The structural integrity of piping systems is crucial to continuous and safe operation of nuclear power plants. Across all designs, the pressure boundary and its related piping and components, form one of the many levels of defense in the continuous and safe operation of a nuclear power plant. It is therefore necessary to identify, understand, evaluate and catalogue all of the various degradation mechanisms and failures that affect various piping systems and components across all nuclear power plants (NPP's). This need was first recognized in 1994 by the Swedish Nuclear Power Inspectorate (SKI) which launched a five-year Research & Development (R&D) project to explore the viability of creating an international pipe failure database (SKI-PIPE) (Riznic, 2007). The project was considered to be very successful and in 2002, the Organization for Economic Co-operation and Development (OECD) Pipe Failure Data Exchange (OPDE) was created. OPDE was operated under the umbrella of the OECD Nuclear Energy Agency (NEA) and was created in order to produce an international database on the piping service experience applicable to commercial nuclear power plants. After the successful completion of OPDE, the OECD, as well as other nternational members, agreed to participate in OPDE's successor: the Component Operational Experience Degradation and Ageing Program (CODAP). The objective of CODAP is to collect information on all possible events related to the failure and degradation of passive metallic components in NPP's. With CODAP winding down to the completion of its first phase in December 2014, this report will focus on the conclusions and the lessons learned throughout the many years of CODAP's implementation. There are currently 14 countries participating in CODAP, many of whom are industry leaders (France, Canada, U.S.A., Germany, Japan, Korea etc.). This cooperation on an international scale provides a library of OPerational EXperience (OPEX) for all participating NPP's (Lydell et. al., 2008). CODAP also allows for the sharing of valuable information on a wide range of reactor types such as Pressurized Heavy Water Reactors (PHWR), Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR). The use of CODAP/OPEX and knowledge databases can and have resulted in a number of significant inspection changes at NPP's worldwide. For example, Canada has already utilized CODAP to address thermal stratification and piping material fatigue issues. Korea has collected and used information for piping failure events to identify sites of potential concern in their in-service inspection programs. Currently there are over 4500 recorded events on pipe failures affecting all ASME Code Classes and non-safety related piping. These events encompass all known modes of damage/degradation and their respective failure modes.

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A four-year research project started in 2012 to develop margin assessment methodologies against external hazards as well as probabilistic risk assessment (PRA) methodologies. In this project, only the decay heat removal function was taken into account assuming no loss of reactor shutdown function. The developed methodology is applied mainly for sodium-cooled fast reactors (SFRs), though it would also be applicable for light water reactors (LWRs) basically. Typical SFR heat sink is air, which is different from the heat sink in LWRs. Therefore, it is important external hazards that influence to air coolers which are located at high elevation. This project addresses extreme weathers (snow, tornado, wind and rainfall), volcanic phenomena and forest fire as representative external hazards. In this study, the external hazard evaluation, the accident sequence and the methodologies of both PRA and margin assessment are developed for each external hazard. This paper describes mainly snow margin assessment methodology development in addition to the project overview including the scope and methodologies against snow, tornado, wind, volcanic eruption and forest fire. For the snow margin assessment, the index is a combination of a snowfall speed and duration. Since snow removal can be expected during the snowfall, the developed snow margin assessment methodology is such that the margin was regarded as the snowfall duration up to the decay heat removal failure which was defined as when the snow removal rate was smaller than the snowfall speed.

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The decommissioning of a nuclear facility is a long term project, handling information which begins from the design, construction and operation. Moreover, the decommissioning project is likely to be extended because of the lack of the waste disposal site especially in Japan. In this situation, because the transfer of knowledge and education to the next generation is a crucial issue, integration and implementation of a system for knowledge management is necessary in order to solve it. We have to arrange, organize and systematize the data and information of the plant design, maintenance history, trouble events, waste management records etc. The collected data, information and records should be organized by computer support system e.g. data base system. It becomes a base of the explicit knowledge. Moreover, measures of extracting tacit knowledge from retiring employees are necessary. The experience of the retirees should be documented as much as possible through effective questionnaire or interview process. The integrated knowledge mentioned above should be used for the planning, implementation of dismantlement or education for the future generation. Finally, the total system of decommissioning knowledge management system (KMS) is proposed.

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Recently, liquid sodium containing titanium nanoparticles (LSnanop) have attracted considerable attention. In this study, suspension state of Ti nanoparticle in liquid sodium was quantum chemically evaluated. The atomic interaction between Ti nanoparticles and sodium atoms in the liquid sodium medium was investigated. There were some literatures which gained quantum chemical insight into a nanoparticle with the surrounding sodium atom. However, liquid sodium medium itself together with a Ti nanoparticle under the realistic temperature has not yet been investigated theoretically. To overcome the problem of conventional theoretical method, we applied computationally low-load Tight Binding Quantum Chemical Molecular Dynamics (TB-QCMD) calculation method to investigate the suspension state of the Ti nanoparticle in liquid sodium metal.

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The RELAP5 code is one of the extensively used thermal hydraulic codes in nuclear reactor safety analysis. However, there exists some controversy on its capability under low pressure natural circulation. To verify its ability to predict the transient behavior of natural circulation under low pressure conditions, the RELAP5/MOD3.2 code is applied for the simulation of a PCS principle test facility (PCS-PTF) of ACP1000 in China. The coolants in PCS-PTF are heated by the condensation process with steam or steam-air-He mixture gas outside the heating section and then get cooled in the upper water tank. In the present study, experiments were carried out under different mixture gas pressure including 0.2 MPa, 0.3 MPa and 0.6 MPa. Calculated results predicted by RELAP5/MOD3.2 are compared with experimental data both in single and two-phase flow conditions. The results indicate that the RELAP5/MOD3.2 code could provide reasonable simulation on main phenomenon occurred in system. For the case of condensation with pure steam, the phenomena of single-phase flow and flashing induced two-phase flow could be well predicted. For the case of condensation with the presence of non-condensable gas, the geyser oscillation frequency from calculation shows good agreement with experimental results, while the mean flow rate in Conditions 3 and 4 are lower than that of experiment by 55.9% and 33.2%, respectively, which are resulted from the underestimate of outer-tube condensation heat transfer coefficient predicted using Colburn-Hougen model in RELAP5 code.

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In the new generation of nuclear reactors design, a series of passive safety systems are adopted to provide significant improvements in plant simplification, safety and reliability. However, when applying the RELAP5 code in predicting the passive residual heat removal system (PRHRS), the capability of RELAP5 code needs further validation. In the present work, in order to validate the RELAP5 code for simulating nucleate pool boiling heat transfer outside vertical tube bundle, a test facility of pool boiling heat transfer outside vertical tube bundle was designed and built. By considering the axial natural convection and the radial gas-liquid heat exchange in large water tank, two kinds of RELAP5 calculation models are adopted in this paper. With the comparison between calculated results and experimental data, it is found that the maximum deviation between the experimental and calculated boiling heat transfer coefficient exceed 50% in the test condition, and the boiling heat transfer coefficient change tendency between them is also obviously different with the increase of heat flux.

