• Nuclear Engineering and Technology
• /
• v.39 no.6
• /
• pp.683-690
• /
• 2007

Because of increasing concerns regarding global warming and the longevity of oil and gas reserves, the importance of nuclear energy as a major source of sustainable energy is gaining recognition worldwide. To make nuclear energy truly sustainable, it is necessary to ensure not only the sustainability of the fuel supply but also the sustained availability of waste repositories, especially those for high-level radioactive waste (HLW). From this perspective, the effort to maximize the waste loading density in a given repository is important for easing repository capacity problems. In most cases, the loading of a repository is controlled by the decay heat of the emplaced waste. In this paper, a comparison of the decay heat characteristics of HLW is made among the various fuel cycle options. It is suggested that, for a future fast breeder reactor (FBR) cycle, the removal and burning of minor actinides (MA) would significantly reduce the heat load in waste and would allow for a reduction of repository size by half.

• Proceedings of the Korean Radioactive Waste Society Conference
• /
• 2005.11b
• /
• pp.17-23
• /
• 2005

• Nuclear Engineering and Technology
• /
• v.23 no.1
• /
• pp.20-29
• /
• 1991

The system analysis for Korean nuclear power reactor option is made on the basis of reliability, cost minimization, finite uranium resource availability and nuclear engineering manpower supply constraints. The reference reactor scenarios are developed considering the future electricity demand, nuclear share, current nuclear power plant standardization program and manufacturing capacity. The levelized power generation cost, uranium requirement and nuclear engineering professionals demand are estimated for each reference reactor scenarios and nuclear fuel cycle options from the year 1990 up to the year 2030. Based on the outcomes of the analysis, uranium resource utilization, reliability and nuclear engineering manpower requirements are sensitive to the nuclear reactor strategy and associated fuel cycle whereas the system cost is not. APWR, CANDU longrightarrow FBR strategy is to be the best option for Korea. However, APWR, CANDU longrightarrow Passive Safe Reactor(PSR）longrightarrowFBR strategy should be also considered as a contingency for growing national concerns on nuclear safety and public acceptance deterioration in the future. FBR development and establishment of related fuel cycle should be started as soon as possible considering the uranium shortage anticipated between 2007 and 2032. It should be noted that the increasing use of nuclear energy to minimize the greenhouse effects in the early 21st century would accelerate the uranium resource depletion. The study also concludes that the current level of nuclear engineering professionals employment is not sufficient until 2010 for the establishment of nuclear infrastructure.

• Proceedings of the Korean Nuclear Society Conference
• /
• 2002.05a
• /
• pp.63-63
• /
• 2002

• Nuclear Engineering and Technology
• /
• v.40 no.1
• /
• pp.9-20
• /
• 2008

Since plutonium accounts for 40-50% of the power produced by uranium fuels, spent fuel contains only residual plutonium. Management of this plutonium is one of the aspects influencing the choice of a fuel cycle back-end option: reprocessing, direct disposal or wait-and-see. Different grades and qualities of plutonium exist depending from their specific generation conditions; all are valuable fissile material. Safeguard authorities watch the inventories of civil plutonium, but access to those data is restricted. Independent evaluations have led to an estimated current inventory of 220t plutonium in total (spent fuel, separated civil plutonium and military plutonium). If used as MOX fuel, it would be sufficient to feed all the PWRs and BWRs worldwide during 7 years or to deploy a FBR park corresponding to 150% of today' s installed nuclear capacity worldwide, which could then be exploited for centuries with the current stockpile of depleted and spent uranium. The energy potential of plutonium deteriorates with storage time of spent fuel and of separated plutonium, due to the decay of $^{241}Pu$, the best fissile isotope, into americium, a neutron absorber. The loss of fissile value of plutonium is more pronounced for usage in LWRs than in FBR. However, keeping the current plutonium inventory for an expected future deployment of FBRs is counterproductive. Recycling plutonium reduce the required volume for final disposal in an underground repository and the cost of final disposal. However, the benefits of utilizing an energy resource and of reducing final disposal liabilities are not the only aspects that determine the choice of a back-end policy.

• Journal of Korean Society of Water and Wastewater
• /
• v.28 no.2
• /
• pp.207-215
• /
• 2014

