• Nuclear Engineering and Technology
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• 제52권3호
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• pp.499-507
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• 2020

Research reactors for radioisotope production, fuel and material testing and research activities are designed, constructed and operated based on the society's needs. In this study, neutronic and thermal hydraulic design of a high neutron flux research reactor core for radioisotope production is presented. Main parameters including core excess reactivity, reactivity variations, power and flux distribution during the cycle, axial and radial power peaking factors (PPF), Pu₂₃₉ production and minimum DNBR are calculated by nuclear deterministic codes. Core calculations performed by deterministic codes are validated with Monte Carlo code. Comparison of the neutronic parameters obtained from deterministic and Monte Carlo codes indicates good agreement. Finally, subchannel analysis performed for the hot channel to evaluate the maximum fuel and clad temperatures. The results show that the average thermal neutron flux at the beginning of cycle (BOC) is 1.0811 × 10¹⁴ n/㎠-s and at the end of cycle (EOC) is 1.229 × 10¹⁴ n/㎠-s. Total Plutonium (Pu₂₃₉) production at the EOC evaluated to be 0.9487 Kg with 83.64% grade when LEU (UO₂ with 3.7% enrichment) used as fuel. This designed reactor which uses LEU fuel and has high neutron flux and low plutonium production could be used for peaceful nuclear activities based on nuclear non-proliferation treaty concepts.

• Nuclear Engineering and Technology
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• 제56권3호
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• pp.959-965
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• 2024

The High Temperature Test Unit (HTTU) was an experimental set-up to conduct separate and integral effects tests of the Pebble Bed Modular Reactor (PBMR) core. The annular core consisted of a randomly packed bed of uniform spheres. Natural convection tests using both nitrogen and helium, and forced convection tests using nitrogen, were conducted. The maximum material temperature achieved during forced convection testing was 1200 ℃. This paper presents the numerical analysis of the flow and temperature distribution for a forced convection test using 3D CFD as well as a 1D systems-CFD computer code. Several modelling approaches are possible, ranging from a fully explicit to a semi-implicit method that relies on correlations of their associated phenomena. For the comparison between codes, the analysis was performed using a porous media approach, where the conduction and radiative heat transfer were lumped together as an effective thermal conductivity and the convective heat transfer was correlated between the solid and gas phases. The results from both codes were validated against the experimental measurements. Favourable results were obtained, in particular by the systems-CFD code with minimal computational and time requirements.

• 비파괴검사학회지
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• 제18권4호
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• pp.292-303
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• 1998

최근 자기적 특성의 변화가 재료의 미세조직적 상태에 민감하다는 보고가 있다. 재료의 미세조직은 상(phase), 결정립의 크기, 석출물 등의 타양한 인자로 구성되며, 본 연구에서는 이러한 인자들을 이용하여 재료의 특성과 Barkhausen noise(BN) 의 관계를 연관시키기 위 한 시도를 하였다. ASTM A 508 Gr. 3 원자로압력용기강을 대상으로 용접열영향부의 미세조직적인 변화를 일곱가지 대표적인 영역으로 구분하였고, 각 영역에서의 미세조직과 기계적특성의 변화를 파악한 후, 자 조건별로 BN을 측정하였다. 통상의 보자력이나 잔류자화 같은 자기적성질은 큰 변화가 없었지만, BN은 미세조직의 상태에 따라 현격한 변화를 보였다. 결정립과 석출물의 크기가 증가함에 따라 BN은 증가하였고. tempered martensite 보다는 tempered bainite 조직에서 높은 BN을 나타냈다.

