• Nuclear Engineering and Technology
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• 제46권5호
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• pp.641-654
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• 2014

A new computer code, named CORONA (Core Reliable Optimization and thermo-fluid Network Analysis), was developed for the core thermo-fluid analysis of a prismatic gas cooled reactor. The CORONA code is targeted for whole-core thermo-fluid analysis of a prismatic gas cooled reactor, with fast computation and reasonable accuracy. In order to achieve this target, the development of CORONA focused on (1) an efficient numerical method, (2) efficient grid generation, and (3) parallel computation. The key idea for the efficient numerical method of CORONA is to solve a three-dimensional solid heat conduction equation combined with one-dimensional fluid flow network equations. The typical difficulties in generating computational grids for a whole core analysis were overcome by using a basic unit cell concept. A fast calculation was finally achieved by a block-wise parallel computation method. The objective of the present paper is to summarize the motivation and strategy, numerical approaches, verification and validation, parallel computation, and perspective of the CORONA code.

• Nuclear Engineering and Technology
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• 제46권5호
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• pp.595-604
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• 2014

SILER (Seismic-Initiated event risk mitigation in LEad-cooled Reactors) is a Collaborative Project, partially funded by the European Commission in the $7^{th}$ Framework Programme, aimed at studying the risk associated to seismic-initiated events in Generation IV Heavy Liquid Metal reactors, and developing adequate protection measures. The project started in October 2011, and will run for a duration of three years. The attention of SILER is focused on the evaluation of the effects of earthquakes, with particular regards to beyond-design seismic events, and to the identification of mitigation strategies, acting both on structures and components design. Special efforts are devoted to the development of seismic isolation devices and related interface components. Two reference designs, at the state of development available at the beginning of the project and coming from the $6^{th}$ Framework Programme, have been considered: ELSY (European Lead Fast Reactor) for the Lead Fast Reactors (LFR), and MYRRHA (Multi-purpose hYbrid Research Reactor for High-tech Applications) for the Accelerator-Driven Systems (ADS). This paper describes the main activities and results obtained so far, paying particular attention to the development of seismic isolators, and the interface components which must be installed between the isolated reactor building and the non-isolated parts of the plant, such as the pipe expansion joints and the joint-cover of the seismic gap.

• 한국자원리싸이클링학회:학술대회논문집
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• 한국자원리싸이클링학회 2006년도 고분자리싸이클링 심포지엄
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• pp.15-32
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• 2006

Biomass had been attracted public attention as eco-friendly resource which not increases the greenhouse gas like carbon dioxide. In this study, it had been collected pyrolytic products such as bio-oil, char and pyrolytic gas from various biomass in a fluidized bed reactor which is one of the fast pyrolysis processes. To understand the characteristics of biomass pyrolysis, the variation of products yield and chemical composition was determined with various operating parameters like temperature, gas velocity($U_{0}/U_{mf}$) and bed height(L/D). In the optimum operating conditions, gas yield and water content was the lowest and concentration of guaiacols and syringols were the highest. The maximum yields of bio-oil was from 55% to 58% at $400^{\circ}C$.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1269-1286
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• 2017

This paper presents a newly developed resonance self-shielding method based on Tone's method in $APOLLO3^{(R)}$ for fast and thermal reactor calculations. The new method is based on simplified models, the narrow resonance approximation for the slowing down source and Tone's approximation for group collision probability matrix. It utilizes mathematical probability tables as quadrature formulas in calculating effective cross-sections. Numerical results for the ZPPR drawer calculations in 1,968 groups show that, in the case of the double-column fuel drawer, Tone's method gives equivalent precision to the subgroup method while markedly reducing the total number of collision probability matrix calculations and hence the central processing unit time. In the case of a single-column fuel drawer with the presence of a uranium metal material, Tone's method obtains less precise results than those of the subgroup method due to less precise heterogeneous-homogeneous equivalence. The same options are also applied to PWR UOX, MOX, and Gd cells using the SHEM 361-group library, with the objective of analyzing whether this energy mesh might be suitable for the application of this methodology to thermal systems. The numerical results show that comparable precision is reached with both Tone's and the subgroup methods, with the satisfactory representation of intrapellet spatial effects.

• Nuclear Engineering and Technology
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• 제55권4호
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• pp.1468-1475
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• 2023

Enhancement of wear resistance of components used in fast reactors is necessary for long service life of the components. Plasma nitriding is a promising surface modification technology to impart high hardness and improved wear resistance of various steel components. This study discusses the characterization of chrome nitrided SS316L casing ring used in secondary sodium pump of fast breeder reactor and its stability under long term sodium exposure. Microstructural and hardness analysis showed that stress relieved component could be chrome nitrided successfully to a thickness of about 100 ㎛. Assessment of in-sodium performance of the chrome nitrided casing ring subjected to long term exposure up to 5000h at 550℃, showed retention of chrome nitrided layer with a case depth almost similar to that before sodium exposure. A slight decrease in the hardness was observed due to prolonged high temperature sodium exposure. Tribological studies indicate very low coefficient of friction indicating the retention of good wear resistance of the coating even after long term sodium exposure.

