• Journal of the Korea Academia-Industrial cooperation Society
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• v.20 no.8
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• pp.362-370
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• 2019

The internal pressure capacity of a modular containment structure requires analysis to prevent the release of radioactive material in the case of an accident. To analyze the capacity, FEM models were prepared while considering the tendon arrangements and the contact surfaces between precast concrete modules, and then static analyses were carried out. The changing characteristics in the displacement and stress under step-wise loading were analyzed, along with the effects of selected parameters. For comparison, the capacity of a monolithic containment structure was also analyzed. Parametric analyses were done to suggest ranges of parameters such as the tendon force, tendon spacing, tendon location in concrete thickness direction, friction coefficient, and concrete thickness. The tendon force and frictional force provide a combined effect between contact surfaces of modules. The same level of internal pressure capacity can be secured even in the modular containment structure as in the monolithic containment structure by increasing the tendon force with additional tendons.

• Proceedings of the Korean Nuclear Society Conference
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• 2004.10a
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• pp.135-136
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• 2004

• Proceedings of the Korean Nuclear Society Conference
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• 1998.10a
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• pp.164-164
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• 1998

• Proceedings of the Korean Nuclear Society Conference
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• 2005.10a
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• pp.9-10
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• 2005

• Nuclear Engineering and Technology
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• v.52 no.12
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• pp.2743-2759
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• 2020

A detailed computational fluid dynamics (CFD) simulation analysis model was developed using ANSYS CFX 16.1 and analyzed to simulate the basic design and internal flow characteristics of a 180 MW small modular reactor (SMR) with a natural circulation flow system. To analyze the natural circulation phenomena without a pump for the initial flow generation inside the reactor, the flow characteristics were evaluated for each output assuming various initial powers relative to the critical condition. The eddy phenomenon and the flow imbalance phenomenon at each output were confirmed, and a flow leveling structure under the core was proposed for an optimization of the internal natural circulation flow. In the steady-state analysis, the temperature distribution and heat transfer speed at each position considering an increase in the output power of the core were calculated, and the conceptual design of the SMR had a sufficient thermal margin (31.4 K). A transient model with the output ranging from 0% to 100% was analyzed, and the obtained values were close to the T_hot and T_cold temperature difference value estimated in the conceptual design of the SMR. The K-factor was calculated from the flow analysis data of the CFX model and applied to an analysis model in RELAP5/MOD3.3, the optimal analysis system code for nuclear power plants. The CFX analysis results and RELAP analysis results were evaluated in terms of the internal flow characteristics per core output. The two codes, which model the same nuclear power plant, have different flow analysis schemes but can be used complementarily. In particular, it will be useful to carry out detailed studies of the timing of the steam generator intervention when an SMR is activated. The thermal and hydraulic characteristics of the models that applied porous media to the core & steam generators and the models that embodied the entire detail shape were compared and analyzed. Although there were differences in the ability to analyze detailed flow characteristics at some low powers, it was confirmed that there was no significant difference in the thermal hydraulic characteristics' analysis of the SMR system's conceptual design.

• Nuclear Engineering and Technology
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• v.31 no.1
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• pp.9-16
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• 1999

A 330 MWt-sized multi-purpose integral-type reactor, SMART is under development in Korea for the use of nuclear energy other than electricity generation. In this study, various radioactive source terms are estimated for SMART. SMART is different from conventional reactor concepts in operation and design. Therefore Specific Calculation method namely recurrence model is used. This model is based on the change rate in the RC radioactivity materials and operational characteristics of SMART Calculation results show tremendously increase of the levels of RC activity because no cleanup of RC and long term operation.

• Nuclear Engineering and Technology
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• v.52 no.7
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• pp.1367-1379
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• 2020

The small modular sodium-cooled fast reactor (SMSFR) is an important component of Generation-IV reactors. The objective of this work is to improve the reactivity control in SMSFR by using innovative systems, including burnable poisons and optimized control rods. SMSFR with MOX fuel usually exhibits high burnup reactivity loss that leads to high excess reactivity and potential fuel melting in control rod withdrawal (CRW) accidents, which becomes an important constraint on the safety and economic efficiency of SMSFR. This work applies two types of burnable poisons in a SMSFR to reduce the excess reactivity. The first one homogenously loads minor actinides in the fuel. The second one combines absorber and moderators in specific assemblies. The influence of burnable poisons on the core characteristics is discussed and integrated into the analysis of CRW accidents. The results show that burnable poisons improve the safety performance of the core in a significant way. Burnable poisons also lessen the demand for the number, absorption ability, and insertion depth of control rods. Two optimized control rod designs with rare earth oxides (Eu₂O₃ and Gd₂O₃) and moderators are compared to the conventional design with natural boron carbide (B₄C). The optimized designs show improved neutronic and safety performance.

• Nuclear Engineering and Technology
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• v.39 no.2
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• pp.95-102
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• 2007

The conceptual design of the MIT modular pebble bed reactor is described. This reactor plant is a 250 Mwth, 120 Mwe indirect cycle plant that is designed to be deployed in the near term using demonstrated helium system components. The primary system is a conventional pebble bed reactor with a dynamic central column with an outlet temperature of 900 C providing helium to an intermediate helium to helium heat exchanger (IHX). The outlet of the IHX is input to a three shaft horizontal Brayton Cycle power conversion system. The design constraint used in sizing the plant is based on a factory modularity principle which allows the plant to be assembled 'Lego' style instead of constructed piece by piece. This principle employs space frames which contain the power conversion system that permits the Lego-like modules to be shipped by truck or train to sites. This paper also describes the research that has been conducted at MIT since 1998 on fuel modeling, silver leakage from coated fuel particles, dynamic simulation, MCNP reactor physics modeling and air ingress analysis.

• Journal of the Korea Academia-Industrial cooperation Society
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• v.21 no.1
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• pp.409-416
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• 2020

The seismic performance of a containment structure should be secured to maintain the structural soundness of a containment structure under various earthquakes that occur globally. Therefore, an analysis of the seismic performance of a modular containment structure for a small modular reactor is also required. To analyze the seismic performance of modular containment, FEM models with contact surfaces between the modules and tendon were prepared and the modal and seismic analyses were performed. The displacement, stress, and gap size of modular containment under earthquake wave were analyzed. The effects of the tendon force, friction coefficient, and earthquake wave on the seismic performance were analyzed. The seismic performance of monolithic containment was also analyzed for comparison. In the 1st and 2nd natural modes, which most likely affect, the modular containment showed horizontal dynamic behavior, which is similar to monolithic containment, because of the combined effects of the tendon force and friction force between modules. When the combined effect is sufficient, the seismic performance of the modular containment is secured over a certain level. An additional increase in seismic performance is expected when some material with a larger friction coefficient is adopted on the contact surface.

• Nuclear Engineering and Technology
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• v.52 no.9
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• pp.2100-2106
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• 2020

This paper newly proposes an efficient analytic springback prediction method to predict the final dimensions of bent cylindrical tubes for a helical tube steam generator in a small modular reactor. Three-dimensional bending procedure is treated as a two-dimensional in-plane bending procedure by integrating the Euler beam theory. To enhance the accuracy of the springback prediction, mathematical representations of flow stress and elastic modulus for unloading are systematically integrated into the analytic prediction model. This technique not only precisely predicts the final dimensions of the bent helical tube after a springback, but also effectively predicts the various target radii. Numerical validations were performed for five different radii of helical tube bending by comparing the final radius after a springback.