In this paper, a method based on ultra-low level liquid scintillation analyzer for measuring uranium radioactivity in low-level radioactive wastewater is proposed. It can easily and quickly measure the radioactivity level of uranium in samples, and even distinguish the main isotopes of uranium. In low-level wastewater samples, if the uranium isotope has the same or similar composition and ratio to the natural uranium isotope, Comparing to the results in units of mass concentration given by laser-fluoremetry method, the liquid scintillation method can give the results in units of activity concentration, and these two results with different units can interconvert to each other. However, interconversion of results on above two measurement methods cannot be applied in low-level wastewater after uranium enrichment, because the proportion of isotopes with high specific activity in above wastewater, such as 234 U and 235 U, is significantly higher than that without enrichment. The measurement results in units of mass concentration will underestimate the uranium radioactivity level of the sample. The liquid scintillation method directly gives the results in units of radioactivity activity concentration, and it is more convenient to compare with relevant national standards to determine whether the emission standards are met. The lowest limit of detection of this method is 0.014 Bq L −1 within 600 min.

In alkaline in-situ leaching uranium, oxygen is the most common oxidizer with bicarbonate as a complexing agent. For those sandstone uranium deposits with strongly reductive capacity or complicated hydrogeological environment, the oxidation by oxygen is low efficiency. An efficient leaching method, therefore, is needed for these uranium deposits. In this study, a typical sandstone uranium deposit which characterizes with high TDS and high chloride content in groundwater and intractable uranium leach is selected to investigate the effects of synergetic oxidation by a strong oxidant with oxygen. Based on the research on batch leach, pressure leach and field trials, the oxidants such as hydrogen peroxide, potassium permanganate and sodium dichloroisocyanurate (NaDCC) are tested. The results of pressure batch leach indicate that synergetic oxidization is achieved by NaDCC in oxygen leaching process. Leaching tests indicate that a minor oxidizer of NaDCC shows good synergetic oxidization with oxygen and leaching effects on uranium minerals. The results also demonstrate that hydrogen peroxide shows no oxidation effects when it is used as a single oxidant. While potassium permanganate shows good oxidation on uranium when it is used as a single oxidant, however, it leads inhibiting effects on oxygen oxidation on uranium minerals. The further field tests are conducted to study the synergetic effects of oxygen with and without sodium dichloroisocyanurate. The preliminary results indicate that a fast leach is observed by the composite oxidants in early stage while no synergetic leach is found after 200 days. Further studies should be conducted in laboratory experiments and pilot scale tests for its potential applications.

The goal of basic research is to gather information and build upon the knowledge base. This type of study identifies a problem and suggests solutions that applied research can consider and employ using empirical methodologies. Basic theory is a valid and accepted research tool, it is often mathematical in nature, and it is formulated to solve a particular problem. The problem is to find the mathematical link between the natural energy of stars and the energy created through fission. An equation is shown here to have a place in the field of nuclear engineering because it provides insight into the question of what makes the binding and decaying of atomic particles mathematically possible. This equation shows that when the two particles of unequal mass move close enough to be caught up in each other’s flow fields, they are entangled in the fields and remain bound together. When two particles are of equal mass however their encounter does not produce a binding mechanism because their flow fields are the same strength and cancel each other out. The fission of uranium-236 into the elements krypton-92 and barium-141 is discussed according to the equation. Then the fusion of deuterium and tritium is analyzed as an example of flow fields at work. Finally the first three steps of the proton-proton chain reaction are compared to the equation.

The Super Fast Breeder Reactor (Super FBR) utilizes supercritical light water as coolant, which changes from liquidlike high density state to gas-like low density state continuously in the core without phase change. In the preceding study (Noda et al., 2017), new concept of axially heterogeneous core with multi-axial fuel shuffling was proposed. The core consisted of two layers of mixed oxide (MOX) fuel and two layers of blanket fuel with depleted uranium (DU), which were arranged alternatively in the axial direction. The study showed that, with independent fuel shuffling in the upper part and lower part of the core, breeding performance could be improved by increasing the upper blanket fuel batch number while keeping the fuel batch number of the rest of the core unchanged, because of increased neutron flux in the upper blanket. However, the study did not consider influence of different coolant density histories in the different axial level of the core on the core neutronics. Hence, this study aims to reveal influence of the different coolant density histories through design and analyses of the multi-axial fuel shuffling core with two MOX layers and three blanket layers. The three levels correspond to the coolant density below, around, and above the pseudo-critical temperature. The neutronics calculations are carried out with SRAC 2006 code and JENDL-3.3 nuclear data library. Unit cell burnup calculations based on collision probability method are carried out for 5 different coolant density histories to consider influence of different neutron spectrum on breeding performance of the core. Influence of instantaneous coolant density changes on the core neutronics are considered by coupling core burnup calculations with thermal-hydraulics calculations based on single channel model. Influence of independent fuel shuffling of the upper blanket on the core neutronics (breeding performance and void reactivity characteristics) is investigated, followed by a similar investigation on the lower blanket. The differences between the two schemes are investigated since coolant density histories are greatly different between the upper blanket and the lower blanket.