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In order to enhance the inherent safety features of fast reactor cores, the effect of duplex fuel, which consist of the central absorber and cylindrical fuel at the peripherals, on reactivity has been studied when the absorber is mixed at the fuel melting accident of the large sodium cooled reactor. The absorber of the duplex fuel is assumed Gadolinium oxide, and the cladding of the pin is a high temperature material such as SiC, in which the melted fuel can be kept even at the fuel melting accident. The insertion reactivity is evaluated as about -6%dk/k by the Monte Carlo Code MVP for the fuel melting condition if the fuel and the absorber are molten into a homogeneous mixture in all of the fuel pins of the core. Though the reactivity is increased to about -2% dk/k if the mixture is compressed to the theoretical density in the cladding, it remained still negative compared to that of the normal operation. This result suggests that the re-criticality problem at severe accidents can be resolved with inherent characteristics of fuel pins.

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As a part of a development of the risk assessment methodologies against external hazards, a new methodology to assess forest fire hazards is being developed. Frequency and consequence of the forest fire are analyzed to obtain the hazard intensity curve and then Level 1 probabilistic safety assessment is performed to obtain the conditional core damage probability due to the challenges by the forest fire. "Heat", "flame", "smoke" and "flying object" are the challenges to a nuclear power plant. For a sodium-cooled fast reactor, a decay heat removal under accident conditions is operated with an ultimate heat sink of air, then, the challenge by "smoke" will potentially be on the air filter of the system. In this paper, numerical simulations of forest fire propagation and smoke transport were performed with sensibility studies to weather conditions, and the effect by the smoke on the air filter was quantitatively evaluated. Forest fire propagation simulations were performed using FARSITE code. A temporal increase of a forest fire spread area and a position of the frontal fireline are obtained by the simulation, and "reaction intensity" and "frontal fireline intensity" as the indexes of "heat" are obtained as well. The boundary of the fire spread area is shaped like an ellipse on the terrain, and the boundary length is increased with time and fire spread. The sensibility analyses on weather conditions of wind, temperature, and humidity were performed, and it was summarized that "forest fire spread rate" and "frontal fireline intensity" depend much on wind speed and humidity. Smoke transport simulations were performed by ALOFT-FT code where three-dimensional spatial distribution of smoke density, especially of particle matters of PM2.5 and PM10, are evaluated. The snapshot outputs, namely "reaction intensity" and "position of frontal fireline", from the sensibility studies of the FARSITE were directly utilized as the input data for ALOFT-FT, whereas it is assumed that the active fire area is along the frontal fireline and all the smoke is blown to leeward direction of the nuclear power plant site. The time-dependent change of particle matter density is utilized to evaluate a cumulative amount of particle matter at the air filter of the decay heat removal system. The total amount of particle matter at the air filter is around several hundred grams per m^2 which is well below the operational limit of the air filter of 15 kg/m^2.

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In order to research the performance of Passive Containment Cooling System (PCS) and verify the containment safety of AP/CAP series pressurized water reactor nuclear power plants (NPPs), an integral test facility for PCS (short name CERT) is designed and constructed in China. In this paper, the CERT test facility is modeled using the T-H program GASFLOW. The steel containment shell and internals, including compartments, flow paths between compartments and heat sinks are considered. The thermal hydraulic response of the facility under scaled mass and energy release of LOCA is studied.

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In the present work it is shown that physically accurate structure of dependences for heat transfer during condensation of vapor on the surface of subcooled jet can be obtained by considering the effect of surface tension at the phase boundary and the unsteady nature of the proceeding thermal hydraulic processes. Currently absent not only the quantity theory of heat transfer in film boiling strongly subcooled liquid, but also convincing qualitative explanation of the high intensity of heat transfer between the superheated solid surface and the subcooled liquid under conditions when direct contact between surface and liquid is impossible. For the case of superheated steam condensation on the surface of the jet strongly subcooled liquid heat transfer are considered similarly to the processes of heat transfer at inverted annular flow and quenching . It is known that the experimental heat transfer coefficients for subcooled water film boiling for 1-2 orders of magnitude more than the observed at the film boiling of saturated liquid. Until now, a reasonable explanation for such strong effect was not found. It is shown that taking into account the wave nature of the process possible to obtain quantitative values of heat transfer coefficients, more or less agreed with the experimental data.

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The coupling of multiple thermal-hydraulic parameters can result in complex flow instability in natural circulation system under rolling motion. A real-time thermal-hydraulic condition prediction is helpful to the operation of systems in such condition. A single hidden layer feedforward neural networks algorithm named extreme learning machine (ELM) is considered as suitable method for this application because of its extremely fast training time, good accuracy and simplicity. However, traditional ELM assumes that all the training data are ready before the training process, while the training data is received sequentially in practical forecasting of flowrate. Therefore, this paper proposes a forecasting method for flowrate under rolling motion based on on-line sequential ELM (OS-ELM), which can learn the data one by one or chunk-by-chunk. The experiment results show that the OS-ELM method can achieve a better forecasting performance than basic ELM method and still keep the advantage of fast training and simplicity.

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This paper describes some general methodological aspects of the assessment of the damage to human life and health caused by a hypothetical nuclear accident at the nuclear power plant (NPP). Probability estimation of death (due to cancer and non-cancer effects of radiation injury), disability and incapacity of individuals were made by taking into account the regulations of Ukraine. According to the assessment, the probability of death due to cancer and non-cancer effects of radiation damage to individuals who received radiation dose of 1 Sv is equal to 0.09. Probability of disability of 1, 2 or 3 group regardless of the radiation dose is 0.009, 0.0054, 0.027, respectively. Probability of temporary disability of the individual who received dose equal to 33 mSv (the level of potential exposure in a hypothetical nuclear accident at the NPP) is equal 0.16. This probability estimation of potential harm to human health and life caused by a hypothetical nuclear accident can be used for NPP in different countries using requirements of regulations in these countries. And also to estimate the amount of insurance payments due to the nuclear damage in the event of a nuclear accident at the NPP or other nuclear industry enterprise.

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AP1000 reactor is designed for 18 month of operating cycle. The core can also be used for 16/20 months of operating cycle. This study is performed to analyze and compare the neutronic parameters of typical AP1000 reactor core for 18 month and 16/20 month alternate cycle lengths. CASMO4E and SIMULATE-3 code package is used for the analysis of initial and equilibrium cores. The key reactor physics safety parameters were analyzed including power peaking factors, core radial and axial power distribution and core reactivity feedback coefficients. Moreover, the analysis of fuel depletion, fission product buildup and burnable poison behaviour with burnup is also analyzed. Full 2-D fuel assembly model in CASMO4E and full 3-D core model in SIMULATE-3 is employed to examine core performance and safety parameters. In order to evaluate the equilibrium core neutronic parameters, the equilibrium core model is attained by performing burnup analysis from initial to equilibrium cycle, where optimized transition core design is obtained so that the power peaking factors remain within designed limits. The MTC for higher concentration of critical boron concentrations is slightly positive at lower moderator temperatures. However, it remains negative at operating temperature ranges. The radial core relative power distribution indicates that low leakage capability of initial and equilibrium cores is reduced at EOC.

