• Proceedings of the Korean Society of Marine Engineers Conference
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• 2006.06a
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• pp.331-332
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• 2006

연료 분사펌프는 600bar 이상의 고압 연료를 분사밸브를 통하여 디젤기관의 연소실에 연속적으로 공급하는 장치이며, PLUNGER&BARREL 은 분사펌프의 핵심 부품으로 작동하고 고온 고압의 환경에서 $4{\mu}m$ 미만의 초정밀 간극을 유지해야 한다. 이러한 정밀한 가공 상태를 유지하기 위해선 내마모성 및 내열성에 적합한 재료와 열처리가 필수 요소이다. 현재 주로 사용되고 있는 방식은 DIAMOND STONE 에 의한 HORNIG 가공을 주로 사용하기 있다. 하지만, 연삭 가공과는 달리 별도의 치수 보정 장치가 없는 관계로, STONE 마모와 그에 따르는 마모량 공차에 의한 정밀 치수 확보의 어려움으로 인해, 고비용 저효율의 약점을 가지고 있다. 그러므로 기존의 HORNING STONE 방식을 대체한 PIN 방식의 HORNING MACHINE 개발을 목표로 한다.

• Nuclear Engineering and Technology
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• v.36 no.6
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• pp.528-539
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• 2004

It is essential in commercial reactors that the safety limits imposed on the fuel pellets and fuel clad barriers, such as the linear power density (LPD) and the departure from nucleate boiling ratio (DNBR), are not violated during reactor operations. In order to accurately monitor the safety limits of current reactor states, a detailed three-dimensional (3D) core power distribution should be estimated from the in-core detector signals. In this paper, we propose a calculation methodology for detailed 3D core power distribution, using in-core detector signals and core monitoring constants such as the 3D Coupling Coefficients (3DCC), node power fraction, and pin-to-node factors. Also, the calculation method for several core safety parameters is introduced. The core monitoring constants for the real core state are promptly provided by the core design code and on-line MASTER (Multi-purpose Analyzer for Static and Transient Effects of Reactors), coupled with the core monitoring program. through the plant computer, core state variables, which include reactor thermal power, control rod bank position, boron concentration, inlet moderator temperature, and flow rate, are supplied as input data for MASTER. MASTER performs the core calculation based on the neutron balance equation and generates several core monitoring constants corresponding to the real core state in addition to the expected core power distribution. The accuracy of the developed method is verified through a comparison with the current CECOR method. Because in all the verification calculation cases the proposed method shows a more conservative value than the best estimated value and a less conservative one than the current CECOR and COLSS methods, it is also confirmed that this method secures a greater operating margin through the simulation of the YGN-3 Cycle-1 core from the viewpoint of the power peaking factor for the LPD and the pseudo hot pin axial power distribution for the DNBR calculation.

• Nuclear Engineering and Technology
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• v.47 no.7
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• pp.791-803
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• 2015

The effective cross sections (XSs) in the direct whole core calculation code nTRACER are evaluated by the equivalence theory-based resonance-integral-table method using the WIMS-based library as an alternative to the subgroup method. The background XSs, as well as the Dancoff correction factors, were evaluated by the enhanced neutron-current method. A method, with pointwise microscopic XSs on a union-lethargy grid, was used for the generation of resonance-interference factors (RIFs) for mixed resonant absorbers. This method was modified by the intermediate-resonance approximation by replacing the potential XSs for the non-absorbing moderator nuclides with the background XSs and neglecting the resonance-elastic scattering. The resonance-escape probability was implemented to incorporate the energy self-shielding effect in the spectrum. The XSs were improved using the proposed method as compared to the narrow resonance infinite massbased method. The RIFs were improved by 1% in $^{235}U$, 7% in $^{239}Pu$, and >2% in $^{240}Pu$. To account for thermal feedback, a new feature was incorporated with the interpolation of pre-generated RIFs at the multigroup level and the results compared with the conventional resonance-interference model. This method provided adequate results in terms of XSs and k-eff. The results were verified first by the comparison of RIFs with the exact RIFs, and then comparing the XSs with the McCARD calculations for the homogeneous configurations, with burned fuel containing a mixture of resonant nuclides at different burnups and temperatures. The RIFs and XSs for the mixture showed good agreement, which verified the accuracy of the RIF evaluation using the proposed method. The method was then verified by comparing the XSs for the virtual environment for reactor applicationbenchmark pin-cell problem, as well as the heterogeneous pin cell containing burned fuel with McCARD. The method works well for homogeneous, as well as heterogeneous configurations.

