Applicants and licensees of nuclear power plants have struggled to address United States Nuclear Regulatory Commission (NRC) expectations to assess if high energy line break (HELB) jet impingement on safety-related structures and components can lead to dynamic amplification. In this paper, evaluation of the potential for such amplification and the occurrence of resonance conclusively demonstrates that the phenomenon does not occur. Specifically, NRC Standard Review Plan Section 3.6.2 Revision 3 identifies a potential for jet load amplification associated with formation of unsteadiness in free jets that induces time varying oscillatory loads on obstacles in the flow path. In a HELB, several physical parameters of steam and two-phase jets issuing from a ruptured pipe – such as non-equilibrium condensation of steam in the jet, unsteady separation between the jet exit and impingement target, non-orthogonal alignment of jet axis to impingement surface, uneven or soft impingement surfaces, or mismatch of jet excitation frequency and impingement target natural frequency – prevent occurrence of the phase lock conditions needed to initiate and maintain a resonance. Justification is provided that the physics, thermodynamics, and geometry of a HELB jet in a light water reactor preclude occurrence of dynamic amplification and resonance of impinged downstream structures and components. Available data from single and two-phase impinging jet experiments and analyses confirm this conclusion.

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The thimble tube, which is made of Zircaly-4, is one of the main components of a PWR fuel assembly. The thimble tube has an important role as a structural member of the skeleton. Another role of the thimble tube is to guide a rod cluster control assembly (RCCA) for insertion during the reactor operation. The function has to be assured not only in normal operation but in a seismic event. In a horizontal seismic event, the fuel assembly vibrates laterally, which gives bending moment to the thimble tube. In addition, axial compression force acts on the thimble tube in a vertical seismic event. The integrity of the thimble tube has to be maintained while this force and moment act. Mitsubishi has confirmed by the elastic stress analysis that the stress of the thimble tube has been lower than the limit value requested for the seismic event. The stress evaluation method is based on the ASME code. The ASME code also describes the limit analysis which is available when the predicted stress is beyond elastic region of the material. In the analysis, the material is assumed to be elastic-perfectly plastic, and the maximum load that the structure can carry is calculated but the load is one direction. In our previous paper, the strength limit of the thimble tube with material plasticity was analyzed considering the bending moment and axial compression force at the same time. The analyses were performed with parameters of material properties. According to the analysis results, the practical strength limit by the analysis with the actual elastic-plastic material was larger than the one with the elastic-perfectly plastic material. This paper describes the experimental verification for the strength limit analysis model of the thimble tube with material plasticity under bending moment and axial compression force. The experiments are conducted to investigate the thimble tube’s strength at room temperature and the elevated temperature, 343^oC. The simulation analyses for the experiments are performed by the 3 dimensional model with elastic-plastic material. According to the comparisons with the experiments, the analysis results show good agreement with the experiment results at both room temperature and 343^oC. Based on the comparison, the analysis model is verified and can be effective for the strength prediction of the thimble tube.

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In the working process of the Nuclear Electric Propulsion, it will generate much waste heat by the power conversion system due to inefficiencies in the thermal-to-electric conversion process. Therefore, how to improve the heat discharge capacity of radiant heat exchanger has become an essential part of the research of space nuclear reactor. In this paper, the correctness of the Monte Carlo Method is verified by the theoretical solution and the Monte Carlo method to calculate the result of the Perpendicular rectangles with a common edge. Monte Carlo Method is used for calculating the radiation shape factor of the radiator and established the radiation heat exchange network between the multi-surfaces to figure out the distribution of heat dissipated on the panel. By analyzing the influence of different positions and temperature on the radiator, some opinion of improving the radiator panels are put forward. These results can serve as a reference for the design of heat rejection subsystem in space nuclear reactors.

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Small Modular Reactor, due to its small power and low source inventory, its off-site radiation impact is mainly concentrated in the near-field of the site. Due to the influence of structures within the field, distribution of the wind field no longer exhibits a standard Gaussian form. It is necessary to correct the atmospheric diffusion parameters to improve the calculation accuracy of the radiation consequence evaluation model. In this paper, the CAP200 SMR is taken as an example, a powerful 3D numerical wind tunnel platform has been developed which is named MSDA-3D, the meteorological combination of 16 wind directions, 6 types of atmospheric stability and annual average wind speed are used. Through the MSDA-3D numerical simulation of the plant area, the correction method of atmospheric diffusion parameters is proposed, and the real atmospheric diffusion parameters that meet the characteristics of the site are also recommended. Based on this, combined with the design basis accident source term of CAP200, research on the impacts of off-site radiation is carried out. Results of this paper show that after considering the buildings within the field, the actual atmospheric diffusion parameters of the site need to be corrected for different atmospheric stability compared to the standard P-G curve. At the same time, for the CAP200 SMR, even using the new atmospheric diffusion parameters, the radiation consequences are limited to 800m within the site, and will not affect the offsite public, and no off-site emergency protection actions are required.

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In this study, a new combined reactivity control pattern is presented to lower the power peaking factor on a floating nuclear power plant. In the pattern the whole cycle length is divided into several boron-adjusting periods. During one boron-adjusting period, the soluble boron concentration remains unchanged and the core reactivity change is controlled by control rods. Boron concentration would only be changed at the end of a boron-adjusting period. In the whole cycle the boron concentration would be changed for two to four times, which is acceptable for a floating nuclear power plant. To get an optimized solution, a model of the combined reactivity control pattern was established, which contained boron concentration and control rod pattern (CRP) as coupled variables. A particle swarm optimization with novel mutation operator (PSONM) was applied to optimize the solution. The model was practiced based on a small modular reactor ACP100 core. The solution was optimized separately for equal division of two, three, four and five boron-adjusting periods. The results show that the combined reactivity control pattern can get a lower power peak factor than the original rod-controlled-only pattern without shortening the cycle length. Therefore, the safety of the floating nuclear power plant can be improved.

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In the secondary circuit of pressurized water reactor (PWR), the pH value of secondary circuit including steam generator water, feedwater and condensate is generally measured after the temperature decreases to 25 degrees Celsius. The pH value measured in this way may deviate from the one at the actual operating temperature. The appropriate pH value can reduce or even avoid corrosion in feedwater and condensate system and steam generator as well as scaling in steam generator. The pH value at the actual operating temperature can be calculated through the software entitled ‘Software for Calculating the Secondary Circuit’s pH value of PWR Basified with Ammonia’. Only the concentration of ammonia, ionization equilibrium constants of ammonia, water, acetic acid, carbonic acid etc. at a certain temperature such as 300 degrees Celsius are required. Therefore, this software for calculating the pH value is very instructive for monitoring the operation of the secondary circuit of PWR.

