• 한국통신학회논문지
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• 제29권8C호
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• pp.1034-1039
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• 2004

저장 장치의 고밀도화로 인한 인접 심벌간의 간섭 현상은 연속되는 천이 길이에 의해 가장 큰 영향을 받게 된다. 따라서, 본 논문에서는 천이 회수를 2, k-구속조건을 8로 제한한 새로운 MTR 변조 코드를 제안하였다. 이 코드에 대해 수직 자기기록 장치에서 저주파 대역에 존재하는 매체 잡음으로 인한 성능 감소를 방지하기 위해서 GS(Guided Scrambling) 방법을 이용한 DC-억압을 수행하였다. 또한, 기존의 자기기록 채널에서 사용되던 코드율 8/9인 코드와의 검출 성능을 컴퓨터 모의 실험을 통해 비교 분석하였다.

• 한국통신학회논문지
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• 제29권12C호
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• pp.1601-1605
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• 2004

연속적인 최대 천이 길이(j)를 2 이하로 제한하는 MTR 코드를 사용할 경우, 4차의 부분 응답 최대 유사도 (PRML) 검출기의 성능은 기존의 일반적인 코드를 사용한 경우에 비해 현저히 향상될 수 있으나, 낮은 코드율로인해 코드율 손실이 생긴다. 본 논문에서는 코드워드 자체에서는 MTR 구속 조건이 2를 만족하면서 코드워드를 연결하였을 때에는 j=3을 허용하는 코드율이 7/8인 코드에 대해 i=2와 j=3인 형태의 트렐리스를 결합시켜 변형된 형태의 PRML 검출 방법을 제안하였다. 이러한 결합된 트렐리스를 갖는 변형된 형태의 4차 PRML 검출 방법은 높은 기록 및도의 수평 또는 수직 자기기록 시스템에서 기존의 코드율 8/9인 코드에 대한 4차 PRML 검출 방법에 비해 10-5 BER에서 최소 2dB 이상의 SNR 성능 이득을 보임을 확인하였다.

• Nuclear Engineering and Technology
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• 제49권5호
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• pp.953-967
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• 2017

In this paper, a complete station blackout (SBO) or complete loss of electrical power supplies is simulated and analyzed in a downward cooling 5-MW pool-type Material Testing Reactor (MTR). The scenario is traced in the absence of active cooling systems and operators. The code nodalization is successfully benchmarked against experimental data of the reactor's operating parameters. The passive heat removal system includes downward water cooling after pump breakdown by the force of gravity (where the coolant streams down to the unfilled portion of the holdup tank), safety flapper opening, flow reversal from a downward to an upward cooling direction, and then the upward free convection heat removal throughout the flapper safety valve, lower plenum, and fuel assemblies. Both short-term and long-term natural core cooling conditions are simulated and investigated using the RELAP5 code. Short-term analyses focus on the safety flapper valve operation and flow reversal mode. Long-term analyses include simulation of both complete SBO and long-term operation of the free convection mode. Results are promising for pool-type MTRs because this allows operators to investigate RELAP code abilities for MTR thermal-hydraulic simulations without any oscillation; moreover, the Tehran Research Reactor is conservatively safe against the complete SBO and long-term free convection operation.

• Nuclear Engineering and Technology
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• 제52권9호
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• pp.1881-1895
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• 2020

The theoretical aspects behind the reactor depletion capability of the Monte Carlo code MCS developed at the Ulsan National Institute of Science and Technology (UNIST) and practical results of this depletion feature for a Material-Testing Reactor (MTR) with plate-type fuel are described in this paper. A verification of MCS results is first performed against MCNP6 to confirm the suitability of MCS for the criticality and depletion analysis of the MTR. Then, the dependence of the effective neutron multiplication factor to the number of axial and radial depletion cells adopted in the fuel plates is performed with MCS in order to determine the minimum spatial segmentation of the fuel plates. Monte Carlo depletion results with 37,800 depletion cells are provided by MCS within acceptable calculation time and memory usage. The results show that at least 7 axial meshes per fuel plate are required to reach the same precision as the reference calculation whereas no significant differences are observed when modeling 1 or 10 radial meshes per fuel plate. This study demonstrates that MCS can address the need for Monte Carlo codes capable of providing reference solutions to complex reactor depletion problems with refined meshes for fuel management and research reactor applications.

• 한국통신학회논문지
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• 제38A권9호
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• pp.744-749
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• 2013

본 논문에서는 수직자기기록 저장장치에서 사용되는 LDPC 부호의 패리티 부분을 (1, 7) 런-길이 제한 부호로 사용할 때, 연판정 값을 입력으로 한 경우의 성능을 조사한다. 사용자 데이터는 최대 천이 런(maximum transition run) 부호로 인코딩된다. 부호율의 손해를 최소화 하기 위하여 LDPC 부호의 패리티에만 (1, 7) 런-길이 제한 부호를 적용한다. 본 논문에서는 성능 향상을 위하여 사용자 데이터 부분에 대하여만 연판정 출력 비터비 알고리즘(soft output Viterbi algorithm, SOVA)을 사용한다. SOVA를 사용한 경우의 성능은 26dB 보다 작은 신호대잡음비에서 좋게 나타난 것에 반하여 26dB 보다 높은 신호대잡음비에서는 나쁘게 나타났다. 이것은 높은 지터 잡음과 LDPC 디코더에 두 가지 다른 형태의 입력에 기인한다.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2852-2859
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• 2020