Laboratories in traditional radiochemistry courses typically involve basic and fundamental understanding in solvent extraction, ion exchange, precipitation, etc. procedures. With the increased focus on nuclear forensics in pre- and post-detonation scenarios different skill sets are now required for the student to learn. At the University of Texas we have developed two independent graduate courses in gamma-ray spectrometry and radiochemistry. Currently, we have amalgamated these two courses to 1. better serve our nuclear engineering graduate students, many of which are involved in nuclear forensics and 2. to attract both undergraduate and graduate students from the Chemistry Department. We incorporated gamma-ray spectrometry laboratories with several others which are nuclear forensics related. The seven laboratory sessions include half-life measurement of 137m Ba (the daughter produce of 137 Cs) and secular equilibrium, basic gamma and beta shielding, and gamma-ray spectrometry calibration, resolution and uncertainty in statistics. These labs have been augmented with four others including uranium fission product identification, 137 Cs soil profile with Compton suppression, tritium analysis in water with a liquid scintillation counter and double replacement reaction.

RELAP/SCDAPSIM is a best-estimate nuclear tool designed to analyze the behaviour of reactor systems during normal and accident conditions. Three main versions of RELAP/SCDAPSIM are currently used by program members and licensed users to support a variety of activities. RELAP/SCDAPSIM/MOD3.4 is the version of the code used by licensed users and program members for critical applications such as research reactors and nuclear power plant applications. Even though the code was initially designed for LWRs (Light Water Reactors), Politehnica University of Bucharest demonstrated the applicability of the RELAP/SCDAPSIM code for CANDU (CANada Deuterium Uranium) reactors analyses, by simulating some of the most important postulated accident transients (i.e. large break LOCA, main steam line break, the natural circulation in the heat transport system). Current trends refer to the BEPU (Best Estimate Plus Uncertainty) approach in the safety analysis of nuclear reactors. BEPU is a modern and technically consistent approach has been built upon best estimate methods including an evaluation of the uncertainty in the calculated results. ISS and UPC started the development of an uncertainty evaluation package to RELAP/SCDAPSIM/MOD4.0 code version which is currently implemented in MOD3.4 version of the code also. The uncertainty evaluation capability is implemented as an alternative run mode, the “uncertainty” mode, which allows the automatic execution of an uncertainty analysis based on the probabilistic approach. A complete uncertainty analysis using RELAP/SCDAPSIM/MOD3.4 code requires the execution of three related phases, namely the “setup” phase, the “simulation” phase consisting of several executions, and the “postprocessing” phase. The uncertainty data has to be supplied for the two types of parameters, the “input treatable” and the “source correlation” quantities. The required information is the probability distribution function and its characteristic parameters. This paper is mainly focused on the application of the uncertainty package in CANDU reactors accident analysis and it describes the steps to perform the uncertainty analysis, the uncertainty selected parameters (input treatable and source correlation parameters), the calculation results with RELAP/SCDAPSIM and some conclusions.

The National Bureau of Standards reactor (NBSR) at the National Institute of Standards and Technology (NIST) is under conversion from high enriched uranium (HEU) to the low enriched uranium (LEU) schema under the Reduced Enrichment for Research and Test Reactors program (RERTR) as a part of the Global Threat Reduction Initiative (GTRI). The conversion of the high performance research reactors (HPRR) such as NBSR is a challenging task due to the high flux need (2.5 × 10 14 n/cm 2 -s for the NBSR), as well as other neutronics performance characteristics requirements without significant changes to the external geometrical configuration. One fuel candidate, the General Atomics (GA) UZrH LEU fuel, has showed particular promise in this regard. The TRIGA LEU fuel was initially developed in the 1980s with particular considerations for fuel conversion for high power regimes such as high density research and test reactors. This study performs a neutronics feasibility study of the UZrH LEU fuel schema for the NBSR, examining the accountability and sustainability of the TRIGA fuel when applying it to the NBSR conversion. To identify the best option to deploy the TRIGA fuel to NBSR in terms of key neutronic performance characteristic, the study is carried out with various considerations in the fuel dimensions, fuel rod layout configurations, and structure material selections. Monte Carlo based computational model is used to assist and facilitate the research procedure. The research findings in this study will determine the viability of the TRIGA fuel type for the NBSR conversion, and provide supporting data for future investigations on this subject.