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A large quantity of hydrogen will be released due to zirconium-water reaction in a pressurized water reactor during severe accident. And the hydrogen will diffuse into containment vessel, and thus increases the risk of explosion. Since the containment vessel is exactly the last barrier for fatal irradiated species, being damaged by hydrogen detonation or an uncontrollable leak of gas are unacceptable. Accordingly, efficient strategies should be adopted to address this issue. Hydrogen elimination and measuring devices, therefore, are widely applied in nuclear power plant all over the world. Because of harsh environment inside the containment vessel during severe accident, always under high temperature, high pressure, and being inundated with steam, irradiated dust and chemical species, it is a big challenge to design a reliable hydrogen concentration measuring system. Generally, the hydrogen concentration measuring systems developed presently can be divided into two categories including non-in situ measurement and in situ measurement; the first integrates with withdraw parts and the hydrogen sensor can only be set outside the containment vessel, while the second based on insitu hydrogen sensor which can be set in the containment vessel. The hydrogen sensors for the non-in situ measurement are always thermal conductivity detectors. Despite lots of controversies on the safety under severe accident, such as a possible leak of the irradiated mixture, the non-in situ measurement has been a mature technique for design basis accident. In-situ measurement is state-of-the-art technique due to direct detection and high accuracy. Hydrogen concentration measuring systems by in-situ measurement and with hydrogen sensor originated from catalytic recombination, thermal conductivity, resistance, and electrochemistry, have produced outstanding performance and attracted a lot of attentions in the world. In this study, we will introduce the most recently development of hydrogen concentration measuring systems designed for severe accident. Furthermore, relevant technique requirements as well as verifications will also be contained.

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An efficient power conversion scheme and its characteristics at part-load conditions are crucial for the thermal system. The combined cycle coupled with HTGR is able to achieve the cascaded utilization of energy, which includes the topping closed Brayton cycle and the bottoming Rankine cycle. This paper provided an effective control method for part load of combined cycle coupled with HTGR. When the power of the combined cycle changes, the topping cycle is controlled by helium inventory, and the bottoming cycle works under constant pressure. The mass flow of steam is changed with helium synchronously. This control method can keep the cycle efficiency at a high level: when the load is changed from 100% to 30%, the cycle efficiency just decreases by 5.1%. This is a new control method, which can ensure the inherent safety of HTGR in a wide range of power.

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In this paper, numerical simulations are performed using several low-Reynolds number k-ε turbulence models due to AKN, MK and V2F for the convection heat transfer of supercritical CO₂ flowing through a heated vertical mini tube with inner diameter of 0.27 mm for various heat fluxes at relatively low Reynolds number of 2900. The predictions are compared with the corresponding experimentally measured values. The performance of various low Reynolds number k-ε turbulence models under normal and deteriorated heat transfer conditions are assessed. Results show that all the three models produce fairly good predictions of local wall temperature variations under normal heat transfer condition. The MK and V2F models give good predictions when the flow acceleration is not very strong. However, they respond insufficiently to the flow acceleration effect when it gets stronger. The V2F model is able to capture the general trends of deteriorated heat transfer due to the combined flow acceleration and buoyancy effects. The AKN model over-responds to the flow acceleration effect in most cases, and distorts the velocity profiles more severely than the V2F model where local heat transfer deterioration occurs. Further improvements of the turbulence models are needed for quantitative predictions. Detailed information obtained from the numerical results using V2F model is qualitatively analyzed to understand the mechanism of the special characteristics of the reduced heat transfer resulted from the buoyancy and flow acceleration. The redistribution of flow field induced by the buoyancy and flow acceleration effects is the main factor leading to the local heat transfer deterioration.

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High temperature gas-cooled reactors and very high temperature reactor (HTGR and VHTR) have a wide range of applications because of their inherent safety and high heat source temperature. Combined cycle can realize the cascade utilization of high-temperature heat source of HTGR and VHTR and improve their energy utilization efficiency. In this article, a theoretical optimization method for the combined cycle coupled with HTGR and VHTR is proposed. By optimizing the topping cycle and the bottoming cycle separately, the multivariate combined cycle efficiency expression is organized into the optimized relationship with three variables (reactor outlet temperature, main steam temperature and main steam pressure). Therefore, the problems of multiple variables, mutual coupling and unclear physical significance of the combined cycle coupled with HTGR and VHTR are solved. The theoretical optimization values can provide guidance for the design of combined cycle of HTGR and VHTR.

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The mission of the U.S. Department of Energy (DOE), Office of Nuclear Energy is to advance nuclear power in order to meet the nation’s energy, environmental, and energy security needs. Advanced high temperature reactor systems such as sodium fast reactors and high and very high temperature gascooled reactors will required compact heat exchangers (CHX) for the next generation of nuclear reactor plant designs (Reference 1). To support this objective, the U.S. Department of Energy is sponsoring research to support the development of the CHX for use in high temperature advanced reactors. The project is being executed by an Integrated Research Project (IRP) and includes team members from the University of Wisconsin–Madison, University of Michigan, Georgia Institute of Technology, University of Idaho, North Carolina State University, Oregon State University, Electric Power Research Institute, MPR Associates, and heat exchanger manufacturers CompRex and Vacuum Process Engineering (References 2 & 3). The objective of the research is to enable the use of compact heat exchanger designs in high temperature advanced reactor service in order to improve plant efficiency and economics. A necessary step for achieving this objective is to ensure that the ASME Boiler and Pressure Vessel Code, Section III has rules for the construction of CHXs for nuclear service. The IRP is conducting the research to support a Code Case that provides those rules. This paper identifies the major technology and research gaps impeding the development of a Code Case for compact heat exchangers for high temperature reactor service constructed in accordance with the ASME Code Section III, Division 5, Class A. The paper will also outline how the ASME Committees can use the basic research to support a Code Case. The major technology gaps include material properties, failure modes and effects, analysis methods, and examination methods. The IRP will study basic material properties of diffusion bonded plate, creep and fatigue models, development of NDE methodology and development of advanced analytical approaches to design.