• Journal of Welding and Joining
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• 제26권2호
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• pp.43-49
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• 2008

Fatigue cracks due to thermal stratification or corrosion in pipelines of nuclear power plants can cause serious problems on reactor cooling system. Therefore, the development of an integrated technology including fabrication of standard specimens and their practical usage is needed to enhance the reliability of nondestructive testing. The test material was austenitic STS 304, which is used as pipelines in the Reactor Coolant System of a nuclear power plants. The best condition for fabrication of thermal fatigue cracks at the notch plate was selected using the thermal stress analysis of ANSYS. The specimen was installed from the tensile tester and underwent continuos tension loads of 51,000N. Then, after the specimen was heated to $450^{\circ}C$ for 1 minute using HF induction heater, it was cooled to $20^{\circ}C$ in 1 minute using a mixture of dry ice and water. The initial crack was generated at 17,000 cycles, 560 hours later (1cycle/2min.) and the depth of the thermal fatigue crack reached about 40% of the thickness of the specimen at 22,000 cycles. As a results of optical microscope and SEM analysis, it is confirmed that fabricated thermal fatigue cracks have the same characteristics as real fatigue cracks in nuclear power plants. The crack shape and size were identified.

• 대한전기학회:학술대회논문집
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• 대한전기학회 2000년도 추계학술대회 논문집 학회본부 A
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• pp.265-268
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• 2000

In this study, a small punch creep (SP-Creep) test using miniaturized specimen has been described for the development of the new creep test method for high temperature structural components such as headers and tubes of boiler, turbine casing and rotor, and reactor vessel. The SP-Creep testing technique has been applied to 1Cr-0.5Mo steel used widely as boiler header material and the creep test temperature are varied at $550^{\circ}C{\sim}600^{\circ}C$. From the experimental results, e.g. SP-Creep curve behaviors, the creep rate in steady state and creep rupture life with test temperature and load, the load exponential value(n, m), the activation energy($Q_{spc}$), the Monkman-Grant relation and the creep life assessment equation etc., it can be summarized that the SP-Creep test may be a useful test method to evaluate the creep properties of the heat resisting material such as boiler header.

• Nuclear Engineering and Technology
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• 제52권12호
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• pp.2725-2732
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• 2020

This paper presents the neutronics benchmark analysis of the first core of the Indonesian multipurpose research reactor RSG-GAS (Reaktor Serba Guna G.A. Siwabessy) calculated by the Serpent Monte Carlo code and the newly released ENDF/B-VIII.0 nuclear data library. RSG-GAS is a 30 MWth pool-type material testing research reactor loaded with plate-type low-enriched uranium fuel using light water as a coolant and moderator and beryllium as a reflector. Two groups of critical benchmark problems are derived on the basis of the criticality and control rod calibration experiments of the first core of RSG-GAS. The calculated results, such as the neutron effective multiplication factor (k) value and the control rod worth are compared with the experimental data. Moreover, additional calculated results, including the neutron spectra in the core, fission rate distribution, burnup calculation, sensitivity coefficients, and kinetics parameters of the first core will be compared with the previous nuclear data libraries (interlibrary comparison) such as ENDF/B-VII.1 and JENDL-4.0. The C/E values of ENDF/B-VIII.0 tend to be slightly higher compared with other nuclear data libraries. Furthermore, the neutron reaction cross-sections of ¹⁶O, ⁹Be, ²³⁵U, ²³⁸U, and S(𝛼,𝛽) of ¹H in H₂O from ENDF/B-VIII.0 have substantial updates; hence, the k sensitivities against these cross-section changes are relatively higher than other isotopes in RSG-GAS. Other important neutronics parameters such as kinetics parameters, control rod worth, and fission rate distribution are similar and consistent among the nuclear data libraries.

