to indirectly incorporate a dry air cooling tower to the PRHRS through an intermediate loop called indefinite PRHRS. The feasibility of the indefinite PRHRS was assessed through a long-term transient simulation using the MARS-KS code. The indefinite PRHRS is expected to remove the residual heat without depleting the stored water. The effect of the environmental temperature on the indefinite PRHRS was confirmed by parametric analysis using comparative simulations with different environmental temperatures.

• Nuclear Engineering and Technology
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• v.53 no.11
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• pp.3580-3596
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• 2021

An integral test facility, PILLAR, was commissioned, aiming to provide valuable experimental results which can be referenced by system and component designers and used for the performance demonstration of liquid-metal-cooled, passive small modular reactors (SMRs) toward their licensing. The setup was conceptualized by a scaling analysis which allows the vertical arrangements to be conserved from its prototypic reactor, scaled uniformly in the radial direction achieving a flow area reduction of 1/200. Its final design includes several heater rods which simulate the reactor core, and a single heat exchanger representing the steam generators in the prototype. The system behaviors were characterized by its data acquisition system implementing various instruments. In this paper, we present not only a detailed description of the facility components, but also selected experimental results of both steady-state and transient cases. The obtained steady-state test results were utilized for the benchmark of a system code, achieving a capability of accurate simulations with ±3% of maximum deviations. It was followed by qualitative comparisons on the transient test results which indicate that the integral system behaviors in passive LBE-cooled systems are able to be predicted by the code.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.4
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• pp.571-576
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• 2023

The initial development plans for the six reactor designs, soon after the release of Generation IV International Forum (GIF) TRM in 2002, were characterized by high ambition [1]. Specifically, the sodium-cooled fast reactor (SFR) and very-high temperature reactor (VHTR) gained significant attention and were expected to reach the validation stage by the 2020s, with commercial viability projected for the 2030s. However, these projections have been unrealized because of various factors. The development of reactor designs by the GIF was supposed to be influenced by events such as the 2008 global financial crisis, 2011 Fukushima accident [2, 3], discovery of extensive shale oil reserves in the United States, and overly ambitious technological targets. Consequently, the momentum for VHTR development reduced significantly. In this context, the aims of this study were to compare and analyze the development progress of the six Gen IV reactor designs over the past 20 years, based on the GIF roadmaps published in 2002 and 2014. The primary focus was to examine the prospects for the reactor designs in relation to spent nuclear fuel burning in conjunction with small modular reactor (SMR), including molten salt reactor (MSR), which is expected to have spent nuclear fuel management potential.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.1100-1107
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• 2024

The nuclear waste attributes of near-term deployable SMRs were assessed using established nuclear waste metrics, which are the DU mass, SNF mass, volume, activity, decay heat, radiotoxicity, and decommissioning LLW volumes. Metrics normalized per unit electricity generation were compared to a reference large PWR. Three SMRs, VOYGR, Natrium, and Xe-100, were selected because they represent a range of reactor and fuel technologies and are active designs deployable by the decade's end. The SMR nuclear waste attributes show both some similarities to the PWR and some significant differences caused by reactor-specific design features. The DU mass is equivalent to or slightly higher than the PWR. Back-end waste attributes for SNF disposition vary, but the differences have a limited impact on long-term repository isolation. SMR designs can vary significantly in SNF volume (and thus heat generation density). However, these differences are amenable to design optimization for handling, storage, transportation, and disposal technologies. Nuclear waste attributes from decommissioning vary depending on design and decommissioning technology choices. Given the analysis results in this study and assuming appropriate waste management system and operational optimization, there appear to be no major challenges to managing SMR nuclear wastes compared to the reference PWR.

• Nuclear Engineering and Technology
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• v.56 no.8
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• pp.3315-3329
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• 2024

Small Modular Reactors (SMRs) adopt the Helically Coiled Once-Through Steam Generators (OTSG) extensively for its compactness and higher heat transfer efficiency. As a heat exchanger between the primary side (reactor coolant system) and the secondary side (feedwater and steam system) of nuclear steam supply system, the inlet/outlet conditions both of shell side and tube side of OTSGs have significant impacts on overall system response. Considering the flexible operation of SMRs and heat application by extracting steam, a simulation tool for accurate prediction of the OTSG dynamic behaviors would be required for optimizing design and control. In this study, the OTSG dynamic simulation model has been developed. Mathematical governing equation has been derived by using moving boundary approach and a simulation module has been developed by using Modelica Language. The developed module has been compared with publicly available experimental results and benchmarked with MARS-KS calculation results. Also, it has been incorporated into the integrated SMR model (i.e., reactor core, primary side, secondary side) and dynamic behaviors with reactivity feedback and heat balancing have been investigated. In both of steady-state and transient conditions, it shows the promising accuracy.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.17 no.2
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• pp.75-82
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• 2021

