• Nuclear Engineering and Technology
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• v.48 no.1
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• pp.43-54
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• 2016

In the present paper, development of the three-dimensional (3D) computational code based on Galerkin finite element method (GFEM) for solving the multigroup forward/adjoint diffusion equation in both rectangular and hexagonal geometries is reported. Linear approximation of shape functions in the GFEM with unstructured tetrahedron elements is used in the calculation. Both criticality and fixed source calculations may be performed using the developed GFEM-3D computational code. An acceptable level of accuracy at a low computational cost is the main advantage of applying the unstructured tetrahedron elements. The unstructured tetrahedron elements generated with Gambit software are used in the GFEM-3D computational code through a developed interface. The forward/adjoint multiplication factor, forward/adjoint flux distribution, and power distribution in the reactor core are calculated using the power iteration method. Criticality calculations are benchmarked against the valid solution of the neutron diffusion equation for International Atomic Energy Agency (IAEA)-3D and Water-Water Energetic Reactor (VVER)-1000 reactor cores. In addition, validation of the calculations against the $P_1$ approximation of the transport theory is investigated in relation to the liquid metal fast breeder reactor benchmark problem. The neutron fixed source calculations are benchmarked through a comparison with the results obtained from similar computational codes. Finally, an analysis of the sensitivity of calculations to the number of elements is performed.

• Proceedings of the Korean Nuclear Society Conference
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• 2005.05a
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• pp.293-294
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• 2005

The safety analyses and evaluation works on the process and facility for ACP demonstration have been performed. The several safety factors, such as the risk, environmental, radiation, structural, criticality, were analyzed. The analysis results confirmed the reliability of the safety on the ACP process and facility during normal and accident conditions.

• Proceedings of the Korean Nuclear Society Conference
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• 1997.10a
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• pp.191-196
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• 1997

As a part of International Criticality Safety Benchmark Evaluation Project (ICSBEP), SAXTON critical experiments were reevaluated. The effects on $K_{eff}$ of the uncertainties in experiment parameters, fuel rod characterization, soluble boron, critical water level, core structure, $^{241}$ Am and $^{241}$ Pu isotope number densities, random pitch error, duplicated experiment, axial fuel position, model simplification, etc., were evaluated and added in benchmark-model $k_{eff}$. In addition to detailed model, the simplified model for Saxton critical experiments was constructed by omitting the top, middle, and bottom grids and ignoring the fuel above water.r.r.

• Nuclear Engineering and Technology
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• v.20 no.1
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• pp.27-34
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• 1988

Since the lack of the spent fuel storage capcity has been expected for all Korean nuclear power plants in the mid-1990s, the maximum density rack (MDR) with consolidated fuels can be proposed to overcome the shortage of the storage capacity in KNU 9 & 10 which have most limited capacities. To ensure the safety when the alternatives are applied in the KNU 9 & 10, the multiplication factor are calculated with varying the rack pitch and the thickness of consolidated storage box by the AMPX-KENO IV codes. The computing system is verified by the benchmark calculation with criticality experiments for arrays of consolidated fuel modules, which was reported by B & W in 1981. Also an abnormal condition, i.e. malposition accident, is simulated. The results indicate that the KNU 9 & 10 storage pools with consolidated fuel are safe in the view of the criticality. Thus the storage capacity can be expanded from 9/3 cores into 27/3 cores even with considering equipments and cooling spaces.

• Resources Recycling
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• v.22 no.5
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• pp.3-12
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• 2013

It is necessary to manage risk of metals which are has rigid supply structures and expected demand expansion, considering to industry structure and resource securing capacity of each country. Countries conducted various criticality analyses and selected mainly rare metals as critical materials(or Critical metals or Critical raw materials). This study examined cases of metals risk evaluation and management which are based on technology changes and imbalance supply-demand. EU and U.S.A evaluated risk on metals needed as supply expansion of renewable energy. Japan forecasted demand of rare metals needed in Japan's growth engine industry. U.K analyzed criticality of metals, considering environmental burden occurred from mining to refining. Critical materials has features such as weak price signal, inelastic supply structure, demand volatility in technology change.

