• Nuclear Engineering and Technology
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• 제55권9호
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• pp.3213-3228
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• 2023

As an important part of the digital reactor, the pin-by-pin wise fine coupling calculation is a research hotspot in the field of nuclear engineering in recent years. It provides more precise and realistic simulation results for reactor design, operation and safety evaluation. CORCA-K a nodal code is redeveloped as a robust pin-by-pin wise neutronics and thermal-hydraulic coupled calculation code for pressurized water reactor (PWR) core. The nodal green's function method (NGFM) is used to solve the three-dimensional space-time neutron dynamics equation, and the single-phase single channel model and one-dimensional heat conduction model are used to solve the fluid field and fuel temperature field. The mesh scale of reactor core simulation is raised from the nodal-wise to the pin-wise. It is verified by two benchmarks: NEACRP 3D PWR and PWR MOX/UO₂. The results show that: 1) the pin-by-pin wise coupling calculation system has good accuracy and can accurately simulate the key parameters in steady-state and transient coupling conditions, which is in good agreement with the reference results; 2) Compared with the nodal-wise coupling calculation, the pin-by-pin wise coupling calculation improves the fuel peak temperature, the range of power distribution is expanded, and the lower limit is reduced more.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1997년도 춘계학술발표회논문집(1)
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• pp.15-20
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• 1997

We present the results HELIOS verification against VENUS PWR critical experiments loaded with high plutonium content mixed oxides fuels. The effective multiplication factors are calculated to be slightly supercritical within an acceptable error bound. In the prediction of power shape, HELIOS results are in close agreement with the measured values. The RMS errors of re-normalized calculated fission rate distribution are less than 1.4 % with either explicit or implicit models or micro tubes/rods in each fuel assembly for both ALL-MOX and GD-MOX mock-up cores.

• 방사성폐기물학회지
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• 제13권4호
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• pp.243-251
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• 2015

국내 원자력발전은 현재 두 번째로 큰 전력 공급 방법이며 원전의 수 역시 증가되는 것으로 계획되어 있다. 그러나, 원자력발전에 의해 발생되는 사용후핵연료에 대해서는 아직 명확한 관리 정책이 확립되어 있지 않다. 원자로 이 후 핵물질 흐름과 관련된 후행 핵연료주기는 사용후핵연료 관리를 위한 기술들의 집합이다. 따라서, 사용후핵연료 관리 정책은 핵연료주기 선택과 함께한다. 핵연료주기 선택의 중요 항목은 경제성으로 이는 사적비용과 함께 외부비용을 더해 결정되어야 한다. 직접비용 인 사적비용과 달리 간접비용인 외부비용에 대한 연구는 원전에 집중되어 있으며 핵연료주기에 대한 연구는 없는 상황이다. 본 연구에서는 핵연료주기에 적용할 수 있는 외부비용 항목들을 도출하고 정량화를 시도하였다. 핵연료주기 외부비용 평가를 위해 고려될 수 있는 핵연료주기로 OT(직접처분), DUPIC(PWR-CANDU 연결), PWR-MOX(PWR 습식재처리), Pyro-SFR (파이로 처리와 고속로 연계)의 네 가지를 선정하였다. 원자력발전의 외부비용 평가에 고려되었던 항목들을 분석하여 핵연료주기에서 에너지 공급 안보비용, 사고위험비용과 수용성 비용을 외부비용 항목으로 도출하고 추산하였다.

• Nuclear Engineering and Technology
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• 제49권6호
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• pp.1339-1345
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• 2017

In the framework of the EU funded project NURESAFE, the subchannel code CTF and the neutronics code DYN3D were integrated and coupled on the NURESIM platform. The developments achieved during this 3-year project include assembly-level and pin-by-pin multiphysics thermal hydraulics/neutron kinetics coupling. In order to test this coupling, a PWR rod ejection transient was simulated on a MOX/UOX minicore. The transient is simulated using two different models of the minicore. In the first simulation, both codes model the core with an assembly-wise resolution. In the second simulation, a pin-by-pin fuel-centered model is used in CTF for the central assembly, and a pin power reconstruction method is applied in DYN3D. The analysis shows the influence of the different models on global parameters, such as the power and the average fuel temperature, but also on local parameters such as the maximum fuel temperature.