Deep penetration problems exist widely in reactor applications, such as SPRR300 (Swimming Pool Research Reactor 300), a light water moderated, enriched uranium fueled research reactor in China. Deterministic transport theory is intrinsically suitable for deep penetration. But there exist some problems when it’s applied in SPRR-300research reactors. First, the reactor core is complicated for geometry description in deterministic theory codes. Monte Carlo method has advantages in complex geometry modeling. And it uses continuous energy cross sections which are independent with specific reactor types and research objections. But usually it’s difficult to converge well enough to deal with deep penetration problems, even though there are a number of variance reduction techniques. Based on the advantages and disadvantages of Monte Carlo and Deterministic method, we proposed a coupled neutron transport calculation method for deep penetration. It combines advantages of these two methods. Firstly, we use Monte Carlo code to finish fine modeling and do the whole reactor core calculation. Domestically developed Monte Carlo code JMCT is used to do the neutron transport calculation. Then homogenized group constants in each mesh are calculated from JMCT output by a self-developed script. Afterwards, we do the whole reactor calculation with deterministic theory code TORT. It directly uses group constants generated by Monte Carlo code. Finally, we can get the deep penetration calculation results from TORT output. Verification is carried out by comparing the group constants of benchmark problem, and by comparing k eff calculated by this method with continuous energy Monte Carlo method. Benchmark calculation is conducted with OECD/NEA slab benchmark problem. The comparison shows that group constants generated by this study are in good agreement with results from published references. Then above group constants are applied to 3-dimensional discrete ordinates deterministic theory transport code TORT. But k eff calculated by TORT is a little lower than that calculated by Monte Carlo code JMCT. To minimize other influence factors, different Sn/Pn order, and different mesh size in TORT has been tried. Unfortunately the k eff difference between these two methods remains. Even though the k eff results in this benchmark are less than k eff calculated by continuous energy MC method, Benchmark results show that all the group constants generated by this method are in good agreement with existing references. So it can be expected that after further verification and validation, this coupled method can be effectively applied to the deep penetration problem in such kind of research reactors.

The 235 U enrichment is one of the most important characteristics of nuclear materials for nuclear safeguards purposes. The multi-group γ-ray analysis method for uranium (MGAU) is an important non-destructive gamma spectroscopy method for 235 U enrichment determination. Using that method, the typical measurement bias is below 3% for uranium material with abundance from 0.3 to 93 %. However, it is not applicable for the samples with thick container or without isotopic decay equilibrium. In this work, the enrichment meter method was studied with two uranium dioxide samples ( 235 U abundance 0.71 % and 3.167 %). The nuclear materials spectra were measured using a planar high-purity germanium detector. Based on the specific gamma peak (185.71 keV) of relative high intensity, this traditional enrichment meter approach gives measurement bias more than 10 %. Thus, this work represents two objects: (1) an alternative approach which was investigated, where the calibration is performed through Monte Carlo simulation (MCNP5) instead of experiment in advance, as the measurement bias was reduced to be around 5 %. Thus, to use this approach, one should have the sample details, such as dimensions, chemical composition and container. (2) The influence of the container wall thickness on the measurement accuracy by Monte Carlo simulation. So, if the container wall thickness is not modeled correctly the measurement accuracy is influenced, which is investigated by simulation.

