• 광학세계
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• 통권122호
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• pp.22-24
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• 2009

광학설계 프로그램 CODE V를 국내에 공급하고 있는 (주)모던하이테크(대표 김명중, www.okdomern.com)이 인하대학교 광기술교육센터 및 미Optical Research Associates(이하 ORA)와 함께 지난 6월 18일 서울 교육문화회관에서 '2009 CODE V User 컨퍼런스'를 개최하였다. CODE V 최신 버전 소개 및 특별강연이 진행된 이 행사에는 ORA사의 조지 베이스 사장이 참석하여 직접 비전을 소개하고 향후 국내 유저들을 위해 적극적인 서비스 지원을 약속했다.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2002년도 하계학술대회 논문집 C
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• pp.1964-1965
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• 2002

This paper reports on the single mode fiber coupling efficiency. There is difference between simulation by CODE-V and theoretical result, we observed CEO which is taken by experiment in lab and found out that CEO taken by CODE-V was closer to real experiment result then theoretical result.

• Nuclear Engineering and Technology
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• 제54권11호
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• pp.4338-4347
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• 2022

The CUPID code was developed and is continuously updated in KAERI. Verification and validation (V&V) is mainly done for light water reactors (LWRs). This paper describes a benchmarking of the detailed mesh level compared with sub-channel level for application to pressurized heavy water reactors (PHWRs), even though component scale comparison for the PHWR moderator system was done once before. We completed a sub-channel level comparison between the CUPID code and the ASSERT code and a CUPID code analysis. Because the ASSERT code has already been validated with numerous experiments, benchmarking with the ASSERT code will offer us more trust on the CUPID code. The target channel has high power and thus high pressure deformation. The high power channel tends to have a high possibility of critical heat flux (CHF), because a high void fraction and quality in channel exit region appear. In this research, after determining the reference grid and T/H model, we compared the sub-channel level results of the CUPID code with those of the ASSERT code.

• 대한수학회지
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• 제59권1호
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• pp.193-204
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• 2022

We present construction methods for free self-orthogonal (self-dual or Type II) codes over ℤ₄[v]/〈v² + 2v〉 which is one of the finite commutative local non-chain Frobenius rings of order 16. By considering their Gray images on ℤ₄, we give a construct method for a code over ℤ₄. We have some new and optimal codes over ℤ₄ with respect to the minimum Lee weight or minimum Euclidean weight.

• 한국해양공학회:학술대회논문집
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• 한국해양공학회 2001년도 추계학술대회 논문집
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• pp.68-72
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• 2001

Two different methods of free span analysis of offshore pipelines by DnV codes were introduced and compared in order to calculate the allowable free span lengths of the offshore pipelines. The allowable span lengths of the offshore pipelines for installation, hydrotest and operation conditions by static and dynamic span analysis were determined. Static analysis was performed by ASME codes and dynamic span analysis was performed by both 1981 DnV code. Comparison of two codes were carried out. A new design procedure to calculate the allowable span lengths was developed with new DnV code.

• 대한기계학회:학술대회논문집
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• 대한기계학회 2007년도 춘계학술대회B
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• pp.3485-3490
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• 2007

The TEXAS-V code tuned for TROI-13 was used for analyzing the parametric findings in TROI experiments. The calculations on the melt composition are relatively similar to the TROI experimental results. The water depth effect in TEXAS-V code seems to be consistent with TROI experiments in some degree. The water area effect of TEXAS-V calculations seems not to be harmonious to that in TROI experiments. This seems to indicate that TEXAS-V as 1-dimensional code or as the numerical steam explosion has a limitation on estimating area effect. Thus, TEXAS-V tuned for TROI-13 seems to have an ability to estimate the parametric effect of TROI experiments. The evaluated TEXAS-V was used for estimating the ex-vessel steam explosion load. The calculated explosion pressure and load were about 40 MPa and 75 kPa.sec, which are not much threatening level for containment integrity, but are arguable value for the integrity.