• Nuclear Engineering and Technology
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• 제39권5호
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• pp.591-602
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• 2007

The results of Reactivity-Initiated Accidents (RIA) experiments have been analysed and the main variables affecting the fuel failure propensity identified. Fuel burn-up aggravates the mechanical loading of the cladding, while corrosion, or better the hydrogen absorbed in the cladding as a consequence of corrosion, may under some conditions make the cladding brittle and more susceptible to failure. Experiments point out that corrosion impairs the fuel resistance for RIA transient occurring at cold conditions, whereas there is no evidence of important embrittlement effects at hot conditions, unless the cladding was degraded by oxide spalling. A fuel failure threshold correlation has been derived and compared with experimental data relevant for BWR and PWR fuel. The correlation can be applied to both cold and hot RIA transients, account taken for the lower ductility at cold conditions and for the different initial enthalpy. It can also be used for non-zero power transients, provided that a term accounting for the start-up power is incorporated. The proposed threshold is easy to use and reproduces the results obtained in the CABRI and NSRR tests in a rather satisfactory manner. The behaviour of advanced PWR alloys and of MOX fuel is discussed in light of the correlation predictions. Finally, a probabilistic approach has been developed in order to account for the small scatter of the failure predictions. This approach completes the RIA failure assessment in that after determining a best estimate failure threshold, a failure probability is inferred based on the spreading of data around the calculated best estimate value.

• 한국방사성폐기물학회:학술대회논문집
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• 한국방사성폐기물학회 2009년도 학술논문요약집
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• pp.55-56
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• 2009

From the OT cycle analysis results, the nuclear power demand grows to ~70 GWe in 2150. The SF and TRU out-core inventories in 2150 will be 186500 t and 2100 t, respectively. The MOX fuel cycle gives 84% and 9% lower values for the SF and out-core TRU inventories, respectively.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1995년도 추계학술발표회논문집(1)
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• pp.29-34
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• 1995

The good features of the analytic function expansion nodal (AFEN) method are utilized to develop a practical scheme jot the multigroup diffusion problems, in combination with the polynomial expansion nodal (PEN) method. The thermal group fluxes exhibiting strong gradients are solved by the AFEN method[1-6], while the fast group fluxes that are smoother than the thermal group fuzes are solved by the PEN method[7-9]. The scheme is applied to a MOX-fuel loaded core with good results.

• 한국원자력학회:학술대회논문집
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• 한국원자력학회 1998년도 추계학술발표회요약집
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• pp.64-64
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• 1998

• Nuclear Engineering and Technology
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• 제54권8호
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• pp.3085-3094
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• 2022

The design concept of a Small High-flux Multipurpose LBE(Lead Bismuth Eutectic) cooled Fast Reactor (SHMLFR) was proposed in the paper. The primary cooling system of the reactor is forced circulation, and the fuel element form is arc-plate loaded high enrichment MOX fuel. The core is cylindrical with a flux trap set in the center of the core, which can be used as an irradiation channel. According to the requirements of the core physical design, a series of physical design criteria and constraints were given, and the steady and transient parameters of the reactor were calculated and analyzed. Regarding the thermal and hydraulic phenomena of the reactor, a simplified model was used to conduct a preliminary analysis of the fuel plates at special positions, and the temperature field distribution of the fuel plate with the highest power density under different coolant flow rates was simulated. The results show that the various parameters of SHMLFR meet the requirements and design criteria of the physical design of the core and the thermal design of the reactor. This implies that the conceptual design of SHMLFR is feasible.

• Nuclear Engineering and Technology
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• 제53권12호
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• pp.3930-3943
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• 2021

To improve the capability of the SAS4A code, which simulates the initiating phase of core disruptive accidents for MOX-fueled Sodium-cooled Fast Reactors (SFRs), the authors have investigated in detail the physical phenomena under unprotected loss-of-flow (ULOF) conditions in a previous paper (Kawada and Suzuki, 2020) [1]. As the conclusion of the last article, fuel stub motion, in which the residual fuel pellets would move toward the core central region after fuel pin disruption, was identified as one of the key phenomena to be appropriately simulated for the initiating phase of ULOF. In the present paper, based on the analysis of the experimental data, the behaviors related to the stub motion were evaluated and quantified by the author from scratch. A simple model describing fuel stub motion, which was not modeled in the previous SAS4A code, was newly proposed. The applicability of the proposed model was validated through a series of analyses for the CABRI experiments, by which the stub motion would be represented with reasonable conservativeness for the reactivity evaluation of disrupted core.