In a de-nitration conversion processes of the nuclear fuel cycle, mixed oxides (MOX) are produced from the reprocessing solution (plutonium and uranium mixed nitrate solution) of used nuclear fuel. The microwave heating (MH) method has various advantages as one of de-nitration conversion techniques, i.e., this can complete rapidly processes, reduce the waste liquid, and operated remotely and easily by the waveguide. Fine crystal powders are thereby generated, and the manufacture of high-density and high-quality pellets is significant advantage. The MH method is accompanied with transient boiling phenomena such as overflow and flashing. From the viewpoint of enhancing mass productivity and cost efficiency, in the future, scaling up for the size of de-nitration vessel and shortening processing time are desired. In addition, the safe design of device and the appropriate conditions in microwave irradiation process are required. Hence the extensive understanding of transient boiling phenomena induced by microwave heating is important, and the detailed mechanism of flushing and overflow should be clarified. The aim of this study is to clarify the transient boiling phenomena and detailed mechanism of flashing and overflow. Flashing are affected by physical factors such as solution properties, vessel characteristics, and input energy. Distilled water and different dielectric constant solution as working fluid are used. Dielectric constant was adjusted by the concentration of potassium chloride aqueous solution. A cylindrical vessel kept enough clean by ultrasonic washing machine before experiments was used. The influence of vessel diameter and initial water depth on flashing is examined. As input energy, amount of power supplied and position of the object to be heated in an oven are varied. A high-speed video camera was installed to observe boiling phenomena, and thermography and fiber optic thermometer were to measure the temperature on vessel walls and in a solution, respectively. As a result, the higher the dielectric constant solution, the less risk of flashing phenomena. In no flashing case, the surface temperature became over 100 °C and the precipitation of potassium chloride was existed. When the high dielectric constant solution was heated by microwave irradiation, the potassium chloride which was precipitated on the inner surface of the vessel was heated to high temperature. On the other hand the temperature of water was lower than that of the deposition of the potassium chloride. As described above, the influence of the high dielectric constant solution on the flashing phenomenon by microwave heating was evaluated. And the mechanism of flashing phenomena was assumed by detailed observation of bubble nucleation. Eventually the mechanism of flashing phenomena was examined.

In nuclear reactors that use plate-type fuel, the fuel plates are thermally managed with coolant flowing through channels between the plates. Depending on the flow rates and sizes of the fluid channels, the hydraulic forces exerted on a plate can be quite large. Currently, there is a worldwide effort to convert research reactors that use highly enriched uranium (HEU) fuel, some of which are plate-type, to low-enriched uranium (LEU). Because of the proposed changes to the fuel structure and thickness, a need exists to characterize the potential for flow-induced deflection of the LEU fuel plates. In this study, as an initial step, calculations of Fluid-Structure Interaction (FSI) for a flat aluminum plate separating two parallel rectangular channels are performed using the commercial code STAR-CCM+ and the integrated multi-physics code SHARP, developed under the Nuclear Energy Advanced Modeling and Simulation program. SHARP contains the high-fidelity single physics packages Diablo and Nek5000, both highly scalable and extensively validated. In this work, verification studies are performed to assess the results from both STAR-CCM+ and SHARP. The predicted deflections of the plate agree well with each other as well as exhibiting good agreement with simulations performed by the University of Missouri utilizing STAR-CCM+ coupled with the commercial structural mechanics code ABAQUS. The study provides a solid basis for FSI modeling capability for plate-type fuel element with SHARP.

In order to maintain the U.S. domestic nuclear capability, its scientific technical leadership, and to keep our options open for closing the nuclear fuel cycle, the Department of Energy, Office of Nuclear Energy (DOE-NE) invests in various R&D programs to identify and resolve technical challenges related to the sustainability of the nuclear fuel cycle. Sustainable fuel cycles are those that improve uranium resource utilization, maximize energy generation, minimize waste generation, improve safety and limit proliferation risk. DOE-NE chartered a Study on the evaluation and screening of nuclear fuel cycle options, to provide information about the potential benefits and challenges of nuclear fuel cycle options and to identify a relatively small number of promising fuel cycle options with the potential for achieving substantial improvements compared to the current nuclear fuel cycle in the United States. The identification of these promising fuel cycles helps in focusing and strengthening the U.S. R&D investment needed to support the set of promising fuel cycle system options and nuclear material management approaches. DOE-NE is developing and evaluating advanced technologies for the immobilization of waste issued from aqueous and electrochemical recycling activities including off-gas treatment and advanced fuel fabrication. The long-term scope of waste form development and performance activities includes not only the development, demonstration, and technical maturation of advanced waste management concepts but also the development and parameterization of defensible models to predict the long-term performance of waste forms in geologic disposal. Along with the finding of the Evaluation and Screening Study will be presented the major research efforts that are underway for the development and demonstration of waste forms and processes including glass ceramic for high-level waste raffinate, alloy waste forms and glass ceramics composites for HLW from the electrochemical processing of fast reactor fuels, and high durability waste forms for radioiodine.