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The U.S. Department of Energy is sponsoring research to support the development of the compact heat exchanger (CHX) for use in high temperature advanced reactors. The project is being executed by an Integrated Research Project (IRP) and includes team members from the University of Wisconsin– Madison, University of Michigan, Georgia Institute of Technology, University of Idaho, North Carolina State University, Oregon State University, Electric Power Research Institute, MPR Associates, and heat exchanger manufacturers CompRex and Vacuum Process Engineering (References 1& 2). The objective of the research is to enable the use of compact heat exchanger designs in high temperature advanced reactor service in order to improve plant efficiency and economics. A necessary step for achieving this objective is to ensure that the ASME Boiler and Pressure Vessel Code, Section III (Reference 3) has rules for the construction of CHXs for nuclear service. However, construction rules alone are not sufficient to deploy a compact heat exchanger in an advanced reactor. ASME Section XI (Reference 4) Rules for the Inservice Inspection (ISI) of a heat exchanger in an operating nuclear reactor will be required as well. The failure mechanisms for compact heat exchanger are unique and are not well understood since they have limited operating experience. The IRP will develop an inservice inspection roadmap to guide IRP research that will enable an owner/operator of an advanced nuclear reactor to implement an appropriate inservice inspection program for a compact heat exchanger. Failure mechanisms and their associated probabilities are the key inputs to a successful inspection program. The inspection roadmap will make use of the soon to be published ASME Section XI, Division 2, Reliability and Integrity Management (RIM) methodology. RIM is a non-deterministic, technology neutral, approach to inservice inspection. A RIM based ISI program is developed by an owner considering the entire reactor technology safety case rather than on the specific component technology. Therefore, owners will need specific information regarding the performance, failure modes, reliability, and inspection options for compact heat exchangers in order to incorporate such a heat exchanger into the plant design. This paper outlines the IRP research as it relates to inputs needed by an owner for the eventual development of an ISI program that includes compact heat exchangers.

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An advanced gas cooled reactor core is constructed using a large number of nuclear graphite components that make up channels for fuel and control rods which interlock through a keying system. The fuel channels are formed with cylindrical bricks which are subjected to fast neutron and radiolytic oxidation during reactor operation, leading to deformation and material properties changes. The deformation in the axial and radial directions of the graphite moderator brick, specifically at the bore surface, is important for the unhindered entry and exit of the fuel rods. Furthermore, deformation at the brick end faces can lead to axial gaps between bricks and can destabilise the fuel channels. Therefore, it is important to understand the deformation and contact behaviour between bricks in the fuel channels. This work focuses primarily on the contact behaviour between bricks and its effects on the ovality of the brick bore. Furthermore effects of changes in material properties on the ovality of the brick bore and contact conditions are also considered. Finite element modelling has been conducted, including the combined effect of irradiation damage dose, irradiation temperature, weight loss (radiolytic oxidation) and column self-weight. Sensitivity studies have also been conducted to assess the parameters which influence ovality and interfacial contact conditions in AGR bricks.

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Verification test to the program of solving the steady state multi-group neutron diffusion equations is very important to ensure the quality of the software. The authors investigate the perturbation formulas for the single-group diffusion equations, and two-group diffusion equations, respectively. In addition, two accurate estimation formulas are derived for the effective multiplication factor of the perturbation system. The estimation formulas not only reflect the properties of the perturbation system, but also can be used to verification test. For this reason, the authors propose a new verification method based on the perturbation theory in this paper. Two numerical examples are presented demonstrating the validity of our theoretical results.

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The layout and the water intakes of a small modular reactor are quite different from previous nuclear power plants (NPPs). There are three main problems with the water intake engineering in this small modular reactor: the water supply problem of the service water system during the plant maintenance; the problem of biofouling in the water intake tunnel; and the high unit energy cost (UEC) of water intake engineering. In order to solve these problems, a variety of solutions are proposed and analyzed in this paper.

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This paper describes the effect of local plastic component on J-integral and crack opening displacement (COD) evaluation of a circumferential penetrated crack, applicable to the leak before break (LBB) assessment for sodium cooled fast reactor (SFR) components. J-integral COD evaluation methods are generally formulated as a summation of elastic and plastic components, and so far many evaluation formulae based on these two components have been proposed. However, strictly, the plastic component consists of local plastic and fully plastic components. Many of the conventional evaluation methods often consider only the fully plastic component as the plastic component. The reason for this is that the effect of the local plastic component is much smaller than that of the fully plastic component excluding materials with extremely small work hardening. In contrast, for materials with high yield stress and small work hardening, such as modified 9Cr-1Mo steel which is one of the candidate materials for SFR piping, the effect of the local plastic component on J-integral and COD cannot be ignored. Therefore, the authors propose formulae taking the effect of local plastic component on J-integral and COD into account, based on finite element analysis (FEA) results, so that it is easy to apply to crack evaluation. The formulae will be employed in the guidelines on LBB assessment for SFR components published from Japan Society of Mechanical Engineers (JSME).

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The safety measures are strengthening at each nuclear power plant in order to restart the power plant, based on the new regulatory standard established by the Nuclear Regulatory Agency after Fukushima No.1 nuclear power plant accident caused by the Grate East Japan Earthquake. As one of the strengthening method against the natural disaster, it is necessary to prevent the water to the area where the important equipment and machines are located which maintain the function of nuclear power plant, from the duct or inlet of outdoor air that composes the air conditioning and ventilation system. Therefore we need to develop the damper which prevents the water through the duct from outside or inside the nuclear power plant. We decided the required detail functions of this damper and started to develop. At first we established the method of stopping the water, the mechanism of the operation and the basic structure, and then we checked the performance of anti-water pressure and earthquake resistance. We improved the damper based on the new knowledges which we found at the verification tests in each steps, and we developed the new type of damper that matched to the all required functions. We developed not only the damper but also the total flood prevention system and we had 48 installation achievement until now.

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The supercritical carbon dioxide Brayton cycle is a key technology in VHTR (Very High Temperature Reactor). The compact heat exchanger, PCHE (printed circuit heat exchanger) is one of the most important devices in supercritical CO₂ power cycle. In this research, 3D PCHE unit model with semicircle zigzag channel with diameter of 1.9 mm is established and computed using CFD methods. The numerical methods and geometric model are firstly verified by comparison with previous experimental results. Numerical simulation is performed with inlet pressure of 7.5 MPa/8.1 MPa, inlet temperature varied from 313.15 K to 373.15 K, and mass flux varied from 400 kg/(m² •s) to 800 kg/(m² •s) in the side of supercritical CO₂. Thermal-hydraulic characteristics of the zigzag PCHE were investigated in this paper. The pitch average local heat transfer coefficient is defined. The relationship between the pitch averaged local heat transfer coefficient and total average heat transfer coefficient is analyzed and discussed based on the numerical results. The effects of the thermo-physical property variations of supercritical CO₂ and the channel structure on the heat transfer and flow are discussed. Visual analysis is applied to show the velocity profile, pressure profile and temperature profile, which are of great help for performance analysis of PCHE. A new correlation is developed based on the numerical results in the present study considering the effects of thermo-physical property variations and channel structures, which is of great significance for the design and optimization of zigzag PCHE.

