• Nuclear Engineering and Technology
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• v.51 no.2
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• pp.345-355
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• 2019

Use of the reconstructed fuel assemblies from partially burnt nuclear fuel pins is analyzed. This reutilization option is a potential candidate technique to make better use of the nuclear resources. Standard two step method is used to calculate node i.e. fuel assembly average burnup and then pin by pin ${\eta}$ values are reconstructed to ascertain the residual reactivity in the used fuel pins. Fuel pins with ${\eta}$ > 1:0 are used to reconstruct to-be-reused fuel assemblies. These reconstructed fuel assemblies are burnt during the cycle 3, 4, 5 and 6 of a 1000 MW PWR core by replacing fresh, once burnt and twice burnt fuel assemblies of the reference core configurations. It is concluded that using reconstructed fuel assemblies for the fresh fuel affect dearly on the cycle length (>50 EFPD) when more than 16 fresh fuel assemblies are replaced. However, this loss is less than 20 days if the number of fresh fuel assemblies is less than eight. For the case of replacing twice burned fuel, cycle length could be increased slightly (10 days or so) provided burnt fuel pins from other reactors were also available. Reactor safety parameters, like axial off set (< ${\pm}10%$), Doppler temperature coefficient (<0), moderator temperature coefficient at HFP (<0) are always satisfied. Though, 2D and 3D pin peaking factors are satisfied (<1:55) and (<2:52) respectively, for the cases using eight or less reconstructed fuel assemblies only.

• Nuclear Engineering and Technology
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• v.45 no.4
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• pp.421-426
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• 2013

The need for extended storage of used nuclear fuel is increasing globally as disposition schedules for used fuel are pushed further into the future. This is creating a situation where dry storage of used fuel may need to be extended beyond normal regulatory licensing periods. While it is generally accepted that used fuel in dry storage will remain in a safe condition, there is little data that demonstrate used fuel performance in dry storage environments for long periods of time. This is especially true for high burnup used fuel. This paper discusses a technical approach that defines a process that develops the technical basis for demonstrating the safety of used fuel over extended periods of time.

• Nuclear Engineering and Technology
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• v.43 no.4
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• pp.329-334
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• 2011

Two conceptual flowsheets were developed for recycling used nuclear fuel. One flowsheet was developed for recycling used oxide nuclear fuel from light water reactors while the other was developed for recycling used metal fuel from fast spectrum reactors. Both flowsheets were developed from a set of design principles including efficient actinide recovery, nonproliferation, waste minimization and commercial viability. Process chemistry is discussed for each unit operation in the flowsheet.

• Nuclear Engineering and Technology
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• v.43 no.5
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• pp.405-412
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• 2011

Used nuclear fuel will likely be stored dry for extended periods of time in the USA. Until a final disposition pathway is chosen, the storage periods will almost definitely be longer than were originally intended. The ability of the important-tosafety structures, systems, and components (SSCs) to continue to meet storage and transport safety functions over extended times must be determined. It must be assured that there is no significant degradation of the fuel or dry cask storage systems. Also, it is projected that the maximum discharge burnups of the used nuclear fuel will increase. Thus, it is necessary to obtain data on high burnup fuel to demonstrate that the used nuclear fuel remains intact after extended storage. An evaluation was performed to determine the conditions that may lead to failure of dry storage SSCs. This paper documents the initial technical gap analysis performed to identify data and modeling needs to develop the desired technical bases to ensure the safety functions of dry stored fuel.

• Nuclear Engineering and Technology
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• v.56 no.7
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• pp.2641-2649
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• 2024

Effective management of slightly used nuclear fuel (SUNF) is crucial for both technical and public acceptance reasons. SUNF management, radiotoxicity risk, and associated financial investment and technological capabilities are major concerns in nuclear power production. Reducing the volume of SUNF can simplify its management, and one possible solution is utilizing small modular reactors (SMR) and advanced fuel designs like those with thorium. This research focuses on studying the neutronic performance and radionuclide inventory of three different thorium fuel configurations. The mass of fissile material in thorium-based fuel significantly impacts Kinf, burn-up, and neutron energy spectrum. Compared to uranium, thorium as a fuel produces far fewer transuranic elements and less long-lived fission products (LLFPs) at the end of the core cycle (EOC). However, certain fission product elements produced from thorium-based fuel exhibit higher radioactivity at the beginning of the core cycle (BOC). Physical separation of thorium and uranium in the fuel block, like seed-and-blanket units (SBU) and duplex fuel designs, generate less radioactive waste with lower radioactivity and longer cycle lengths than homogeneous or mixed thorium-uranium fuel. Furthermore, the SBU and duplex feel designs exhibit comparable neutron spectra, leading to negligible differences in SUNF production between the two.

• Nuclear Engineering and Technology
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• v.47 no.1
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• pp.85-93
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• 2015

Background: $^{241}Am$, $^{243}Am$, and $^{237}Np$ isotopes are among the most radiotoxic components of spent nuclear fuel. Recently, researchers have planned different incineration scenarios for the highly radiotoxic elements of nuclear waste in critical reactors. Computational methods are widely used to predict burnup rates of such nuclear wastes that are used under fuel matrixes in critical reactors. Methods: In this work, the Monte Carlo N-particle transport code was used to calculate the neutronic behavior of a transuranic (TRU)-bearing CANada Deuterium Uranium 6 reactor. Results: The computational data showed that the 1.0% TRU-containing thorium-based fuel matrix presents higher proliferation resistance and TRU depletion rate than the other investigated fuel Matrixes. The fuel matrix includes higher negative temperature reactivity coefficients as well. Conclusion: The investigated thorium-based fuel matrix can be successfully used to decrease the production of highly radiotoxic isotopes.

