In this study, the well understood single round jet in crossflow is compared to an array of three jets aligned on the central axis of the crossflow, all of which are distinct with independent temperature and flow control. The benefit of this study is to better understand how an array of jets mix with a crossflow in turbulent conditions in an effort to further validate computation models for a variety of multiple-jet applications. The test section involves jets issuing vertically downward into a cross flow provided by a low speed wind tunnel. The jets are spaced two diameters apart, providing for an upstream, middle, and downstream jet. The hydraulic interactions are tracked via a stereoscopic particle image velocimetry (S-PIV) system. Several reference measurements and the uncertainty of the results are discussed to aid analogous computational fluid dynamics (CFD) models in the future. First order statistics between the single and triple jet cases with fixed flow rates are compared. Temporal analysis yields dominant frequencies at distinct regions within the flow. Insights into flow control of multiple jets in a cross flow is discussed. The experimental data sets compare the classical single jet in cross flow to the extension of multiple jets. Several differences in the flow behavior are found. The data sets serve as benchmark cases for future CFD models that will aim to replicate these flow types in different real-world applications such as coolant flows in a gas cooled reactor.

The offsite consequence study on High Temperature Gas Cooled Reactor Pebble-bed Module (HTR-PM) shows that its emergency planning zone (EPZ) for whether plume exposure pathway or ingestion exposure pathway can both be limited within the exclusion area boundary (EAB) of nuclear power plant (NPP), which is about a few hundred meters away from the reactor. This conclusion provides theoretical basis and strong technical support for potential industrial applications of HTR-PM. By this premise, the coupling risk for the industrial application of HTR-PM is quantitatively evaluated in this paper. The results indicate that the risk increments for both HTR-PM and industrial facility due to the colocation on one site could ensure they still meet their corresponding risk criteria. On this basis, the current proper management mode is recommended for the ally of HTR-PM and industrial facilities from technical perspective, which is the onsite of this HTR-PM-industry ally can adopt the mode of separated management, while the offsite co-management. Further, the final suggestions on offsite emergency preparedness for this HTR-PM-industry ally are also given out co-considering the current policy, practice and public acceptability. This work will also support the industrial application of other advanced reactors with inherent safety features.

In the ESNII+ EU FP7 project, a reactor physics benchmark aiming at the whole core calculation with the reflectors and detailed description of the structural elements was specified. This benchmark is based on the 2009 CEA concept of the ALLEGRO core. Fixed nominal technological data at nominal reactor state (geometry, composition) were prescribed which had to be modified in specified calculation branches according to different types of the thermal expansion and control rod positions. The parameters of the point kinetic model to be applied in a system thermal hydraulic code had to be determined this way. Static mechanical models of the expansion processes were specified by the benchmark. The goal of the calculation exercise was to verify the reactor physics codes, namely to get information about the modelling uncertainties and — after — their influence on the calculated results of the safety analyses. The obtained deviations between the participants are characterizing the user effects, the modelling uncertainties and the influence of the nuclear data differences all, without the possibility of their separation because of the complexity of the benchmark problem. A conclusion could be drawn that a step by step procedure starting from simple problems (homogenous material, Wigner-Seitz cell or subassembly in asymptotic approach) is necessary if we wish to identify the reasons of the deviations. For the Doppler effect, a decision was made in this direction already in the ESNII+ project where an infinite regular lattice problem without any leakage had to be solved. This approach of the simplicity is followed by the present benchmarks (one rod and one assembly), but extending the simple benchmarks with burnup calculations and taking into account leakage in asymptotic approximation by neglecting the complicated processes necessary in the reflector regions.

