Cesium 137 discharged on ground surface sorbs tightly on fine particles made from silts and clays, and resulting particulate species sorbing ^<137>Cs migrates into deeper soil layer by rainfall penetration. In this research, generation of particulate ^<137>Cs species at ground surface and its migration behavior were examined under simulated environmental conditions. Migration experiments were carried out by a column method, in which deionized water was fed intermittently at the drying interval for 7 days into a sand layer contaminated with ^<137>Cs. A portion of the ^<137>Cs in the upper surface region, which formed particulate species by sorbing on fine particles, migrated into the deeper layer. Fine particle itself also was generated at the sand surface by weathering. The sand was weathered during the drying period, so that small amount of fine particles including ^<137>Cs was newly dissociated from the sand. Such particulate ^<137>Cs species may be transported downward and accumulated very slowly by repeated cycles of rainfall and drying, during long term.

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In the seismic probabilistic risk assessment for nuclear power plants, methods based on Monte Carlo simulation of response analyses with variously simulated ground-motion time histories have been conducted. In some methods, ground motions are generated to fit specified response spectra, e.g., the uniform hazard spectra or the response spectra via an attenuation relationship. However in these methods, the relation between detailed seismic-source characteristics and structural damages cannot be clarified. In some other methods, fault rupture models are used with specified source characteristics, while the occurrence probability of specified rupture is not clear. In the preceding study by Nishida et al. (2013), the methodology for reproducing ground-motion time histories were proposed, where ground motions are consistent with seismic hazard of the reference site and are associated with source characteristics. Ground motions were generated with fault rupture models with stochastic parameters of source characteristics, so that earthquake occurrence probabilities for parameters can be assumed. In this paper, results for response analyses of structures with some sets of these generated ground motions are presented. The response analyses were conducted for a simple RC structure and for Pressurized Water Reactor building. The relation between source characteristics and seismic intensities as well as the relation between sou

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In order to improve the safety of Boiling Water Reactor (BWR), it is required to know the behavior of the plant when an accident occurred as can be seen at Fukushima Daiichi nuclear power plant accident. Especially, it is important to estimate the behavior of molten core jet in the lower part of the reactor pressure vessel at a severe accident. In the BWR lower plenum, the flow characteristics of molten core jet are affected by many complicated structures, such as control rod guide tubes, instrument guide tubes and core support plate. However, it is difficult to evaluate these effects on molten core jet experimentally. Therefore, we considered that multi-phase computational fluid dynamics approach is the best way to estimate the effects on molten core jet by complicated structure. The objective of this study is to develop the evaluation method for the flow characteristic of molten core jet including the effects of the complicated structures in the lower plenum. So we are developing a simulation method to estimate the behavior of molten core jet falling down through the core support plate to the lower plenum of the BWR. The simulation method is based on interface tracking method code TPFIT (Two Phase Flow simulation code with Interface Tracking). To verify and validate the applicability of the developed method in detail, it is necessary to obtain the experimental data that can be compared with detailed numerical results by the TPFIT. Thus, the authors are carrying out experimental works by use of multi-phase flow visualization technique. In the experiments, time series of interface shapes are observed by high speed camera and velocity profiles in/out of the jet are measured by the PIV method. In this paper, we carried out a numerical simulation of the jet breakup phenomena in the multi-channels with various simulant molten materials to evaluate the influence of properties on the jet breakup phenomena. As a result, it was confirmed that density and surface tension affected on the falling down velocity of the simulant materials and the interface behavior of the molten jet. However, viscosities of the simulant materials have small effects on jet breakup phenomena, including the interface shape and size of fragments.

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In accidents when the water level of the reactor core descends below the top of the active fuel, the cooling limit height is a key factor in determining the accident mitigation procedure and the boiling two-phase flow in a fuel-rod bundle exhibits multi-dimensional and complex flow structures during such boil-off process. A rod bundle boil-off experiment was conducted to determine the three-dimensional void-fraction distribution and axial profile of the rod-surface temperature during the boil-off process under atmospheric pressure conditions. The 5×5 rod bundle, featuring a heated length of 2 m, had an axially and radially uniform power profile, with eight pairs of sheath thermocouples embedded on the heated rod to monitor its surface temperature distribution. The void-fraction distribution was acquired with five pairs of SubChannel Void Sensor (SCVS) as time-series data. The experimental results showed the relationship between an effective cooling level and boiling two-phase flow dynamics in the rod bundle.

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At the Korea Atomic Energy Research Institute (KAERI), the Korea Research Reactor (KRR-2) and one uranium conversion plant (UCP) were decommissioned. A project was launched in 1997, for the decommissioning of KRR-2 reactor with the goal of completion by 2008. Another project for the decommissioning of the UCP was launched in 2001. The physical dismantling works were started in August 2003 and the entire project was completed by the end of 2010. KAERI has developed a computer information system, named DECOMMIS, for an information management with an increased effectiveness for decommissioning projects and for record keeping for the future decommissioning projects. This decommissioning information system consists of three sub-systems; code management system, data input system (DDIS) and data processing and output system (DDPS). Through the DDIS, the data can be directly inputted at sites to minimize the time gap between the dismantling activities and the evaluation of the data by the project staff. The DDPS provides useful information to the staff for more effective project management and this information includes several fields, such as project progress management, man power management, waste management, and radiation dose control of workers and so on. The DECOMMIS was applied to the decommissioning projects of the KRR-2 and the UCP, and was utilized to give information to the staff for making decisions regarding the progress of projects. It is also to prepare the reference data for the R&D program which is for the development of the decommissioning engineering system tools and to maintain the decommissioning data for the next projects. In this paper, the overall system will be explained and the several examples of its utilization, focused on waste management and manpower control, will be introduced.

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The weld pass sequence generally affects residual stress profiles. In this study, the profiles of bead welded specimens were experimentally and analytically examined. The material used in this study was the low alloy steel SQV2A (JIS G 3120), which is equivalent to ASTM A533 Type B Class 1. Phase transformation usually occurs around the welding area in SQV2A and also affects the residual stress profiles. Two specimens were fabricated. One was single pass bead welded (single pass specimen) and the other was five pass bead welded (five pass specimen). Residual stresses of these specimens were measured by using the strain relief method. The measurement results show that longitudinal residual stress on the welded surface of the single pass specimen was lower than that of the opposite side surface due to phase transformation. Longitudinal residual stress around the 5th pass area of the five pass specimen was similar to that around the 1st pass area in the single pass specimen. The residual stress in the previous passes area, where passes from the 1st to the 4th were welded, was higher than that of the 5th pass area. These findings show that residual stress occurred around the welding area of these specimens and that the subsequent weld passes affect the profiles. To estimate the residual stress profiles numerically, temperature history analysis and residual stress analysis were conducted. Phase transformation strain was taken into consideration in these analyses as a component of thermal strain. The analysis results are consistent with the experiment results.

