The intercooled cycle (IC) is a simplified novel proposal for generation IV nuclear power plants (NPP) based on studies demonstrating efficiencies of over 45%. As an alternative to the simple cycle recuperated (SCR) and the intercooled cycle recuperated (ICR), the main difference in configuration is no recuperator, which reduces its size. It is expected that the components of the IC will not operate at optimum part power due to seasonal changes in ambient temperature and grid prioritization for renewable sources. Thus, the ability to demonstrate viable part load performance becomes an important requirement. The main objective of this study is to derive off-design points (ODPs) for a temperature range of −35 °C to 50 °C and core outlet temperatures (COTs) between 750 °C and 1000 °C. The ODPs have been calculated using a tool designed for this study. Based on the results, the intercooler changes the mass flow rate and compressor pressure ratio (PR). However, a drop of ∼9% in plant efficiency, in comparison to the ICR (6%) was observed for pressure losses of up to 5%. The reactor pressure losses for IC have the lowest effect on plant cycle efficiency in comparison to the SCR and ICR. Characteristic maps are created to support first-order calculations. It is also proposed to consider the intercooler pressure loss as a handle for ODP performance. The analyses brings attention to the IC an alternative cycle and aids development of cycles for generation IV NPPs specifically gas-cooled fast reactors (GFRs) and very-high-temperature reactors (VHTRs), using helium.

The effect of structural state (solution annealed (SA) and after 40% cold work (CW)) and surface finishing (turning, grinding, and polishing) on the corrosion behavior of austenitic 1.4970 (15-15 Ti) steel in flowing (2 m/s) Pb-Bi eutectic containing 10 −7 mass% dissolved oxygen at 400 °C and 10 −6 mass% O at 500 °C is investigated. At 400 °C for ∼13,000 h, the corrosion losses are minor for steel in both structural states and for surfaces finished by turning and grinding—a thin Cr-based oxide film is formed. In contrast, the polished surface showed initiation of solution-based corrosion attack with the formation of iron crystallites and preferential propagation along the grain boundaries. The depth of corrosion attack does not exceed 10 μ m after ∼13,000 h. At 500 °C for 2000 h, the samples in both structural states showed general slight oxidation. Cold-worked steel underwent a severe groove-type and pit-type solution-based attack of 170 μ m in maximum depth, while the SA sample showed only sporadic pit-type corrosion attack to the depth of 45 μ m in maximum.

The continuous generation of graphite dust particles in the core of a high-temperature reactor (HTR) is one of the key challenges of safety during its operation. The graphite dust particles emerge from relative movements between the fuel elements or from contact to the graphitic reflector structure and could be contaminated by diffused fission products from the fuel elements. They are distributed from the reactor core to the entire reactor coolant system. In case of a depressurization accident, a release of the contaminated dust into the confinement is possible. In addition, the contaminated graphite dust can decrease the life cycle of the coolant system due to chemical interactions. On one hand, the knowledge of the behavior of graphite dust particles under HTR conditions using helium as the flow medium is a key factor to develop an effective filter system for the discussed issue. On the other hand, it also provides a possibility to access the activity distribution in the reactor. The behavior can be subdivided into short-term effects like transport, deposition, remobilization and long-term effects like reactions with material surfaces. The Technische Universität Dresden has installed a new high-temperature test facility to study the short-term effects of deposition of graphite dust particles. The flow channel has a length of 5 m and a tube diameter of 0.05 m. With helium as the flow medium, the temperature can be up to 950 °C in the channel center and 120 °C on the sample surface, the Reynolds number can be varied from 150 up to 1000. The particles get dispersed into the accelerated and heated flow medium in the flow channel. Next, the aerosol is passing a 3 m long adiabatic section to ensure homogenous flow conditions. After passing the flow straightener, it enters the optically accessible measurement path made from quartz glass. In particular, this test facility offers the possibility to analyze the influence of the thermophoretic effect separately. For this, an optionally cooled sample can be placed in the measuring area. The thickness of the particle layer on the sample is estimated with a three-dimensional laser scanning microscope. The particle concentration above the sample is measured with an aerosol particle sizer (APS). Particle image velocimetry (PIV) detects the flow-velocity field and provides data to estimate the shear velocity. In combination with the measured temperature-field, all necessary information for the calculation of the particle deposition and particle relaxation times are available. The measurements are compared to results of theoretical works from the literature. The experimental database is relevant especially for computational fluid dynamics (CFD)-developers, for model development, and model verification. A wide range of phenomena like particle separation, local agglomeration of particles with a specific particle mass, and selective remobilization can be explained in this way. Thus, this work contributes to a realistic analysis of nuclear safety.

