• Nuclear Engineering and Technology
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• v.37 no.1
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• pp.91-100
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• 2005

Kyung-hee Thorium Fuel (KTF), a heterogeneous thorium-based seed and blanket design concept for pressurized light water reactors, is being studied as an alternative to enhance proliferation resistance and fuel cycle economics of PWRs. The proliferation resistance characteristics of the KTF assembly design were evaluated through parametric studies using neutronic performance indices such as Bare Critical Mass (BCM), Spontaneous Neutron Source rate (SNS), Thermal Generation rate (TG), and Radio-Toxicity. Also, Fissile Economic Index (FEI), a new index for gauging fuel cycle economy, was suggested and applied to optimize the KTF design. A core loaded with optimized KTF assemblies with a seed-to-blanket ratio of 1: 1 was tested at the Korea Next Generation Reactor (KNGR), ARP-1400. Core design characteristics for cycle length, power distribution, and power peaking were evaluated by HELIOS and MASTER code systems for nine reload cycles. The core calculation results show that the KTF assembly design has nearly the same neutronic performance as those of a conventional $UO_2$ fuel assembly. However, the power peaking factor is relatively higher than that of conventional PWRs as the maximum Fq is 2.69 at the M$9^{th}$ equilibrium cycle while the design limit is 2.58. In order to assess the economic potential of a heterogeneous thorium fuel core, the front-end fuel cycle costs as well as the spent fuel disposal costs were compared with those of a reference PWR fueled with $UO_2$. In the case of comprising back-end fuel cycle cost, the fuel cycle cost of APR-1400 with a KTF assembly is 4.99 mills/KWe-yr, which is lower than that (5.23 mills/KWe-yr) of a conventional PWR. Proliferation resistance potential, BCM, SNS, and TG of a heterogeneous thorium-fueled core are much higher than those of the $UO_2$ core. The once-through fuel cycle application of heterogeneous thorium fuel assemblies demonstrated good competitiveness relative to $UO_2$ in terms of economics.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.15 no.4
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• pp.391-402
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• 2017

The Korea Radioactive Waste Agency (KORAD) has developed a concrete cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. During long-term storage of spent nuclear fuel in concrete casks kept in dry conditions, the integrity of the concrete cask and spent nuclear fuel must be maintained. In addition, the radiation dose rate must not exceed the storage facility's design standards. A suitable shielding design for radiation protection must be in place for the dry storage facilities of spent nuclear fuel under normal and accident conditions. Evaluation results show that the appropriate distance to the annual dose rate of 0.25 mSv for ordinary citizens is approximately 230 m. For a $2{\times}10$ arrangement within storage facilities, rollover accidents are assumed to have occurred while transferring one additional storage cask, with the bottom of the cask facing the controlled area boundary. The dose rates of 12.81 and 1.28 mSv were calculated at 100 m and 230 m from the outermost cask in the $2{\times}10$ arrangement. Therefore, a spent nuclear fuel concrete cask and storage facilities maintain radiological safety if the distance to the appropriately assessed controlled area boundary is ensured. In the future, the results of this study will be useful for the design and operation of nuclear power plant on-site storage or intermediate storage facilities based on the spent fuel management strategy.

• Nuclear Engineering and Technology
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• v.25 no.3
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• pp.361-370
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• 1993

Low power physics tests should be peformed for the domestic pressurized light water reactors (PWRs) after refueling. The tests are peformed to ensure that operating characteristics of the core are consistent with predictions and that the core can be operated as designed. But in some low power physics tests, slow but steady reactivity increasing phenomena were noticed after step reactivity insertion by the control rod movement. These reactivity increasing phenomena are due to the low flux level and the gamma background because an uncompensated ion chamber (UIC) is used as the ex-core neutron detector. The gamma background may affect the results or the lour power physics tests. The aims or this paper are to analyze the grounds of such phenomena, to simulate a reference bank worth measurement test and to present a resolution quantitatively. In this study, the gamma background level was estimated by numerically solving the point kinetics equations accounting the gamma background effect. The reactivity computer check test was simulated to verify the model. Also, an appropriate neutron flux level was determined by simulating the reference bank worth measurement test. The determined neutron flux level is approximately 0.3 of the nuclear heating flux. This level is about 3 times as high as the current test upper limit specified in the test procedure. Then, the findings from this work were successfully applied to Kori unit 4 cycle 7 and Yonggwang unit 1 cycle 7 physics tests.