Differing from the adoption of helium as working fluid of closed Brayton cycle (CBC) for terrestrial high temperature gas cooled reactor (HTGR) power plants, helium-xenon mixture with a proper molar weight was recommended as working fluid for space nuclear reactor power with CBC conversion. It is essential to figure out how the component of helium-xenon mixture affects the net system efficiency, in order to provide reference for the selection of appropriate cycle working fluid. After a discussion of the physical properties of different helium-xenon mixtures, the related physical properties are studied to analyze their affection on the key parameters of CBC, including adiabatic coefficient, recuperator effectiveness and normalized pressure loss coefficient. Then the comprehensive thermodynamics of CBC net system efficiency is studied in detail considering different helium-xenon mixtures. The physical properties study reveals that at 0.7 MPa and 400 K, the adiabatic coefficient of helium-xenon mixture increases with increased molar weight, from 0.400 (pure helium) to 0.414 (pure xenon), while recuperator effectiveness firstly increases and then decreases with the increase of molar weight, and the normalized pressure loss coefficient increases monotonically with molar weight increases. The thermodynamic analysis results show that the adiabatic coefficient has less effect on the net system efficiency, while the net system efficiency increases with increased recuperator effectiveness, and the net system efficiency decreases with normalized pressure loss coefficient increases. Finally, the mixture of helium-8.6% xenon was adopted as working fluid, instead of pure helium, for ensuring less turbine mechanicals (turbine and compressor) stages, and resulting maximum recuperator effectiveness. At the given cold / hot side temperature of 400 / 1300 K, the net system efficiency can reach 29.18% theoretically.

High Temperature Reactor-Pebblebed Modules (HTR-PM) is a typical high-temperature gas cooled reactor (HTGR) [1]. Tritium is one of the most important radionuclides in reactors owing to its very harmful β-radiation and long half-life. In the HTR-PM, Silicon carbide (SiC) is the main barrier of triisotropic (TRISO) particles to prevent the diffusion of Tritium into the primary circuit [2]. When Tritium into the primary circuit and circulate to Steam Generator (SG), the Incoloy800H alloy is another important material to prevent the diffusion of Tritium into the secondary circuit [3]. When analyzing the source term of Tritium in HTR-PM primary and secondary circuit, it is important and necessary to know the diffusion behavior of Tritium in SiC and Incoloy800H and furthermore, the detail mechanism of diffusion is also essential, which could not be obtained from traditional phenomenological analysis and conservative estimation. In order to solve this challenge, a framework with ab-initio methods is established. In this paper, the detail theory of ab-initio theory and the actual usage in the calculation of the diffusion path, barrier energy are given firstly. And then, the most physical path and the minimum energy barrier will be determined, which can be considered as the diffusion activation energy. The calculated results of activation energy of Tritium in SiC and Incoloy800H are 0.442eV and 0.757eV respectively. Furthermore, the theoretical results are compared with the experimental data, and it is found that both are in agreement with each other. These results are very helpful for understanding the diffusion behaviors of Tritium in HTR-PM materials and can be used to guide the tritium source term analysis in HTR-PM, which are first studied from a micro perspective.

The vast majority of Nuclear Power Plants (NPPs) are equipped with water- and heavy-water-cooled reactors. Such NPPs have lower thermal efficiencies (30–36%) compared to those achieved at NPPs equipped with Advanced Gas-cooled Reactors (AGRs) (∼42%) and Sodium-cooled Fast Reactors (SFRs) (∼40%), and, especially, compared to those of modern advanced thermal power plants, such as combined cycle with thermal efficiencies up to 62% and supercritical-pressure coal-fired power plants — up to 55%. Therefore, NPPs with water- and heavy-water-cooled reactors are not very competitive with other power plants. Therefore, this deficiency of current water-cooled NPPs should be addressed in the next generation or Generation-IV nuclear-power reactors / NPPs. Very High Temperature Reactor (VHTR) concept / NPP is currently considered as the most efficient NPP of the next generation. Being a thermal-spectrum reactor, VHTR will use helium as a reactor coolant, which will be heated up to 1000°C. The use of a direct Brayton helium-turbine cycle was considered originally. However, technical challenges associated with the direct helium cycle have resulted in a change of the reference concept to indirect power cycle, which can be also a combined cycle. Along with the VHTR, Gas-cooled Fast Reactor (GFR) concept / NPP is also regarded as one of the most thermally efficient concept for the upcoming generation of NPPs. This concept was also originally thought to be with the direct helium power cycle. However, technical challenges have changed the initial idea of power cycle to a number of options including indirect Brayton cycle with He-N 2 mixture, application of SuperCritical (SC)-CO 2 cycles or combined cycles. The objective of the current paper is to provide the latest information on new developments in power cycles proposed for these two helium-cooled Generation-IV reactor concepts, which include indirect nitrogen-helium Brayton gas-turbine cycle, supercritical-pressure carbon-dioxide Brayton gas-turbine cycle, and combined cycles. Also, a comparison of basic thermophysical properties of helium with those of other reactor coolants, and with those of nitrogen, nitrogen-helium mixture and SC-CO 2 is provided.