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Decommissioning Engineering Center, which is called FUGEN, has started dismantling works based on its decommissioning program since 2008. Now is in initial stage, the purpose is accumulating of experiences and obtaining of various data in dismantling works. The dismantling works was launched in turbine system whose contamination was relatively low level. Feed-water-heaters and main-steam-pipes had been dismantled already, and now, main-condensers have been dismantling. Approximately 1000 tons of dismantled waste was produced from the dismantling works so far. Dismantling work needs to be done safely and effectively with appropriate cutting devices depending on the situations that is its method, quality and shape of the materials to be dismantled. Therefore, in FUGEN, varieties of conventional cutting devices, which are thermal cutting method (Gas cutting, Gasoline firearms cutting, Plasma-Arc cutting) and mechanical cutting method, are used in the dismantling works to evaluate their applicability depending on the work situations. Obtained cutting data are summarized and evaluated in order to reflect to the following own works and other decommissioning plants.

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Zircalloy has been widely employed as an excellent material covering the fuel rod. The mechanical and thermal properties have been explored by various experiments [1]. In terms of its use as the fuel cladding, its response to oxidation reactions is an important topic when it is exposed to high temperature and high pressure steam during severe accidents. Especially, the hydrogen production accompanied by the oxidation is critical because it can lead to the crisis of the hydrogen explosion, as observed in the Fukushima nuclear power plant accidents. Silicon carbide (SiC) has been considered as an alternative cladding material owing to an advantage that hydrogen production is much suppressed in the equivalent condition compared to Zircalloy. Therefore, we simulate the oxidation reaction for both materials, i.e. Zirconium metal and SiC in atomistic level by using the ReaxFF reactive force field method to simulate the chemical reaction molecular dynamics. Through such comparative studies between Zirconium and SiC in the same condition, we clarify how the temperature and the steam pressure accelerates the oxidation reaction and the resultant hydrogen production in both materials at typical severe accident conditions. The advantage using ReaxFF is that it allows us to directly trace the oxygen diffusion inside the Zirconium metal and SiC depending on the temperature and vapor pressure together with the oxidation reaction. We can compare the reaction processes in both materials. Especially, we paid attention to the rate of hydrogen production in both materials.

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This paper describes experimental and analytical works to confirm that the design standard for Sodium cooled Fast Reactor (SFR) components sufficiently covers possible failure mechanisms. Present design standards for SFR components have been proposed by Japan Society of Mechanical Engineers (JSME). For economical reasons, Mod.9Cr-1Mo ferritic steel, equivalent to ASME Gr.91, will be widely adopted to coolant systems in future SFRs. Creep-fatigue damage evaluation method in JSME design standard for SFR components has been constructed based on experiments and/or numerical analyses of conventional austenitic stainless steels, such as 304SS. Since the material characteristics of Mod.9Cr-1Mo steel are substantially different from those of austenitic stainless steels, it is required to verify the applicability of the design standards to the SFR components made of Mod.9Cr-1Mo steel. However, few structural test data exist to verify the applicability of the conventional design standards. Therefore, to investigate creep-fatigue failure aspect of the several types of structural discontinuities, a series of uni-axial creep-fatigue tests were conducted using double-ended notch bar specimens made of Mod.9Cr-1Mo steel under displacement controlled condition with 30 minute holding. In these tests, imposed axial displacement was controlled to result in 0.7% axial strain range at the minimum section area. The curvature radii of the specimens were 1.6 mm, 11.2 mm and 40.0 mm. The specimen having 1.6 mm notch and 11.2 mm notch failed from outer surface but the specimen having 40.0 mm notch showed obvious internal crack nucleation. In addition, though total duration time of the creep-fatigue test was only 2,000 hours, a lot of creep voids and inter granular crack growth were observed. To clarify the cause of such peculiar failure, some additional experiments were performed, as well as some numerical analyses. We could point out that such a peculiar failure aspect might result from corresponding stress distribution in the cross section. Since radial temperature distribution resulted from high-frequency induction heating and local multiaxiality also might produce such peculiar failure, analytical investigations were carried out. As a result of a series of investigations, possible causes of such peculiar failure could be narrowed down. A future investigation plan was proposed to clarify the most significant cause.

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The CONTAIN-LMR code is being developed in the Japan Atomic Energy Agency (JAEA) since 1980's for the purpose to utilize for the quantitative assessment of accident consequences considered in sodium-cooled fast reactor (SFR) plant. It is a best-estimate integrated analysis tool to predict lots of thermal- hydraulic behaviors and radioactive materials transfer to the environment in the case of hypothetically postulated ex-vessel severe accident progression. Out of various physical and chemical behaviors treated in the code, the present paper describes sodium fire related study issues such as computational modeling and its validation activities with focusing on important evaluating targets. Major objective in using sodium leak and fire modules in the evaluation of SFR severe accident is not necessarily to clarify the detailed or localized phenomena, but rather to quantify the longer-term or overall dominant thermal consequences affecting to the plant building. The related computational models implemented in CONTAIN-LMR are therefore typical ones to reproduce sodium spray and pool fires, which include relatively simplified model for eliminating higher computational cost. The authors present in this paper representative code validation practices for fundamental sodium combustion and consequential heat and mass transfer behavior models through the numerical analyses of sodium leak and fire experiments performed in the SAPFIRE facility.