This paper deals with the simulation-based design optimization and experimental validation of the characteristics of an in-core silver Self-Powered Neutron Detector (SPND). Optimized dimensions of the SPND are determined by combining Monte Carlo simulations and analytical methods. As a first step, the Monte Carlo transport code MCNPX is used to follow the trajectory and fate of the neutrons emitted from an external source. This simulation is able to seamlessly integrate various phenomena, including neutron slowing-down and shielding effects. Then, the expected number of beta particles and their energy spectrum following a neutron capture reaction in the silver emitter are fetched from the TENDEL database using the JANIS software interface and integrated with the data from the first step to yield the origin and spectrum of the source electrons. Eventually, the MCNPX transport code is used for the Monte Carlo calculation of the ballistic current of beta particles in the various regions of the SPND. Then, the output current and the maximum insulator thickness to avoid breakdown are determined. The optimum design of the SPND is then manufactured and experimental tests are conducted. The calculated design parameters of this detector have been found in good agreement with the obtained experimental results.

• Nuclear Engineering and Technology
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• 제51권1호
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• pp.54-59
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• 2019

The main objective of nuclear reactor safety is to maintain the nuclear fuel in a thermally safe condition with enough safety margins during normal operation and anticipated operational occurrences. In this research, core flow bypass is studied under the conditions of the unavailability of safety systems. As core bypass occurs, the core flow rate is assumed to decrease exponentially with a time constant of 25 s to new steady state values of 20, 40, 60, and 80% of the nominal core flow rate. The thermal hydraulic code PARET is used through these calculations. Reactor thermal hydraulic stability is reported for all cases of core flow bypass.

• Nuclear Engineering and Technology
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• 제52권3호
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• pp.469-476
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• 2020

Research reactors in-core experimental facilities are designed to provide the highest steady state flux for user's irradiation requirements. However, fuel conversion from highly enriched uranium (HEU) to low enriched uranium (LEU) driven by the ongoing effort to diminish proliferation risk, will impact reactor physics parameters. Preserving the reactor capability to produce the needed flux to perform its intended research functions, determines the conversion feasibility. This study investigates the neutron flux in the central experimental facility of two material test reactors (MTR), the IAEA generic10 MW benchmark reactor and the 22 MW s Egyptian Test and Research Reactor (ETRR-2). A 3D full core model with three uranium enrichment of 93%, 45%, and 20% was constructed utilizing the OpenMC particle transport Monte Carlo code. Neutronics calculations were performed for fresh fuel, the beginning of life cycle (BOL) and end of life cycle (EOL) for each of the three enrichments for both the IAEA 10 MW generic reactor and core 1/98 of the ETRR-2 reactor. Criticality calculations of the effective multiplication factor (K_eff) were executed for each of the twelve cases; results show a reasonable agreement with published benchmark values for both reactors. The thermal, epithermal and fast neutron fluxes were tallied across the core, utilizing the mesh tally capability of the code and are presented here. The axial flux in the central experimental facility was tallied at 1 cm intervals, for each of the cases; results for IAEA 10 MW show a maximum reduction of 14.32% in the thermal flux of LEU to that of the HEU, at EOL. The reduction of the thermal flux for fresh fuel was between 5.81% and 9.62%, with an average drop of 8.1%. At the BOL the thermal flux showed a larger reduction range of 6.92%-13.58% with an average drop of 10.73%. Furthermore, the fission reaction rate was calculated, results showed an increase in the peak fission rate of the LEU case compared to the HEU case. Results for the ETRR-2 reactor show an average increase of 62.31% in the thermal flux of LEU to that of the HEU due to the effect of spectrum hardening. The fission rate density increased with enrichment, resulting in 34% maximum increase in the HEU case compared to the LEU case at the assemblies surrounding the flux trap.

• Nuclear Engineering and Technology
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• 제49권4호
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• pp.700-709
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• 2017

This work looks at the effect of changes in kinetic parameters on simultaneous reactivity insertions and beam tube flooding in a typical material testing reactor-type research reactor with low enriched high density ($U_3Si_2-Al$) fuel. Using a modified PARET code, various ramp reactivity insertions (from $0.1/0.5 s to $1.3/0.5 s) plus beam tube flooding ($0.5/0.25 s) accidents under uncontrolled conditions were analyzed to find their effects on peak power, net reactivity, and temperature. Then, the effects of changes in kinetic parameters including the Doppler coefficient, prompt neutron lifetime, and delayed neutron fractions on simultaneous reactivity insertion and beam tube flooding accidents were analyzed. Results show that the power peak values are significantly sensitive to the Doppler coefficient of the system in coupled accidents. The material testing reactor-type system under such a coupled accident is not very sensitive to changes in the prompt neutron life time; the core under such a coupled transient is not very sensitive to changes in the effective delayed neutron fraction.