Tristructural isotropic (TRISO) fuel particles are chosen as the major fuel type of High temperature gas cooled reactor (HTGR). The TRISO coated particle also acts as the first barrier for radioactivity retention. The performance of the TRISO coated particle has a significant influence on the safety of HTGR. A set of fuel performance analysis codes have been developed during the past decades. The main functions of these codes are conducting stress calculation and failure probability prediction. PANAMA is a widely used German version fuel performance analysis code, which simulates the mechanical performance of TRISO coated particle under normal and accident conditions. In this code, only a simple pressure vessel model is considered, which is insufficient in stress analysis and fuel failure rate prediction. Nowadays, efforts have been done to update the fuel performance model utilized in PANAMA code, and a new TRISO fuel performance analysis code, FFAT, is under developed. This paper describes the newly updated TRISO fuel performance model and presents some first results based on the updated model.

Based on the tri-isotropic (TRISO) fuel technology of high temperature gas cooled reactor, a new fuel concept for improving the accident tolerance of light water reactor (LWRs) named inert matrix dispersion pellet (IMDP) was proposed. Through the silicon carbide matrix and embedded TRISO fuel particles, the safety of the nuclear fuel could be enhanced. Recently, dummy IMDPs were fabricated by China General Nuclear Power Corporation (CGN) and thermal conductivity was tested. According to the tested data, a FEA model using ABAQUS combined its secondary development function was developed and benchmarked. Several influence factors of the effective thermal conductivity (ETC) of the IMDP were studied by the FEA model, such as burn-up, TRISO packing fraction and temperature. The heat transfer behaviors of IMDP and UO2 under typical normal PWR operating condition were also studied.

High Temperature Gas cooled Reactor (HTGR) that is one of GEN-IV reactor types enhances its safety adopting a passive cooling system, Reactor Cavity Cooling System (RCCS). When the active cooling system for the reactor core in a HTGR doesn’t work, the decay heat from the reactor core is transferred to the reactor vessel and the concrete wall of the reactor cavity, which are cooled by the RCCS. The RCCS consists of the vertical rectangular ducts, called risers, surrounding reactor vessel at a certain distance and chimneys that are connected to risers. Risers receiving the decay heat from the reactor vessel, the air inside the riser is heated and flows up to the chimney which accelerates exhalation of the air to the external atmosphere. The RCCS performance depends on the heat transfer rate inside the riser ducts, but the turbulent mixed convection that may occur in the riser ducts can complicate the verification. In this study, an experimental facility had been constructed to investigate heat transfer phenomena inside a riser duct and experiments with various heat flux and flow rate conditions were carried out. The experimental results showed that the turbulent mixed convection occurred for certain experimental conditions in the riser duct, which leads to heat transfer deterioration. Therefore, a correlation was suggested to predict how much the heat transfer could decrease compared to the forced convection. Large eddy simulation was carried out for CFD analysis. The heat transfer coefficients from LES showed consistent results with the correlation. The unique velocity profile and secondary flow was observed when the heat transfer was deteriorated. This study will contribute to the evaluation of the RCCS performance.

