• Proceedings of the Korean Nuclear Society Conference
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• 1996.05c
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• pp.423-428
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• 1996

The feasibility study on burning Am-241, 243 and Cm-244 nuclides in a conventional PWR (Pressurized Water Reactor) was carried out by using the TRIFON code that was developed by the Institute of Theoretical and Experimental Physics in Russia in 1992. TRIFON code uses updated ABBN Russian nuclear cross section library. The reference reactor is the Korea nuclear power plant unit 8 (YGN 2). The burning effect of Am-241, 243 and Cm-244 nuclides was studied with UO$_2$(3.5 w/o)fuel assembly and MOX (4.44 w/o) fuel assembly. The loaded mass ratio of Am-241, 243 and Cm-244 nuclides was obtained from the mass ratio of Am-241, 243 and Cm-244 nuclides in 10 year cooling spent fuel with average discharge burnup of 33 GWD/MTU. The effective transmutation rates of Am-241, 243 and Cm-244 nuclides in UO$_2$ fuel assembly were found to be higher than those in MOX fuel assembly. The result from TRIFON code was compared to that from CASMO-3/NEM-3D code system. For more reliable calculation of transmutation for MA(Minor Actinides) more sophisticated decay chain scheme of MA should be investigated and nuclear cross section library of MA should be considerably improved.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2005.11b
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• pp.156-164
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• 2005

The challenges China is facing in energy security are briefly discussed. Then, the development of nuclear power in China in the first half of 21 st century is envisioned, and it is expected that Generation-3 PWR nuclear power plants (NPPs) would be the leading units of nuclear power in the coming $30\~40$ years. As part of the nuclear power program, the R&D work on nuclear fuel cycle is generally proposed.

• Proceedings of the Korean Nuclear Society Conference
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• 1996.05a
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• pp.46-51
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• 1996

In this study we calculate the infite multiplication factors ($k_{\infty}$) and the Doppler temperature coefficients (DTC) of two mixed-oxide (MOX) fuel rods with different plutonium contents by using PHOENIX-P, HELIOS and CASMO-3 codes. The results were compared against the reference values obtained by MCNP-3A continuous-energy Monte Carlo code. The purpose of this study is to benchmark the accuracy of these lattice codes. The PHOENIX-P's Doppler coefficients calculated were in good agreement with the MCNP results within the Monte-Carlo uncertainty band which is in the order of $\pm$ 10% for the Doppler coefficients..

• Nuclear Engineering and Technology
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• v.53 no.6
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• pp.1731-1735
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• 2021

The operation of the nuclear reactor requires accurate and fast methods and techniques for analysing its kinetics. These techniques become even more important when the MOX-fuel is used due to the lower value of delayed neutron fraction 𝛽 for ²³⁹Pu. Based on a Birth-and-Death process review, the mathematical model of thermal reactor core has been proposed different from existing ones. The analytical method for thermal point-reactor parameters evaluation is described within this work. The proposed method is applied for analysis of the unsteady transient processes taking place in a thermal reactor at its start-up or shutdown power change, as well as during small accidental power variation from the rated value. Theoretical determination of MASURCA reactor core reactivity through the analysis of experimental data on neutron time spectra was made.

• Nuclear Engineering and Technology
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• v.55 no.3
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• pp.884-894
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• 2023

Design and safety assessment of fuel pins for application in innovative Generation IV fast reactors calls for a dedicated nuclear fuel modelling and for the extension of the fuel performance code capabilities to the envisaged materials and irradiation conditions. In the INSPYRE Project, comprehensive and physics-based models for the thermal-mechanical properties of U-Pu mixed-oxide (MOX) fuels and for fission gas behaviour were developed and implemented in the European fuel performance codes GERMINAL, MACROS and TRANSURANUS. As a follow-up to the assessment of the reference code versions ("pre-INSPYRE", NET 53 (2021) 3367-3378), this work presents the integral validation and benchmark of the code versions extended in INSPYRE ("post-INSPYRE") against two pins from the SUPERFACT-1 fast reactor irradiation experiment. The post-INSPYRE simulation results are compared to the available integral and local data from post-irradiation examinations, and benchmarked on the evolution during irradiation of quantities of engineering interest (e.g., fuel central temperature, fission gas release). The comparison with the pre-INSPYRE results is reported to evaluate the impact of the novel models on the predicted pin performance. The outcome represents a step forward towards the description of fuel behaviour in fast reactor irradiation conditions, and allows the identification of the main remaining gaps.