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A coupled neutronics/thermal-hydraulics three dimensional code system SNTA is developed for SCWR core steady state analysis by modular coupling the improved neutronics nodal methodological code and SCWR thermal-hydraulic sub-channel code. The problem of CSR1000 core is studied to verify SNTA. The results calculated by SNTA are agreed well with those by CASIR and SRAC. SNTA is more efficient than CASIR and SRAC whose neutronics modules are based on the Finite Difference Method. The numeric results show that SNTA can be well applied to SCWR core steady state analysis and core concept design.

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An experiment study was conducted on convection heat transfer process in vertical heating tube on natural circulation conditions. The investigation result shows that the convection heat transfer process on natural circulation conditions has a violent characteristic of mixed convection heat transfer. On low flow-rate conditions, the natural convection process in heating tube induced by the intense temperature difference between inner surface of heating tube and bulk of cooling water dominates the whole heat transfer process inside heating tube, and heat transfer coefficients in this region are distinctly higher than calculation values of forced circulation formula, due to the existence of secondary flows induced by natural convection process. And on high flow rate conditions, the convection heat transfer process in heating tube is strongly influenced by the heating-surface thermal buoyancy, which leads to the laminarization of the fluid in the boundary layer. Therefore, the convection heat transfer coefficients in high flow-rate region are lower than the calculation results. The primary causations leading the above-mentioned characteristics of convection heat transfer are the relatively high heating power compared with low natural circulation flow rate and the self-adaption coupling between circulations parameters. Geometrical structure of heating tube also affects convection heat transfer significantly, namely the convection heat transfer coefficients of natural circulation decrease with the diminution of the tube caliber. Finally, an empirical correlation for calculation of heat transfer coefficients in heating tubes on natural circulation conditions is putted forward, by using the non-dimensional parameter Gr/Re^2 to modifier the Gnielinski Formula. The fitting result shows good agreements with experiment results.

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Binary droplet's collision is one important phenomenon, which exists in the two phase flow within some nuclear power plant equipments, such as the steam generator, the steam separator and the vane dryer. Numerical simulation was carried out by using VOF method to study equal-sized binary saturated water droplet's head-on collision in high pressure steam. Two typical collision processes, coalescence and reflexive separation were investigated. Droplets' shapes and internal velocity field during the collision process were observed. The transformation between kinetic energy and surface energy, and the energy dissipation were analyzed to achieve better knowledge about the collision process. By simulating equal-sized head-on collisions with various We number, critical We between coalescence and reflexive separation was found and compared with Qian and Law's model. For collisions at 6.75 MPa (i.e. steam generator and steam separator's operating pressure), Qian and Law's model was revised to achieve better prediction.

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In this paper, the time-dependent dynamic behavior of Molten Salt Reactor (MSR) is investigated by analyzing the dynamic response induced by a localized perturbation. Due to the small effect of the perturbation on the radial direction, the MSR is simplified as a one-dimensional system. To obtain more realistic case, the thermal feedback effect is taken into account. The theoretical models are established with the diffusion equations and the conservation equations for mass and energy. The group constants for various temperatures are processed with a HELIOS code to build the thermal feedback manner. Equations for fluctuations induced by perturbations are derived with linear perturbation theory. All the static and dynamic equations are solved numerically. The main conclusions are as follows. First, the assumption of a homogeneous system overestimates the effect of fuel circulation on the neutron flux distribution. Next, linear perturbation theory works well in large scope and can be regarded as an effective method to study the behavior of reactor noise. In addition, as the fuel velocity increases or the core height decreases, the reactor period becomes shorter and MSR quickly reaches the new steady state, indicating the stronger self-stabilization ability. Moreover, the Doppler reactivity feedback effect is dominated due to the absence of the graphite.

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This paper focus on the Pellet-Cladding Interaction (PCI) evaluating by a self-organizing Radial Basis Function Neural Network (RBFNN) during the operation of Zircaloy clad fuel in the Water cooled Reactors. Using the neural network, which built through the analyzing of the existing data but the physical process of PCI, is a suitable way to reduce the calculation complexity. In this sense, the stage of PCI can be evaluated online. A self-organized RBFNN is used, which can vary its structure dynamically in order to maintain the prediction accuracy. The hidden neurons in the RBF neural network can be added or removed online based on the neuron activity and mutual information, to achieve the appropriate network complexity and maintain overall computational efficiency. The PCI experiment data from literatures is used to test this method, and the results demonstrate its effectiveness.

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Propagation characteristics of hydrogen-air deflagration need to be understood for an accurate risk assessment. Especially, flame propagation velocity is most important factor. Propagation velocity of spherical flame has been estimated based upon the propagation velocity of flat flame; however, existing method is not enough to estimate the actual propagation velocity because cellular flame front is formed in hydrogen-air premixed flame. In this study, we investigated the propagation characteristics of hydrogen-air deflagration in explosion tests using a closed chamber which has large windows, 300mm in diameter. Explosion tests were conducted at atmospheric pressure and room temperature and in the range of equivalent ratio from 0.3 to 1.0. By using Schlieren photography, flame propagation phenomena were observed, and flame radius and flame propagation velocity were measured. These results showed that the flame propagation velocity of hydrogen-air mixture is accelerated owing to developing of cellular flame in a huge space. Based upon these results, an estimation method of flame propagation characteristics was proposed.

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As the wide adoption of the Digital Instrument and Control system (D-I&C) for Nuclear Power Plants (NPPs), the maintenance of reliability and the safety of the D-I&C is becoming more and more concerned due to the inherent complexity introduced by the D-I&C. Wherein the main difficulty to evaluate the reliability and safety of the D-I&C lies in how to model the system including both the hardware and software. The main objective of this paper is to conduct on a feasible study to model the D-I&C by taking notice of the advantage of the Colored Petri Net (CPN). A function-centered modeling approach is proposed for formal modeling and analysis of D-I&C by using CPN. The proposed model uses the hierarchical modeling capability of CPN, by which different levels of abstraction can be treated. Therefore, the proposed CPN model represents the dynamic behavior of a large scale D-I&C system used in NPPs. Using the proposed CPN model builder, the analyst can choose the specific level of abstraction for modeling and verifying the D-I&C system design. In the paper, a digital main feed water control system of a conventional Pressurized Water Reactor (PWR) plant is modeled as the example of the whole D-I&C system.

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To improve the thermal efficiency of intermediate heat exchanger (IHX) in high temperature gas-cooled reactor, the influence of artificial roughness on heat transfer enhancement is simulated by CFD software. From the aspects of flow distribution, axial velocity, transverse velocity, turbulence kinetic energy, temperature field and wall heat transfer coefficient, we made a preliminary research comparing the smooth tube and rough tube. The results showed that artificial roughness would destroy the adhesive layer of turbulent boundary layer, and the roughness would enhance lateral disturbance. Though the efficiency of heat convection was improved, the frictional resistance was also increased because the decrease of axial velocity. As a result, the research can provide necessary theoretical support to design a rough tube having the best heat transfer efficiency.