The 3D printing technology is one of the fourth industrial revolution technology that drives innovation in the manufacturing process, and should be applied to nuclear industry for various purposes according to the manufacturing trend change. In nuclear industry, it can be applied to manufacture obsolete items and new designed parts in advanced reactors or small modular reactors (SMRs), replacing the traditional manufacturing technologies. A gate valve body was manufactured, which was obsolete in nuclear power plant, using DED(Directed Energy Deposition) metal 3D printing technology after restoring design characteristics including 3D design drawing by reverse engineering. The 3D printed valve body was assembled with commercial parts such as seat-ring, disk, stem, and actuator for performance test. For the valve assembly, including 3D printed valve body, several tests were performed, including pressure test, end-loading test, and seismic test according to KEPIC MGG and KEPIC MFC. In the pressure test, hydraulic pressure of 391kgf/cm² was applied to 3D printed valve body, and no leak was detected. Also the 3D printed valve assembly was performed well in end-loading and seismic tests.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• v.20 no.1
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• pp.49-55
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• 2024

Stainless 630 (or 17-4PH) is a precipitation-hardening martensitic stainless steel that has excellent mechanical properties and corrosion resistance. These characteristics make the STS630 to be used as a consisting material for various components such as spider, pin, spring, and spring retainer, of the control rod drive mechanism (CRDM) in pressurized water reactors (PWRs). In general, it is well known that the oxide layer of stainless steel consists of a duplex layer, a compact inner layer of FeCr₂O₄ spinel, and a coarse-grained outer layer of Fe₃O₄ spinel in PWR primary coolant condition. However, the characteristics of the oxide layer can be sensitively influenced by various water chemistry conditions such as temperature, dissolved oxygen, dissolved hydrogen, pH, pH adjuster type, and exposure time. In this work, we investigate the corrosion properties of the STS630 as a function of coolant temperature in an NH3 alkaline solution for its boron-free application in a small modular reactor, to confirm the feasibility for usage as a boron-free SMR structural material. As a result, oxide layer of corroded STS630 is consist of double-layer oxides consisting of a Cr-rich dense inner oxide and a Fe-rich polyhedral outer particles like as that in commercial PWR primary coolant. The corrosion rate of STS630 increases with increase in test time and temperature and the corrosion rate-time model equation was developed based on experimental data. Overall, it is expected that the results in this study provides useful data for the corrosion behavior of STS630 in alkaline environments, contributing to the development of selecting suitable materials for SMRs.

• Nuclear Engineering and Technology
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• v.56 no.3
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• pp.803-811
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• 2024

Nuclear accidents such as Fukushima Daiichi have highlighted the potential of passive safety systems to replace or complement active safety systems as part of the overall prevention and/or mitigation strategies. In addition, passive systems are key features of Small Modular Reactors (SMRs), for which they are becoming almost unavoidable and are part of the basic design of many reactors available in today's nuclear market. Nevertheless, their potential to significantly increase the safety of nuclear power plants still needs to be strengthened, in particular the ability of computer codes to determine their performance and reliability in industrial applications and support the safety demonstration. The PASTELS project (September 2020-February 2024), funded by the European Commission "Euratom H2020" programme, is devoted to the study of passive systems relying on natural circulation. The project focuses on two types, namely the SAfety COndenser (SACO) for the evacuation of the core residual power and the Containment Wall Condenser (CWC) for the reduction of heat and pressure in the containment vessel in case of accident. A specific design for each of these systems is being investigated in the project. Firstly, a straight vertical pool type of SACO has been implemented on the Framatome's PKL loop at Erlangen. It represents a tube bundle type heat exchanger that transfers heat from the secondary circuit to the water pool in which it is immersed by condensing the vapour generated in the steam generator. Secondly, the project relies on the CWC installed on the PASI test loop at LUT University in Finland. This facility reproduces the thermal-hydraulic behaviour of a Passive Containment Cooling System (PCCS) mainly composed of a CWC, a heat exchanger in the containment vessel connected to a water tank at atmospheric pressure outside the vessel which represents the ultimate heat sink. Several activities are carried out within the framework of the project. Different tests are conducted on these integral test facilities to produce new and relevant experimental data allowing to better characterize the physical behaviours and the performances of these systems for various thermo-hydraulic conditions. These test programmes are simulated by different codes acting at different scales, mainly system and CFD codes. New "system/CFD" coupling approaches are also considered to evaluate their potential to benefit both from the accuracy of CFD in regions where local 3D effects are dominant and system codes whose computational speed, robustness and general level of physical validation are particularly appreciated in industrial studies. In parallel, the project includes the study of single and two-phase natural circulation loops through a bibliographical study and the simulations of the PERSEO and HERO-2 experimental facilities. After a synthetic presentation of the project and its objectives, this article provides the reader with findings related to the physical analysis of the test results obtained on the PKL and PASI installations as well an overall evaluation of the capability of the different numerical tools to simulate passive systems.