• Nuclear Engineering and Technology
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• 제55권7호
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• pp.2419-2431
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• 2023

Experiments related to Boron Neutron Capture Therapy (BNCT) accomplished at the Institute of Nuclear Techniques (INT), Budapest University of Technology and Economics (TUB) are presented. Relevant investigations are required before designing BNCT for vivo applications. Samples of relevant boron compounds (H₃BO₃, BDTPA) usually employed in BNCT were investigated with neutron beam. Channel #5 in the research reactor (100 kW) of INT-TUB provides the neutron beam. Boron samples are mounted on a carrier for neutron irradiation. The particle attenuation of several carrier materials was investigated, and the one with the lowest attenuation was selected. The effects of boron compound type, mass, and compound phase state were also investigated. To detect the emitted charged particles, a traditional ZnS(Ag) detector was employed. The neutron beam's interaction with the detector-detecting layer is investigated. Graphite (as a moderator) was employed to change the neutron beam's characteristics. The fast neutron beam was also thermalized by placing a portable fast neutron source in a paraffin container and irradiating the H₃BO₃. The obtained results suggest that the direct measurement approach appears to be insufficiently sensitive for determining the radiation dose committed by the Alpha particles from the ¹⁰B (n,α) reaction. As a result, a new approach must be used.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(2)
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• pp.515-519
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• 1997

The purpose of this study is to reduce the fast neutron fluence at the reactor pressure vessel(RPV) and to provide a basis for plant-life extension. In this study, different neutron absorbers were employed in the core outer assemblies of Kori Unit 1 Cycle 14. The modified assemblies were used to calculate fast neutron fluence at the RPV and to evaluate reduction of outer assembly power and total power in core. By comparison with the case of no suppression fixture, the fast neutron fluence of a case with two rows stainless steel around the assembly with natural uranium pins is decreased by 85.8%. It is noted that the modification of outer assembly is more efficient than the previous low leakage loading pattern (LLLP) applied to Kori Unit 1. Also, compared fast neutron fluence in Cycle 1 with Cycle 14, fast neutron fluence at the RPV between Cycle 1 and Cycle 14 is not significantly different. It is found that LLLP applied to the Kori Unit 1 has not contributed to fast neutron fluence reduction at the RPV.

• 한국생산제조학회지
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• 제8권4호
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• pp.123-129
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• 1999

Although type 316 stainless steel is widely used such as in reactors of petrochemical plants and pipes of steam power plants and s attracting attention as potential basic material for the fast breeder reactor structure alloys in nuclear power plants and is attracting attention as potential basic material for the fast breeder reactor structure alloys in nuclear power plants the effect of precipitates which form during the long term exposure at service temperature on creep properties is not known sufficiently. In this study to investigate the creep properties and the influence of prior aging on the microstructure to form precipitates specimens were first solutionized at 113$0^{\circ}C$ for 20 minutes and then aged for different times of 0 hr, 100 hrs, 1000 hrs and 2200 hrs at 75$0^{\circ}C$ After heat treatments tensile tests both at room temperature and $650^{\circ}C$ and constant load creep ruptuere tests were carried out.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.495-501
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• 1997

A preliminary and parametric sensitivity study on recriticality risk of fast reactor cores after a hypothetical total core meltdown accident was performed. Only neutronic aspects of the accident were considered, independent of the accident scenario, and efforts were made to estimate the quantity of molten fuel which must be ejected out of the core to assure a sub-critical state after the accident. Two types of parameters were examined : characteristic parameters of molten core such as geometry, molten pool type (homogenized or stratified), fuel temperature, environment, and relative parameters to normal core such as core size(small or large), and fuel type (oxide, nitride, metal). The first type of parameters was found to intervene more directly in the recriticality risk than the second type of parameters.

• Nuclear Engineering and Technology
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• 제28권5호
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• pp.488-499
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• 1996

We have extended the source projection analytic nodal discrete ordinates method (SPANDOM) for more flexible applicability in analysis of hexagonal assembly cores. The method (SPANDOM-FH) does not invoke transverse integration but instead solves the discrete ordinates equation analytically after the source term is projected and represented in hybrid form of high-order polynomials and exponential functions. SPANDOM-FH which treats a hexagonal node as one node is applied to two fast reactor benchmark problems and compared with TWOHEX. The results of comparison indicate that the present method SPANDOM-FH predicts accurately $k_eff$ and flux distributions in hexagonal assembly cores. In addition, SPANDOM-FH gives the continuous two dimensional intranodal scalar flux distributions in a hexagonal node. The reentering models between TWOHEX and SPANDOM were also compared and it was confirmed that SPANDOM's model is more realistic. Through the results of benchmark problems, we conclude that SPANDOM-FH has the sufficient accuracy for the nuclear design of fast breeder reactor (FBR) cores with hexagonal assemblies.