• Nuclear Engineering and Technology
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• v.21 no.3
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• pp.186-192
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• 1989

As a part of the design of the top end piece(TEP) for the 14$\times$14 reload fuel, various models of top end piece structure were analysed, using the ANSYS code, under fuel assembly shipping and handling load conditions. The 3-dimensional isoparametric elements were used in each model. By rearrangement of slots and holes on the adapter plate, without violating the design requirements, and also by changing the enclosure attachment method used on the adapter plate from pin joints to through-weld, the load carving capacity of the adapter plate was greatly strengthened. These concepts were adopted for the design of the 14$\times$14 reload fuel.

• Proceedings of the KSME Conference
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• 2008.11b
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• pp.3175-3180
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• 2008

A metal-fueled pool-type liquid metal fast reactor (LMFR) provides large margins to sodium boiling and fuel damage under accident conditions. The favorable passive safety results are obtained by both a reactivity feedback mechanism in the core and a passive decay heat removal system. Among the various reactivity feedbacks, the ones by a thermal expansion of a radial dimension of the core and by the control rod drivelines are strongly dependent on the flow conditions in the core and the hot pool, respectively. The effects of multidimensional thermal hydraulic characteristics on these reactivity feedbacks are investigated by the system-wide safety analysis code SSC-K with advanced thermal hydraulics models. Particularly a detailed three dimensional thermal hydraulics reactor core model is integrated into SSC-K for use in a whole system analysis of the passive safety aspects of LMR designs. The model provides fuel and cladding temperatures for every fuel pin in a reactor and coolant temperatures for every coolant sub-channel in the reactor.

• Nuclear Engineering and Technology
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• v.53 no.1
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• pp.1-18
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• 2021

The criticality analysis of VVER-1000 mock-up benchmark experiments from the LR-0 research reactor operated by the Research Center Rez in the Czech Republic has been conducted with the MCS Monte Carlo code developed at the Computational Reactor Physics and Experiment laboratory of the Ulsan National Institute of Science and Technology. The main purpose of this work is to evaluate the newest ENDF/B-VIII.0 nuclear data library against the VVER-1000 mock-up integral experiments and to validate the criticality analysis capability of MCS for light water reactors with hexagonal fuel lattices. A preliminary code/code comparison between MCS and MCNP6 is first conducted to verify the suitability of MCS for the benchmark interpretation, then the validation against experimental data is performed with both ENDF/B-VII.1 and ENDF/B-VIII.0 libraries. The investigated experimental data comprises six experimental critical configurations and four experimental pin-by-pin power maps. The MCS and MCNP6 inputs used for the criticality analysis of the VVER-1000 mock-up are available as supplementary material of this article.

• Nuclear Engineering and Technology
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• v.55 no.2
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• pp.756-768
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• 2023

Neutron transport calculations are extremely challenging due to the high computational cost of large and complex problems. A multilevel octree grid algorithm (MLTG) of discrete ordinates method was developed to improve the modeling accuracy and simulation efficiency on 3-D Cartesian grids. The Balakovo-3 VVER-1000 neutron dosimetry benchmark is calculated to verify and validate this numerical technique. A simplified S₂ synthetic acceleration is used in the MLTG calculation method to improve the convergence of the source iterations. For the triangularly arranged fuel pins, we adopt a source projection algorithm to generate pin-by-pin source distributions of hexagonal assemblies. MLTG provides accurate geometric modeling and flexible fixed source description at a lower cost than traditional Cartesian grids. The total number of meshes is reduced to 1.9 million from the initial 9.5 million for the Balakovo-3 model. The numerical comparisons show that the MLTG results are in satisfactory agreement with the conventional S_N method and experimental data, within the root-mean-square errors of about 4% and 10%, respectively. Compared to uniform fine meshing, approximately 70% of the computational cost can be saved using the MLTG algorithm for the Balakovo-3 computational model.