Wastewater containing heavy metals such as copper (Cu) and nickel (Ni) is harmful to humans and the environment due to its high toxicity. Crystallization in a fluidized bed reactor (FBR) has recently received significant attention for heavy metal removal and recovery. It is necessary to find optimum reaction conditions to enhance crystallization efficacy. In this study, the effects of crystallization reagent and pH were investigated to maximize crystallization efficacy of Cu-S and Ni-S in a FBR. CaS and $Na_2S{\cdot}9H_2O$ were used as crystallization reagent, and pH were varied in the range of 1 to 7. Additionally, each optimum crystallization condition for Cu and Ni were sequentially employed in two FBRs for their selective removal from the mixture of Cu and Ni. As major results, the crystallization of Cu was most effective in the range of pH 1-2 for both CaS and $Na_2S{\cdot}9H_2O$ reagents. At pH 1, Cu was completely removed within five minutes. Ni showed a superior reactivity with S in $Na_2S{\cdot}9H_2O$ compared to that in CaS at pH 7. When applying each optimum crystallization condition sequentially, only Cu was firstly crystallized at pH 1 with CaS, and then, in the second FBR, the residual Ni was completely removed at pH 7 with $Na_2S{\cdot}9H_2O$. Each crystal recovered from two different FBRs was mainly composed of CuxSy and NiS, respectively. Our results revealed that Cu and Ni can be selectively recovered as reusable resources from the mixture by controlling pH and choosing crystallization reagent accordingly.

• Nuclear Engineering and Technology
• /
• v.13 no.1
• /
• pp.22-35
• /
• 1981

The U$_3$O$_{8}$ requirements are estimated for the high, intermediate, and low growth projections of nuclear power in Korea. To each projection, four illustrative reactor-mix strategies and four fuel cycle options are applied for estimating the requirements. The reactor types considered are PWR, PHWR. and FBR. The fuel cycles considered are once-through cycle, U/Pu recycle, and improved once-through cycle. Also the amount of Pu-fissile recovered from U recycle is estimated. The maximum cumulative (to the year 2000) requirements of U$_3$O$_{8}$ occupy about 4 to 5 percent of the WOCA requirements and are about 23 times larger than the U$_3$O$_{8}$ resources in Korea. For the high nuclear power growth projection, the cumulative amount of Pu-fissile recovered from U recycle is sufficient for the startup of 2 units of 1200 MWe fast reactors by the year 2000. 2000.

• Nuclear Engineering and Technology
• /
• v.53 no.12
• /
• pp.3930-3943
• /
• 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.

• Proceedings of the Materials Research Society of Korea Conference
• /
• 2011.05a
• /
• pp.15-15
• /
• 2011

As you know, boron compounds, borax ($Na_2B_4O_5(OH)_4{\cdot}8H_2O$) etc. were known thousands of years ago. As for natural boron, it has two naturally occurring and stable isotopes, boron 11 ($^{11}B$) and boron 10 ($^{10}B$). The neutron absorption $^{10}B$ is included about 19~20% with 80~81% $^{11}B$. Boron is similar to carbon in its capability to form stable covalently bonded molecular networks. The mass difference results in a wide range of ${\beta}$ values between the $^{11}B$ and $^{10}B$. The $^{10}B$ isotope, stable with 5 neutrons is excellent at capturing thermal neutrons. For example, it is possible to decrease a thermal neutron required for the nuclear reaction of uranium 235 ($^{235}U$). If $^{10}B$ absorbs a neutron ($^1n$), it will change to $^7Li+^1{\alpha}$ (${\alpha}$ ray, like $^4He$) with prompt ${\gamma}$ ray from $^{11}B$ $^{11}B$ (equation 1). $$^{10}B+^1n\;{\rightarrow}\;^{11}B\;{\rightarrow}\; prompt \;{\gamma}\;ray (478 keV), \;^7Li+4{\alpha}\;(4He)\;\;\;\;{\cdots}\; (1)$$ If about 1% boron is added to stainless steel, it is known that a neutron shielding effect will be 3 times the boron free steel. Enriched boron or $^{10}B$ is used in both radiation shielding and in boron neutron capture therapy. Then, $^{10}B$ is used for reactivity control and in emergency shutdown systems in nuclear reactors. Furthermore, boron carbide, $B_4C$, is used as the charge of a nuclear fission reaction control rod material and neutron cover material for nuclear reactors. The $B_4C$ powder of natural B composition is used as a charge of a control material of a boiling water reactor (BWR) which occupies commercial power reactors in nuclear power generation. The $B_4C$ sintered body which adjusted $^{10}B$ concentration is used as a charge of a control material of the fast breeder reactor (FBR) currently developed aiming at establishment of a nuclear fuel cycle. In this study for new boron compound, silicon boride ceramics for capturing thermal neutrons, preparation and characterization of both silicon tetraboride ($SiB_4$) and silicon hexaboride ($SiB_6$) and ceramics produced by sintering were investigated in order to determine the suitability of this material for nuclear power generation. The relative density increased with increasing sintering temperature. With a sintering temperature of 1,923 K, a sintered body having a relative density of more than 99% was obtained. The Vickers hardness increased with increasing sintering temperature. The best result was a Vickers hardness of 28 GPa for the $SiB_6$ sintered at 1,923K for 1 h. The high temperature Vickers hardness of the $SiB_6$ sintered body changed from 28 to 12 GPa in the temperature range of room temperature to 1,273 K. The thermal conductivity of the SiB6 sintered body changed from 9.1 to 2.4 W/mK in the range of room temperature to 1,273 K.