Laser cleaning study was performed on prepared samples using a nanosecond pulsed ytterbium fiber laser. To prepare samples, AISI 304L stainless steel samples were oxidized and implemented with non-radioactive contaminants in a controlled manner. In order to validate the cleaning process for metallic equipment polluted in nuclear installations, two types of contamination with europium (Eu) and with cobalt (Co) were studied. Eu was used as a simulator-product resulting from uranium fission, while Co — as an activation-product of nickel, which is a composing element of a primary coolant system of a reactor. The oxide layers have suffered laser scanning which was followed by the furnace treatment to obtain thicknesses in the range of 100 nm to 1 μm depending on the oxidation parameters [1] with a mean weight percentage of 1% of Eu and 1 % of Co in the volume of the oxide layer. Glow Discharge Optical Emission (GD-OES) and Mass Spectrometry (GD-MS) analyses have been performed to assess the efficiency of the cleaning treatment and to follow the distribution of residual contamination with a detection limit of 0.1mg/kg of Eu and Co. Decontamination rates up to 95.5 % were obtained. One of the identified reasons for this limitation is that the radionuclides are trapped in surface defects like micro cracks [2, 3]. Therefore, cleaning treatments have been applied on surface defects with controlled geometry and a micrometric aperture obtained by laser engraving and juxtaposition of polished sheets of AISI 304L stainless steel. The goal of this study is surface decontamination without either welding or inducing penetration of contamination into the cracks. GD-MS analysis and Scanning Electron Microscopy (SEM) were performed to analyze the efficiency of the treatment and the diffusion of contaminants in this complex geometry.

The influence of gas introduction on the critical safety of the nuclear fuel system under the condition of cold condition, given reactor material and geometry structure is studied. Refer to bubble effect test experiment on nuclear critical safety test device (YSR) and considering solid-liquid two-phase nuclear fuel system with uranyl nitrate solution - uranium dioxide fuel element as the experimental platform, the dynamic process of the real behavior of bubbles in uranyl nitrate solution has been simulated in the quasi-static way by replacing bubble generator with aluminous bubble simulation elements. Bubble effect is the reactivity change caused by the change of volume of solution, neutron leakage and absorption property in the nuclear fuel system due to the bubbles generated in the solution. In the dissolving process of spent fuel, oxygen or nitrogen are usually added to accelerate the dissolution of fuel element shear section, and some other bubble production are also caused by the heat released during the dissolution process. Here, the bubble production caused by the heat is omitted and only artificial gas introduction is considered in my study. When there are enough bubbles in the uranium solution system, the volume of the solution will increase, which will inevitably influence the absorption and leakage property of the neutrons, and further influence the reactivity of the nuclear fuel system. The corresponding relationship between the bubble-intake rate and the bubble equivalent diameter, arising velocity and bubble share is determined through fluid dynamics modeling to manufacture the aluminous bubble simulation elements. The theoretical calculation by MONK9A and the critical experimental measurements are also compared and analyzed in this paper. The results showed that the reactivity caused by bubbles was negative, and the greater the intake rate, the greater the negative effect. Meanwhile the theoretical calculated value was in good agreement with the experimental value and the maximum deviation was 63.4 pcm.

The radon from uranium tailings spreads fast and has a wide range of pollution, which poses a potential radiation hazard to the environment and the public in downwind region. In this paper, the open and naked uranium tailings are selected as research object. By setting up multiple Gaussian plume models with single point source, the diffusion of radon in the uranium tailings is simulated with different atmospheric stability, average wind speed, height and downwind distance. The results show that the maximum radon concentration increases while the related downwind distance decreases as the atmospheric becoming stable. The higher wind speed does not affect the downwind distance where the maximum radon concentration occurs, but it decreases the maximum radon concentration. The concentration of radon in residential area decreases but the decreasing rate speeds up with height going up. The distribution of radon in vertical and horizontal direction tends to be homogeneous while the near-surface area concentration decreases rapidly as farther downwind distance.

The closed chamber method is widely used in measuring radon exhalation rate, which can avoid the error caused by the leakage and anti-diffusion phenomena. Firstly, considering the actual situation that uranium ore is difficult to obtain and have a high radioactivity, the uranium-like rock was made according to the similarity theory. Secondly, the diffusion length and intrinsic radon exhalation rate were obtained by using the closed chamber method. Thirdly, the theoretical values of radon exhalation rate made by uranium-like ore block were calculated, compared with the measured values. This study shows that the uranium-like rock block made by the best mass ratio is helpful for the subsequent experiment, and the error between the theoretical calculation and the measured value is no more than 9.14%. This indicates it is reliable to estimate radon exhalation rate by diffusion length and intrinsic radon exhalation rate and can also provide a foundation for rapidly gaining radon exhalation rate of the same type material by the closed chamber method. This study can further promote the study of the radon exhalation rate under the complex physical conditions and then better guide the protection work of radon radiation in underground mining.