• Nuclear Engineering and Technology
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• v.55 no.6
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• pp.2230-2245
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• 2023

KANT (KAIST Advanced Nuclear Tachygraphy) is a PWR core simulator recently developed at Korea Advance Institute of Science and Technology, which solves three-dimensional steady-state and transient multigroup neutron diffusion equations under Cartesian geometries alongside the incorporation of thermal-hydraulics feedback effect for multi-physics calculation. It utilizes the standard Nodal Expansion Method (NEM) accelerated with various Coarse Mesh Finite Difference (CMFD) methods for neutronics calculation. For thermal-hydraulics (TH) calculation, a single-phase flow model and a one-dimensional cylindrical fuel rod heat conduction model are employed. The time-dependent neutronics and TH calculations are numerically solved through an implicit Euler scheme, where a detailed coupling strategy is presented in this paper alongside a description of nodal equivalence, macroscopic depletion, and pin power reconstruction. For validation of the steady, transient, and depletion calculation with pin power reconstruction capacity of KANT, solutions for various benchmark problems are presented. The IAEA 3-D PWR and 4-group KOEBERG problems were considered for the steady-state reactor benchmark problem. For transient calculations, LMW (Lagenbuch, Maurer and Werner) LWR and NEACRP 3-D PWR benchmarks were solved, where the latter problem includes thermal-hydraulics feedback. For macroscopic depletion with pin power reconstruction, a small PWR problem modified with KAIST benchmark model was solved. For validation of the multi-physics analysis capability of KANT concerning large-sized PWRs, the BEAVRS Cycle1 benchmark has been considered. It was found that KANT solutions are accurate and consistent compared to other published works.

• Nuclear Engineering and Technology
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• v.34 no.5
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• pp.455-467
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• 2002

RELAP5/MOD3.2 was modified to perform the thermal-hydraulic safety analysis for HANARO transients. Several aspects of RELAP5/MOD3.2 were modified or replaced by new features to properly simulate the unique HANARO characteristics such as the finned fuel element, the cooling mechanisms by both plate type heat exchanger and the natural circulation. Especially, the heat transfer packages were modified to be more appropriate for the safety analysis and the heat transfer models were developed for the plate type heat exchanger as well as natural circulation through the pool water. This modified version of RELAP5/MOD3.2 is renamed as RELAP5/HANARO. The thermal-hydraulic simulations of the single fuel pin test and plate type heat exchanger were peformed to assess the realistic predicting capabilities of RELAP5/HANARO and compared with experimental results and manufacturer＇s data in this paper. In addition, the natural circulation experiment using the scaled bundle was simulated to validate the capability of RELAP5/HANARO. The simulation results show almost similar trend with experimental data. Therefore, it is proved that RELAP5/HANARO has a confidence to use for the safety analyses of HANARO.

• Nuclear Engineering and Technology
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• v.42 no.1
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• pp.47-54
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• 2010

The SuperCritical Water-cooled Reactor (SCWR) pursues high power density to reduce its capital cost. The fast spectrum SCWR, called a super fast reactor, can be designed with a higher power density than thermal spectrum SCWR. The mechanism of increasing the average power density of the super fast reactor is studied theoretically and numerically. Some key parameters affecting the average power density, including fuel pin outer diameter, fuel pitch, power peaking factor, and the fraction of seed assemblies, are analyzed and optimized to achieve a more compact core. Based on those sensitivity analyses, a compact super fast reactor is successfully designed with an average power density of 294.8 W/$cm^3$. The core characteristics are analyzed by using three-dimensional neutronics/thermal-hydraulics coupling method. Numerical results show that all of the design criteria and goals are satisfied.