• Nuclear Engineering and Technology
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• v.55 no.9
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• pp.3464-3471
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• 2023

The Uranium/Molybdenum metallic fuel has been proposed as promising advanced fuel concept especially in the dispersed fuel geometry. The fuel is manufactured in the form of small fuel droplets (particles) placed in a fuel pin covered by a matrix. In addition to fuel particles, the pin contains voids necessary to compensate material swelling and release of fission gases from the fuel particles. When investigating this advanced fuel design, two important questions were raised. Can the dispersed fuel performance be analyzed using homogenization without significant inaccuracy and what size of fuel drops should be used for the fuel design to achieve optimal utilization? To answer, 2D burnup calculations of fuel assemblies with different fuel particle sizes were performed. The analysis was supported by an additional 3D fuel pin calculation with the dispersed fuel particle size variations. The results show a significant difference in the multiplication factor between the homogenized calculation and the detailed calculation with precise fuel particle geometry. The recommended fuel particle size depends on the final burnup to be achieved. As shown in the results, for lower burnup levels, larger fuel drops offer better multiplication factor. However, when higher burnup levels are required, then smaller fuel drops perform better.

• Journal of Radiation Industry
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• v.17 no.3
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• pp.283-292
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• 2023

Spent nuclear fuel corresponds to high-level radioactive waste that has high decay heat and radioactivity. Accordingly, Spent nuclear fuel withdrawn from the reactor core is primarily stored and managed in a spent nuclear fuel pool in the nuclear power plant to reduce decay heat and radioactivity. In Korea, most nuclear power plant store all spent nuclear fuel in a spent nuclear fuel pool. For wet storage, there are no defense in depth different with reactor core. The study related to spent nuclear fuel pool accident should be carried out to ensure safety. Therefore, it is necessary to analyze previous study cases related to safety of spent nuclear fuel pool and accident cases to build foundational knowledge. The Objective of this study is to analyze study cases of safety assessment and cases for spent nuclear fuel pool accident. For analyzing study cases of safety assessment, possible phenomena when spent nuclear fuel pool accident occurring identified, Subsequently, study cases for safety assessment about each phenomena were investigated, and materials & methods and results for each study are analyzed. For analyzing cases for spent nuclear fuel pool accident, we analyzed accident cases caused by loss of cooling and loss of coolant in spent nuclear fuel pool. Subsequently, causes and change of water level and temperature by each accident case are analyzed. As a result of the analysis on study cases of spent nuclear fuel pool accident, the results of the study conducted by each research institute were vary depending on the computer code, materials & methods of experiment and major assumptions used in the study. As a result of analyzing cases for spent nuclear fuel pool accident, it was found that accident cases for loss of cooling is more than cases for loss of coolant accident. Even though the types of accident in spent nuclear fuel pool were similar, the specific causes were different by each accident case. All the accident cases analyzed did not lead to severe accidents, such as nuclear fuel being exposed to the air. The result of this study will be used as fundamental data for study on spent nuclear fuel pool accident that will be conducted in the future.

• Nuclear Engineering and Technology
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• v.43 no.4
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• pp.335-342
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• 2011

As part of the Department of Energy's Fuel Cycle Research and Development Program an electrochemical technology employing molten salts is being developed for recycle of metallic fast reactor fuel and treatment of light water reactor oxide fuel to produce a feed for fast reactors. This technology has been deployed for treatment of used fuel from the Experimental Breeder Reactor II (EBR-II) in the Fuel Conditioning Facility, located at the Materials and Fuel Complex of Idaho National Laboratory. This process is based on dry (non-aqueous) technologies that have been developed and demonstrated since the 1960s. These technologies offer potential advantages compared to traditional aqueous separations including: compactness, resistance to radiation effects, criticality control benefits, compatibility with advanced fuel types, and ability to produce low purity products. This paper will summarize the status of electrochemical development and demonstration activities with used nuclear fuel, including preparation of associated high-level waste forms.

• Nuclear Engineering and Technology
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• v.56 no.9
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• pp.3668-3685
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• 2024

The operation of a nuclear power plant relies on precalculated nuclear design predictions based on core calculations of various reactor states. The fuel temperature is a crucial factor in determining the reactor fuel behavior, but assessing the temperature variation in a fuel pellet taking into account neutron transport is challenging. Detailed simulation of the temperature behavior within the fuel pellet can be obtained by coupling of Monte Carlo neutron transport codes with thermal-hydraulics solvers. However, this approach is not practical for standard nuclear design calculations, and computationally cheaper and faster methods must be used. In nuclear core simulators, a concept of a single "effective temperature" that yields the same neutron response as in the case of the actual temperature shape is mainly applied. This paper evaluates various fuel temperature models used in nuclear core simulation calculations, ultimately recommending a new effective temperature model that considers the burnup correction.