• Journal of Radiation Protection and Research
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• v.21 no.1
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• pp.59-72
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• 1996

The continuous expansions and development of nuclear power have led to generation of the significant volume of spent fuels and radioactive wastes. And so, safe and effective management of the spent fuel has been becoming internationally sensitive and significant issue since the early 1990s. Especially, more importance would be added in the view point of international politics, because of recent political changes in the countries of Eastern Europe including dissociation of the former Soviet Union and the difficulties faced by the nuclear industries worldwide. Accordingly, this paper is proposed to show an overview of national strategies and Policies on the spent fuel management, that are being assessed and carried out worldwide at this time. The overview is based on recent developments of the national strategies, their implementations and some related experiences presented in IAEA International meetings and some technical papers.

• Nuclear Engineering and Technology
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• v.55 no.10
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• pp.3648-3658
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• 2023

The early localization of a fuel subassembly with a failed (wet rupture) fuel pin is very important in reactors to limit the associated radiological and operational consequences. This requires a fast and reliable system for failure detection and their localization in the core. In the Prototype Fast Breeder Reactor, the system specially designed for this purpose is Failed Fuel Location Modules (FFLM) housed in the control plug region. It identifies a failed sub-assembly by detecting the presence of delayed neutrons in the sodium from a failed sub-assembly. During the commissioning phase of PFBR, it is mandatory to demonstrate the FFLM effectiveness. The paper highlights the engineering and physics design aspects of FFLM and the integrated simulation towards its function demonstration with a source assembly containing a perforated metallic fuel pin. This test pin mimics a MOX pin of 1 cm² of geometrical defect area. At 10% power and 20% sodium flow rate, the counts rate in the BCCs of FFLM system range from 75 cps to 145 cps depending upon the position of DN source assembly. The model developed for the counts simulation is applicable to both metal and MOX pins with proper values of k-factor and escape coefficient.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2009.11a
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• pp.53-54
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• 2009

The preliminary quantitative analysis of proliferation resistance for the five nuclear fuel cycles demonstrated that the thermal MOX fuel cycle is most vulnerable to proliferation due to the presence of pure $PuO_2$ in the fuel cycle, while the once-through fuel cycle has the highest proliferation resistance. The innovative next generation fuel cycles such as Pyro-SFR and Wet-SFR were found to have similar levels of proliferation resistance to that of the DUPIC fuel cycle which is believed to have proliferation resistance strong enough for commercial deployment. The sensitivity analysis also demonstrated the effectiveness of the proposed methodology in applying to existing and/or newly developing nuclear fuel cycles so as to improve the proliferation resistance characteristic of the fuel cycle systems.

• Nuclear industry
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• v.36 no.1
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• pp.75-76
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• 2016

일본은 사용후연료의 재처리 정책을 진행하고 있으나 고속증식로 계획이 지연됨에 따라 그 처리의 목표가 확실하지 못하다. MOX(Mixed-Oxide Fuel : 혼합 산화물 연료)에 의한 처리가 계획되어 있으나 상당히 어려운 상황이다. 또 2018년에 미 일 원자력협력협정의 갱신이 예정되어 있는데 이 플루토늄의 취급이 문제가 될 예정이다. 미국은 핵무기의 원료나 테러에 사용할지도 모르는 플루토늄의 확산을 경계하고 있기 때문이다.

• Proceedings of the Korean Nuclear Society Conference
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• 2001.10a
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• pp.225-225
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• 2001