• Nuclear Engineering and Technology
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• 제41권8호
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• pp.1115-1126
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• 2009

As one of the ways to enhance the economical features in sodium-cooled fast reactor development, the concept of an integrated steam generator and pump system (ISGPS) is proposed from a structural point of view. And the related intermediate heat transfer system (IHTS) piping layout compatible with the ISGPS is described in detail. To assure the creep design lifetime of 60 years, the structural integrity is investigated through high temperature structural evaluation procedures by the SIE ASME-NH computer code, which implements the ASME-NH design rules. From the results of this study, it is found that the proposed ISGPS concept is feasible and applicable to a commercial SFR design.

• Transactions of the Korean Society of Pressure Vessels and Piping
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• 제12권1호
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• pp.126-133
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• 2016

Recently, the research and development of Sodium-cooled Fast Reactors (SFRs) have made progresses. However, liquid sodium, the coolant of an SFR, is chemically unstable and sodium fire can be occurred when liquid sodium leaks from sodium pipe. To reduce the damage by the sodium fire, many fire walls and fire extinguishers are needed for SFRs. LBB concept in SFR might reduce the scale of sodium fire and decrease or eliminate fire walls and fire extinguishers. Therefore, LBB concept can contribute to improve economic efficiency and to strengthen defense-in depth safety. The LBB assessment procedure has been well established, and has been used significantly in light water reactors (LWRs). However, an LBB assessment of an SFR is more complicated because SFRs are operated in elevated temperature regions. In such a region, because creep damage may occur in a material, thereby growing defects, an LBB assessment of an SFR should consider elevated temperature effects. The procedure and method for this purpose are provided in RCC-MRx A16, which is a French code. In this study, LBB assessment was performed for PGSFR IHTS hot-leg pipe according to RCC-MRx A16 and the applicability of the code was discussed.

• Journal of the Korean Society for Nondestructive Testing
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• 제28권4호
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• pp.364-371
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• 2008

Ultrasonic waveguide sensor has been developed for under-sodium viewing of reactor internal structures of a sodium-cooled fast reactor (SFR). The structure design concept of a waveguide sensor assembly was suggested and evaluated for the application in SFR. A 10 m long ultrasonic waveguide sensor assembly has been manufactured and the experimental feasibility tests were carried out. The 10 m long distance propagation performance of zero-order antisymmetric $A_0$ Lamb wave has been verified. The feasibility of ultrasonic waveguide sensor has been demonstrated by the C-scanning resolution performance test.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1528-1539
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• 2023

Since the leakage of sodium in an SFR (sodium-cooled fast reactor) causes an explosion upon reaction with air and water, sodium leakages represent an important safety issue. In this study, a novel technique for improving the reliability of sodium leakage detection applying DDVR (dynamic data validation and reconciliation) is proposed and verified to resolve this technical issue. DDVR is an approach that aims to improve the accuracy of a target system in a dynamic state by minimizing random errors, such as from the uncertainty of instruments and the surrounding environment, and by eliminating gross errors, such as instrument failure, miscalibration, or aging, using the spatial redundancy of measurements in a physical model and the reliability information of the instruments. DDVR also makes it possible to estimate the state of unmeasured points. To validate this approach for supporting sodium leakage detection, this study applies experimental data from a sodium leakage detection experiment performed by the Korea Atomic Energy Research Institute. The validation results show that the reliability of sodium leakage detection is improved by cooperation between DDVR and hardware measurements. Based on these findings, technology integrating software and hardware approaches is suggested to improve the reliability of sodium leakage detection by presenting the expected true state of the system.

• Proceedings of the KSME Conference
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• 대한기계학회 2007년도 춘계학술대회A
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• pp.285-290
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• 2007

KALIMER-600 is a sodium cooled fast reactor with a fast spectrum neutron reactor core. The NSSS design has three heat transport systems of a PHTS (Primary Heat Transport System), a IHTS (Intermediate Heat Transport System) and a SGS (Steam Generation System). PHTS is a pool type and has a large amount of sodium in the pool. The mechanical design targets are maintaining the enough structural integrity for a seismic load of SSE 0.3g and the thermal and mechanical loads by the high temperature environments and an economical competitiveness when compared with other reactor types.

• Journal of computational fluids engineering
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• 제21권3호
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• pp.8-14
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• 2016

A numerical analysis is conducted to estimate the core expansion and the energy behaviors induced by a core disruptive accident in a sodium-cooled fast reactor. The numerical formulation based on underwater explosion theory is carried out to simulate the core explosion inside the reactor vessel. The transient pressure, temperature and expansion of the core are examined by solving the equation of state and nonlinear governing equation of momentum conservation in one-dimensional spherical coordinates. The energy balance inside the computation domain is examined during the core expansion process. Heat transfer between the core and the sodium coolant, and the bubble rise during the expansion process are briefly investigated.

• 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.

• Korean Journal of Metals and Materials
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• 제48권5호
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• pp.410-416
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• 2010

Studies were carried out to establish the technology for sodium-clad compatibility and to analyze the compatibility behavior of the Sodium-cooled Fast Reactor (SFR) cladding material under a flowing sodium environment. The natural circulation facility caused by the thermal convection of the liquid sodium was constructed and the 316-series stainless steels were exposed at $650{^{\circ}C}$ liquid sodium for 1458 hours. The weight change and related microstructural change were analyzed. The results showed that the quasi-dynamic facility represented by the natural convection exhibited similar results compared to the conventional dynamic facility. Selective leaching and local depletion of the chromium, re-distribution of the carbide, and the decarburization process took place in the 316-series stainless steel under a flowing sodium environment. This process decreased as the sodium flowed along the channel, which was caused by the change in the dissolved oxygen and carbon activity in the liquid sodium.

• Journal of computational fluids engineering
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• 제21권1호
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• pp.19-29
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• 2016

A Prototype Gen-IV Sodium-cooled Fast Reactor which is one of the $4^{th}$ generation nuclear reactors is in development by Korea Atomic Energy Research Institute. The reactor is composed of four main fluid systems which are categorized by its functions, i.e., Primary Heat Transport System, Intermediate Heat Transport System, Decay Heat Removal System and Sodium-Water Reaction Pressure Relief System. The coolant of the reactor is liquid sodium and sodium-to-sodium heat exchangers are installed at the interfaces between two fluid systems, Intermediate Heat Exchangers between the Primary Heat Transport System and the Intermediate Heat Transport System and Decay Heat Exchangers between the Primary Heat Transport System and the Decay Heat Removal System. For the design and performance analysis of the Intermediate Heat Exchanger and the Decay Heat Exchanger, a computer code was written during previous step of research. In this work, the computer code named "SHXSA" has been validated preliminarily by computational fluid dynamics simulations.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• 제10권1호
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• pp.21-26
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• 2012

Microstructural and mechanical property were evaluated at the medium-sized HT9 (12Cr-1MoWV) forged steel which was considered as primary candidate for the fuel cladding in sodium-cooled fast reactor (SFR). Material was forged at $1170^{\circ}C$ after the induction melting to make round bar as 160mm diameter, 7000mm length then the radial distribution of microstructure as well as microhardness was evaluated. The results showed that overall microstructure exhibited as ferrite-martensite structure, where small amount (2~3%) of delta ferrite was formed throughout the specimen and maximum 15% of transformed ferrite was formed at the center, where it gradually decreased toward the radial direction. Sensitivity analysis of the cooling curve and Time-Temperature-Transformation (TTT) diagram revealed that formation of transformed ferrite could be avoided when the diameter was decreased down to 120mm.