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Loss of Coolant Accidents (LOCAs) inside and outside containment are typical initiating events considered in Probabilistic Safety Assessment (PSA). The LOCA events on individual lines are grouped by similar success criteria including availability of the mitigation systems. While keeping the total LOCA frequency unchanged, there could be two approaches to the LOCA modeling in PSA: (1) a LOCA event on each line is represented by guillotine break, or (2) partial fracture(s) is separately considered from guillotine break (e.g., Small and Medium LOCAs caused by cracks or hole on pipes of larger size with different success criteria). In this paper, the modelling uncertainty associated with the relatively simplified approach (1) is characterized by sensitivity analyses based on the detailed approach (2), using the ABWR (Advanced Boiling Water Reactor) PSA. LOCAs inside containment are analyzed by the internal events at power PSA while LOCAs outside containment are analyzed by the internal flooding at power PSA.

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The decommissioning of TEPCO’s Fukushima Daiichi nuclear power plant accident is one of the most important issues in Japan. The criticality safety of fuel debris is imperative to prevent radiation exposure of workers. To prevent the criticality accident, the estimation of criticality of fuel debris is required for the fuel debris retrieval. Though the expert knowledge of reactor physics is necessary to estimate the criticality of fuel debris, many people who make a plan of fuel debris retrieval may not know well about criticality analysis. We developed a handy criticality analysis tool HAND to quickly estimate the criticality of fuel debris without expert knowledge of reactor physics. Since the input data of HAND is so simple and users can intuitively understand the calculation results, this tool is expected to be an effective tool to estimate the criticality of fuel debris. This paper explains the overview and calculation example of HAND.

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In order to elevate the resistance against irradiation-assisted stress corrosion cracking (IASCC) of reactor pressure vessel internals (RINs) in the environment of boiling water reactor (BWR), austenitic stainless steels (SSs) with tantalum (Ta) addition have been examined. The oversized element Ta is considered to reduce the concentration of free vacancy due to trapping effect. It is expected that radiation-induced grain boundary (GB) segregation of constituent element such as chromium (Cr) can be suppressed by Ta addition. In this study the irradiation tests were performed with Fe2+ ion in a dose range from 0.1 to 3 dpa on the specimens sampled from base materials and heat-affected zone (HAZ) of welding joint plates. Considering the Cr segregation might also occur during the heat history of welding process, we first confirmed and discussed the effect of Ta addition to SUS310S on irradiation resistance properties in the HAZ. The radiationinduced segregation (RIS) on random GBs in HAZ after the irradiation tests was evaluated by scanning transmission electron microscope (STEM) and X-ray energy dispersive spectroscopy (XEDS). The improvement of resistance against RIS by Ta addition was evaluated. Further the effect of Ta addition on radiation induced hardening was investigated on the base materials, but from the results of nanoindentation test the obvious relief of radiation-induced hardening was not confirmed. The corrosion properties after heavy ion irradiation were also discussed on the base materials in the conditions of electrochemical potentiokinetic reactivation (EPR) test. GB corrosion was observed on the ion-irradiated surface of commercial material of SUS310S. However, no GB corrosion was observed on the irradiated surface of SUS310S with 0.4 % Ta addition, indicating the effect of Ta addition on the improvement of resistance against IASCC.

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After the Fukushima Dai-ichi accident, regulations for nuclear power plant have been strengthened in Japan and the following assessment is required in case of the decommissioning of Spent Fuel Pool (SFP); that is “Prevention and mitigation measures for an accident of SFP should be maintained until spent fuel assemblies are removed from the SFP, or it should be demonstrated that these measures are not necessary”. From this background, the simple (but conservative) assessment of spent fuel integrity in large LOCA event has been carried out in Japan. Although the safety margin of the simple assessment is considered to be sufficient, it is important to capture it quantitatively. In this study, CFD analysis has been conducted for the assessment of spent fuel integrity in large LOCA event and the maximum temperature of spent fuel assemblies has been evaluated. Then, it has been compared with the result of the simple assessment method. As a case study, additional CFD analysis has been conducted, where water level in SFP decreases to the Bottom of Active Fuel (BAF) due to boil-off. Since this scenario might be more severe than large LOCA scenario, the number of spent fuel assemblies, their decay heat and loading pattern to maintain spent fuel integrity are investigated. A summary of the result and the insight from the study are provided in this paper.

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Grid-to-rod fretting wear resulting from fluid-induced vibrations in nuclear reactors has attracted great attention for its economy and safety concerns. To mitigate potential losses due to such fretting wear, the structural dynamics in the fluid-structure interaction system is required to be well understood and accurately estimated. Conventional approaches to describing the structural dynamics in such a system include analytical models and empirical correlations which were derived from fitting with experimental data. Meanwhile, a novel approach through numerical simulations to addressing the problem has been progressively developed and optimized, which features no requirements for force coefficients obtained from experiments as an input to analytical models. In the present paper, the dynamics of a flexible clamped-clamped cylindrical shell subjected to a turbulent external axial flow was studied numerically. The cylindrical shell was made of aluminium, 1 m in length, 14.7 / 14.16 mm in O.D. / I.D., and confined by a rigid tube of 40 mm in I.D.. In the fluid-structure interaction system, fluid and structural domains were spatially discretized by finite volume method (FVM) and finite element method (FEM), respectively. The mesh scale in the fluid domain was comparable with that in the literature. For the discretized sub-domains of study, the fluid was described by the Navier-Stokes equations, and the structure was described by the Euler-Bernoulli beam equation, the fluid and structure systems were coupled on their interface by exchanging force information at each calculation time step. The solvers to the two systems herein were CFX Ansys and Mechanical Ansys, respectively. Thus no third party code such as a commercial package MpCCI, is required to couple the two calculation schemes. The simulation results show good agreement with experimental data in the magnitude of oscillating amplitude, and witness less difference with the experimental data than that by many empirical correlations. The time series oscillating displacement fluctuates around its original equilibrium state as periodically dissipating energy to and obtaining energy from the surrounded fluid. It is also found that the fundamental frequency obtained from the simulations is in quantitatively consistent with that by empirical correlations. The dimensionless displacement distribution along the cylinder span is in quantitatively consensus with the mode shape in the literature. The simulation approach also features its miscellaneous capabilities in predicting the fluid dynamics in fluid-structure interaction systems, such as fluid flow field, pressure distribution, turbulence intensity, etc. Experimental laboratories will also benefit from comparison opportunities.