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The innovative CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burnup strategy has been successfully applied to both fast and thermal reactors. In particular for thermal reactor applications, CANDLE block-type high temperature gas-cooled reactors (HTGRs) with either uranium or thorium fuel cycle had been proposed and investigated for their simple and safe reactor operation, and the ease of designing a long life reactor. Small sized long life CANDLE HTGRs with thorium fuel shows superior burnup performance than the ones with uranium fuel but their axial power peaks are relatively higher which may not be advantageous during a depressurization accident. In this work, we proposed and investigated the use of Pa-231 mixed homogeneously in the (Th-232/U-233)O_2 fuel kernel of the TRISO particles to obtain lower axial power peaks. Addition of Pa-231 decreases the required amount of natural gadolinium burnable poison in the fresh fuel for establishing a valid CANDLE HTGR design since Pa-231 has a large thermal absorption cross section. Besides the role as a burnable poison nuclide, Pa-231 also serves as a fertile nuclide during the CANDLE burning since Pa-231 is finally transmuted to a fissile U-233 nuclide. A promising analysis result shows that for U-233 enrichment of 15 w/o and Pa-231 addition of 7.50 w/o, the axial power peak is decreased from 5.9 to 3.6 W/cm^3 while the averaged burnup is increased from 138 to 149 GWd/t. This extends the core life time about 16 %, i.e. from 35 to 41 years with CANDLE active core height of 800 cm and reactor thermal power of 30 MWth.

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The Nuclear safety instrument and control system is directly related to the safety of the reactor. So redundant and diversity design is used to ensure the system's security and reliability. This make the traditional safety system large, more cabinets and wiring complexity. To solve these problem, we can adopt new technology to make the design more simple. The simplify conceptual design can make the system less cabinets, less wiring, but high security, strong reliability.

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Source term for severe accident analysis of molten salt reactors(MSRs) has been investigated as part of preliminary efforts to develop MSRs. As a severe accident of MSRs, exposure of heated fluoride fuel molten salt to atmosphere was assumed to take place. Vaporization of fluoride molten salt was studied by means of the two methods, the Knudsen effusion mass spectrometry as well as the transpiration method. The former was applied to pseudo-binary fluoride systems to clarify the behaviors of cesium and iodine in the fluoride molten salt. The latter was applied to the mixture of CsI and FLiNaK. These experiments were carried out as the first step of the source term studies, so that interaction with air components has not been covered yet. From this study, useful information related to the source term for MSRs have been obtained. This work suggests how to solve the problem to establish the source term for severe accident analysis of MSRs.

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Once a severe accident occurs, to estimate the future behavior and risks of the plant, it is important to know the real-time fuel information such as decay heat and nuclide inventories, which is usually different from the safety evaluation with conservative conditions. In order to fulfill this requirement, Mitsubishi Heavy Industries, Ltd. (MHI) has been developing a fuel management system - M-SAVIOR, which analyses and manages the real-time fuel information such as source term, decay heat and radioactivity. Nuclear power electric utilities and their fuel suppliers have to manage the holding uranium and plutonium inventories properly as the accounting records. Using our new system, the nuclides relating to radioactivity can be added to the accounting records easily and consistently with the uranium and plutonium inventories presently maintained. In this paper, we describe our new system - M-SAVIOR and its evaluation method of the individual fuel information.

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Reactor imaging using scattering of cosmic-ray muon is planned to assess the damages to the reactors at Fukushima Daiichi in 2015. The technique was demonstrated at a research reactor, Toshiba Nuclear Critical Assembly. Simulation studies also showed feasibility of the reactor imaging using muon-scattering method.

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There are many differences between the flow and heat transfer characteristics of nuclear reactors under ocean and land-based conditions for the effects of ocean waves. In this paper, thermal hydraulic characteristics of a passive residual heat removal system (PRHRS) for an integrated pressurized water reactor (IPWR) in ocean environment were investigated theoretically. A series of reasonable theoretical models for a PRHRS in an IPWR were established. A transient analysis code in FORTRAN 90 format has been developed to analyze the thermal-hydraulic characteristics of the PRHRS under ocean conditions. The code was implemented to analyze the effects of different ocean motions on the transient thermal-hydraulic characteristics of PRHRS. It is found that the oscillating amplitudes and periods of the system parameters are determined by those of the ocean motions. The obtained analysis results are significant to the improvement design of the PRHRS and the safety operation of the IPWR.

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By the method of numerical simulation and the geometric modeling based on existing experiment table, the heat transfer characteristics of water, sodium and lead-bismuth in a 2mm vertical rectangular channel were studied. The effect of different boundary conditions on the thermal characteristics was also studied. The effect of inlet velocity and heat flux density on the heat transfer coefficient and the pressure drop of different fluids in the rectangular channel were studied. It provides a reference for selection of working fluid in the narrow rectangular channel.

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Severe core disruptive accidents in sodium-cooled fast reactor and core molten materials and the coolant sodium interaction (Fuel-Coolant Interaction, FCI), are one of the international key and difficult problem on the core safety studies of reactor, especially for FCI process, drastic changes of the mult iphase flow and heat transfer between cold and hot molten fluid interface, and molten material forms, deformation and fragmentation and so on. Based on so many difficulties with large deformation and numerical incontinuous problems in mesh method, a meshless moving particle semi-implicit (MPS) method is introduced in this study. This paper focuses on the simulation of freezing and fragmentation behavior of a single molten Type 304 stainless steel droplet penetrating into sodium pool. An effective viscosity model is introduced to take account of the viscosity change during the phase change. The improved MPS algorithm with surface tension model is adopted to a series of simulation of a single molten stainless steel droplet penetrating into the sodium pool in the range of 1 < V_0 < 5m/s , 1430 < T_<drop> < 1700℃ and 300 T_<na> = 300℃ . The penetrating process is simply clarified in this study. The D_<max> value has a tendency to be smaller with increasing T_i and We_a.

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Research and development on three-dimensional vibration simulation technologies for nuclear facilities is one of the missions of the Center for Computational Science and e-Systems, Japan Atomic Energy Agency. Until now, three-dimensional building and equipment models of High Temperature engineering Test Reactor (HTTR) have been constructed and validated using comparisons with seismic observation records. In this paper, we report the results of a seismic observation simulation of the 2011 off the Pacific coast of Tohoku Earthquake that occurred on 3/11/2011 using three-dimensional models of the HTTR building. The simulation results show good agreement with the observation data.