The effective thermal diffusivity and conductivity of pebble bed in the high temperature gas-cooled reactor (HTGR) are two vital parameters to determine the operating temperature and power in varisized reactors with the restriction of inherent safety. A high-temperature heat transfer test facility and its inverse method for processing experimental data are presented in this work. The effective thermal diffusivity as well as conductivity of pebble bed will be measured at temperature up to 1600 °C in the under-construction facility with the full-scale in radius. The inverse method gives a global optimal relationship between thermal diffusivity and temperature through those thermocouple values in the pebble bed facility, and the conductivity is obtained by conversion from diffusivity. Furthermore, the robustness and uncertainty analyses are also set forth here to illustrate the validity of the algorithm and the corresponding experiment. A brief experimental result of preliminary low-temperature test is also presented in this work.

One of the key elements in probabilistic risk assessment is the identification and characterization of uncertainties. This paper suggests a procedure to identify influencing factors for uncertainty in source term evaluation, which are important to risk of public dose. We propose the following six steps for the identification in a systematic manner in terms of completeness and transparency of the results using both a logic diagram based on basic equations and expert opinions: (1) identification of uncertainty factors based on engineering knowledge of accident scenario analysis; (2) derivation of factors at the level of physical phenomena and variable parameters by expansion of dynamic equation for the system and scenario to be investigated, (3) extraction of uncertainties in variable parameters; (4) selection of important factors based on sensitivity study results and engineering knowledge; (5) identification of important factors for uncertainty analysis using expert opinions; and (6) integration of selected factors in the aforementioned steps. The proposed approach is tested with a case study for a risk-dominant accident scenario in direct cycle high-temperature gas-cooled reactor (HTGR) plant. We use this approach for evaluating the fuel temperature in terms of reactor dynamics and thermal hydraulic characteristics during a depressurized loss-of-forced circulation (DLOFC) accident with the failure of mitigation systems such as control rod systems (CRS) in a representative HTGR plan. In total, six important factors and 16 influencing factors were successfully identified by the proposed method in the case study. The selected influencing factors can be used as input parameters in uncertainty propagation analysis.

This work focuses on the effect of dissolved oxygen concentration in liquid lead-bismuth eutectic (LBE) on the onset of dissolution corrosion in a solution-annealed 316 L austenitic stainless steel. Specimens made of the same 316 L stainless steel heat were exposed for 1000 h at 450 °C to static liquid LBE with controlled concentrations of dissolved oxygen, i.e., 10 −5 , 10 −6 , and 10 −7 mass%. The corroded 316 L steel specimens were analyzed by scanning electron microscopy (SEM) and energy dispersive X-ray spectroscopy (EDS). A complete absence of dissolution corrosion was observed in the steel specimens exposed to liquid LBE with 10 −5 and 10 −6 mass% oxygen. In the same specimens, isolated “islands” of FeCr-containing oxides were also detected, indicating the localized onset of oxidation corrosion under these exposure conditions. On the other hand, dissolution corrosion with a maximum depth of 59 μ m was detected in the steel specimen exposed to liquid LBE with 10 −7 mass% oxygen. This suggests that the threshold oxygen concentration associated with the onset of dissolution corrosion in this 316 L steel heat lies between 10 −6 and 10 −7 mass% oxygen for the specific exposure conditions (i.e., 1000 h, 450 °C, static liquid LBE).