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After the Fukushima severe accident, nuclear power development has been in stagnation in all over the world. The Chinese nuclear industry has a slowdown on new NPP construction. As a result, high level technique on safety and effective communication are required. For nuclear power engineering company with EPC mode, high quality on training and technical communication is the principal investment in order to achieve better service on engineering design, environmental impact assessment, environmental engineering design, and equipment supervision and so on. EPC mode requires wide range knowledge on almost every field related to nuclear on nuclear power engineering. In this paper, the author investigated the case of the only nuclear power engineering EPC company (CNPE) in China and present an overview on its training and technical communication both domestic and abroad. Basically, there are 4 main branches of training. The internal training focuses on specifically task (both management and technique), such as HSE training, QC training and quality and safety training. Long term education in the university is organized by cooperated mechanism. Code and platform training is partly carried out by international organization or company, and the experienced engineers coach makes up the other part. The communication is a large part since the EPC mode needs the information and requirements from the NPP entity, authority, and the other institutes, international organizations (like IAEA, NINE, IRSN, OECD, NRC and CEA etc.) and sometimes the public. The overview of the training and communication of the EPC company prevails the outline of its advantage on domestic communication and disadvantage on international technical communication. The paper can be a tool on the soft strength construction of company under EPC mode to broaden its business like consultation and training. Some advice is given by the author on the consultation and global communication in the future.

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This study was conducted to investigate the transient heat transfer process between the solid surface and the coolant (helium gas) in Very High Temperature Reactor (VHTR). Forced convection transient heat transfer for helium gas flowing over a twisted plate with different length was experimentally and theoretically studied. The heat generation rate of the twisted plate was increased with a function of Q^^.= Q_0 exp(t/τ)(where t is time, τ is period). Experiment was carried out at various periods ranged from 35 ms to 14 s and gas temperature of 303 K under 500 kPa. The flow velocities ranged from 4 m/s to 10 m/s. Platinum plates with a thickness of 0.1 mm and width of 4 mm were used as the test heaters. The plates were twisted with the same helical pitch of 20 mm, and length of 26.8 mm, 67.8 mm and 106.4 mm (pitch numbers of 1, 3 and 5), respectively. Based on the experimental data, it was found that the average heat transfer coefficient approaches the quasi-steady-state value when the dimensionless period τ* (τ* = _τU/L, U is flow velocity, and L is effective length) is larger than about 100 and it becomes higher when τ* is small. The heat transfer coefficient decreases with the increase of twisted plate length under the same period of heat generation rate. According to the experimental data, the distribution for heat transfer coefficient along the heater is nonlinear. Numerical simulation results were obtained for average surface temperature difference, heat flux and heat transfer coefficient of the twisted plates with different length and showed reasonable agreement with experimental data. Based on the numerical simulation, mechanism of local heat transfer coefficient distribution was clarified.

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Extended surface on heat transfer is the most method adopted. To acknowledge the key factor in heat transfer, the mechanism of laminar convective heat transfer enhancement has been experimentally investigated by measuring heat transfer data in spirally integral bent fin tubes at various Graetz numbers. To maintain laminar flow status, lube oil is used as the working fluid and water as the coolant. Graetz number ranged from 50 to 400. For the smooth annular tube, the result of the analysis is different from Sieder-Tate equation as expected. The influence of different cross sections using hydraulic diameter is no longer suitable for heat transfer in annular pipe. The effects of fin height and pitch on heat transfer and friction factor are discussed in detail. The experimental results indicate that the effect of fin height appears to be more crucial on heat transfer and high fins are not recommended at low Graetz number. There is an inflection point along the curve of thermohydraulic performance which is proved a maximum point and tends to be affected by geometric parameters. A novel heat transfer formula is proposed based upon collecting data, above geometry parameters are taken into account.

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In the present paper, we verify the effectiveness of the two-relaxation-time (TRT) collision operator in reducing boundary slip of temperature computed by the immersed boundary-thermal lattice Boltzmann method (IB-TLBM). In the linear collision operator of the TRT, we decompose the distribution function into symmetric and antisymmetric components and define the relaxation parameters for each part. The Chapman-Enskog expansion indicates that one relaxation time for the antisymmetric component is related to the thermal conductivity. We derive the theoretical relation between a temperature slip at the boundary and reveal that the relaxation time for the symmetric part controls the temperature at the boundary and boundary slip of temperature computed by the IB-TLBM. We apply the IB-TLBM based on the implicit correction method with two relaxation times for the natural convection in a square enclosure containing a circular cylinder. The streamline, isotherms, and average Nusselt number calculated by the proposed method agree well with those of previous numerical studies.

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A large amount of hydrogen is generated in the event of a severe accident in a light water reactor. It is difficult to reduce the hydrogen concentration in the low oxygen concentration environment in a Boiling Water Reactor (BWR) condition. In order to deplete hydrogen under such conditions, we have been studying the application of metal oxides as a source of oxygen atoms. The reaction of hydrogen and a metal oxide produces steam, and the metal oxide changes to metal. We considered that this newly formed solid metal inhibits a rapid reaction between the hydrogen and metal oxide by increasing the hydrogen diffusion distance to the metal oxide. We conducted some experiments to select effective metal oxides. The weight loss of the metal oxide in the experiment was used as a measure of the efficacy of hydrogen treatment. From the results, CuO and MnO2 were selected as promising metal oxides. Furthermore, we observed a change in their crystalline structure with in-situ X-ray diffraction analysis (XRD) at a hydrogen concentration of 3% and a temperature of 523 K. CuO peaks were observed in the XRD profile in the initial stage of the reaction, and then Cu peaks appeared after 10 minutes, from which it was verified that the CuO reacted with the hydrogen. We confirmed that the Cu formed by the reaction of CuO with hydrogen acted as an inhibitor to a rapid reaction. We also conducted experiments to clarify the effects of the hydrogen and water vapor concentrations and the temperature on the reaction rate of hydrogen and CuO. It was confirmed that the reaction rate increased with increasing hydrogen concentration and temperature. On the other hand, the reaction rate of hydrogen decreased with increasing vapor concentration.

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Corrosion behavior of 1.4970 (15-15 Ti) austenitic steel in Pb-Bi eutectic (LBE) was investigated depending on the structural state of material. Samples after solution annealing and subsequent 20 and 40 % reduction by cold rolling, as well as sample in as-received state were exposed to static LBE at 550°C for 1100 and 2200 h. Concentration of oxygen during tests ranged from 10-8 to 5×10^<-11> mass%O. After 1100 h, samples revealed selective leaching of Ni and Cr resulting in formation of a ferrite layer penetrated by LBE. The maximum depth of corrosion attack increases in the following sequence: solution annealed state (〜12 μm); 20% reduction (〜110 μm); as-received state (〜115 μm) and 40% reduction (〜170 μm). Similar corrosion trend was observed after 2200 h with maximum depth of attack increasing gradually from solution annealed sample (〜220 μm) to 20% cold working (〜450 μm) and then to 40% (〜620 μm). However, maximum corrosion depth of sample in as-received state does not exceed 〜60 μm. Effect of cold working on the corrosion response of steels is discussed taking into account existing literature data.

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Subcooled flow boiling for low pressure conditions is important in nuclear reactors. Many studies are concerned with safety analyses of safety systems operating at low pressure, such as the external reactor vessel cooling (ERVC) system which is widely adopted as a part of in-vessel retention (IVR) in severe accident management strategies. The best-estimate thermal-hydraulic computer code ATHELT (Analysis of Thermal-Hydraulics of Leaks and Transients) is developed by the GRS (Gesellschaft fur Anlagen-und Reaktorsicherheit). The set of available correlations in the ATHLET code is mostly based on experiments at high pressure conditions. So, the use of some of these correlations may be questionable. The present study identifies the most important influential parameters in the ATHLET code which affect the void fraction course at low pressure conditions.