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Design study of an advanced loop type sodium cooled fast reactor (JSFR) has been performed in Japan. A column type upper internal structure (UIS) is installed in the upper plenum of reactor vessel in JSFR. This UIS is composed of control rod guide tubes and several perforated horizontal plates. Further, each horizontal plate has a radial slit for operation of the fuel handling system. This unique structure of the UIS permits existence of the fluids from the fuel assembly outlets in the UIS. The maximum temperature difference between cold fluid and hot fluid is approximately 100 K as a tentative condition. Therefore, high cycle thermal fatigue may occur at the bottom plate (CIP) of the UIS where the hot sodium from the fuel subassembly can mix with the cold sodium from the control rod channel and the blanket fuel subassembly. We have been conducted a water experiment using a reactor upper plenum model to grasp the thermal-hydraulic phenomena around control rod (CR) channels, and radial blanket assemblies and to obtain countermeasures for significant temperature fluctuation on the CIP. The experimental apparatus has 1/3 scale and 60 degree sector model of the reactor upper plenum. The temperatures around the CR channels and the blanket assemblies were measured with thermocouples. By the experiment, characteristics of fluid temperature fluctuation between the handling head of the assemblies and the CIP are measured and countermeasure, for the significant temperature fluctuation generation will be discussed on the influence of the distance from the handling head outlet to the lower surface of the CIP. It was confirmed that temperature fluctuations could be mitigated by expanding a distance between a handling head and CIP.

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As the "central nerve system", the highly reliable Instrumentation & Control (I&C) systems, which provide the right functions and functions correctly, are always desirable not only for the end users of NPPs but also the suppliers of I&C systems. The Digitalization of nuclear I&C system happened in recent years brought a lot of new features for nuclear I&C system. On one side digital technology provides more functionalities, and it should be more reliable and robust; on the other side, digital technology brings new challenge for nuclear I&C system, especially the software running in the hardware component. The software provides flexible functionalities for nuclear I&C system, but it also brings the difficulties to evaluate the reliability and safety of it because of the complexity of software. The reliability of software, which is indispensable part of I&C system, will have essential impact on the reliability of the whole system, and people definitely want to know what the reliability of this intangible part is. The methods used for the evaluation of reliability of system and hardware hardly work for software, because the inherent difference of failure mechanism exists between software and hardware. Failure in software is systematically induced by design error, but failure in hardware is randomly induced by material and production. To continue the effort on this hot topic and to try to achieve consensus on the potential methodology for software reliability evaluation, a cooperative research project called RAVONSICS (Reliability and Verification and Validation of Nuclear Safety I&C Software) is being carried on by 7 Chinese partners, which includes University, research institute, utility, vendor, and safety regulatory body. The objective of RAVONSICS is to bring forwards the methodology for the software reliability evaluation, and the software verification technique. RAVONSICS works cooperatively with its European sister project HARMONICS, both will target the software reliability evaluation and testing methodologies and techniques.

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In order to precisely investigate molten core relocation behavior in the Fukushima Daiichi nuclear power station, we have developed the detailed and phenomenological numerical simulation code named JUPITER for predicting the molten core behavior including solidification and relocation based on the three-dimensional multiphase thermal-hydraulic simulation models. In this paper, in order to carry out the realistic simulation which uses five components, i.e., fuel rod, channel box, control rod and absorber, we improved the multicomponent analysis method and implemented the simple radiation model to the code and carried out to check effectiveness of the models based on the numerical simulation of melting behavior of the simulated core internals. From the present numerical results, it was confirmed that a newly developed multicomponent analysis method appropriately can predict the relocation behavior of molten materials in complicated structures, that is relocation and solidification behavior for fuel components including the heat source, i.e., simulated decay heat, and the simplified radiation heat transfer.

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The influence of dislocation density on yield strength, which is a key factor in the anomalous deformation behavior in UFG metals, was investigated by huge scale atomistic simulations. Polycrystalline models with intragulanular Frank-Read sources were constructed to elucidate the relationship between the inter- and intra-granular plastic deformation processes and the mechanical properties. Then the uniaxial tension and compression were applied to the polycrystalline copper. Consequently it was found that Frank-Read sources were activated prior to intergranular dislocation emission, and the yield event of the whole system seems to occur when some dislocation sources activated. The yield stress is strongly influenced by the number of intragranular dislocation sources, i.e., dislocation density. Additionally, the Bauschinger effect of UFG metals is caused by the change in dislocation density in the process of forward and backward deformation.

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This paper describes the robust interpolation method for axial reaction rate (RR) distributions measured by fixed in-core detectors (FID) employed in Mitsubishi online core monitoring system VISION. Contrast to the movable detector (M/D), axial power distributions measured with FIDs have to be interpolated in order to obtain power profiles, because FID's signals are discretely measured. Therefore, the interpolation method is very important for the FID system. Besides, the lack of measurement RR data due to detector failures affects strongly the power profiles. Thus, the FID system has to use the robust interpolation method. The developed interpolation method is based on the calibration method of calculated axial RR distributions with measurement values. This means, the measured RR distribution is obtained by calibrating calculated RRs to fit discrete FIDs measured values. Applicability of the present method to PWRs was confirmed by the virtual FID data, which were made from M/D measurements in a typical PWR. Then we applied the present method to the virtual FID data to reconstruct the RR distribution of M/D. The reconstructed distributions were in good agreement with the measured M/D RR distributions. We confirmed that this method is applicable to the FID system in PWR.

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A computer code, CONTAIN-LMR, has been developed in Japan Atomic Energy Agency (JAEA) for application to a probabilistic risk assessment (PRA) of liquid metal fast reactors (LMFRs) since the original CONTAIN code had been introduced from Sandia National Laboratories (SNL) of U.S. in 1982. The CONTAIN-LMR code is a best-estimate, integrated analysis tool for predicting the physical, chemical and radiological conditions inside a containment building of LMFRs following a severe accident with reactor vessel melt-through. The code is also able to predict the source term to the environment in the accident. This code can treat many important phenomena consistently such as sodium fire, radioactive aerosol behavior, a water release from heated concrete, hydrogen burn, sodium-concrete reaction and core debris-concrete interaction occurred in the accident with inter-cell heat and mass flow under the multiple cell geometry. This paper describes the chronology of the code development in JAEA briefly as an introduction, and after that, the outline of computational models in the code, the examples of the code validation, and the future plan of the code application are described.

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Hybrid control rod is a passive decay heat removal device, which is a new concept with combination of heat pipe and control rod by inserting neutron absorber inside heat pipe as inner structure. Heat pipe is an excellent heat transfer device using the principle of efficient convection heat transfer by phase change of working fluid and convection through a wick structure inside closed metal tube. Since heat pipe can operate without external power source, it can achieve totally passive heat transfer. It has the strength when refueling water for core cooling by ECCS cannot be injected due to impossible depressurization inside reactor vessel. In that situation, hybrid control rod can remove the decay heat from core directly as a heat pipe inserted into reactor vessel as well as shutdown of reactor as a control rod. Therefore, it can significantly reduce the possibility of core meltdown and release of radioactive material by assuring sufficient core cooling. For evaluating the concept of the hybrid control rod and its application, design features of hybrid control rod were analyzed using a CFD code. In this paper, first a conventional heat pipe was analyzed and compared with experimental results obtained in same conditions to validate the current heat pipe analysis model. The CFD predictions were in a good agreement with experimental results in terms of wall temperature. Different from general heat pipes using vacuum condition, the hybrid control rod was designed to be pressurized inside the heat pipe considering working condition of high temperature inside reactor vessel. The hybrid control rod reflecting a reactor environment was analyzed for wall temperature and thermal characteristics of conventional heat pipe. As a result, it found that the current hybrid control rod concept can remove 18.20 kW per rod at reactor condition.