• Nuclear Engineering and Technology
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• v.39 no.1
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• pp.31-42
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• 2007

Energy supply is increasingly showing up as a major issue for electricity supply, transportation, settlement, and process heat industrial supply including hydrogen production. Nuclear power is part of the solution. For electricity supply, as exemplified in Finland and France, the EPR brings an immediate answer; HTR could bring another solution in some specific cases. For other supply, mostly heat, the HTR brings a solution inaccessible to conventional nuclear power plants for very high or even high temperature. As fossil fuels costs increase and efforts to avoid generation of Greenhouse gases are implemented, a market for nuclear generated process heat will be developed. Following active developments in the 80's, HTR have been put on the back burner up to 5 years ago. Light water reactors are widely dominating the nuclear production field today. However, interest in the HTR technology was renewed in the past few years. Several commercial projects are actively promoted, most of them aiming at electricity production. ANTARES is today AREVA's response to the cogeneration market. It distinguishes itself from other concepts with its indirect cycle design powering a combined cycle power plant. Several reasons support this design choice, one of the most important of which is the design flexibility to adapt readily to combined heat and power applications. From the start, AREVA made the choice of such flexibility with the belief that the HTR market is not so much in competition with LWR in the sole electricity market but in the specific added value market of cogeneration and process heat. In view of the volatility of the costs of fossil fuels, AREVA's choice brings to the large industrial heat applications the fuel cost predictability of nuclear fuel with the efficiency of a high temperature heat source tree of Greenhouse gases emissions. The ANTARES module produces 600 MWth which can be split into the required process heat, the remaining power drives an adapted prorated electric plant. Depending on the process heat temperature and power needs, up to 80% of the nuclear heat is converted into useful power. An important feature of the design is the standardization of the heat source, as independent as possible of the process heat application. This should expedite licensing. The essential conditions for success include: ${\bullet}$ Timely adapted licensing process and regulations, codes and standards for such application and design ${\bullet}$ An industry oriented R&D program to meet the technological challenges making the best use of the international collaboration. Gen IV could be the vector ${\bullet}$ Identification of an end user(or a consortium of) willing to fund a FOAK

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.16 no.4
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• pp.455-472
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• 2018

It is important to study how to manage dry storage casks of spent nuclear fuels (SNF), because wet storage spaces for SNF will shortly be at full capacity in the Republic of Korea. The US has operated a dry storage cask system for several decades, and has carried out significant studies into how to successfully manage dry storage cask for SNF. This type of expertise and experience is currently lacking in the Republic of Korea. The degradation of dry casks is an important issue that must be considered. In particular, chloride-induced stress corrosion cracking (CISCC) is known to lead to the release of radioisotopes from canisters. The U.S. Department of Energy, U.S. Nuclear Regulatory Commission, and the Electric Power Research Institute have undertaken research into the CISCC mechanism. In addition, Sandia National Laboratories (SNL) has extensively researched CISCC and how to manage it in dry storage canisters. In this review paper, the probabilistic model proposed by the SNL is analyzed and, based on this model, US-based CISCC research is reviewed in detail. This paper will inform the management of dry cask storage of SNF from light water reactors in austenite stainless steel canisters in the Republic of Korea.

• Nuclear Engineering and Technology
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• v.25 no.1
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• pp.91-101
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• 1993

To predict accurately the thermal hydraulic behavior of light water reactors during normal or abnormal operation, the accurate estimation of the void distribution is required. Up to date, many techniques for predicting void fraction of two-phase flow systems have been suggested. Among these techniques, the drift-flux model is widely used because of its exact calculation ability and simplicity. However, to get more accurate prediction of void fraction using drift-flux model, slip and flow regime effects must be considered more properly In the drift-flux method, these two effects are accounted for by two drift-flux parameters ; $C_{o}$ and (equation omitted). At earlier stage, $C_{o}$ is measured in a circular tube. In this study, $C_{o}$ is experimentally determined by measuring local void fraction and vapor velocity distribution in a rectangular subchannel having 4 heating rods which simulates nuclear subchannels. The measurements are peformed with two-electrical conductivity probes which are known to be adequate for measuring local parameters. The experiments are performed at low flow rate and the system pressure less than 3 atmo spheric pressure. In this experiment, (equation omitted), is not measured, but quoted from well-known empirical correlation to formulate $C_{o}$. Finally, $C_{o}$ is expressed as a function of channel averaged void fraction. fraction.