• Nuclear Engineering and Technology
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• v.34 no.3
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• pp.187-201
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• 2002

The KN-12 cask is designed to transport 12 PWR spent nuclear fuels and to comply with the requirements of Korea Atomic Energy Act, IAEA Safety Standards Series No.57-1 and US 10 CFR Part 71 for a Type B(U)F package. It provides containment, radiation shielding, structural integrity, criticality control and heat removal for normal transport and hypothetical accident conditions. W.H 14$\times$14, 16$\times$16 and 17$\times$17 fuel assemblies with maximum allowable initial enrichment of 5.0 wt.%, maximum average burn-up of 50,000 MWD/MTU and minimum cooling time of 7 years being used in Korea will be loaded and subsequently transported under dry and wet conditions. A forged cylindrical cask body which constitutes the containment vessel is closed by a cask lid. Polyethylene rods for neutron shielding are arranged in two rows of longitudinal bore holes in the cask body wall. A fuel basket to accommodate up to 12 PWR fuel assemblies provides support of the fuels, control of criticality and a path to dissipate heat. Impact limiters to absorb the impact energy under the hypothetical accident conditions are attacked at the top and at the bottom side of the cask during transport. Handling weight loaded with water is 74.8 tons and transport weight loaded with water with the impact limiters is 84.3 tons. The cask will be licensed in accordance with Korea Atomic Energy Act 3nd fabricated in Korea in accordance with ASME B&PV Code Section 111, Division 3.

• Nuclear Engineering and Technology
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• v.24 no.3
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• pp.319-328
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• 1992

The Galerkin formulation of the finite element method is applied to the integral law of the first-order form of the one-group neutron transport equation in one-dimensional spherical geometry. Piecewise linear or quadratic Lagrange polynomials are utilized in the integral law for the angular flux to establish a set of linear algebraic equations. Numerical analyses are performed for the scalar flux distribution in a heterogeneous sphere as well as for the criticality problem in a uniform sphere. For the criticality problems in the uniform sphere, the results of the finite element method, with the use of continuous finite elements in space and angle, are compared with the exact solutions. In the heterogeneous problem, the scalar flux distribution obtained by using discontinuous angular and spatical finite elements is in good agreement with that from the ANISN code calculation.

• Nuclear Engineering and Technology
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• v.10 no.2
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• pp.87-92
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• 1978

Multigroup constant calculation system for TRIGA-type reactor analysis was provided. Calculations for initial criticality, temperature coefficient, flux and power distributions of TRICA-Mark III reactor were carried out by using diffusion code CITATION. And some of results were compared with the values of start-up experiments and design values. It could be confirmed that the prepared computation system is very useful for TRIGA-type reactor analysis.

• Nuclear Engineering and Technology
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• v.25 no.3
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• pp.457-469
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• 1993

JEF-1-based 50-group cross section set for fast reactor calculations was generated using NJOY system. The set was then examined by analyzing measured integral quantities such as criticality and central reaction rate ratios for 27 fast critical assemblies. The calculated results using the new set were also compared with those of ENDF/B-IV or-V-based fast set. In general, the JEF-1-based set shows an improvement in predicting measured integral quantities in comparison with the previous set. With a few exceptions, JEF-1 results are comparable to those of ENDF/B-V.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.58-63
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• 1996

This paper is devoted to evaluating the accuracy of general 2D geometric transport code, HELIOS, and determining the order of merit in modeling for some important HELIOS input parameters. Benchmark test for 12 critical lattices show that HELIOS predicts criticality accurately within experimental uncertainties, showing only 250 pcm overestimation with a standard deviation of 450 pcm. The sensitivity test suggest that current coupling order, neutron group library, geometrical modeling, and resonance options must be considered carefully to obtain accurate results. Especially, current coupling order and sub-rings in fuel regions turn out to be most critical in HELIOS modeling. For MOX loaded cores, it is additionally necessary to pay attention to the resonance option and the validity of small group neutron library.