The understanding of the radial distribution of temperature in a fuel pellet, under normal operation and accident conditions, is important for a safe operation of a nuclear reactor. Therefore, in this study, we have solved the steady-state heat conduction equation, to analyze the temperature profiles of a 12 mm diameter cylindrical dispersed nuclear fuels of U 3 O 8 -Al, U 3 Si 2 -Al, and UN-Al operating at 597 ° C. Moreover, we have also derived the thermal conductivity correlations as a function of temperature for U 3 Si 2 , uranium mononitride (UN), and Al. To evaluate the thermal conductivity correlations of U 3 Si 2 , UN, and Al, we have used density functional theory (DFT) as incorporated in the Quantum ESPRESSO (QE) along with other codes such as Phonopy, ShengBTE, EPW (electron-phonon coupling adopting Wannier functions), and BoltzTraP (Boltzmann transport properties). However, for U 3 O 8 , we utilized the thermal conductivity correlation proposed by Pillai et al. Furthermore, the effective thermal conductivity of dispersed fuels with 5, 10, 15, 30, and 50 vol %, respectively of dispersed fuel particle densities over the temperature range of 27–627 °C was evaluated by Bruggman model. Additionally, the temperature profiles and temperature gradient profiles of the dispersed fuels were evaluated by solving the steady-state heat conduction equation by using Maple code. This study not only predicts a reduction in the centerline temperature and temperature gradient in dispersed fuels but also reveals the maximum concentration of fissile material (U 3 O 8 , U 3 Si 2 , and UN) that can be incorporated in the Al matrix without the centerline melting. Furthermore, these predictions enable the experimental scientists in selecting an appropriate dispersion fuel with a lower risk of fuel melting and fuel cracking.

A fast breeder reactor (FBR) is considered as the promising technology in terms of load reduction on the environment, because the FBR has capability to improve usage efficiency of uranium resources and can reduce high-level radioactive waste which needs to be managed for millions of years. A cold trap is one of the important components in the FBR to control the impurity concentration of the liquid sodium. For accurate evaluation of the cold trap performance, we have been proposing the three-dimensional (3D) numerical analysis method of the cold trap. In this method, the evaluation of the impurity precipitation phenomena on the surface of the mesh wire of the cold trap is the key. For this, the numerical analysis method which is based on the lattice kinetic scheme (LKS) has been proposed. In order to apply the LKS to the impurity precipitation simulation of the cold trap, two models (the low Reynolds number model and the impurity precipitation model) have been developed. In this paper, we focused on the validation of these models. To confirm the validity of the low Reynolds number model, the Chapman–Enskog analysis was applied to the low Reynolds number model. As a result, it has been theoretically confirmed that the low Reynolds number model can recover the correct macroscopic equations (incompressible Navier–Stokes equations) with small error. The low Reynolds number model was also validated by the numerical simulation of two-dimensional (2D) channel flow problem with the low Reynolds number conditions which correspond to the actual cold trap conditions. These results have confirmed that the error of the low Reynolds number model is ten times smaller than that of the original LKS. The validity of the impurity precipitation model was investigated by the comparison to the precipitation experiments. In this comparison, the mesh convergence study was also conducted. These results have confirmed that the proposed impurity precipitation model can simulate the impurity precipitation phenomena on the surface of the mesh wire. It has been also confirmed that the proposed impurity precipitation model can simulate the impurity precipitation phenomenon regardless of the cell size which were tested in this investigation.

Studies on debris bed formation behavior are important for improved evaluation of core relocation and debris bed coolability that might be encountered in a core disruptive accident (CDA) of sodium-cooled fast reactors (SFR). Motivated to clarify the flow-regime characteristics underlying this behavior, both experimental investigations and empirical-model development are being performed at the Sun Yat-sen University in China. As for the experimental study, several series of simulated experiments are being conducted by discharging various solid particles into water pools. To obtain a comprehensive understanding, a variety of experimental parameters, including particle size (0.000125– 0.008 m), particle density (glass, aluminum, alumina, zirconia, steel, copper, and lead), particle shape (spherical and nonspherical), and water depth (0–0.8 m) along with the particle release pipe diameter (0.01–0.04 m) were varied. It is found that due to the different interaction mechanisms between solid particles and water pool, four kinds of flow regimes, termed, respectively, as the particle-suspension regime, the pool-convection dominant regime, the transitional regime, and the particle-inertia dominant regime, were identifiable. As for the empirical-model development, aside from a base model which is restricted to predictions of spherical particles, in this paper considerations on how to cover more realistic conditions (esp. debris of nonspherical shapes) are also discussed. It is shown that by coupling the base model with an extension scheme, respectable agreement between experiments and model predictions for regime transition can be achieved for both spherical and nonspherical particles given our current range of conditions.

