• Nuclear Engineering and Technology
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• v.43 no.6
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• pp.489-498
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• 2011

UK National Nuclear Laboratory's (NNL's) version of the ENIGMA fuel performance code is described, including details of the development history, the system modelled, the key assumptions, the thermo-mechanical solution scheme, and the various incorporated models. The recent development of ENIGMA in the areas of whole core analysis and dry storage applications is then discussed. With respect to the former, the NEXUS code has been developed by NNL to automate whole core fuel performance modelling for an LWR core, using ENIGMA as the underlying fuel performance engine. NEXUS runs on NNL's GEMSTONE high performance computing cluster and utilises 3-D core power distribution data obtained from the output of Studsvik Scandpower's SIMULATE code. With respect to the latter, ENIGMA has been developed such that it can model the thermo-mechanical behaviour of a given LWR fuel rod during irradiation, pond cooling, drying, and dry storage - this involved: (a) incorporating an out-of-pile clad creep model for irradiated Zircaloy-4; (b) including the ability to simulate annealing out of the clad irradiation damage; (c) writing of additional post-irradiation output; (d) several other minor modifications to allow modelling of post-irradiation conditions.

• Nuclear Engineering and Technology
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• v.49 no.1
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• pp.216-223
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• 2017

In this work, linear and mass attenuation coefficients, effective atomic number and electron density, mean free paths, and half value layer and $10^{th}$ value layer values of barium-bismuth-borosilicate glasses were obtained for 662 keV, 1,173 keV, and 1,332 keV gamma ray energies using MCNP-4C code and XCOM program. Then obtained data were compared with available experimental data. The MCNP-4C code and XCOM program results were in good agreement with the experimental data. Barium-bismuth-borosilicate glasses have good gamma ray shielding properties from the shielding point of view.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.19 no.3
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• pp.297-306
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• 2021

In this study, basic strategies for the decommissioning and site remediation of the Fukushima Daiichi Nuclear Power Station (FDNPS) were investigated. Six scenarios were formulated based on two of the three decommissioning strategies of nuclear power plants defined by the International Atomic Energy Agency (IAEA): immediate dismantling and deferred dismantling. A multicriteria decision analysis was performed to analyze the preferences of the options from the viewpoints of the timeframe to complete decommissioning, the resulting waste, the site usability, and the availability of the radioactive waste disposal route. The same six scenarios were applied to both the FDNPS and the nuclear power plants that ceased operation after a normal plant life cycle for comparison. For the FDNPS, the decommissioning project involved fuel debris retrieval, dismantling, and site remediation. The analysis results suggest that the balance between the amount of waste and the time to achieve the end state may be one of the most critical factors to consider when planning the decommissioning and site remediation of the FDNPS.

• Nuclear Engineering and Technology
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• v.38 no.4
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• pp.359-374
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• 2006

Since 1991, Korea, Canada and United States have performed the direct use of spent pressurized water reactor (PWR) fuel in the Canada deuterium uranium (CANDU) reactors (DUPIC) fuel development project. Unlike the Tandem fuel cycle, which requires a wet reprocessing, the DUPIC fuel technology can directly refabricate CANDU fuels from the PWR spent fuel and, therefore, is recognized as a highly proliferation-resistant fuel cycle technology, which can be adopted even in non-proliferation treaty countries. The Korea Atomic Energy Research Institute (KAERI) has fabricated DUPIC fuel elements in a laboratory-scale remote fuel fabrication facility. KAERI has demonstrated the fuel performance in the research reactor, and has confirmed the operational feasibility and safety of a CANDU reactor loaded with the DUPIC fuel using conventional design and analysis tools, which will be the foundation of the future practical and commercial uses of DUPIC fuel.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.18 no.spc
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• pp.1-19
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• 2020