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The zirconium alloys are used in nuclear engineering for decades, as a fuel cladding or other parts of the fuel assembly. It is an extraordinary material thanks to its good mechanical and neutronic properties. Unfortunately due to the high temperature oxidation it is necessary to protect parts of reactor against the cracking of material which could be severe for whole reactor. One of the promising protecting material is a zirconium dioxide which is forming on a pure zirconium in contact with air. But this dioxide, zirconia, is not stable because of destabilization and transformation of phases upon the cooling. To stabilize zirconia many other oxides may be added (MgO, Y_2O_3, CeO_2, etc.). Especially the yttrium stabilized zirconia (YSZ) shown significant advantages. In order to enhance the stability of YSZ beyond 1200℃, the further doping of system with other elements is required. So far the Ta_2O_5 addition, creating TaYSZ, is the best candidate with the best hot corrosion resistance properties, so the most promising material for protection of fuel cladding against corrosion degradation.

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Independent deterministic analyses of system response and severe accident progression after a station blackout scenario for CANDU single and multi-unit reactors have unveiled a number of design vulnerabilities that cause uncontrolled pressure boundary ruptures; premature expulsion of coolant from main loops and from the moderator heat sink; direct exposure of core debris and fission product releases to the containment; thermo-mechanical failure of the thin shell Calandria vessel welds; accelerated production of hydrogen with containment boundary failures by steaming as well as by sparsely populated PARS units potentially exposed to high concentration deuterium/hydrogen they are unable to adequately mitigate. A number of design enhancements can however be undertaken to minimize risk from a severe accident by eliminating or avoiding some of the undesirable system responses. Engineered design measures to also avoid some the undesirable accident progression paths have been summarized in the paper. The Canadian regulatory body CNSC prepared a number of design enhancement requests as part of post Fukushima reviews but then accepted measures and submissions that do not meet the public expectations for risk reduction after Fukushima on all counts. Without thoughtful and timely design changes, consequences of a severe core damage in a CANDU reactor can pose risks that are still unacceptable especially after the hype that surrounded the utility 'Stress Tests' and regulatory 'Action Items' that followed Fukushima disaster in 2011. It is stressed that operating CANDU and other similar PHWRs in India are unique; have an enviable safety record; have maintained very high operational loads and the observations in this paper relate to requirements for mitigating a range of potential severe accidents whose probability has been demonstrated to be relatively high given that less than 15000 reactor years of operation have resulted in 3 severe accidents that have caused more off-site damage than ever anticipated. That is something the founding fathers did not consider within the design basis and the current regulating bodies have a difficult time wrapping their head around. Utilities that operate the plants have shown no visible leadership in self assessments; have rejected any attempts to enlighten them and the risk posed by continued operation of these plants can only be minimized by an open discussion of the totality of issues as attempted in this paper.

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Severe accident progression in a PHWR is strongly influenced by the large variability that exists amongst various fuel channels and fuel entities within channels. A best effort prediction necessitates consideration of all risk sensitive phenomena and processes in a detail that reduces uncertainties in rates as well as magnitudes of source terms. A review of available computational codes points to a need of a much greater than hitherto undertaken, detailed modelling of the horizontal PHWR reactor channels, fuel bundles, end fittings and feeders with an advanced, more reactor geometry specific consideration of solid debris behaviour in the Calandria vessel. A new computer code ROSHNI improves upon existing computational aides firstly in detail in which the reactor is modelled and further in its fuller consideration of important PHWR related severe accident progression pathways and phenomena. The paper details the code modelling approach used in capturing the important system response parameters. It is expected that the source terms of Deuterium gas and fission products can now be evaluated with lesser uncertainty and with greater degree of control available to the analysts. Users are able to perform parametric and uncertainty analyses with greater ease. Initial modelling of progression of a severe core damage accident is being targeted for a CANDU 6 reactor. However, the code structure is generalized to extend the ability to other PHWR designs including multi-unit CANDU stations and Indian PHWRs. After Fukushima, serious concerns have been raised about severe accident mitigation capabilities of all operating nuclear power plants including CANDU reactors whose design concept presents specific vulnerabilities not common with other reactor types. While operating organizations and their supportive regulating bodies are cautious, to the point of lethargy, about design upgrades to operating reactors, only a best effort modelling of realistic accident progression pathways can empower them with the information required to justify the required risk reduction upgrades to reactors that were never designed with severe accidents within their design basis. Reduction of uncertainties by better modelling can also assist in the development of more effective severe accident management options and better operator training as the lessons learnt from Fukushima must be taken seriously. Otherwise decisions on hardware upgrades will remain handicapped by larger than necessary uncertainties inherent in the available but obsolete PHWR reactor severe accident modelling techniques. This paper concentrates on the rationale for development of a new code; its overall modelling approach and details of reactor core modelling. In summary, the code models transient behaviour of each and all fuel channels; models all individual fuel bundles with each represented by 16 fuel and Zircaloy rings; computes the response of the associated end fittings and carbon steel feeders; considers fuel heatup during staggered boiloff in channels followed by differentiated dry heatup of each fuel channel in steam and deuterium with consideration of moderator depletion. In all, fuel temperatures are evaluated at about 80,000 locations with separate consideration of thermal behaviour of 380 pairs of end fittings and the same number of feeders whose thermo-chemical response to severe accident conditions has a potential of generating additional combustible gas, not previously evaluated. It considers disassembly and thermo-chemical behaviour of individual fuel bundles to multiple fields of debris first suspended over underlying intact channels and gradually relocated to the Calandria vessel with consideration of impact of about 100 in-core devices. It also considers effect of failure of the Calandria vessel under thermomechanical loads and evaluates energy and fission product source terms for containment response. Deuterium gas source

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High temperature gas-cooled reactor (HTGR) with direct gas turbine cycle has the potential advantages of high cycle efficiency. It uses the high temperature gas generated by the reactor core to work directly, breaking the limitation of the low gas temperature of Rankine cycles in traditional reactors. In most former studies, the working medium used in the cycle is Helium (He), and there are few researches about the influence of mixed working medium on the cycle efficiency. In this paper, the thermal cycle efficiency is calculated under the condition of different mixing ratio, to analyze the influence from different parameters such as compression ratio. The results show that the optimum molar fraction of CO_2 can be made reasonably through considering all the influence factors, to maximize the cycle efficiency. And increasing the ratio of CO_2 and Helium can significantly improve the dropping trend of the cycle efficiency under the high compression ratio. In the meantime, however, the cycle pressure loss caused by the increase of CO_2 should be considered as well.