Nuclear reactor codes require validation with appropriate data representing the plant for specific scenarios. The thermal-hydraulic data is scattered in different locations and in different formats. Some of the data is in danger of being lost. A relational database is being developed to organize the international thermal hydraulic test data for various reactor concepts and different scenarios. At the reactor system level, that data is organized to include separate effect tests and integral effect tests for specific scenarios and corresponding phenomena. The database relies on the phenomena identification sections of expert developed PIRTs. The database will provide a summary of appropriate data, review of facility information, test description, instrumentation, references for the experimental data and some examples of application of the data for validation. The database has expandable platform and will accommodate different types of reactors and thermal-hydraulic tests. The current database platform includes scenarios for PWR, BWR, VVER, and specific benchmarks for CFD modelling data and is to be expanded to include references for molten salt reactors. There are place holders for high temperature gas cooled reactors, CANDU and liquid metal reactors. This relational database is called The International Experimental Thermal Hydraulic Systems database – TIETHYS and currently resides at Nuclear Energy Agency (NEA) of the OECD and is freely open to public access. Going forward the database will be extended to include additional links and data as they become available. The organizations conducting thermal-hydraulic tests are encouraged to submit their test data for inclusion in database for wider exposure and for preservation. https://www.oecd-nea.org/tiethysweb/

Helium blower is the core component of high temperature gas cooled reactor, which rotor is supported by active magnetic bearings (AMBs). The special advantage of AMB is that there is no contact between bearing and rotor, and this permits operation with small friction, long service life, no lubrication system, and no pollution to the helium environment. [1–3] Helium blower rotor is mainly composed of rotating shaft, impeller, motor, cooling blower and so on, which runs in an uneven temperature field that the impeller runs in a helium chamber of 250 degrees centigrade, and the motor housing’s outer surface temperature is 65 degrees centigrade. The temperature rises from standstill to stable operation will cause the thermal deformation of rotor and bearing, leading to the change of gap between rotor and bearing, which will lead to the change of electromagnetic force of AMB. The electromagnetic force determines the bearing stiffness and bearing damping of the AMB, so the change of temperature is the most important to the stiffness and damping of the AMB, which can affect the dynamic characteristics of the rotor. Through finite element method (FEM) to calculate the temperature field and displacement field of helium blower, the change of the gap of AMB and rotor is calculated. The rotor radial displacement orbits are obtained through numerical simulation, which are affected by thermal deformation. Finally, the results of numerical simulation are verified by experiments. The simulation and experiments both show that temperature rise can increase the vibration amplitude of rotor, so the influence of thermal deformation should be considered when designing the active magnetic bearings.

With the continuous development of the nuclear power technology in the world, all countries in the world are becoming more and more interested in the inherent safety of nuclear power technology, while the research and development of the spherical bed type high temperature gas cooled reactor nuclear power technology in China has formally catered to this demand. As a major national science and technology project, since the construction of the high temperature gas cooled reactor demonstration project (HTR-PM) since 2012, the civil construction of the nuclear island has been basically completed, the installation of equipment has been carried out orderly, and many process systems have entered debugging and operation stage gradually. As an important auxiliary process system, fuel handling and storage system for online refueling of the pebble bed high temperature gas cooled reactor, plays an important role in relation to the stable operation of the reactor. The main functions of the fuel handling and storage system are loading the fresh fuel elements and unloading the spent fuel elements which has reached its target burnup continuously for reactor operation, the spent fuel elements would be discharged into the spent fuel canister firstly, when the spent fuel storage canister is full of spent fuel, the canister would be sealed through welding method, and then the spent fuel canister would be transferred and stored in the spent fuel storage silo with the ground crane system. The fuel element of the pebble bed high temperature gas cooled reactor is spherical fuel element with graphite matrix, the fuel elements will have friction and collision with the inner wall of the pipeline in transporting process, which will produce graphite dust, the graphite dust should be removed continuously though filtration method, so as not to affect the fuel elements transportation in pipeline. This article focus on the production mechanism and filtering method of the graphite dust in graphite matrix fuel element transporting process in pipeline, to study the graphite dust removal technology, and then we could provide theoretical guidance for the design and operation of the key system and equipment for HTR-PM.