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In this paper, a Mechanical Shim (MSHIM) control system simulation platform for the AP1000 reactor which was said as a Generation III+ pressurized water reactor was developed. Based on this platform, simulation studies for three MSHIM control strategies were performed. It has been demonstrated by the analysis results that the improved MSHIM control strategy purposed in this paper not only can provide much tighter axial offset (AO) control but also can reduce the control rod movement, indicating its superiority than the original MSHIM control strategy. The study can be used as a reference for engineering practices.

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Reactor coolant temperature control plays a central role among all reactor control systems because the effect of temperature control is directly related to the load tracking capability and also to the reactor protection margins, such as the over-temperature and over-power restraints. Special for the Mode-A reactor, the capacity of reactor is depending on the accurate and quality of temperature control. For example, in most of the time, all the Mode-A reactors in China are operating at basic load, so the temperature control is always worked in an simple way. But the temperature control system must also consider the quicker response and capacity to the wide range change of load, this paper based on point-kinetic model and heat transfer model in reactor, designs a suitable temperature model for control system. Considering that the traditional PID control method is difficult to solve the robustness and stability of high-order systems for large range operation, this paper uses Multi-model Predictive Control to solve this problem. According to the mechanism model, and establish nine single models, this paper online calculation model -matching degree to get recursive bayesian probability weighting, then get global prediction model. According to the global model, design Multi-model predictive control strategy. The simulation results indicate that under large conditions change, also robust and stable control performance is obtained.

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Aiming at the small nuclear power unit with electric power below 300MWe defined by the International Atomic Energy Agency, this paper makes a relatively complete research on main small reactors developed by domestic and foreign research institutes, and as for the engineered safety feature (ESF) characteristics of each type of small reactor, some reasonable analyses are carried out to alleviate the accidents. Combined with CAP200 compact reactor features developed by SNERDI/SPIC, China, this paper puts forward some effective measures to alleviate non-loss-of-coolant accident (Non-LOCA), loss of coolant accident (LOCA) and containment pressure. And to comb these limited accidents that challenge the ESF, this paper carries out some quantitative accident evaluation, to ensure that the ESF of CAP200 compact reactor has been designed to make sure the integrity of the reactor core and containment.

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The integrated system validation (ISV) of a nuclear power plant is very important for human factor engineering (HFE) to ensure a good design. To accomplish the ISV, operational conditions and scenarios should be sampled and identified before the verification and validation (V&V) activities. As a novel design, besides the regular sample dimensions of the scenario indentificaiton, some special considerations are made for HTR-PM (high-temperature gas-cooled reactor-pebble bed module). Similar with other NPPs, the sampling dimension of the V&V includes the consideration of plant conditions and personnel tasks. Different plant conditions are sampled for ISV, such as normal operation of startup, shutdown and changes of plant power. Many types of personnel tasks are sampled, such as important human actions (HAs), manual initiation of protective actions, etc. Some special HFE aspects of HTR-PM are considered in the ISV, such as the new staffing level and new staffing model(John O, 2002), operation of the two reactor modules in different states, operation of systems shared between the two reactor modules, operation of non-LWR (light water reactor) process and reactivity control, etc. The scenario identified are incorporated in the later ISV activites.

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It is thought that nuclear power plants in Japan have prevented external leaks from valve packing for more than ten years due to re-torqueing. However, it remains to be elucidated how external leaks are prevented by re-torqueing in practice. Here we show the stress relaxation behavior of packing and devise the requirement of re-torqueing to prevent external leaks from gland packing. Firstly, a suitable gland packing material is selected for use in high leak potential locations. Secondly, re-torqueing is performed twenty four hours after fastening. Thirdly, coned disc springs are used to maintain bolt preloads as required. Our results provide the new insight that re-torqueing is a simple and effective way to reduce external leaks from valves, in particular, when using carbon fiber gland packing or combination packing with expanded graphite.

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The single droplet phase change model during motion is developed based on the phenomena description and mechanism comprehension, which including the droplet phase change model as well as the droplet motion model. Then, the calculation of the droplet phase change characteristics during moving in the uniform flow in the gravity separation space is conducted. The results show that when the droplet are evaporating during its moving, the radius will decrease continuously and it will be carried more easily by the steam vapor, which will lead to the larger separation radii of the droplets and the reduced the gravity separation efficiency. In addition, this paper shows the three-dimensional map for the critical separation radii over the pressure difference and the steam vapor flow velocity, which can contribute to forecast the influence of the droplet phase change on the separation characteristics. The results can be applied in the design of the steam-water separation plants.

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Zirconium hydride is a new type of shielding material. It has a low density, a high hydrogen content and a high thermal stability. Russia and Japan conducted related research work on it as the shielding material. In the shield design of small sodium-cooled fast reactor (called SSFR for short) in China, reference was made to the shield design of sodium-cooled fast reactors currently. Boron carbide and stainless steel shields were used on the stack side with a thickness of 101 cm. However, the SSFR is a portable reactor system with limited shielding space. The shielding space is large in the current shield design scheme, and the shield design needs to be optimized to reduce the thickness of the SSFR stack side shield. Zirconium hydride is selected as the shielding material on the SSFR stack side. The shielding properties of zirconium hydride under the energy spectrum of the SSFR core region were calculated using the one-dimensional discrete ordinate method code (ANISN code). The zirconium hydride has a small neutron absorption cross section, and slowed-down thermal neutrons are not effectively absorbed, which may increase the stack side shield thickness. Therefore, boron carbide with a higher thermal neutron absorption cross-section was added to the zirconium hydride, and the thickness of the stack side shielded with zirconium hydride and boron carbide was calculated. The results indicate that compared with the original design scheme, the SSFR stack side is shielded with zirconium hydride and boron carbide (the volume ratio of boron carbide is less than 0.3), and the shielding thickness is reduced by about 20%. It can be seen that the mixed shielding material of zirconium hydride and boron carbide has good shielding performance and can be used as a shielding material for SSFR. Zirconium hydride and boron carbide mixed shielding materials have high engineering application value for small reactors, especially small sodium-cooled fast reactors. However, there is no research on the development of mixed shielding materials for zirconium hydride and boron carbide at home and abroad. Therefore, the next step will be to study the preparation methods of zirconium hydride and boron carbide shielding materials and carry out relevant shielding tests.