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A computer code, CONTAIN-LMR, is an integrated analysis tool to predict the consequence of severe accident in a liquid metal fast reactor. Because a sodium-concrete reaction behavior is one of the most important phenomena in the accident, a Sodium-Limestone Concrete Ablation Model (SLAM) has been developed and installed into the original CONTAIN code at Sandia National Laboratories (SNL) in the U.S. The SLAM treats chemical reaction kinetics between the sodium and the concrete compositions mechanistically using a three-region model, containing a pool (sodium and reaction debris) region, a dry (boundary layer (B/L) and dehydrated concrete) region, and a wet (hydrated concrete) region, the application is limited to the reaction between sodium and limestone concrete. In order to apply SLAM to the reaction between sodium and siliceous concrete which is an ordinary structural concrete in Japan, the chemical reaction kinetics model has been improved to consider the new chemical reactions between sodium and silicon dioxide. The improved model was validated to analyze a series of sodium-concrete experiments which were conducted in Japan Atomic Energy Agency (JAEA). It has been found that relatively good agreement between calculation and experimental results is obtained and the CONTAIN-LMR code has been validated with regard to the sodium-concrete reaction phenomena.

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A district heating and snow melting system utilizing waste heat from Gas Turbine High temperature Gas Reactor of 300 MWe (GTHTR300), a heat-electricity cogeneration design of high temperature gas-cooled reactor, was analyzed. Application areas are set in Sapporo and Ishikari, the heavy snowfall cities in Northern Japan. The heat transport analyses are carried out by modeling the components in the system; pipelines of the secondary water loops between GTHTR300s and heat demand district and heat exchangers to transport the heat from the secondary water loops to the tertiary loops in the district. Double pipe for the secondary loops are advantageous for less heat loss and smaller excavation area. On the other hand, these pipes has disadvantage of more electricity consumption for pumping. Most of the heat demand in the month of maximum requirement can be supplied by 2 GTHTR300s and delivered by 9 secondary loops and around 5000 heat exchangers. Closer location of GTHTR300 site to the heat demand district is largely advantageous economically. Less decrease of the distance from 40 km to 20 km made the heat loss half and cost of the heat transfer system 22% smaller.

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Markov's chains (MC) are well-know and widely applied in dependability and performability analysis of safety-critical systems, because of the flexible representation of system components dependencies and synchronization. There are few roadblocks for greater application of the MC: accounting the additional system components increases the model state-space and complicates analysis; the non-numerically sophisticated user may find it difficult to decide between the variety of numerical methods to determine the most suitable and accurate for their application. Thus obtaining the high accurate and trusted modeling results becomes a nontrivial task. In this paper, we present the metric-based approach for selection of the applicable solution approach, based on the analysis of MCs stiffness, decomposability, sparsity and fragmentedness. Using this selection procedure the modeler can provide the verification of earlier obtained results. The presented approach was implemented in utility MSMC, which supports the MC construction, metric-based analysis, recommendations shaping and model solution. The model can be exported to the well-known off-the-shelf mathematical packages for verification. The paper presents the case study of the industrial NPP I&C system, manufactured by RPC Radiy. The paper shows an application of metric-based approach and MSMC tool for dependability and safety analysis of RTS, and procedure of results verification.

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Equivalent non-Gaussian excitation method is proposed to obtain the moments up to the fourth order of the response of systems under non-Gaussian random excitation. The excitation is prescribed by the probability density and power spectrum. Moment equations for the response can be derived from the stochastic differential equations for the excitation and the system. However, the moment equations are not closed due to the nonlinearity of the diffusion coefficient in the equation for the excitation. In the proposed method, the diffusion coefficient is replaced with the equivalent diffusion coefficient approximately to obtain a closed set of the moment equations. The square of the equivalent diffusion coefficient is expressed by the second-order polynomial. In order to demonstrate the validity of the method, a linear system to non-Gaussian excitation with generalized Gaussian distribution is analyzed. The results show the method is applicable to non-Gaussian excitation with the widely different kurtosis and bandwidth.

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MELCOR contains a coupling interface based on the MPI-Standard, which enables the communication to other codes such as RELAP5 or GASFLOW. However a detailed knowledge of this coupling interface in MELCOR is necessary to use this possibility. Therefore, at the KIT the software tool DINAMO (Direct Interface for Adding Models) has been developed. This program contains the coupling routines as well as an interface to communicate with other programs. Using DINAMO it is also possible to utilize new or enhanced models for phenomena, which occur during a severe accident in a nuclear power plant, in MELCOR without modification of the MELCOR source code. In the present work MELCOR calculations of experiments in the LIVE-Facility are presented. The LIVE-Facility is used to simulate the behavior of a melt in the lower plenum of a reactor pressure vessel (RPV). For these calculations we coupled MELCOR via DINAMO with the Phase-Change Effective Convectivity Model (PECM), which has been developed at the KTH in Stockholm. Using the PECM it is possible to improve the prediction of a core melt in the lower plenum of a RPV in case of a core melt accident.

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This paper presents an overview of public relations activities of Nuclear Waste Management Organization of Japan (NUMO), the Japanese implementer of the geological disposal of high-level radioactive waste, which is operated for 14 years since its establishment. The first part of the paper outlines brief history of its public relations activities to date, and describes important events over the period. In the second part, it discusses the quantitative results of NUMO's public opinion surveys on its geological disposal project, conducted on a yearly basis, and assesses what NUMO needs to focus on for obtaining public consensus for implementing this project in the country. Then, in the third part of the paper, it introduces nationwide consensus-building activities and current challenges in the field of education. Lastly, the paper concludes with introduction of current Government's movement and its further perspective and with discussion of how NUMO will pursue its mission in cooperation with stakeholders.