The Canadian super critical water-cooled reactor (SCWR) concept requires materials to operate at higher temperatures than current generation III water-cooled reactors. Materials performance after radiation damage is an important design consideration. Materials that are both corrosion resistant and radiation damage tolerant are required. This paper summarizes the operating conditions including temperature, neutron flux, and residence time of in-core Canadian SCWR components. The focus is on the effects of irradiation on in-core components, including those exposed to a high neutron flux in the fuel assembly, the high pressure boundary between coolant and moderator, as well as the low-temperature, low-flux calandria vessel that contains the moderator. Although the extreme conditions and the broad range of SCWR in-core operating conditions present significant materials selection challenges, candidate alloys that can meet the performance requirements under most in-core conditions have been identified. However, for all candidate materials, insufficient data are available to unequivocally ensure acceptable performance and experimental irradiations of candidate core materials will be required. Research programs are to include out-of-pile tests on un-irradiated and irradiated alloys. Ideally, in-flux studies at appropriate temperatures, neutron spectrum, dose rate, duration, and coolant chemistry will be required. Characterization of the microstructure and the mechanical behavior including strength, ductility, swelling, fracture toughness, cracking, and creep on each of the in-core candidate materials will ensure their viability in the Canadian SCWR.

In recent years, the morphological characteristics and stabilization methods of free interface in liquid windowless target become hot research topics in accelerator driven subcritical system (ADS). Based on the structure design of a certain windowless spallation target, computational fluid dynamics (CFD) software of CFX was used to simulate and analyze its free interface character. The method of k – ε turbulence, cavitation, and volume of fluid (VOF) model was used to study the flow characteristic of liquid Lead-Bismuth eutectic (LBE) alloy with cavitation phase change and to analyze the free interface morphology characteristics of coolant in the target area. It is concluded that the target region forms two stable free interfaces when fluid outlet pressure is in the range of 10–40 kPa and fluid entrance velocity is in the range of 0.5–1.2 m/s. The flow field near the free interface structure is complex. The vortex region appears, and the disorders in the vortex flow pattern lead to fluctuation of the free interface. After the study of stable free interface morphology establishing process, heat transfer characteristic of windowless target was further analyzed.

The Canadian supercritical water-cooled reactor (SCWR) design is part of Canada's Generation IV reactor development program. The reactor uses batch fueling, light water above the thermodynamic critical point as a coolant and a heavy water moderator. The design has evolved considerably and is currently at the conceptual design level. As a result of batch fueling, a certain amount of excess reactivity is loaded at the beginning of each fueling cycle. This excess reactivity must be controlled using a combination of burnable neutron poisons in the fuel, moderator poisons, and control blades interspersed in the heavy water moderator. Recent studies have shown that the combination of power density, high coolant temperatures, and reactivity management can lead to high maximum cladding surface temperatures (MCST) and maximum fuel centerline temperatures (MFCLT) in this design. This study focuses on improving both the MCST and the MFCLT through modifications of the conceptual design including changes from a 3 to 4 batch fueling cycle, a slightly shortened fuel cycle (although exit burnup remains the same), axial graded fuel enrichment, fuel-integrated burnable neutron absorbers, lower reactivity control blades, and lower reactor thermal powers as compared to the original conceptual design. The optimal blade positions throughout the fuel cycle were determined so as to minimize the MCST and MFCLT using a genetic algorithm and the reactor physics code PARCS. The final design was analyzed using a fully coupled PARCS-RELAP5/SCDAPSIM/MOD4.0 model to accurately predict the MCST as a function of time during a fueling cycle.