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According to the relevant safety standards and requirements, before radioactive materials transport packages put into use, it need to through a series of safety tests such as water immersion test, thermal test, and mechanical test and so on. In general, we need to load a series of sensors on the package to monitor the data when conducting experiments. In this paper, video analysis method is used to analyze the acceleration during a free drop experiment, in the test, a package fell freely from a height of 9 meters and fell on a rigid target, and we use two high speed cameras to record the process of the package colliding with the rigid target from different directions. And we analyze the package according to the video content of the high-speed camera after the experiment, we can get the acceleration data when the package colliding with the rigid target, next, we compare the analysis results with the experimental data. The results show that the results obtained by video analysis are comparable to the actual measured data, this method can be used as an auxiliary analysis method in the mechanical test, and it can prove the measured result to some extent.

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The study on the flux of atmospheric radionuclide into the ocean under nuclear power plant accidents is significant for emergency decision and nuclear accident mitigating. Based on the atmospheric diffusion model of Gaussian plume, coupled with the characteristics of dry and wet deposition of radionuclide under accident conditions, a model of atmospheric radionuclide flux into the ocean under accident conditions is established, and the effects of dry and wet deposition on the concentration of radionuclide in the ocean are analyzed. The results show that, under hypothetical accidents, the ground concentration in the downwind direction of radionuclide in the atmosphere rises first and then decreases, reaching the maximum value at about 300 meters under the postulated accident. The dry deposition flux is three to four orders of magnitude lower than the ground concentration, and the wet deposition flux is two orders of magnitude larger than the dry deposition flux. Comparing to dry deposition,wet deposition has a greater impact on the radionuclide flux into the ocean. radionuclide of dry deposition can act as a surface source of radionuclide in the ocean, and radionuclide of wet deposition can act as a point source.

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Waste minimization of air filter generated from nuclear power plant can be realized by clearance and design of new recyclable structure. The metal framework can get clearance by several steps to reduce the waste volume. According to some framework clearance of scrap air filter, we put forward some improvement suggestions on the air filter structure . So that the framework can be easily dismantled, secondary contamination can be avoided. Waste minimization goal can be further achieved by reuse of the outside metal framework.

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There are varieties of the design of Small Modular Reactors(SMR) all around the world. Researching and Analysing the advanced design of SMR of other countries, the innovative SMR is brought forward. The innovations are made in many aspects, such as Reactor, Specialized Safety System, Containment and so on. As for actively responding the innovative design requests of ACP100+ of CNNC, This paper focuses on the advanced, safe and economical design and improvement, providing significant references for the standardization and commercialization of SMR in future. Innovative assumes of future SMRs are brought forward, depending on the advancement of Science technique and Component manufacture, for realizing the safe and effective energy supply and some other requests for human’s better lives.

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In the issue of IVR (In-Vessel Retention) during severe accidents in nuclear reactors, the corium pool will finally evolve into two-layer configuration with oxide layer in bottom and metal layer on top, as well as the crust formation at interface. In case of the interface crust became molten, the strong turbulence in the oxide layer may entrain the top metal layer and break the interface stability. In this paper, the numerical simulations were performed based on the WMLES turbulence model, phase change model and VOF model to describe thermal behaviors, as well as the turbulent flow interaction at interface. Furthermore, the model based on KHI theory was developed. The results can be used to evaluate the interface instability of the two-layer corium pool and its influence on the heat transfer behavior.

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As the parallel implementation of nuclear power projects with multi-technologies at multi-sites, the balance issues of different resources including but not limited to manpower and machines, materials among projects become more and more serious. It is critical for EPC (Engineering Procurement Construction) contractors to allocate and predict project resources in a scientific and reasonable way. According to the life cycle planning with standard curve of resources for projects, the probability model of resource configuration and prediction based on Monte Carlo Simulation(MCS) is proposed, which is combined with project schedule, risk and resource, to achieved the optimization control and prediction objectives plus the transformation of from static to dynamic and qualitative to quantitative. This method takes into account the dynamic progress and risks, the short-term and long-term resource needs, which achieves the coordination and flexible allocation of resources, and reduces cost effectively. Finally, a case study is performed in this paper， and a more scientific multiple Nuclear Power Plant (NPP) manpower reserve plan is developed based on the result to achieve the maximization of resources benefit for decision makers.

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The total cost of project management human resources for large nuclear EPC projects is high. Compared to the manual labor, it’s more difficult to evaluate the management staffing level. Based on the construction experience of historical nuclear EPC projects and by applying some allocating methods suitable for EPC contractor, this paper develops the standard allocation baseline of project management human resources including the staffing curve throughout the project life cycle. This baseline has been applied on developing the staffing plans for several large standard nuclear EPC projects. Aiming at applying the standard allocation on non-standard projects, this paper also develops an adaptive adjustment method based on VFT and AHP methodology. By analyzing the difference between the standard project and non-standard ones on project scale and difficulty, the method adjusts the standard allocation to offer reference for the total amount of human resources of non-standard projects accordingly.

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In marine nuclear power plants (MNPP) which contain several reactors and turbines working in parallel, the power outputs of each turbine interfere with each other because of a common steam header that they are connected to causes the coupling effect among the turbine inlet flowrates. In order to avoid the mutual influence between the turbines, this paper presents a decoupling header outlet flowrate controlling method based on neural network inverse system (NNIS) composed of a static neural network with state feedbacks of the original multi-reactor and multi-load system. The existence of the inverse system for the header is proved by the reversibility derivation of interactor algorithm. The simulation results indicate that the proposed decoupling controller has excellent decoupling performance, which can turn the multivariable system into a number of single input and single output systems, and eliminate the mutual influence. Furthermore, it has a faster tracking capability than conventional PID controller which can improve the safe and stable operation of the multi-reactor and multi-load system.

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Boiling water reactor is the direct cycle that delivers the steam generated in the reactor to the turbine, so that it is important to detect leakage of main steam in terms of minimizing public and worker doses if an accident occurs. Leak detection system has several detection methods to be able to detect leakage when various kinds of leakage such that from complete break to partial break of main steam piping occurs. In those methods, it is considered that the main steam tunnel room temperature high signal or the turbine building temperature high signal is effective also for partial break of main steam piping that is more difficult to detect than complete break. Thus, this study focuses on the phenomenological response for the detection signal.

