• Nuclear Engineering and Technology
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• v.33 no.6
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• pp.566-584
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• 2001

In this paper, the progressive deformation and the creep-fatigue damage for the conceptually designed reactor internals of KALIMER(Korea Advanced Liquid MEtal Reactor) are carried out by using the elastic analysis method in the RCC-MR code for normal operating conditions including the thermal load, seismic load (OBE) and dead weight. The maximum operating temperature of this reactor is 53$0^{\circ}C$ and the total service lifetime is 30 years. Thus, the time- dependent creep and stress-rupture effects become quite important in the structural design. The effects of the thermal induced membrane stress on the creep-fatigue damage are investigated with the risk of the elastic follow-up. To calculate the thermal stress, detailed thermal analyses considering conduction, convection and radiation heat transfer mechanisms are carried out with the ANSYS program. Using the results of the elastic analysis, the progressive deformation and creep-fatigue damages are calculated step by step using the RCC-MR in detail. This paper ill be a very useful guide for an actual application of the high temperature structural design of the nuclear power plant accounting for the time-dependent creep and stress-rupture effects.

• Computational Structural Engineering
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• v.19 no.1 s.71
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• pp.7-18
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• 2006

• Proceedings of the Korean Nuclear Society Conference
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• 1999.05a
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• pp.374-374
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.416-416
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 2002.10a
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• pp.137.2-137
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• 2002

• Proceedings of the Korean Nuclear Society Conference
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• 1999.10a
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• pp.158.2-158
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• 1999

• Proceedings of the Korean Nuclear Society Conference
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• 2001.05a
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• pp.202-202
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• 2001

• Transactions of the Korean Society of Mechanical Engineers A
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• v.26 no.3
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• pp.479-486
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• 2002

In this study, the progressive inelastic deformation, so called, thermal ratchet phenomenon which can occur in high temperature structures of liquid metal reactor was simulated with thermal ratchet structural test facility and 316L stainless steel test cylinder. The thermal ratchet deformation at the reactor baffle cylinder of the liquid metal reactor can occur due to the moving temperature distribution along the axial direction as the sodium free surface moves up and down under the cyclic heat-up and cool-down transients. The ratchet deformation was measured with the laser displacement sensor and LVDTs after cooling the structural specimen which is heated up to 55$0^{\circ}C$ with steep temperature gradients along the axial direction. The temperature distribution of the test cylinder along the axial direction was measured with 28 channels of thermocouples and was used for the ratchet analysis. The thermal ratchet deformation was analyzed with the constitutive equation of nonlinear combined hardening model which was implemented as ABAQUS user subroutine and the analysis results were compared with those of the test. Thermal ratchet load was applied 9 times and the residual displacement after 9 cycles of thermal load was measured to be 1.79mm. The ratcheting deformation shapes obtained by the analysis with the combined hardening model were in reasonable agreement with those of the structural tests.

• Nuclear Engineering and Technology
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• v.48 no.4
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• pp.952-964
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• 2016

The prototype generation IV sodium-cooled fast reactor (PGSFR) has been developed by the Korea Atomic Energy Research Institute. This reactor uses sodium as a reactor coolant to transfer the core heat energy to the turbine. Sodium has chemical characteristics that allow it to violently react with materials such as a water or steam. When a sodium-water reaction (SWR) occurs due to leakage or breakage of steam generator tubes, high-pressure waves and corrosive reaction products are produced, which threaten the structural integrity of the components of the intermediate heat-transfer system (IHTS) and the safety of the primary heat-transfer system (PHTS). In the PGSFR, SWR events are included in the design-basis event. This event should be analyzed from the viewpoint of the integrities of the IHTS and fuel rods. To evaluate the integrity of the IHTS based on the consequences of the SWR, the behaviors of the generated high-pressure waves are analyzed at the major positions of a failed IHTS loop using a sodium-water advanced analysis method-II code. The integrity of the fuel rods must be consistently maintained below the safety acceptance criteria to avoid the consequences of the SWR. The integrity of the PHTS is evaluated using the multidimensional analysis of reactor safety-liquid metal reactor code to model the whole plant.

• Nuclear Engineering and Technology
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• v.53 no.2
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• pp.423-429
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• 2021

The investigation of the utilization of enriched ²⁰⁸Pb as a coolant to enhance the performance of a long-life fast reactor with a Modified CANDLE (Constant Axial shape of Neutron flux, nuclide densities, and power shape During Life of Energy production) burnup scheme has performed. The analyzes were performed on a reactor with thermal power of 800 MegaWatt Thermal (MWTh) with a refueling process every 15 years. Uranium Nitride (enriched ¹⁵N), ²⁰⁸Pb, and High-Cr martensitic steel HT-9 were employed as fuel, coolant, and cladding materials, respectively. One of the Pb-nat isotopes, ²⁰⁸Pb, has the smallest neutron capture cross-section (0.23 mb) among other liquid metal coolants. Furthermore, the neutron-producing cross-section (n, 2n) of ²⁰⁸Pb is larger than sodium (Na). On the other hand, the inelastic scattering energy threshold of ²⁰⁸Pb is the highest among Na, ^natPb, and Bi. The small inelastic scattering cross-section of ²⁰⁸Pb can harden the neutron energy spectrum. Therefore, ²⁰⁸Pb is a better neutron multiplier than any other liquid metal coolant. The excess neutrons cause more production than consumption of ²³⁹Pu. Hence, it can reduce the initial fuel loading of the reactor. The selective photoreaction process was developing to obtain enriched ²⁰⁸Pb. The neutronic was calculated using SRAC and JENDL 4.0 as a nuclear data library. We obtained that the modified CANDLE reactor with enriched ²⁰⁸Pb as coolant and reflector has the highest k-eff among all reactors. Meanwhile, the ^natPb cooled reactor has the lowest k-eff. Thus, the utilization of the enriched ²⁰⁸Pb as the coolant can reduce reactor initial fuel loading. Moreover, the enriched ²⁰⁸Pb-cooled reactor has the smallest power peaking factor among all reactors. Therefore, the enriched ²⁰⁸Pb can enhance the performance of a long-life Modified CANDLE fast reactor.