During the normal operation of a pebble bed gas-cooled reactor (PBR), the fuel pebbles undergo a multi-circulation on the basis of online burnup assay. In our last ICONE paper, we proposed a model to describe the relationship between online burnup assay and economy and safety of PBR. It was concluded that improvements on burnup assay accuracy could reduce fuel cost as well as the possibility that excessive burnup of fuel pebble results in unexpected radioactive discharge. However further work was expected on the burnup distribution of pebbles in and out of the core to precisely quantify the relationship. In this paper, the methodology to construct the burnup distribution of fuel pebbles in and out of the core is proposed. Firstly a model for pebble flow circulation is developed to provide a basic simulation framework. Then the irradiation history of fuel pebble could be tracked by combining pebble flow model and burn up calculation. The representative kinematic model and discrete element method (DEM) are introduced to numerically simulate the profiles of pebble flow. The classical batch-tracking methods as well as our newly-introduced DEM-tracking method are presented to perform the time-dependent analysis of pebble burnup. Overall with the burnup data obtained after going through multiple cycles, the burnup distribution of fuel pebbles in and out of the core could be reconstructed through the statistics result according to the pebble circulation model. Finally the quantification of the relationship between the pebble burnup assay and the economy and safety of the PBR would be more precise, thus providing implications on proposing reasonable requirements for accuracy of online burnup assay.

Performance of methane steam reforming reactor heated by helium for hydrogen production has been studied by numerical method. Results show with the increasing of reactant gas inlet velocity, temperature in the reactor drops, leading to the decreasing of methane conversion and hydrogen production rate. Methane conversion, hydrogen production and hydrogen production rate rise with the increasing of reactant gas inlet temperature, while the increasing degree of system thermal efficiency reduces. Besides, with helium inlet velocity rising, temperature in the reactor increases and reaction in the reactor becomes more sufficient. Therefore, methane conversion and hydrogen production also increase when helium inlet temperature of rises, but its influence is weaker compared to that of helium inlet velocity. In the process of methane steam reforming heated by high temperature gas cooled reactor (HTGR) for hydrogen production, lower reactant gas inlet velocity, suitable inlet temperature, higher inlet velocity and higher HTGR outlet temperature of helium are preferable.

The UK advanced gas cooled reactors (AGRs) use a graphite core with carbon dioxide gas as the primary coolant. There is a diaphragm above the core which separates re-entrant gas at lower temperature and higher pressure from that leaving the channel guide tubes at reactor outlet temperature. This diaphragm is known as the hot box dome. The dome is perforated to facilitate the passage of fuel and control rods into the core. The dome is fabricated in carbon-manganese steel and incorporates a number of full penetration welds which are post-weld heat treated (PWHT). The dome’s upper surface is insulated to protect it from gas at high temperature, intended to maintain the dome at a temperature of below 380°C. Since dome failure could conceivably result in gas by-passing and, hence, failing to adequately cool the core the original safety case claims that gross failure of the dome is incredible. More recently potential failure modes of the dome have been reviewed and various dome weld failure scenarios have been analysed and assessed to demonstrate a tolerance to the consequences of complete failure of certain welds. On this basis the dome could be shown to satisfy a lesser classification of high integrity, although no claim to reclassify the region has been made. Through-life temperature monitoring is carried out to demonstrate that the peak dome temperature remains below 380°C. This has shown evidence of rising temperatures, believed to arise from a reduction in the effectiveness of the upper surface insulation, an effect that was acknowledged by the original design. Work to investigate this effect has developed the understanding of the dome thermal environment which is far more complex than previously thought. The hottest parts of the dome are far smaller and more localised than previously thought, and lie immediately above the monitored locations. In order to support a case to operate for an extended life, it is now proposed that the upper temperature limit could safely be increased to 390°C. Structural integrity analyses and assessments have been carried out to support the proposed increase to 390°C and include a demonstration of the absence of a cliff-edge effect by assessing cases with the hottest parts of the dome at temperatures of 400 and 410°C. The work seeks to demonstrate adequate margins of safety against all potential failure modes. ASME III code assessment against primary stress limits has been used to guard against failure by plastic collapse and/or creep rupture. Creep-fatigue initiation assessments have been used to demonstrate margins against the formation of defects using the EDF Energy high temperature assessment procedure, R5. This has enabled the consideration of potential defects to be confined to those that might have formed during welding or PWHT and have been missed by extensive pre-service inspections. Notwithstanding the low likelihood that any such defects exist, with high confidence it may be postulated that any that do would be located in welds and be of limited size. Defect tolerance assessments have been carried out, including the calculation of limiting defect sizes in accordance with the EDF Energy R6 procedure and the growth of postulated defects by creep and fatigue using R5. Other failure and degradations mechanisms have been considered and eliminated as a potential threat by drawing on reviews of relevant operating experience on other reactors with similar materials and environments, and material property data from long term tests. This paper describes how the multi-facetted programme of work, which proposes a modification to an existing safety case, has been devised to explicitly address all conceivable modes of failure and demonstrate a robust argument against each one.