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In order to understand the influence of flow on the measurement of ¹⁶N monitor，the fine structure of HFETR core and different irradiation areas were modeled with MCNP software and the ratio of the measurement of ¹⁶N monitor to power is calculated considering the distribution of the fast neutron flux ,energy spectrum and the decay of ¹⁶N in different areas interested. The calculation show that the if flow range was under the normal reactor operation condition, the ratio of the measurement of ¹⁶N monitor to power is decreased with the increase of flow but if the choice of the measurement position was appropriate, the response change of ¹⁶N monitor was very small.Meanwhile, the validation of calculation with the measurement of ionization chamber show that the simulation is reliable and some suggestions could be offered.

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The source terms of fission products in the airborne effluents of the CPR1000 and HPR1000 (Hualong One) nuclear power plants (NPPs) in China were estimated for cycles without fuel failure. We used both the operational data for the primary coolants of domestic and foreign NPPs as well as the latest research on reactor source terms for this purpose. The results indicated that the activity concentrations of fission product nuclides in the gaseous effluents were two orders of magnitude lower than the detection limit of laboratory instruments. After comparing the domestic and foreign regulations on gaseous effluent monitoring of NPPs and reviewing the domestic gaseous effluent monitoring practices, recommendations on gaseous effluent monitoring were offered in this paper. This study could provide a technical basis for gaseous effluent management for NPPs and help revise the relevant Chinese regulations and standards in the future.

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In severe accidents (SAs) of nuclear power plants, release of gas containing fission products (FPs) from the reactor vessel is thought to be a major issue. As to reduce the leakage of FPs into the environment, gas containing FPs are generally discharged though the wet-well and decontaminated by the transfer effect of FPs from the gas-phase to the liquid phase. This effect is called pool scrubbing. In SA analysis codes such as MELCOR, it is predicted in the model that FP particle motion inside a single bubble, created by the bubbly flow inside the wet well as a major factor in decontamination. However, there are few experimental evidences to support the modeling of this phenomena. Therefore, in this paper, the experiment to observe particle decontamination in order to understand the relationship of particle motion and decontamination. In the decontamination behavior measurement, by using Maki-interferometer microscope, density distribution of particle decontaminated from an air bubble was observed. From the results, we succeed in visualizing the phase distribution around the air bubble due to optical path difference of decontaminated particle. From the results, strong occurrence of decontamination is observed in the major axis and the bottom of bubble. By comparing the results to the previously obtained particle motion, we confirmed that the decontamination in the major axis is related to the inner flow in terms of centrifugal force due to the peak of interface velocity, and the decontamination in the bottom of could be explained by the gravitational sedimentation.

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There are many kinds of purification devices in nuclear power plant reactor water loops. Multilevel purification device designed of the Venturi effect is one of them and needs to be investigated. This paper mainly introduces hydraulic experimental investigations on the multilevel Venturi purification device. Hydraulic experimental investigations focus on flowing resistance characteristic of Multilevel Purification Device. A large number of experimental data are obtained from hydraulic tests, including the resistance coefficient of every umbrella purification unit and the total resistance coefficient of the device. The total resistance coefficient of the device is 22.55~23.66. Test research demonstrates that the hydraulic characteristic of the multilevel Venturi purification device can meet the engineering needs well.

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Flow instability is frequently discovered in two-phase flow. It can cause mechanic vibration and periodical thermal stress in nuclear power reactor facilities. Natural circulation is widely used in modern reactor design because of its passive safety characteristics. As a self-feedback dynamic system, natural circulation flow instability can be regarded as a chaotic phenomenon with noise-interfering inherent vibration frequencies. Thus flow rate and system pressure oscillations under natural circulation can cause unpredictable problems to reactor operation and control. As a typical kind of flow instability in reactor thermal-hydraulics, pressure drop oscillations (PDO) are widely studied by experiment and simulation methods. A series of experiments in a natural circulation loop with a rod bundle heating channel shows a chaotic evolution route. For different connection positions between pressurizer and natural circulation loop, flow rate and system pressure vibration show different chaotic and fractal characteristic in PDO phenomenon. Several nonlinear characteristics parameters such as and fractal dimensions are calculated to analyze thermal-hydraulic characteristics of PDO. Natural circulation flow rate maintains a constant under a low heat flux. When the heat flux increases, flow rate becomes a quasiperiodic oscillation and the chaotic vibration occurs.

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Various researches have indicated the relative advantages of ferritic/martensitic (FM) steels over austenitic and other steels currently in use in light water reactors, and have regarded them as the most promising structural materials in both present and future reactors. Fe-Cr are the model alloys of these steels. Continuous exposure to severe irradiation environment will no doubt affect the microstructure of these materials significantly, which can subsequently manifest as observable changes in the materials’ physical properties. Various experimental and simulation literatures were reviewed in this paper with the aim of further understanding the irradiation effects of Fe-Cr alloys, which are the reference model systems for high chromium steels. Factors such as; solute type and concentration, irradiation dose, temperature, number of displacements per atom, dpa (fluence/dose), as well as type and energy of irradiating particle, among other things, affect the production and evolution of radiation damage in Fe-Cr alloys. Researchers’ findings on the way some of these factors affect the alloy under irradiation condition are presented in the paper. While the addition of Chromium (Cr) was found to improve the performance of FM steels by strongly suppressing void swelling, minimizing the effect of increase in irradiation hardening with decrease in dose rate, and so on, the effect of its content on Fe-Cr alloys under irradiation condition is however not clearly understood by the experimental studies, perhaps due to factors such as the nonmonotonic variation of its content with properties like void swelling. Primary radiation damage production and evolution, as presented in various Molecular Dynamics (MD) simulation literatures reviewed, determine the macroscopic response of a material to irradiation, thus making it paramount to understand. Insight into the atomic processes leading to changes in mechanical properties of materials can be gained through multiscale computer simulations. Rigorous efforts are therefore needed in this regard to help enhance our understanding of the effects of irradiation on materials and the way they can best be mitigated during design in order to help ensure safe and reliable operation of nuclear power plants.

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Two-phase flow instabilities have aroused considerable attention and been extensively studied in nuclear engineering. Among them, density wave oscillations are the foremost ones possibly encountered in boiling water reactors and steam generators. Because reaching flow instability is usually a gradual process with increasing oscillation amplitude, a variety of flow instability criteria have been put forward from different perspectives. With various flow instability threshold calculation methods, flow instability boundaries with considerable discrepancy could be obtained. In this work, density wave oscillations in a vertical single boiling channel with constant pressure drop are simulated by RELAP5/MOD4.0 code. The cross-sectional areas of timedependent volume components are varied to ensure valid pressure boundary conditions. Different heating power growth rates are tested to achieve flow instabilities by increasing heating power linearly in order to check whether the onset of the flow instabilities is influenced by different disturbances introduced. Flow instability boundary data obtained from the combinations of three typical flow instability criteria and three flow instability threshold calculation methods are compared to systematically investigate the influence of these two factors. Finally, the effect of inlet subcooling is studied to explain the above observations. This work could serve as a reference for the selection of flow instability criteria and flow instability threshold calculation methods to obtain accurate and conservative flow instability boundaries.