This work attempts to investigate the thermal hydraulic safety of lithium lead ceramic breeder (LLCB) test blanket system (TBS) in International Thermonuclear Experimental Reactor (ITER) with the help of modified thermal hydraulic code relap/scdapsim/mod 4.0. The design basis accidents, in-vessel and ex-vessel loss of coolant of first wall (FW) of test blanket module (TBM) are analyzed for this safety assessment. The sequence of accidents analyzed was started with postulated initiating events (PIEs). A detailed modeling of first wall helium cooling system (FWHCS) loop and lithium lead cooling system (LLCS) is presented. The analysis of steady-state normal operation along with 10 s power excursion before the accident is also discussed in order to better understanding of initial condition of accidents. The analysis discusses a number of safety concerns and issues that may result from the TBM component failure, such as vacuum vessel (VV) pressurization, TBM FW temperature profile, passive decay heat removal capability of TBM structure, pressurization of port cell and Tokomak cooling water system vault annex (TCWS-VA) and to check the capability of passive safety system (vacuum vessel pressure suppression system (VVPSS)). The analysis shows that in these accident scenarios, the critical parameters have reasonable safety margins.

On Mar. 11, 2011, the 2011 off the Pacific coast of Tohoku Earthquake of magnitude 9.0 occurred. When the great earthquake occurred, the high temperature engineering test reactor (HTTR) had been stopped under the periodic inspection and maintenance of equipment and instruments. A comprehensive integrity evaluation was carried out for the HTTR facility because the maximum seismic acceleration observed at the HTTR exceeded the maximum value of design basis earthquake. The concept of comprehensive integrity evaluation is divided into two parts. One is the “visual inspection of equipment and instruments.” The other is the “seismic response analysis” for the building structure, equipment and instruments using the observed earthquake. All equipment and instruments related to operation were inspected in the basic inspection. The integrity of the facilities was confirmed by comparing the inspection results or the numerical results with their evaluation criteria. As the results of inspection of equipment and instruments associated with the seismic response analysis, it was judged that there was no problem for operation of the reactor, because there was no damage and performance deterioration. The integrity of HTTR was also supported by the several operations without reactor power in cold conditions of HTTR in 2011, 2013, and 2015. Additionally, the integrity of control rod guide blocks was also confirmed visually when three control rod guide blocks and six replaceable reflector blocks were taken out from reactor core in order to change neutron startup sources in 2015.

The whole core model of China experimental fast reactor (CEFR) is established according to the parameters of China experimental fast reactor, which are given by technical publication from the International Atomic Energy Agency (IAEA-TECDOC-1531), and the physical parameters of CEFR are simulated with the Monte Carlo N-particle code (MCNP4a). The calculation results are compared with the data contained in the safety analysis report of CEFR. The calculation results are consistent with the design values, which successfully demonstrate the acceptable fidelity of the MCNP model. The MCNP model will be further refined and applied for nuclear safety review of the CEFR in the future.

Due to the advantages of small volume, light weight, and long-time running, nuclear reactor can provide an ideal energy source for space crafts. In this paper, two small compact prismatic nuclear reactors with different core block materials are presented, which have a thermal power of 5 MW for 10 years of equivalent full power operation. These two reactors use Mo-14%Re alloy or nuclear grade graphite IG110 as core block material, loaded with 50% and 39.5% enriched uranium nitride (UN) fuel and cooled by helium, whose inlet/outlet temperature of the reactor and operational pressure are 850/1300 K and 2 MPa, respectively. High temperature helium flowing out of the reactor can be used as the working medium for closed Brayton cycle power conversion with high efficiency (more than 20%). Neutronics analyses of reactors for the preliminary design in this paper are performed using reactor Monte Carlo (RMC) code developed by Tsinghua University. Both the reactors have enough initial excess reactivity to ensure 10 years of full power operation without refueling, have safety margin for reactor shutdown with one control drum failed, and remain subcritical in the submersion accident. Finally, the two reactors are compared in aspect of the 235 U mass and the total reactor mass.