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The critical heat flux (CHF) limit is a great concern by many in the field of heat removal technology through pool boiling system. External cooling of pressure reactor vessel by in-vessel retention (IVR) method will involve boiling process in order to remove decay heat from the molten core through the lower head of the vessel. Increasing CHF could give extra safety margin for nuclear power plant to operate. Many researchers have shown that CHF is significantly enhanced by nanofluids compared to pure water. Nanoparticle deposited on heated surface improves surface wettability in which sustain more liquid to heat transfer surface. Therefore, dry out regions is delayed and further CHF enhancement is observed. On the other hand, surface modification by attaching honeycomb porous plate on heated surface have shown CHF enhancement approximately twice in comparison with plain surface. This is due to automatic liquid supply by capillary action and separation of liquid and vapor path contributed by the honeycomb structure. In the present study, the effects of surface modification by water-based nanofluid concentrations and honeycomb porous plate were investigated experimentally in saturated pool boiling at atmospheric pressure. Experimental result for combination of honeycomb porous plate and water-based nanofluids concentration of 4.0 g/L (0.110% by volume) shows the most enhanced CHF compared to other surface modification.

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Recently, the thermal emission issued from many components and devices in engineering fields has exceeded 10 MW/m^2. The common cooling methods have not met the requirement. Microbubble emission boiling (MEB) is expected for being employed to solve the problem due to its high heat transfer performance. A high-speed video camera (Fastcam SA5) was employed to study bubble dynamic behavior in MEB visually for improving the understanding of it. Experimental results showed that the bubble behavior of MEB differed from both nucleate boiling and film boiling. In the regime of MEB, a large irregular bubble would form on the heating surface and break up rapidly afterwards, but not departure from the heating surface. The period of bubble ebullition was shorter and the change rate of bubble radius was higher in MEB than those in other boiling modes. A dimensionless analysis showed that the wall superheat and heat flux had great impact on the variation of dimensionless bubble radius in MEB, while had slight influence on that in other boiling modes. The inertia effects increasingly dominated the condensation and collapse process with the increase in wall superheat and heat flux for MEB.

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The purpose of this study is to derive a CCFL (counter-current flow limitation) correlation and its uncertainty for steam generator (SG) U-tubes in a pressurized water reactor. Pressure and temperature are very high in actual U-tubes. Hence, in this paper, we evaluated effects of pressure and temperature on CCFL characteristics using numerical simulations. Results computed with the k-ω SST turbulence model gave a trend opposite to the ROSA-IV/LSTF data in the pressure range of 1.0-7.0 MPa, and the computed falling water flow rates decreased as pressure increased. Because computations with the k-ω SST were unstable at lower pressures than 1.0 MPa, the laminar flow model was used even though it significantly overestimated falling water flow rates. The results showed that: (1) the flooding under steam-water conditions was mitigated more than that under air-water conditions; (2) the falling water flow rate had a maximum value at about 1.0 MPa; and (3) the laminar flow model resulted in an opposite trend to the ROSA-IV/LSTF data in the pressure range of 1.0-7.0 MPa, as the k-ω SST turbulence model did. Thus, we concluded that accurate measurements should be made in a wide range of pressures using a single vertical pipe in order to confirm effects of fluid properties on CCFL.

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The most important kinetic parameters are effective delayed neutron fraction (β_<eff>), mean neutron generation time (Λ) and reactivity coefficient (α), which play a key role in nuclear reactor control and safety analysis. In this work, these kinetic parameters for the accelerator driven system (ADS) cores are calculated by using Monte Carlo method. The calculated values of βeff and Λ agree well with the reference data. In addition, the effect of minor actinides (MA) on kinetic parameters is investigated. Both uranium and plutonium mixed oxides (MOX) fuel and (U,Pu,MA)O_2 fuel are taken into account to load in the different cores. The results show that the values of β_<eff> and Λ decreases with increasing the content of MA in the (U,Pu,MA)O_2 fuel. The results also demonstrate that prompt negative feedback effects, including Doppler feedback effect and Coolant feedback effect, would be weakened after loading a large amount of MA into the core.

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A large scale pulsation that oscillates in a narrow gap that connects two rectangular channels is observed. The phenomenon is similar to the mixing of cooling fluid in rod bundles of power plants from one sub-channel to another through the gaps between the rods. The fluid mixing in rod bundles has been studied from 1960s to the present time to predict the temperature distributions of the rods and the cooling fluid in a nuclear reactor that were used when a power plant was designed for its safety assessment. The nature of the fluid mixing has been accounted for by turbulence and secondary flows, which also occur in the turbulent flow field. Recently, hot wire measurement revealed that the fluid mixing was periodic and coupled with Reynolds number and the gap geometry. In addition, the visualization of the flow in a rectangular channel containing a cylindrical rod showed the existence of a vortex in the narrow gap between the rod and the channel and it also accounted for the fluid mixing. The amplitude of the visualized vortex seems to become larger as it flows downstream, however the cause of the amplitude increase has not discussed yet. In this study, a numerical analysis has been performed in a composed channel consisting of two rectangular channels which are connected by a narrow gap near a wall. An algebraic Reynolds stress model is used to predict Reynolds stresses in the channel precisely. Calculated results for axial mean velocity, streamwise and gap-parallel turbulence intensities are compared to the experimental data to validate the proposed numerical model. The proposed model can predict characteristic isocontours of axial mean velocity, streamwise turbulence intensity and gap-parallel turbulent shear stress, although quantitative differences are observed in several predictions from experimental data. Calculated planar turbulent shear stress and secondary flows near the gap are shown to compare with the result by large eddy simulation. In the gap region, velocity vectors show the large scale pulsation that is also observed in the experiment. Streamwise variations of axial mean velocity and axial velocity fluctuation, turbulence intensity and turbulent shear stresses of the gap center are discussed as an indication of the leading point of the large scale pulsation.

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According to a guideline by the regulatory body, materials, which irradiated by neutron less than 6.25μSv/h, can be treated as "Non-Radioactive Waste (NR)" (Advisory Committee 2007). And we call a boundary of radioactive waste and Non-radioactive waste "NR Boundary". We improved an estimation procedure of NR boundary. In our previous works (Tanaka 2014), we had estimated NR Boundary around Main Steam line (MS) of the BWR using "Sn and Sn Coupling technique". In the case of Boiling Water Reactor (BWR) MARK I type power plant, NR Boundary is in the vicinity of an area where MS is situated on. This Coupling technique provides a means of coupling fluxes calculated with two-dimensional (2D) Sn code "DORT" (ORNL 1998) to three-dimensional (3D) Sn code "TORT" (ORNL 1998) through construction of external sources for TORT. We had set neutron-flux on the inner surface of Biological Shielding Wall (BSW), which had been calculated using DORT in the previous works, as external neutron source for this TORT calculation. We had also measured neutron-flux using activation foils at 9 locations around MS. Comparing the calculation results and the measured flux at ten locations, we had found that underestimation had occurred at some locations around MS. We supposed that one of cause of this underestimation is that Sn method tends to calculate neutron diffusion more excessive under the local phenomena in neutron-flux distribution like streaming occurs. To improve this underestimation, we decided to apply MCNP-5 (Monte Carlo N-Particle transport code) (LANL 2003) to the calculation around MS because MCNP-5 simulates the local phenomena such as streaming better than 3D Sn code. We also performed TORT calculation with alternative directional mesh such as biased directional mesh, which have mesh arrangement that leaned along the streaming. We have implemented NR boundary estimation around MS in TORT calculation and MCNP-5 calculation. We compared both calculation results with the measured fluxes.