The increasing of terror menace in recent years led the international community to enhance the efforts to minimize threats to people in everyday life by developing devices, techniques, and procedures targeted to improve the collective security. In this framework the Italian National Agency for New Technologies, Energy and Sustainable Economic Development (ENEA) developed a new device to improve CBRNe resilience, the Neutron Active Interrogation system (NAI). It has been conceived and optimized to identify transuranic-based Radioactive Dispersal Devices potentially hidden in packages, envisaging its utilization in field applications. NAI is based on the detection of neutrons from induced fission on small amount, of the order of a few grams, of fissile material. The device exploits a portable neutron generator based on d-t fusion reaction, a polyethylene structure for reducing the neutron energy in order to maximize the fission cross-section, and an array of 3 He proportional counters. Fissile material detection is made using the Differential Die-Away time Analysis (DDAA), an active neutron technique based on the difference among the die-away times of fast interrogation neutrons and prompt fission neutrons induced by thermal neutrons in the moderating system. The original experimental setup was tested on the field during the live demo open to the public at the EDEN Project Demonstration occurred in September 2015 at ENEA Frascati Research Centre in Rome. Since then, the setup has been modified to improve the device detection capabilities. NAI performances have been tested within different environmental conditions, e.g. open field geometry vs. bunker-like geometry, to study the effects of scattering phenomena. The optimized configuration here presented is transportable, lightweight, and able to detect 2 grams of 235 U contained in a salt of depleted uranium in real time, independently from the measurement environmental conditions.

Uranium hexafluoride (UF 6 ) accident is a typical accident in the nuclear fuel cycle. It combine radioactive and chemical hazards, so it is necessary to attach great importance to the UF6 accident. This paper analyzed and summarized the possible accident scenarios, causes and consequences, and classified the UF 6 accident risk factors, and put forward corresponding preventive and emergency measures. Preliminary hazard analysis of the accident can help us better understand the accident process, so we can takes steps for corresponding risk factors in advance, and prevent it will not be developed into an accident, so than we can obtain the effect of nip in the bud.

Based on the principle of similarity, uranium ore samples were prepared from raw materials such as uranium tailings, quartz sand and refined iron powder, and then the samples were treated with different packages for measuring the sample accumulated radon concentration. In the actual measurement process, due to the characteristics of radionuclide decay, instrument reasons and human factors, the data will be a certain deviation. Therefore, the method of wavelet analysis is used to denoise the accumulated radon concentration and obtain radon exhalation rate. The results of the study show: the correlation coefficient of cumulative radon concentration fitted by wavelet denoising is improved greatly, and all of them are above 0.99, the recalculated radon exhalation rate of the single side of the sample is decreased by 0.06Bq · m −2 · s −1 , and double-edged is decreased by 0.02Bq · m −2 · s −1 . The experiment proved that wavelet theory can be used to correct calculated value of radon exhalation rate of uranium-like rock. At the same time, it provides a new method for further study of uranium mine radiation protection parameters.

A method to measure an amount of nuclear materials in fuel debris by using muon tomography has being developed for proceeding with decommissioning of Fukushima Daiichi nuclear power plant. As a result of the Fukushima Daiichi nuclear disaster, the molten fuels were mixed with reactor structures and accumulated as fuel debris in the reactor buildings. There is still a large amount of fuel debris remained in each reactor. Fuel debris removal is planned in the near future and the debris will be taken out in this process. The debris need to be inspected from a viewpoint of nuclear material control. Since the debris is a mixture of fuel and other structures, it is hard to quantitate nuclear materials in debris by existing measurement method. Muons are cosmic-ray particles which have high energies, therefore, they are highly penetrative. This feature makes muon tomography sensitive to find heavy materials such as uranium or plutonium. We conducted a simulation study of applying muon tomography to measure fuel debris by using a Monte-Carlo method. A simulation model which includes muon detectors, shielding container and fuel debris was constructed to reproduce a measurement situation at the site. In conclusion, muon tomography quantitate the nuclear materials, therefore, this method should be useful for the fuel debris removal of Fukushima Daiichi reactors.