• 대한수학회지
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• 제58권1호
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• pp.29-44
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• 2021

The set D of column vectors of a generator matrix of a linear code is called a defining set of the linear code. In this paper we consider the problem of constructing few-weight (mainly two- or three-weight) linear codes from defining sets. It can be easily seen that we obtain an one-weight code when we take a defining set to be the nonzero codewords of a linear code. Therefore we have to choose a defining set from a non-linear code to obtain two- or three-weight codes, and we face the problem that the constructed code contains many weights. To overcome this difficulty, we employ the linear codes of the following form: Let D be a subset of ��2n, and W (resp. V ) be a subspace of ��2 (resp. ��2n). We define the linear code ��D(W; V ) with defining set D and restricted to W, V by $${\mathcal{C}}_D(W;V )=\{(s+u{\cdot}x)_{x{\in}D^{\ast}}|s{\in}W,u{\in}V\}$$. We obtain two- or three-weight codes by taking D to be a Vasil'ev code of length n = 2m - 1(m ≥ 3) and a suitable choices of W. We do the same job for D being the complement of a Vasil'ev code. The constructed few-weight codes share some nice properties. Some of them are optimal in the sense that they attain either the Griesmer bound or the Grey-Rankin bound. Most of them are minimal codes which, in turn, have an application in secret sharing schemes. Finally we obtain an infinite family of minimal codes for which the sufficient condition of Ashikhmin and Barg does not hold.

• Nuclear Engineering and Technology
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• 제37권6호
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• pp.587-594
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• 2005

In an effort to develop a thermal-hydraulic (TH) safety analysis code for Gas-cooled Reactors (GCRs), the MARS code, which was primarily developed for TH analysis of water reactor systems, has been extended here for application to GCRs. The modeling requirements of the system code were derived from a review of major processes and phenomena that are expected to occur during normal and accident conditions of GCRs. Models fur code improvement were then identified through a review of existing MARS code capability. Among these, the following priority models necessary fur the analysis of limiting high and low pressure conduction cooling events were evaluated and incorporated in MARS-GCR/V1 : 1) Helium (He) and Carbon Dioxide ($CO_2$) as main system fluids, 2) gas convection heat transfer, 3) radiation heat transfer, and 4) contact heat transfer models. Each model has been assessed using various conceptual problems for code-to-code benchmarks and it was demonstrated that MARS-GCR/V1 is capable of capturing the relevant phenomena. This paper describes the models implemented in MARS-GCR/V1 and their verification and validation results.

• Nuclear Engineering and Technology
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• 제51권2호
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• pp.356-368
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• 2019

This paper presents the verification and validation (V&V) of the STREAM/RAST-K 2.0 code system for a pressurized water reactor (PWR) analysis. A lattice physics code STREAM and a nodal diffusion code RAST-K 2.0 have been developed by a computational reactor physics and experiment laboratory (CORE) of Ulsan National Institute of Science and Technology (UNIST) for an accurate two-step PWR analysis. The calculation modules of each code were already verified against various benchmark problems, whereas this paper focuses on the V&V of linked code system. Three PWR type reactor cores, OPR-1000, three-loop Westinghouse reactor core, and APR-1400, are selected as V&V target plants. This code system, for verification, is compared against the conventional code systems used for the calculations in nuclear design reports (NDRs) and validated against measured plant data. Compared parameters are as follows: critical boron concentration (CBC), axial shape index (ASI), assembly-wise power distribution, burnup distribution and peaking factors. STREAM/RAST-K 2.0 shows the RMS error of critical boron concentration within 20 ppm, and the RMS error of assembly power within 1.34% for all the cycles of all reactors.

• Nuclear Engineering and Technology
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• 제54권9호
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• pp.3494-3515
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• 2022

This paper presents the development of a nodal diffusion code, RAST-V, and its verification and validation for VVER (vodo-vodyanoi energetichesky reactor) analysis. A VVER analytic solver has been implemented in an in-house nodal diffusion code, RAST-K. The new RAST-K version, RAST-V, uses the triangle-based polynomial expansion nodal method. The RAST-K code provides stand-alone and two-step computation modes for steady-state and transient calculations. An in-house lattice code (STREAM) with updated features for VVER analysis is also utilized in the two-step method for cross-section generation. To assess the calculation capability of the formulated analysis module, various verification and validation studies have been performed with Rostov-II, and X2 multicycles, Novovoronezh-4, and the Atomic Energy Research benchmarks. In comparing the multicycle operation, rod worth, and integrated temperature coefficients, RAST-V is found to agree with measurements with high accuracy which RMS differences of each cycle are within ±47 ppm in multicycle operations, and ±81 pcm of the rod worth of the X2 reactor. Transient calculations were also performed considering two different rod ejection scenarios. The accuracy of RAST-V was observed to be comparable to that of conventional nodal diffusion codes (DYN3D, BIPR8, and PARCS).