The deposition of graphite dust produced by the collision between graphite components would cause the security issue in HTGR and need to be analyzed. In this paper, a numerical calculation about the graphite dust’s deposition in the pebble bed reactor core of high-temperature gas cooled reactor was conducted. The three-dimensional steady-state solver was employed for the calculation of flow field and temperature field during simulation. The discrete phase model (DPM) and Lagrange method were applied for the simulation of graphite dust. Effects of parameters such as particle diameter, pebble bed layer numbers, inlet velocity and surface temperature of fuel elements on deposition of graphite dust are analyzed. The results indicated that a majority of particles deposit on the first layer pebble because of first layer’s shielding effect on nether layers. Moreover, deposition efficiency of graphite dust increases with increasing particles diameter and increasing inlet velocity due to greater motion inertia of particles. Compared with fewer layers, more layers structure would lead to larger deposition efficiency because of more opportunities for collision between graphite dust and pebbles, but the difference is not obvious. In addition, the higher surface temperature of fuel elements would cause lower deposition efficiency due to larger thermophoretic force which would drive particles to deviate from pebbles.

Investigations and numerical modelling are performed on the heat transfer to CO 2 at supercritical pressure under buoyancy affected conditions during heating in a vertical tube with inner diameter of 2 mm. Numerical modelling are carried out using several low Reynolds number (LRN) k-ε models, including the model due to Launder and Sharma (LS), Abe, Kondoh and Nagano (AKN), Myong and Kasagi (MK) models. The numerical results are compared with the corresponding experimental data and the predicted values using the semi-empirical correlation for convection heat transfer of supercritical fluids without deterioration. The abilities of various LRN models to predict the heat transfer to fluids at supercritical pressures under normal and buoyancy affected heat transfer conditions are evaluated. Detailed information related to the flow and turbulence is presented to get better understanding of the mechanism of the heat transfer deterioration due to buoyancy, as well as the different behavior of various LRN turbulence models in responding to the buoyancy effect, which gives clues in future model improvement and development to predict the buoyancy affected heat transfer more precisely and in a broader range of conditions as they come to be used to simulate the flow and heat transfer in various applications such as in the supercritical pressure water-cooled reactor (SCWR) and the supercritical pressure steam generator in the high temperature gas cooled reactor (HTR).

Online burnup measurement is a unique feature for pebble bed gas-cooled reactor and the fuel balls undergo a multi-circulation on the basis of the online burnup assay. It is ascertained that the accuracy of the online burnup assay is related with the economy and safety of pebble bed reactor. In the economical perspective, the burnup assay accuracy allow some part of pebbles that are below the burnup limit in the orifice to be discharged out of the core. In the safety view, the burnup assay allow some part of pebbles in the reactor core to exceed the burnup limit. In this paper, a mathematical model is proposed to establish the relationship. The model is implemented based on some reasonable theoretical hypothesis, and the influence of assay accuracy on the reactor safety and fuel cost issues are discussed based on the simulated results given by different assay accuracy. It is ascertained that improvements on burnup assay accuracy could save the fuel cost and improve the PBR economical efficiency as well as reduce the probability of radioactive release due to over-irradiation and enhance the safe reliability of PBR. Further research on the burnup distribution of pebbles in and out of the core and the burnup assay model are expected to provide some implications on proposing reasonable requirements for accuracy of online burnup assay.