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Natural convection heat transfer from vertical rod bundles in liquid sodium was numerically analyzed for three types of the bundle geometry (two parallel, equilateral triangle and equilateral square arrays). The unsteady laminar three dimensional basic equations for natural convection heat transfer caused by a step heat flux were numerically solved until the solution reaches a steady-state. The PHOENICS code was used for the calculation considering the temperature dependence of thermo-physical properties concerned. The 2 to 4 test rods for diameter (D=7.6 mm), heated length (L=200 mm) and L/d (=26.32) were used in this work. The surface heat fluxes for each cylinder were equally given for a modified Rayleigh number, R_<f,L>, ranging from 3.06×10^4 to 3.14×10^7 (q=1×10^4〜7×10^6 W/m^2) in liquid temperature (T_L=673.15 K). The values of S/D for the rod bundle were ranged from 1.4 to 3 on each bundle geometry. The spatial distributions of local and average Nusselt numbers, Nu_<θ,z> and (Nu_<av,B>)_N, on vertical rods of a bundle were clarified. The average values of Nusselt number, (Nu_<av,B>)_<N,S/D>, for three types of the bundle geometry with various values of S/D were calculated to examine the effect of the bundle geometry, S/D and R_<f,L> on heat transfer. The bundle geometry for the higher (Nu_<av,B>)_N value under the condition of S/D=constant was examined. The correlation for (N_<uav,B>)_<N,S/D> for three types of bundle geometry above mentioned including the effects of R_<f,L> and S/D were developed. The correlations can describe the theoretical values of (Nu_<av,B>)_<N,S/D> for three types of the bundle geometry for S/D ranging from 1.4 to 3 within-7.44 to 10.73 % difference.

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PSA (Probabilistic Safety Assessment) method has been widely used in nuclear power plant safety designs. Thermal hydraulic analysis is a basic and important issue in PSA. During the process of thermal hydraulic analysis, the response of reactor coolant system and performance of the reactor core can be simulated, so that the safety functions can be assured. Also, safety equipment demands can be defined, including the success criterion of useful equipment and time window for operator to action. The damage state of reactor core will be judged, too. Nuclear power plant operates under certain conditions, such as full power condition, hot standby, hot shutdown, cold shutdown. LPSD condition includes nuclear power plant conditions when NPP operate under low power and shutdown conditions. LPSD is divided to several plant operational state (POS), based on important parameters of nuclear power plant, such as, core power and decay heat, water level of primary circus, temperature and pressure of reactor coolant system, state of the containment, and etc. Under POSC condition, the plant is subcritical; the primary pressure is between 2.4MPa and 13.9MPa; the average coolant temperature is between 160℃ and 284℃. MBLOCA (middle break loss-of-coolant accident) under a certain LPSD condition of POSC for typical three loop nuclear power plant is analyzed in this article. The break equivalent diameter of a MBLOCA may be between 50.8mm and 152.4mm. Thus, the spectrum of MBLOCA is calculated firstly and a limiting break area is selected to envelope all break sizes. Then success criterions of relevant systems and time windows for operators were calculated and selected. The results show that, after MBLOCA under POSC condition takes place, the operator needs to open one train of HHSI (high pressure safety injection system) or open one GCT-a (turbine bypass to atmosphere) valve on intact loop to decrease reactor coolant system's pressure and temperature within 1800s, so that the reactor core will be safe. The results could be used for Level 1 PSA.

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It is important to investigate the molten jet behavior in fuel coolant interaction (FCI) during reactor severe accidents, which affects the subsequent phenomena, such as vapor explosion in light water reactor (PWR) or the molten core coolability in fast breeder reactor (FBR). The focus of the present study is placed on the numerical simulation of hydrodynamic behavior of molten jet including the deformation and surface waves in nonboiling conditions in order to isolate the jet breakup phenomenon from the complex interactions caused by coolant boiling. A multi-phase code with the Volume of Fluid Method (VOF) is used. Experimental data (Iwasawa Y, et al., 2014) is used as a benchmark to examine the capability of current approach. In this paper, effect of inject velocity and inject diameter on jet breakup length and breakup time are quantitatively analyzed. Effect of solidification is also discussed.

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A prototype 3D simulation system for reactor decommissioning based on the software of DELMIA, and VIRTOOLS is built. This system is composed of the database, real simulating scene, decommissioning process simulating, decommissioning data analyzing, project management, evaluation and expert assistant decision-making and other sub-systems. On the basis of this simulation system, 3D radiation field computation algorithm is developed based on the point-kernel method and the computing errors of the radiation level at key points are less than ten times of the measuring data. The computation software of 3D radiation field is programmed based on the software of VS and SQL server. This software is built in the system and dynamic calculation and data refresh of 3D radiation field are realized. The mapping formula between color and radiation data is proposed and visual display of 3D radiation field is realized. The calculating method of person exposure dose is worked out based on the decommissioning procedures by acquiring related parameters and real-time display of the worker's dose is realized. Virtual cutting algorithm is developed by segmentation and reorganization of three digital models. Cutting process can be simulated for any pipes at any position.

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The advanced mechanical shim (MSHIM) core control strategy is implemented in the AP1000 reactor for core reactivity and axial power distribution control simultaneously. In this paper, the Particle Swarm Optimization (PSO) algorithm has been used to optimize the MSHIM control system parameters for AP1000 reactor. First, a nodal core model is adopted to describe the dynamic behavior of the reactor core. Then, a simulation platform of the AP1000 reactor including the nodal core model and the MSHIM control strategy is developed in Matlab/Simulink. Based on the simulation platform, the typical 10% step load decrease from 100% to 90% full power is simulated to generate necessary data and the PSO algorithm is implemented successfully for the optimization of MSHIM control system parameters. It has been demonstrated by the calculation results that the optimized MSHIM control system parameters can improve the reactor power control capability without compromising axial offset control.

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Alpha radiations from nuclear waste produced at nuclear power plants is checked and classified according to the radiation level. In the case of conventional detectors that use ZnS (Ag) scintillators, detectors are placed almost in close contact to the nuclear waste because of the short range of alpha particles in the air. Those ZnS detectors have nearly 100-% detection efficiencies to alpha radiation, however, they are only capable of measuring objects with flat surfaces. We have developed alpha-detection systems that detect alpha-induced UV lights from some distance away. To operate the systems in the high-gamma-ray environment, we developed Alpha Camera that improved UV sensitivity while reducing the gamma-ray background drastically. We have carried out demonstrations of Alpha Camera where an Am-241 alpha source of 0.2 kBq/100 cm^2 was imaged from 1-m away. Alpha Camera successfully identified a 0.2-kBq/100 cm^2 alpha emitter with a 60-sec measurement.

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In-Vessel Retention (IVR) is a method to maintain molten debris in a reactor vessel (RV) by RV outer surface cooling. Structural integrity of RV and cooling capacity on RV outer surface are important to verify IVR strategy. Critical Heat Flux (CHF) data is necessary to estimate cooling capacity on the RV outer surface. And there are some CHF data to estimate cooling capacity on the RV outer surface. However, these data were obtained for specific plants. Thus, the objective of this study is developing a CHF correlation for various PWR plants. The objectives of this paper are developing test equipment and testing plan for the CHF correlation. Firstly, plant conditions during severe accidents were organized. Then, ranges of testing parameters were estimated with the plant conditions. And specifications of the test equipment were set to cover the range of parameters. Secondly, testing cases were set based on design of experiments. The test cases are suitable to develop experimental correlations.