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We report characterization and radio assay of the surface run-off substances obtained (or collected) at the forest observation plots in the Ogi district of Kawauchi-mura and the Yamakiya district of Kawamata-machi, and at the soil-saving dam in the Ogaki district of Namie-machi, Fukishima prefecture, Japan. Both Cs-134 and Cs-137 were detected from soil samples taken in these areas, and their concentrations were high particularly in the fine substances (silts and clay). The radioactive caesium concentration of surface run-off soils which flowed out from Kawauchi, and Kawamata observation plot were on the same level with that of the splash erosion soils. The radioactive caesium concentrations of the Ogaki soil saving dam were increased with finer-sized particle contents, mainly silt fraction in the sample soils.

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In 2006, Japan Nuclear Fuel Limited (JNFL) began "the delivery lesson on radiation" in order to promote understanding of radiation for the elementary and junior high schools students. Currently, "the delivery lesson on radiation" is aimed mainly for the junior high school students. About 1800 students per year take the lesson. We will report the effects and tasks of "the delivery lesson on radiation" based on the results of the questionnaire filled out before and after the lesson.

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The pebble bed water-cooled reactor (PBWR) provides another prospective energy source with advantages of traditional pressurized water reactor (PWR) and pebble bed reactor (PBR). Nevertheless, researches on PBWR are too scarce to advance its further development. This research investigates the thermal hydraulics of body-centered cubic (BCC) and face-centered cubic (FCC) PBWR core through computational fluid dynamics (CFD). Bridge model is adopted to modify the contact points between adjacent pebbles, thus making it possible to generate high quality meshes in narrow spaces. The comparison between BCC and FCC arrangement is conducted to analyze their influence on water flow and heat transfer. Results indicate that the average value of Nusselt number ( Nu )for BCC arrangement is 50% lower than FCC arrangement, whereas the pressure drop coefficient for FCC arrangement is 4.4 times as that of BCC arrangement. Furthermore, vortices are identified in void space for both arrangements but their contribution to improving heat transfer is different. In addition, the maximum temperature of fuel pebbles for both arrangements is about 1400K lower than melting point of Zirconium. Last, the simulation results are compared to KTA, which shows good correspondence generally.

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Development of damaged fuel debris cutting technique in Fukushima Daiichi nuclear plant is one of the important issues for the defueling operations such as removing from the reactor, handling and storage, and so on. Previous knowledge in the Three Mile Island Unit 2 (TMI-2) nuclear plant accident is reported that fuel debris was made of uranium oxides, stainless steel, zircaloy and so on. Therefore, it is difficult to cut fuel debris by machine processing. Laser cutting is convenient to cut complex materials. However, it is difficult to know the thickness of fuel debris in advance, so it is hard to apply the conventional laser cutting method. In this study, we have been developed the cutting technique for fuel debris by using laser gouging method. Test pieces that simulate fuel debris ware fabricated by embedding ceramic pellet on stainless steel. Laser gouging test was performed by cutting surface of the test pieces with high power fiber laser, whose maximum irradiation power was 10kW. According to the test result, laser gouging process has the ability to grind the simulated fuel debris under atmospheric condition. In addition, same kind of cutting effect could be performed underwater.

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The plant simulator is required to be capable of simulating accident events stably and accurately, particularly for applications to an operation support system for plant accident events. The simulation system has a three-dimensional neutron calculation, a two-phase flow calculation, a fuel rod temperature calculation, and a control-block calculation. The two-phase flow calculation uses a two-phase flow stabilized calculation model to cover a wide pressure range. The two-phase flow calculation includes the virtual mass term and an explicit diffusion term to remove numerical instability. And in particular for the low pressure and large diameter flow channel conditions, the interfacial shear model was developed using experimental data and it was applied in the two-phase flow calculation. We evaluated effect of the stabilization by the steady-state two-phase flow analysis in a vertical pipe and confirmed that the stability of the analysis was sufficiently good and the void fraction was predicted accurately in the pressure range from 0.5MPa to 7.2MPa by introducing the explicit diffusion term and the developed interfacial shear model for the low pressure and the large diameter flow channel.

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The Korea Hydro & Nuclear Power Co., Ltd., (KHNP) has investigated and evaluated various efficient thermal treatment technologies for the LILW. In 1994 and 1995, the feasibility of several melter technologies was assessed from technical and economic perspectives. Finally, the R&D project to develop the vitrification technology using CCIM (Cold Crucible Induction Melter) and PTM (Plasma Torch Melter) was launched in 1997. This R&D project had been completed from 1997 to 2002. KHNP started the project to construct the commercial facility using the results of the R&D project in 2002. The HanUl Vitrification Facility (UVF), to be used for the vitirification of low-and intermediate-level radioactive waste (LILW) generated by nuclear power plants (NPPs), is the world's first commercial facility using CCIM technology. The design of UVF had been conducted from 2002 to 2005. The construction was begun in 2005 and was completed in 2007. From 2007 to 2009, all key performance tests, such as the system functional test, the cold test, the hot test, and the real waste test, were successfully carried out. The UVF commenced the commercial operation in October 2009. Based on the successful construction and operation of UVF, the advanced R&D project has been started to develop the large-scale vitrification facility.

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To contain the impact of core disruptive accidents (CDAs) of sodium-cooled fast reactors in the reactor vessels, it is important to retain molten core material within the vessel during CDAs. This concept is called in-vessel retention (IVR). Molten core material discharged into the lower sodium plenum has the potential to impose a significant thermal load on the lower structures of the reactor vessels and thus may compromises IVR. However, if the molten core material is fragmented into smaller particles well before it reaches the lower structures, such the thermal load should be significantly reduced by enhanced quenching of this core material. Hence, the fragmentation of molten core material is crucial for achieving IVR. In this paper, based on the experimental results of a series of fragmentation tests (FR tests), in which around 10 kg of molten alumina (melting point: approximately 2300 K) was discharged into a sodium pool (depth: 1 m, diameter: 0.4 m, sodium temperature: 673 K) through a duct (inner diameter: from 40mm to 63 mm) by using an experimental facility in National Nuclear Center of the Republic of Kazakhstan, dominant fragmentation mechanisms of molten core material discharged into sodium are discussed. In the FR tests, molten alumina was finely fragmented in the sodium pool, thereby forming a debris bed. The mass median diameters of solidified alumina particles were around 0.3 mm, which are comparable to particle sizes predicted by hydrodynamic instability theories such as Kelvin-Helmholtz instability. However, even though hydrodynamic instability theories predict that particle size decreases with the increase of Weber number (We), such the dependence of particle size on We was not observed in the FR tests. Considering that in the tests, the distances for fragmentation of the molten alumina were approximately from 60 % to 70 % (i.e., around 65 %) below the values predicted using an existing representative correlation that regards hydrodynamic instabilities as a dominant fragmentation mechanism, it can be interpreted that the tendency of measured mass median diameters (i.e., non-dependence on We) suggests that before hydrodynamic instabilities sufficiently grow to induce fragmentation, thermal phenomena such as local coolant vaporization and resultant vapor expansion accelerate fragmentation.