Burnable poison (BP) is used to control excess reactivity in supercritical water cooled reactor (SCWR). It helps reduce the number of control rods. Over all BP designs, the design in which rare-earth oxide mixes with fuel is widely used in SCWR. BP has influence on fuel assembly neutronics performance. After comparing four kinds of rare-earth oxide, Er 2 O 3 is chosen as BP for the annular fuel assembly. The effect of different BP loading patterns on assembly power distribution is analyzed. The safety of annular fuel assembly is estimated with different BP contents. Core performance with and without BP is compared. The results had shown that the core radial power peaking factor decreased after introducing BP. It was also shown that the core axial power peaking factor increased, and the power peak moved toward the top of the core. The reason of this effect was studied. Two optimizations were given based on this study: decreasing the temperature of lower plenum and increasing the gradients of axial enrichments. By applying these optimizations, core axial power peaking factor and maximum cladding surface temperature decreased.

Supercritical fluids (SCFs) become more and more important in various engineering applications. In nuclear power systems, SCFs are considered as coolant of the reactor core such as the supercritical water-cooled reactor (SCWR), superconducting magnets and blankets in the fusion reactors, or as fluid in the energy conversion systems of the next generation nuclear reactors. Accurate determination of heat transfer and the temperature of the structural material (e.g., fuel rod cladding) is of crucial importance for the system design. Thus, extensive studies on heat transfer to SCFs have been carried out in the past five decades and are still ongoing worldwide. However, no breakthrough is recognized or expected in the near future. In this paper, the status, main challenges, and future R&D needs are briefly reviewed. Three aspects are taken into consideration, i.e., experimental studies, numerical analysis, and model development for the prediction of heat transfer coefficient (HTC). Several key challenges and also the important subjects of the future R&D needs are identified. They are (a) data base for turbulence quantities, (b) multisolution of wall temperature, (c) extensive Reynolds-averaged Navier–Stokes (ERANS) method, and (d) new prediction method for HTC.

While supercritical water is a perfect coolant with excellent heat transfer, a temporary decrease of the system pressure to subcritical conditions, either during intended transients or by accident, can easily cause a boiling crisis with significantly higher cladding temperatures of the fuel assemblies. These conditions have been tested in an out-of-pile experiment with a bundle of four heated rods in the supercritical water multipurpose loop (SWAMUP) facility coconstructed by CGNPC and SJTU in China. Some of the transient tests have been simulated at KIT with a one-dimensional (1D) matlab code, assuming quasi-steady-state flow conditions, but time dependent temperatures in the fuel rods. Heat transfer at supercritical and at near-critical conditions was modeled with a recent look-up table of Zahlan (2015, “Derivation of a Look-Up Table for Trans-Critical Heat Transfer in Water Cooled Tubes,” Ph.D. dissertation, University of Ottawa, Ottawa, ON, Canada.), and subcritical film boiling was modeled with the look-up table of Groeneveld et al. (2003, “A Look-Up Table for Fully Developed Film Boiling Heat Transfer,” Nucl. Eng. Des., 225 (1), pp. 83–97.). Moreover, a conduction controlled rewetting process was included in the analyses, which is based on an analytical solution of Schulenberg and Raqué (2014, “Transient Heat Transfer During Depressurization From Supercritical Pressure,” Int. J. Heat Mass Transfer, 79 (12), pp. 233–240.). The method could well reproduce the boiling crisis during depressurization from supercritical to subcritical pressure, including rewetting of the hot zone within some minutes, but the peak temperature was somewhat under-predicted. Tests with a lower heat flux, which did not cause such phenomena, could be predicted as well. In another test with increasing pressure, however, a boiling crisis was also observed at a heat flux, which was significantly lower than the critical heat flux (CHF) predicted by the CHF look-up table of Groeneveld et al. (2007, “The 2006 CHF Look-Up Table,” Nucl. Eng. Des., 237 (15–17), pp. 1909–1922.). The paper is summarizing the physical models and the numerical approach. Comparison with experimental data is used to discuss the applicability of the method for the design of supercritical water-cooled reactors (SCWR).