• Journal of Korean Society of Environmental Engineers
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• v.32 no.4
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• pp.333-340
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• 2010

Wastewater treatment using algal communities and biodiesel production from wastewater-cultivated algal biomass is a promising green growth technology. In literature, there are many studies providing information on algal species producing high content of lipid. However, very little is known about adaptability and wastewater treatability of such high-lipid algal species. In this study, we attempted to characterize algal biomass production and wastewater treatability of high-lipid algal species under municipal wastewater condition. For this, four known high-lipid algal strains including Chlorella vulgaris AG 10032, Ankistrodesmus gracilis SAG 278-2, Scenedesmus quadricauda, and Botryococcus braunii UTEX 572 were individually inoculated into municipal wastewater where its indigenuous algal populations were removed prior to the inoculation, and the algae-inoculated wastewater was incubated in the presence of light source (80${\mu}E$) for 9 days in laboratory batch reactors. During the incubations, algal biomass production (dry weight) and the removals of dissolved organics (COD), nitrogen and phosphorous were measured in laboratory batch reactors. According to algal growth results, C. vulgaris, A. gracilis and S. quadricauda exhibited faster growth than indigenuous wastewater algal populations while B. braunii did not. The wastewater-growing strains exhibited efficient removals of total-N, ${NH_4}^+$-N, Total-P and ${PO_4}^{3-}$-P which satisfy the Korea water quality standards for effluent from municipal wastewater treatment plants. A. gracilis and S. quadricauda exhibited efficient and stable treatability of COD but C. vulgaris showed unstable treatability. Taken together with the results, A. gracilis and S. quadricauda were found to be suitable species for biomass production and wastewater treatment under municipal wastewater condition.

• Journal of Nuclear Fuel Cycle and Waste Technology(JNFCWT)
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• v.14 no.4
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• pp.343-356
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• 2016

The Korea Radioactive Waste Agency (KORAD) has developed a dual-purpose metal cask for the dry storage of spent nuclear fuel that has been generated by domestic light-water reactors. The metal cask was designed in compliance with international and domestic technology standards, and safety was the most important consideration in developing the design. It was designed to maintain its integrity for 50 years in terms of major safety factors. The metal cask ensures the minimization of waste generated by maintenance activities during the storage period as well as the safe management of the waste. An activation evaluation of the main body, which includes internal and external components of metal casks whose design lifetime has expired, provides quantitative data on their radioactive inventory. The radioactive inventory of the main body and the components of the metal cask were calculated by applying the MCNP5 ORIGEN-2 evaluation system and by considering each component's chemical composition, neutron flux distribution, and reaction rate, as well as the duration of neutron irradiation during the storage period. The evaluation results revealed that 10 years after the end of the cask's design life, $^{60}Co$ had greater radioactivity than other nuclides among the metal materials. In the case of the neutron shield, nuclides that emit high-energy gamma rays such as $^{28}Al$ and $^{24}Na$ had greater radioactivity immediately after the design lifetime. However, their radioactivity level became negligible after six months due to their short half-life. The surface exposure dose rates of the canister and the main body of the metal cask from which the spent nuclear fuel had been removed with expiration of the design lifetime were determined to be at very low levels, and the radiation exposure doses to which radiation workers were subjected during the decommissioning process appeared to be at insignificant levels. The evaluations of this study strongly suggest that the nuclide inventory of a spent nuclear fuel metal cask can be utilized as basic data when decommissioning of a metal cask is planned, for example, for the development of a decommissioning plan, the determination of a decommissioning method, the estimation of radiation exposure to workers engaged in decommissioning operations, the management/reuse of radioactive wastes, etc.