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During the recirculation after postulated loss of coolant accident (LOCA) scenario in PWR, debris that generated in the containment might challenge the core flow path and cause the head-loss across the fuel assemblies. Experimental loop and test section are designed for the debris-loading head-loss tests of fuel assemblies. The goal of the testing is to study the relationship of debris, flow rates and head-loss. Pre-simulation of the blockage in fuel flow path is performed using CFD code. A simplified 2×2 grid model of rod bundles is used to compute and compare the flow conditions under varies flow velocities and blockage percentages. The results demonstrate that the pressure will sharply drop after the flow past the mixing vane of the grid, and the head-loss will increase with the increasing blockage rate and flow velocity. The results match the objective law that the velocity is high in flow channels and low in fuel rods whether the blockage is occurred or not.

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This study aims to experimentally investigate buckling failure of stainless steel tube columns under external pressure which was laterally applied on the tubes. In present study, buckling failure temperature was measured for three different tubes (denoted as Tube 1, Tube 2, and Tube 3) in a wide range of pressure conditions. The relation between the external pressure and the buckling failure temperature was obtained for each tube. The results show different behaviors of the buckling failure temperature depending on the buckling mode. For the first mode of buckling (Tube 1 and Tube 2), the buckling failure temperature increased with decrease of pressure applied but it did not change after a certain pressure condition. However, the buckling failure temperature was linearly increased with the decrease of the pressure for the second mode of buckling (Tube 3). Additionally, the effect of tube dimensions (radius to thickness ratio, R/t) on the buckling failure temperature was also examined for four pressure conditions. Almost uniform buckling failure temperature was observed for tube columns with relatively small R/t values, but the failure temperature drastically decreased with increase of R/t values. Buckling processes under external pressure were visualized by using two high speed camera. These results will help understanding the Fukushima accident, particularly the drastic depressurization in Unit 1 of the Fukushima Daiichi Nuclear Power Plants.

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The concentration of radioactive aerosols in the containment increases rapidly during the serious nuclear accidents. The common aerosol filters use mesh structures or filter papers, which can significantly increase the flow resistance in the course of aerosol deposition and retard the containment pressure relief following the nuclear accidents. In addition, filter papers replacement is also a technological problem, and the used filter papers will turn into radioactive solid wastes. This paper proposes and studies MEMS inertial impactor filter that filtrates and collects one to three microns atmospheric particles without introducing much flow resistance and replacing filter papers. The model is developed and the analysis is made with help of the calculation results.

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In the event of a nuclear power plant accident such as a core meltdown and a cooling system failure, the containment contains radioactive materials released from the reactor pressure vessel to reduce the activity of the radioactive materials and the effects of radiation in the vicinity of the plant. Since high sealing performance and high pressure resistance are required of the containment, a silicone or EPDM rubber gasket with high heat and radiation resistance is used for the sealing of the sealing boundary of the containment. In recent years, it has been shown that a large amount of steam is released into the containment in the case of a severe accident. Consequently, radiation resistance at high temperature as well as steam resistance is required of the rubber gasket placed at the sealing boundary. However, the steam resistance of silicone rubber is not necessarily as good as that of EPDM rubber. Therefore, it is necessary to evaluate the sealing characteristics of rubber gaskets in such a degrading environment in a severe accident. O. Kato et al. [1] conducted a study on the degradation status of rubber gaskets and their application limits at high temperature. However, few studies have evaluated rubber gaskets in high-temperature radiation and steam environments. In this study, we degraded silicone rubber and EPDM rubber used for the containment in the high-temperature radiation and steam environments expected to occur in a severe accident and evaluated the useful life of the rubber as a sealing material by estimating the change in its performance as a sealing material from the change in permanent compressive strain in the rubber.

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We report the study of supercritical flow of carbon dioxide in a vertical tube under non-uniform heat flux. The investigation of supercritical fluid is crucial for advanced generation IV (GIF) nuclear reactors since both supercritical carbon dioxide and supercritical water are planned to be used in those designs. The supercritical carbon dioxide is going to be used in secondary loop of liquid metal fast reactors, the supercritical water will be used as a coolant itself. Thanks to scaling parameters for fluid-to-fluid modelling at supercritical conditions based on inlet conditions approach it is possible to predict response for different supercritical fluids at different scaled conditions. Hence, the results can be used for investigation of other supercritical fluids if it is needed. Since, the heat flux in particular fuel element varies axially it is imperative to investigate the influence of non-uniform flux in this direction. Current literature provides experimental results only when uniform heat flux is applied, whereas non-uniform heat flux can show the thermal response at different stage of fuel cycle. Hence, it is possible to simulate how the flow behaves from the beginning of cycle (BOC) when the shape of the curve is a cosine distributed to the end of cycle (EOC) moment when the peaking factor is skewed to the bottom of the fuel element. The numerical investigation is performed for a 2-D axisymmetric model of a tube with use of CFD code. First, the model is validated for a case with uniform heat flux against results found in the literature. Then, non-uniform heat flux is represented with two parameter equation to describe the variation along axial direction. Sensitivity study related to influence of pressure, temperature and average heat flux in order to better understand the phenomena are conducted. Obtained results provide information about heat transfer coefficient (HTC), heat transfer deterioration (HTD) in supercritical conditions and which parameters can influence it. With better understanding of this phenomena it is possible to prevent or to avoid it, since HTD may cause overheating of the fuel elements inside reactor core and lead to an accident.