There are more than 400 reactors in operation to generate electricity in the world, most of them are pressurized water reactors and boiling water reactors, which generate great amount of spent fuel every year. The residual heat power of the spent fuel just discharged from the reactor core is high, it is required to store the spent fuel in the spent fuel storage pool at the first 5 years after discharged from the reactor, and then the spent fuel could be moved to the interim storage facility for long term storage, or be moved to the factory for final treatment. In the accident of the Fukushima in 2011, the spent fuel pool ruptured, which led to the loss of coolant accident, it was very danger to the spent fuel assemblies stored in the pool. On the other hand, the spent fuel stored in the dry storage facility was safe in the whole process of earthquake and tsunami, which proved inherent safety of the spent fuel dry storage facility. In china, the High Temperature gas cooled Reactor (HTR) is developing for a long time in support of the government. At the first stage, HTR-10 with 10MW thermal power was designed and constructed in the Institute of Nuclear Energy Technology (INET) of Tsinghua University, and then the High Temperature Reactor-Pebble bed Modules (HTR-PM) is designed to meet the commercial application, which is in constructing process in Shandong Province. HTR has some features of the generation four nuclear power plant, including inherent safety, avoiding nuclear proliferation, could generate high temperature industrial heat, and so on. Spherical fuel elements would be used as fuel in HTR-PM, there are many coating fuel particles separated in the fuel element. As the fuel is different for the HTR and the PWR, the fuel element would be discharged into the appropriate spent fuel canister, and the canister would be stored in the appropriate interim storage facility. As the residual power density is very low for the spent fuel of HTR, the spent fuel canister could be cooled with air ventilation without water cooling process. The advantage of air cooling mode is that it is no need to consider the residual heat removal depravation due to loss of coolant accident, so as to increase the inherent safety of the spent fuel storage system. This paper introduced the design, arrangement and safety characteristics of the spent fuel storage well of HTR-PM. The spent fuel storage wells have enough capacity to hold the total spent fuel canisters for the HTR-PM. The spent fuel storage facility includes several storage wells, cold intake cabin, hot air discharge cabin, heat shield cylinders, well lids and so on. The cold intake cabin links the inlets of all the wells, which would be used to import cold air to every well. The hot air discharge cabin links the outlets of all the wells, which would be used to gather heated air discharged from every well, the heated air would be discharged to the atmosphere through the ventilating pipe at the top of the hot air cabin. The design of the spent fuel storage well and the ventilating pipe could discharge the residual heat of the spent fuel canisters in the storage wells, which could ensure the operating safety of the spent fuel storage system.

Coal is burnt in actual Fischer-Tropsch process to provide heat for steam generation required for coal gasification in a coal conversion-to-liquid (CTL) fuel process. In light of the world energy demand growth and consumption rate of the world reserves, the use of alternative energy resource for heat supply to a CTL process seems to be more attractive. It is expected that the use of alternative energy on the current process will not only contribute to the extension of the coal reserves life-time but will also significantly contribute to reduce the carbon foot-print. In this study, a high-temperature gas cooled reactor is considered to supply heat to a 240 MWth CTL process and a HPHE is conceptually designed for physical dimensions and its dynamic behaviour is observed. The process heat superheated steam is generated in a once-through heat pipe heat exchanger steam generator using sodium as working fluid. A heat pipe heat exchanger once through steam generator is modelled and simulated with the aim of predicting the transient and dynamic behaviour of the heat transfer process occurring in the HPHE.