Based on the current high-level radioactive waste management basic plan and the analysis results of spent nuclear fuel characteristics, such as dimensions and decay heat, an improved geological disposal concept for spent nuclear fuel from domestic nuclear power plants was proposed in this study. To this end, disposal container concepts for spent nuclear fuel from two types of reactors, pressurized water reactor (PWR) and Canada deuterium uranium (CANDU), considering the dimensions and interim storage method, were derived. In addition, considering the cooling time of the spent nuclear fuel at the time of disposal, according to the current basic plan-based scenarios, the amount of decay heat capacity for a disposal container was determined. Furthermore, improved disposal concepts for each disposal container were proposed, and analyses were conducted to determine whether the design requirements for the temperature limit were satisfied. Then, the disposal efficiencies of these disposal concepts were compared with those of the existing disposal concepts. The results indicated that the disposal area was reduced by approximately 20%, and the disposal density was increased by more than 20%.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.21 no.4
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• pp.571-576
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• 2023

The initial development plans for the six reactor designs, soon after the release of Generation IV International Forum (GIF) TRM in 2002, were characterized by high ambition [1]. Specifically, the sodium-cooled fast reactor (SFR) and very-high temperature reactor (VHTR) gained significant attention and were expected to reach the validation stage by the 2020s, with commercial viability projected for the 2030s. However, these projections have been unrealized because of various factors. The development of reactor designs by the GIF was supposed to be influenced by events such as the 2008 global financial crisis, 2011 Fukushima accident [2, 3], discovery of extensive shale oil reserves in the United States, and overly ambitious technological targets. Consequently, the momentum for VHTR development reduced significantly. In this context, the aims of this study were to compare and analyze the development progress of the six Gen IV reactor designs over the past 20 years, based on the GIF roadmaps published in 2002 and 2014. The primary focus was to examine the prospects for the reactor designs in relation to spent nuclear fuel burning in conjunction with small modular reactor (SMR), including molten salt reactor (MSR), which is expected to have spent nuclear fuel management potential.

• Nuclear Engineering and Technology
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• v.40 no.1
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• pp.1-8
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• 2008

One of the major current challenges to nuclear energy lies in its competitiveness. To stay competitive the industry needs to reduce maintenance and fuel cycle costs, while enhancing safety features. Extended burnup is one of the methods applied to meet these objectives However, there are a number of potential fuel failure causes related to increased burnup, as follows: l) Corrosion of zirconium alloy cladding and the water chemistry parameters that enhance corrosion; 2) Dimensional changes of zirconium alloy components, 3) Stresses that challenge zirconium alloy ductility and the effect of hydrogen (H) pickup and redistribution as it affects ductility, 4) Fuel rod internal pressure, 5) Pellet-cladding interactions (PCI) and 6) pellet-cladding mechanical interactions (PCMI). This paper discusses current and potential failure mechanisms of these failure mechanisms.

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2018.11a
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• pp.45-46
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• 2018

• Proceedings of the Korean Radioactive Waste Society Conference
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• 2017.10a
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• pp.63-64
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• 2017

• Nuclear Engineering and Technology
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• v.37 no.4
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• pp.317-326
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• 2005

The burning-out of excess plutonium from the reprocessing of spent nuclear fuel and from the dismantlement of nuclear weapons is recently emphasized due to the difficulties in securing the final repository for the spent fuel and the necessity to consume the ex-weapons plutonium. An irradiation test in the Halden reactor was launched by the OECD Halden Reactor Project (HRP) to investigate the in-pile behavior of plutonium-embedded fuel as a form of mixed oxide (MOX) and of inert matrix fuel (IMF). The first cycle of irradiation was successfully accomplished with good integrity of test fuel rods and without any undesirable fault of instrumentations. The test results revealed that the MOX fuel is more stable under irradiation environments than IMF. In addition, MOX fuel shows lower thermal resistance due to its better thermal conductivity than IMF. The on-line measured in-pile performance data of attrition milled MOX fuel are used in the analysis of the in-pile performance of the fuel with the fuel performance code, COSMOS. The COSMOS code has been developed for the analysis of MOX fuel as well as $UO_2$ fuel up to high burnup and showed good capability to analyze the in-reactor behavior of MOX fuel even with different instrumentation.