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With excellent capability for heat removing and compression resistance, condensers with horizontal in-tubes condensation have been used in some current designs for nuclear plant safety systems, especially for passive containment cooling system (PCCS). To confirm the thermal-hydraulic behavior of the horizontal condensers in PCCS when LOCA happens, the study on condensation of steam in the presence of non-condensable gases in tubes is prior to be investigated. An experimental investigation on forced convection condensation of steam/air mixtures in a horizontal tube was presented. The experimental data was obtained under the inlet mixtures velocity ranging from 17.27-36.88m/s, inlet noncondensable gas mass fraction ranging from 5%-32%, total pressure ranging from 0.15-0.4MPa and local inner wall subcooling ranging from 20.5-78.7K. The influences of inlet air mass fraction, velocity and pressure of the bulk mixture gases and inner wall subcooling on local condensation heat transfer were analyzed. A correlation for evaluating the local condensation heat transfer coefficient along the tube has been developed which shows a great agreement with the experiment results with an error of ±20%.

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Li_4SiO_4 (lithium orthosilicate) is a promising ceramic tritium breeder for the future D-T fusion reactor, which is also considered as a candidate material for Helium Cooled Pebble Bed Test Blanket Module (HCPB TBM) of China. Simple and efficient microwave-induced solution combustion synthesis (MSCS) approach was adopted to prepare Li_4SiO_4 powder by using lithium nitrate and tetraethyl orthosilicate as raw materials. In MSCS, microwave power is a key factor, which controls the combustion mode, flame temperature, phase composition, and microstructure of the product. The synthesized Li_4SiO_4 particles are homogeneous distributed with an average crystalline size of 20 nm when the microwave power is 1100 W. The Li_4SiO_4 powder shows comparable sinterability to the conventional solid state reaction, which reaches 89.2% of theoretical density at 1173 K with a small grain size of 2 μm.

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In 2006, the author published his research on characteristics of boron co-deposition with crud experienced in AOA and crud deposition anomaly (CDA). These anomalous crud deposition phenomena are contrary to the general theory of mineralogy where hydrothermal growth of crystals was widely used and therefore, the crud should not deposit on fuels where the reactor water is at the highest temperature. Significant quantities of meta-ZrO_2 and nickel iron oxyborates (bonaccordite), notably Ni_2FeBO_5 have also been found in deposits on cores with AOA. On the basis of the general characterization information released by EPRI, the author has constructed a potential-pH diagram of Ni_2FeB(OH)_<10>, which is a hydrated state of FeNi_2(BO_3)O_2. However, the theoretical results were not consistent with the reported phenomenology since the removal of excess electrons is needed for metallic cations (e.g. Fe^<++>, Ni^<++> and Cr^<++>) to deposit as CRUD on the fuel's surface. This indicates that the reactor fuels must be cathodic, thus demonstrating that further investigation of the radiation effects on electrochemical reactions are indispensable. This is the main agenda of this paper. Since the author's last paper the author has confirmed that 'long-cell action' corrosion plays a pivotal role in practically all unresolved corrosion issues for all types of water-cooled reactors. Some of these unresolved issues are IGSCC, PWSCC, AOA and FAC. In conventional corrosion science, established by US NBS in 1958, it is well known that 'long cell action' can seriously accelerate or suppress the local cell corrosion activities. The author believes that the omission of this basic corrosion mechanism is the root cause for practically all un-resolved corrosion issues. The long-cell action corrosion mechanism is induced in water-cooled reactors as demonstrated with potential differences between irradiated- and un-irradiated reactor water. The author recently developed a unified theory, which enables the estimation of redox potential differences induced by radiation. The author's previous calculations through modifying the Nernst equation revealed a reasonable agreement with the published in-pile experimental results when a "fitting parameter" is introduced. Based on these new findings this paper revisits previous E-pH diagrams to investigate the effects of radiation. The Nernst equation, applied to reactor water, determines the actual in-core operational environment with a potential shift from the electrochemical potential due to solute species such as dissolved hydrogen and oxygen in the reactor water.

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This paper describes the capability of an air cooling system (ACS) to remove decay heat from the core of an Advanced Boiling Water Reactor (ABWR). The motivation of the present research is the Fukushima Severe Accident (SA).The plant suffered damage due to the tsunami and entered a state of Station Blackout (SBO) during which (sea) water cooling was not available. To prevent this kind of situation, we propose a passive decay heat removal system (DHRS). The plant behavior during the SBO is calculated using the system code NETFLOW++ assuming an ABWR which has a steam-driven ACS. Two types of air coolers (ACs) are applied for the calculation model of an ABWR, i.e., a steam condensing air cooler (SCAC) of which the intake for the heat transfer tubes is provided in the steam region, and a single-phase type of which the intake is provided in the water region. Plant transients with ACS are calculated under conditions of forced circulation and also under conditions of natural convection. The results of the calculations indicate that the decay heat can be removed safely with a reasonably sized ACS if the heat transfer tubes are cooled with forced air circulation. However, the decay heat can be removed by air natural convection after safety relief valves are activated for a couple of days. The heat removal rate is evaluated as a function of air mass velocity at the inlet of heat transfer tube banks.

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Since the scientific cause for a series of hydrogen explosions during the Fukushima accident has not been established, the author investigated his basic theory named "radiation-induced electrolysis" by applying the estimation of the amounts of H_2 generation during the active phase of the Fukushima accident. The author's theory was originally developed by including Faraday's Law of electrolysis into the basic time-dependent material balance equation of radiation-chemical species for his study on accelerated corrosion phenomena which is widely observed in aged plants. Through this mechanism as much as 5,300 m^3-STP of accumulated hydrogen gas is estimated to be inside the PCV just prior to the hydrogen explosion which occurred a day after the reactor trip in Unit 1. With this large volume of hydrogen gas the explosion was a viable possibility upon the "venting" operation as stated in previous reports. In view of this, this paper intends to continue the investigation of the Fukushima accident while focusing on hydrogen generation from the spent fuel pools, as the author's previous estimation for 1F4 SFP indicated a rapid initiation of hydrogen generation when the pool water temperature exceeded 40 degrees centigrade. These observations lead to investigate the potential mechanism by changing the pool water temperature and flow velocity in the spent fuels. During the trial calculations it was discovered by chance that SBO induced a rapid initiation of electrolysis when the pool water temperature surpassed 40°C with a range of low water flow velocity through the spent fuels as reported in NPC-2014. Due to a difference in the absorbed dose rate of water through γ-decay heat, an electrochemical potential difference should exist in the highly radioactive region of spent fuels stored by evacuating the core and less radioactive fuels stored for several years. All these spent fuels were stored in spent fuel racks placed at the bottom of the pool covered with a stainless steel lining. The metallic contacts enabled electric conduction between the highly radioactive fuel assemblies and cooled spent fuels. This hypothesis leads to a simple solution of inserting ceramic insulator to prevent direct metallic contact of the spent fuel racks to the liner and each other in order to disconnect the closed circuit. However, the author has reservation of this hypothesis, without first having the appropriate experimental verification. The estimation of the amount of hydrogen was calculated without experimental evidence of the electric current going through the spent fuel racks.

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