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The Pebble Module High Temperature gas-cooled Reactor demonstration project (HTR-PM) has entered the process system commissioning stage after years of design and construction. As one of the fourth generation nuclear power plants with inherent safety technology, HTR-PM adopted dry in-site storage technology for spent fuel based on unshielded canister and dry storage silo, which is different with the wet storage method used in PWR. So all of the design, installation and commissioning of the spent fuel residual heat removal system for HTR-PM are innovative technologies. From the point of view of commissioning technology, this paper will study the key technologies and solutions in commissioning process of the spent fuel residual heat removal system. The spent fuel elements of HTR-PM are stored in the stainless steel spent fuel canister, and the transfer and hoisting of the spent fuel canister are realized through the operation of the ground crane and shield cover equipment set. The storage canister full of spent fuel would be stored in spent fuel storage silos, each silo could lay five spent fuel canisters. The spent fuel repository is composed of buffer storage region and intermediate storage region. In normal operation, active ventilation is adopted in the spent fuel buffer storage region, and active ventilation is carried out through the centrifugal blower to discharge the spent fuel residual heat. The main test object described in this paper is the residual heat removal system in the spent fuel buffer storage region. The spent fuel residual heat in the intermediate storage region would be removed through natural ventilation completely. All of the spent fuel residual heat removal system in the spent fuel buffer storage region and intermediate storage region has inherent safety characteristics. The main purpose of the commissioning test of the spent fuel storage residual heat removal system in HTR-PM is to investigate the ventilating system commissioning characteristics for the spent fuel storage silo under the centrifugal blower operating condition. The ventilation test in the silo includes all kinds of operating conditions with different number of spent fuel canisters in each silo. The flow measurement in silo under the condition of non-uniform flow in silo is realized through the commissioning test. The movement and hoisting of the spent fuel storage canisters in silo are realized through the operation of the ground crane, so as to realize the wanted distribution condition of different number of storage canisters in the spent fuel storage silo. The residual heat removal centrifugal blower would be in operating condition, and the air flow rate in the storage silo and the typical differential pressure between inlet and outlet of the silo would be measured. All of the working conditions of the test could be realized through the operation of the canister and other equipment, or through the usage of special designed equipment, to achieve the simulation of the real operating conditions. The commissioning technology of the spent fuel storage residual heat removal system of high temperature gas cooled reactor demonstration project is the key to verify the safety of large-scale spent fuel residual heat removal ventilation cooling technology. As one of the basic test for spent fuel dry storage system, it will provide important basis for the design, commissioning and operation in subsequent developing stage of the spent fuel dry storage system.

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Based on the phenomenon that radon concentration in water has been influencing on people live deeply, the water in the surrounding villages of the tailings reservoir was taken as research object, in order to investigate the distribution of radon concentration in the surrounding water of a uranium tailings reservoir. Samples of five wells in two villages and tap water in another village were taken respectively, the radon concentrations were measured after each sampling by RAD7. And the variation of radon concentration in water at different temperatures were studied by using constant-temperature heating experiment. The study shows that the maximum concentration of water radon in five wells is 43,606 Bq/m³ , the minimum is 742 Bq/m³ , and the average is 12,089 Bq/m³ . The maximum concentration of water radon in tap water is 1,537 Bq/m³ , the minimum is 38 Bq/m³ , and the average is 511 Bq/m³ . The concentration of radon in well water is about 24 times that of tap water. The concentration of water radon is inversely proportional to temperature. While over 50℃, the decrease of water radon is more than 97%. According to this study, it is concluded that the sampling area has a higher level of radon concentration than standard, and it should arouse the concern of local governments and regulators. And heating is an effective way to reduce radon in water which can be promoted and popularized.

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The aim of this research is to produce materials that will be useful in complex applications such as nuclear reactor core, fuel rods especially in Gen-IV reactors which operate at higher temperatures than the present reactors, as well as in waste immobilization, high temperature, stress and fatigue applications which are tolerant to a number of radiation fluencies over a period of time. 10 g samples of five different polycrystalline compounds which includes Y₄Zr₃O₁₂, Y₄Hf₃O₁₂, Yb₄Zr₃O₁₂, Yb₄Hf₃O₁₂ and Sc₄Hf₃O₁₂ were produced using conventional ceramic process by solid state sintering for 7days at 15000C in high temperature-furnace with a 50C/min ramp up and down. Three of the samples Yb₄Zr₃O₁₂, Yb₄Hf₃O₁₂ and Sc₄Hf₃O₁₂ were found to be characterized by an ordered, fluorite derivative structure called delta (δ) phase while the other two, Y₄Zr₃O₁₂ and Y₄Hf₃O₁₂ has fluorite phase structure. An in-situ ion beam irradiation experiment was performed on the polycrystalline δ-phase compound with stoichiometry Yb₄Hf₃O₁₂ by Transmission Electron Microscopy (TEM) process. The machine was operated at 200 kV with a light 6 keV He⁺ ions to a maximum fluencies of 1.3 x 10¹⁷ ions/cm² to study the radiation damage in the sample. The sample was prepared for TEM by crushing the pellet into powder using isopropanol and collecting the suspension on carbon coated copper grid. The irradiation was performed from room temperature to 6000C respectively. TEM resolvable gas bubbles were observed which seem to be identical to thermally induced at a temperature of 3000C-5500C in the compound. The result suggests that He⁺ is retained during irradiation, but requires heat to form bubbles. This structural observation in the δ-phase composition can be used to understand the radiation damage effects.

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Abu Dhabi Polytechnic Reactor Simulator (ADPRES) is an open reactor simulator designed particularly as a learning and research tool for nuclear engineering students. ADPRES is written in pure FORTRAN 90 language from scratch. It solves static and transient three-dimensional, multi-group neutron diffusion equation in Cartesian geometries. Standard fourth order Nodal Expansion Method (NEM) is implemented in ADPRES. Currently, the simple adiabatic approach is used for transient problems without thermal-hydraulics feedbacks. It had been tested on several well-known benchmarks to evaluate its accuracy: IAEA-3D, DVP BWR, for static problems and LMW transient benchmark for time-dependent problem. The results showed that ADPRES results agree very well against reference solutions. The ADPRES’ source codes can be found in the author’s GitHub repository.

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