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The JAERI's Reprocessing Test Facility (JRTF) was the first engineering-scale reprocessing facility constructed in Japan. The JRTF was operated from 1968 to 1969 to reprocess spent fuels from the Japan Research Reactor No.3 (JRR-3). As a result of the operation (total 3 runs) by PUREX process, 200 g of highly purified plutonium (Pu) were extracted. In this operation, about 70 m^3 of liquid waste was generated and part of this waste, which including Pu, with relatively high radioactivity, was stored in six large tanks. After shutdown of the facility, the JRTF decommissioning program was started in 1990 to develop decommissioning technologies and to obtain experiences and data on dismantling of fuel cycle facilities. Liquid waste in the tanks was treated from 1982 to 1998. Dismantling of tanks started in 2002. The tanks were installed in narrow concrete cells and inside of the cell was high dose area. Dismantling method for the tank is important factor to decide manpower and time for dismantlement. In this paper, in-situ dismantling of the liquid waste storage tank and its preparation work are discussed.

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For the purpose of the early realization of fusion energy, construction of JT-60 Super Advanced (JT-60SA), a superconducting tokamak facility, is steadily proceeding in JAEA Naka Fusion Institute located in Naka city, Ibaraki prefecture, Japan, toward the first plasma in March 2019. This project is conducted under the Satellite Tokamak Programme of the Broader Approach Agreement and the Japanese national programme. It contributes to a wide range of fusion research and development, especially in conducting supportive researches for the ITER project to accomplish its technical targets, and conducting complementary researches to the ITER project necessary to design and construct a demonstration fusion power plant (DEMO). It also has an essential role to train domestic scientists and technicians, especially those in younger generation, who are expected to play leading roles in ITER and DEMO. JT-60SA, upgrade of JT-60U which used normal conducting coils, commands high-temperature and high-pressure deuterium plasmas in the breakeven condition for a long pulse duration (typically 100 s). It has powerful auxiliary heating and current drive tools using Neutral Beam Injection system and Electron Cyclotron Resonance Frequency system, several kinds of plasma stability control coils and flexible plasma shaping capability. JT-60SA is a powerful device for exploring and optimizing operation scenarios of ITER and DEMO. Major components of JT-60SA have been procured by EU and Japan; e.g. toroidal field coils, poloidal field coils, vacuum vessel (VV), thermal shields, power supply system, cryostat, cryogenic system and so forth. Some facilities previously used by JT-60U are reused with refurbishment and repair; e.g. central substation, motor generators, and so forth, to reduce overall construction cost. Performance of auxiliary heating and current drive facilities are also improved. Disassembly of JT-60U in the torus hall of Naka Fusion Institute was completed in October 2012. Assembly of JT-60SA started in January 2013. Cryostat base and three poloidal field coils were already installed in the tokamak hall, and seven 40-degree VV sectors as well as two 30-degree VV sectors put on the cryostat base are being welded on by one. Quench Protection Circuit, a power supply system fabricated by EU, was also delivered to Naka. On site assembly work has been set forward in a robust manner. In parallel with construction of JT-60SA, research plan of JT-60SA has been intensively discussed involving members of research communities in EU and Japan. In the eight principal research areas, research proposals from European and Japanese researchers have been elaborated and put together as "JT-60SA Research Plan", which is open to public and will be updated on a periodic basis in the future.

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Kinetic analysis in criticality accident is necessary for the estimation of the integral power and the dose. The analysis must be performed even for the geometry of several regions which are weakly coupled, i.e. each region is apart from other regions or/and moderating material exists between the regions. A kinetic analysis method for weakly coupled systems has been developed to analyze the first pulse in super prompt criticality accidents. In the method, time dependent fission probability density functions after a fission in each region are calculated by Monte Carlo method from the time between the collisions. By using the fission probability density functions, kinetic behavior of each region can be calculated. The difference of the functions between the different systems is discussed based on the difference of compositions and geometry of the systems.

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During the normal operation of the ITER tokamak, few hundred kilograms of dust containing beryllium (Be) and tungsten (W) will be produced due to the erosion of the walls of the vacuum chamber by the plasma. During a loss of coolant accident (LOCA) or a loss of vacuum accident by air ingress (LOVA), hydrogen could be produced by dust oxidation with steam. Evaluation of the risk of dust and hydrogen explosion that may lead to a loss of containment, requires studying the physical processes involved in the dust re-suspension and its distribution in the tokamak chamber. This experimental study is conducted by the Institut de Radioprotection et de Surete Nucleaire (IRSN) to simulate dust re-suspension phenomena induced by high velocity jet under low pressure conditions. Tests are conducted in a large scale facility (TOSQAN, 7 m^3) able to reproduce primary vacuum conditions (1 mbar). Optical diagnostics such as PIV technique (Particles Image Velocimetry) are implemented on the facility to provide time resolved measurements of the dust re-suspension in terms of phenomenology and velocity. This paper presents the TOSQAN facility with its configuration for studying dust re-suspension under low pressure conditions and preliminary underway experiments showing the mechanism of dust re-suspension by sonic and supersonic flows.

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Grey correlation degree calculation model which reflects the positive and negative relation of sequences is applied to calculate the relevant influence factors on the natural circulation and heat transfer of LBE(Lead Bismuth Eutectic). Results show that in the range of selected parameters: temperature difference, height difference and tube diameter show positive correlation with the velocity and heat transfer of LBE. And within a certain range, the influence of temperature difference is larger than the influence of height difference, and the influence of height difference is larger than the influence of tube diameter, but there is little difference about the numerical of grey correlation degree, it means that the impact almost equal.

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The RELAP/SCDAPSIM/MOD4.0 code is being developed as part of the international SCDAP Development Training Program (SDTP). SDTP consists of more than 90 organizations in 30 countries supporting the development of technology, software, and training materials for the nuclear industry. RELAP/SCDAPSIM/MOD4.0, which is the first version of RELAP5 completely rewritten to FORTRAN 90/95/2000 standards, has a number of unique options relative to previous versions of RELAP/SCDAPSIM including advanced numerical techniques and options, integrated uncertainty analysis, advanced graphical user interfaces, standardized interface for user supplied 3D reactor kinetics package, and advanced models and correlations for water and other fluids including Na, LiPb, PbBi, and molten salts. MOD4.0 has been used (a) for reactor simulation and analysis (LWR/HWRs, research reactors, Generation IV reactors), (b) for the design and analysis of experimental facilities and advanced fluid systems, and (c) as the general framework for advanced fluid systems model development activities. This paper describes the unique features of MOD4.0, discusses the validation and verification of the code, provides an overview of current applications of the code, and provides a more detailed discussion of the ongoing model development activities to support the design and analysis of alternative fluid systems such as the LiPb and Helium loops used in the thermal blanket modules for ITER being built in France and PbBi loops used in the MYRRHA accelerator-driven system proposed for Belgium.

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