• 비파괴검사학회지
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• 제20권3호
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• pp.183-190
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• 2000

본 연구에서는 고속 중성자가 조사된 원자로 용기 재료의 자기적 성질 변화를 $M{\ddot{o}}ssbauer$ 분광법을 이용하여 측정하였으며, X-선 회절실험을 이용하여 중성자 조사재료의 결정성 변화를 평가하였다. 시편은 크기가 $23mm{\times}18mm{\times}70{\mu}m$로 제작되었으며, 343K에서 $10^{12}n/cm^2{\sim}10^{18}n/cm^2$의 범위에서 중성자 조사를 하였다. X-선 회절실험 결과로부터, $10^{16}n/cm^2$의 중성자가 조사된 시료에서부터 결정성이 변화가 시작되고, $10^{17}n/cm^2$ 이상의 중성자가 조사된 시료에서 결정성이 심각하게 손상되는 것이 관찰되었다. 또한 $M{\ddot{o}}ssbauer$ 분광실험으로부터 중성자 조사량이 $10^{16}n/cm^2$ 이하인 시료에서는 자기적성질의 변화가 관찰되지 않았으나, $10^{17}n/cm^2$ 이상의 중성자가 조사된 시료에서 자기완화 현상이 일어나는 것이 관찰되었다. 따라서 두 실험 모두 비파괴적 실험방법에 따른 중성자 조사취화 평가로 활용이 가능한 것으로 평가되었다.

• 한국분말야금학회:학술대회논문집
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• 한국분말야금학회 2006년도 Extended Abstracts of 2006 POWDER METALLURGY World Congress Part2
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• pp.1161-1162
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• 2006

In order to overcome the recrystallization embrittlement and irradiation embrittlement of Mo, which are major problems for its fusion applications, internally nitrided Mo alloys were prepared by a novel multi-step internal nitriding. Neutron irradiation was performed in the Japan Material Testing Reactor (JMTR). After irradiation, nitrided Mo alloys exhibited $\iota$ ower ductile-brittle transition temperature than irradiated TZM. These results suggested that multi-step internal nitriding was effective to the improvement in the embrittlement by irradiation. Transmission electron microscope observation revealed that TiN particles precipitated by nitriding acted as a sink for irradiation-induced defects.

• Journal of Magnetics
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• 제22권1호
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• pp.34-42
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• 2017

Material defects in the Steam Generator Tubes (SGT) of sodium cooled fast breeder reactor (PFBR) can lead to leakage of water into sodium. The water and sodium reaction will lead to major accidents. Therefore, the examination of steam generator tubes for the early detection of defects is an important requirement for safety and economic considerations. In this work, the Magnetic Flux Leakage (MFL) based Non Destructive Testing (NDT) technique is used to perform the defect detection process. The rectangular notch defects on the outer surface of steam generator tubes are modeled using COMSOL multiphysics 4.3a software. The obtained MFL images are de-noised to improve the integrity of flaw related information. Grey Level Co-occurrence Matrix (GLCM) features are extracted from MFL images and taken as input parameter to train the neural network. A comparative study on characterization have been carried out using feed-forward back propagation (FFBP) and cascade-forward back propagation (CFBP) algorithms. The results of both algorithms are evaluated with Mean Square Error (MSE) as a prediction performance measure. The average percentage error for length, depth and width are also computed. The result shows that the feed-forward back propagation network model performs better in characterizing the defects.

• 비파괴검사학회지
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• 제20권5호
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• pp.373-380
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• 2000

파괴적인 방법으로 재료나 설비의 기계적 성질을 평가하기 위해서는 많은 경비와 시간이 소요되므로 비파괴적인 방법을 이용한 재료의 기계적 성질 평가가 주목을 받고 있다. 본 논문에서는 초음파 특성을 이용하여 열처리 온도와 조건에 따른 기계적 성질 변화를 평가하였다. 초음파의 감쇠계수는 열처리 온도가 높을수록 결정립의 크기가 커지므로 증가하였다. 초음파의 감쇠계수는 열처리(담금질, 템퍼링, 용접후열처리)가 진행됨에 따라 감소하였다. 초음파 속도는 담금질과 템퍼링/PWHT한 시편에서 40 m/s의 차이를 보이고 있었다. 감쇠계수와 인성과의 상관성은 매우 좋으며 이러한 상관관계는 단조한 압력용기의 비파괴 평가에 이용할 수 있으며, 더욱이 현장 적용에 효율적으로 응용할 수 있을 것이다.