• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.558-563
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• 1997

A thermal sizing code, named as HSGSA (Helical coil Steam Generator Sizing Analyzer), for a sodium heated helical coil steam generator is developed for KALIMER (Korea Advanced LIquid MEtal Reactor) design. The theoretical modeling of the shell and tube sides is described and relevant correlations are presented. For assessment of HSGSA, a reference plant design case is compared to the calculational outputs from HSGSA simulation.

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.460-474
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• 2023

A simple but accurate mathematical model is crucial for dynamic simulations and controller design of helical coil once-through steam generator (OTSG). This paper presents a three-region movable boundary dynamic model of the helical coil OTSG. Based on the secondary side fluid conditions, the OTSG is divided into subcooled region (two control volumes), two-phase region (two control volumes) and superheated region (three control volumes) with movable boiling boundaries between each region. The nonlinear dynamic model is derived based on mass, energy and momentum conservation equations. And the linear model is obtained by using the transfer function and state space transformation, which is a 37-order model of five input and three output. Validations are made under full-power steady-state condition and four transient conditions. Results show good agreements among the nonlinear model, linear model and the RELAP5 model, with acceptable errors. This model can be applied to dynamic simulations and controller design of helical coil OTSG with constant primary-side flow rate.

• Proceedings of the Korean Nuclear Society Conference
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• 2002.05a
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• pp.143.1-143
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• 2002

• 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.

• Transactions of the Korean Society of Mechanical Engineers A
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• v.27 no.6
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• pp.930-937
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• 2003

The extension of design life among power plants is increasingly becoming a world-wide trend. Kori #1 unit in Korea is operating two cycle. It has two man-ways for tube inspection in a steam generator which is one of the important components in a nuclear power plant. Especially, stud bolts fur man-way cover have damaged by disassembly and assembly several times and degradation for bolt materials for long term operation. It should be evaluated and compared by ASME Code criteria for integrity evaluation. Integrity evaluation criteria which has been made by the manufacturer is not applied on the stud bolts of nuclear pressure vessels directly because it is controlled by the yield stress of ASME Code. It can apply evaluation criteria through FEM analysis to damaged female threads and to evaluated safety fer helical-coil method which is used according to Code Case-N-496-1. From analysis results, we found .that it is the same results between stress intensity which got from FEM analysis on damaged female threads over 10％ by manufacture integrity criteria and 2／3 yield strength criteria on ASME Code. It was also confirmed that the helical-coil repair method would be safe.

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2699-2708
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• 2020

Helical-coil steam generator (HCSG) technology is a major design candidate for small modular reactors due to its compactness and capability to produce superheated steam with high generation efficiency. In this paper, we investigate the feasibility of the passively autonomous power maneuvering by coupling the 3-D transient multi-physics of a soluble-boron-free (SBF) core with a time-dependent HCSG model. The predictor corrector quasi-static method was used to reduce the cost of the transient 3-D neutronic solution. In the numerical system simulations, the feedwater flow rate to the secondary of the HCSGs is adjusted to extract the demanded power from the primary loop. This varies the coolant temperature at the inlet of the SBF core, which governs the passively autonomous power maneuvering due to the strongly negative coolant reactivity feedback. Here, we simulate a 100-50-100 load-follow operation with a 5%/minute power ramping speed to investigate the feasibility of the passively autonomous load-follow in a 450 MW_th SBF PWR. In addition, the passively autonomous frequency control operation is investigated. The various system models are coupled, and they are solved by an in-house Fortran-95 code. The results of this work demonstrate constant steam temperature in the secondary side and limited variation of the primary coolant temperature. Meanwhile, the variations of the core axial shape index and the core power peaking are sufficiently small.

• Nuclear Engineering and Technology
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• v.45 no.3
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• pp.347-360
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• 2013

This paper presents an experimental investigation of the small-break loss-of-coolant accident (SBLOCA) and the loss-of-feedwater accident (LOFW) in a scaled integral test facility of REX-10. REX-10 is a small integral-type PWR in which the coolant flow is driven by natural circulation, and the RCS is pressurized by the steam-gas pressurizer. The postulated accidents of REX-10 include the system depressurization initiated by the break of a nitrogen injection line connected to the steam-gas pressurizer and the complete loss of normal feedwater flow by the malfunction of control systems. The integral effect tests on SBLOCA and LOFW are conducted at the REX-10 Test Facility (RTF), a full-height full-pressure facility with reduced power by 1/50. The SBLOCA experiment is initiated by opening a flow passage out of the pressurizer vessel, and the LOFW experiment begins with the termination of the feedwater supply into the helical-coil steam generator. The experimental results reveal that the RTF can assure sufficient cooldown capability with the simulated PRHRS flow during these DBAs. In particular, the RTF exhibits faster pressurization during the LOFW test when employing the steam-gas pressurizer than the steam pressurizer. This experimental study can provide unique data to validate the thermal-hydraulic analysis code for REX-10.

• Nuclear Engineering and Technology
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• v.45 no.4
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• pp.439-458
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• 2013

REX-10 is a fully-passive small modular reactor in which the coolant flow is driven by natural circulation, the RCS is pressurized by a steam-gas pressurizer, and the decay heat is removed by the PRHRS. To confirm design decisions and analyze the transient responses of an integral PWR such as REX-10, a thermal-hydraulic system code named TAPINS (Thermal-hydraulic Analysis Program for INtegral reactor System) is developed in this study. Based on a one-dimensional four-equation drift-flux model, TAPINS incorporates mathematical models for the core, the helical-coil steam generator, and the steam-gas pressurizer. The system of difference equations derived from the semi-implicit finite-difference scheme is numerically solved by the Newton Block Gauss Seidel (NBGS) method. TAPINS is characterized by applicability to transients with non-equilibrium effects, better prediction of the transient behavior of a pressurizer containing non-condensable gas, and code assessment by using the experimental data from the autonomous integral effect tests in the RTF (REX-10 Test Facility). Details on the hydrodynamic models as well as a part of validation results that reveal the features of TAPINS are presented in this paper.