In this study, we analyze different methods for the generation of homogenized few-group cross sections (XS) using continuous-energy Monte Carlo codes Serpent 2.1.29 and MCU–PTR for full-core diffusion calculation of small-size light water research reactors. Few-group reaction cross sections, scattering matrices and diffusion coefficients generated in an infinite lattice model with and without B1-leakage correction of spectra, and in full-scale 3D core calculations, are compared. Few-group XS are used in the TIGRIS diffusion nodal simulator to calculate the simplified cores with IRT-3 M fuel and a light water reflector. Special attention is paid to the impact of diffusion coefficients on the results of diffusion calculations. We compare the following different methods for the generation of diffusion coefficients: the usual “out-scatter” approximation, the transport correction for hydrogen implemented in Serpent, and the migration method implemented in MCU–PTR.

The gradual corrosion and activation of the cooling loop with time in water-cooled reactor is one of the safety and environmental issues. Activated corrosion products (ACPs) are the dominant radiation hazard in water-cooled reactor under normal operation conditions, and directly determine Occupational Radiological Exposure (ORE) during operation and maintenance. ACPs have different phases: ions, particles, oxides layer, deposits layer. Among them, particles are an important movable source item in coolant due to their unique behavioral characteristics and they can deposit on pipe surface to form deposits layer. In this paper, code CATE is upgraded to V3.0 based on CATE V2.1 and a four-phase three-node model is proposed for calculating radioactivity of multi-phase ACPs. For code testing, the blanket cooling loop of International Thermonuclear Experimental Reactor (ITER) is simulated, and the results obtained by CATE V3.0 agree reasonably well with published data calculated by other international codes PACTITER and TRACT, which means CATE V3.0 is available and credible on ACPs analysis of water-cooled reactor. Then ACPs in the primary loop of a typical PWR is analyzed using CATE V3.0 and the sensitivity of some important parameters on ACPs is analyzed. The results showed that ions play an important role in corrosion products mass of coolant and oxides layer plays a decisive role in corrosion products mass of pipe surface. The main contributors to radioactivity of coolant, in-flux pipe, out-flux pipe are separately ions, oxides layer and deposits layer. Long-lived nuclides such as Co-58 and Co-60 are major contributors to total radioactivity after shutdown. From the sensitivity analysis, it can be seen that using low-cobalt materials and controlling pH of coolant rationally can help reduce source term of ACPs.

This paper presents the execution and the results of a rod ejection benchmark inspired by the UAM benchmarks II-2b: cross sections uncertainties propagation during a rod ejection transient in a mini-core (3 × 3 PWR-type assemblies), without feedback. Performing transient computations in two groups time-dependent neutron diffusion approximation while starting from a steady-state power of 0.1409 MW, we obtained a mean power peak around 0.25 MW with a standard deviation of, nearly, 1.5%. The time-dependent standard deviation curve follows the time-dependent mean power curve and the relative standard deviation, in a percentage of the mean power, increases with the mean power peak value. We found, too, that the first-group diffusion coefficient and the first-group absorption cross section are, practically, the only contributors to the total power peak variance. Drawbacks and findings from the exercise provide essential input for uncertainty propagation in a further multi-physics simulation of a postulated rod ejection accident.

The proliferation resistance (PR) of an inert matrix fuel (IMF) in the transuranic nuclear fuel cycle (NFC) of a high temperature gas cooled reactor is evaluated relative to the uranium and plutonium mixed-oxide (MOX) NFC of a light water reactor using PRAETOR code and sixty-eight input attributes. The objective is to determine the impacts of chemical stability of IMF and fuel irradiation on the PR. Specific material properties of the IMF, such as lower plutonium content, carbide ceramics coating, and absence of 235U, contribute to enhance its relative PR compared to MOX fuel. The overall PR value of the fresh IMF (an unirradiated direct use material with a one-month diversion detection timeliness goal) is nearly equal to that of the spent MOX fuel (an irradiated direct use nuclear material with a three-month diversion detection timeliness goal). Final results suggest a reduced safeguards inspection frequency to manage the IMF.

Tritium management is a vital factor that needs to be considered before a molten salt reactor (MSR) or fluoride-salt-cooled high-temperature reactor (FHR) is constructed. The analysis of tritium transport characteristics under steady and transient conditions is essential in the conceptual design of MSRs and FHRs. In this study, tritium transport characteristics analysis is conducted for a fluoride-salt-cooled pebble bed reactor, i.e., thorium molten salt reactor with solid fuel (TMSR-SF) under accident conditions. The tritium transport characteristics analysis (TAPAS) code is improved with a power model and heat-and-mass transfer model to adapt to TMSR-SF. The transient responses of the tritium transport characteristics of TMSR-SF under the unprotected reactivity insertion accident (URIA) and unprotected overcooling accident (UOC) are determined. It is found that the amount of tritium fluoride (TF) adsorbed on graphite varies, indicating the dynamic equilibrium between the TF adsorption and desorption on graphite. Due to the nuclear density variations of Li-6 and Be-9 in the fluoride salt under the accident conditions, the tritium production rate increases and decreases under the UOC and URIA conditions, respectively. Generally, under these two accident conditions, the tritium permeation rates are lower than that of the normal operation condition. In these cases, it is unnecessary to develop extra safety facilities for tritium control. This study contributes to the tritium management of TMSR-SF.

Passive autocatalytic recombiners (PARs) have become one of the primary measures to mitigate the hydrogen risk for severe accidents in current and advanced water-cooled nuclear power plants (NPPs). The pgm (platinum group metals) coating of the catalyst inside the PAR is capable to oxidize hydrogen into water vapor as well as to convert carbon monoxide into carbon dioxide. The optimal efficiency of conversion requires a significant surplus of oxygen compared to the stoichiometric oxidization. A test series under containment typical boundary conditions was performed in the THAI facility (Thermal-hydraulics, Hydrogen, Aerosols and Iodine) to investigate the PAR performance under the presence of carbon monoxide. Ratios of injection mass flow rates of hydrogen and carbon monoxide into the test facility are investigated between 4:1 and 7:1, based on hypothetical release of CO by the molten core concrete interaction (MCCI). The findings are related to database on PAR performance carried out in previous OECD/NEA projects, THAI (2007–2009), THAI-2 (2011–2014), and THAI-3 (2016–2019). Special emphasis is placed on the investigation of the PAR under conditions of oxygen starvation, taking also into account a potential PAR poisoning by carbon monoxide, and under potential PAR induced ignitions at elevated hydrogen and carbon monoxide concentrations.

The design and deployment of integrated modular water reactors (IMRs) and evolutionary reactors whose operations are based on passive systems are on the rise. The natural circulation (NC) based passive systems being adopted rely on natural convection, buoyancy driven-flow and vapour condensation. In this work, a numerical approach that can determine the required conditions for establishing this widely applied thermal hydraulic phenomenon in simple turbulent two phase nuclear passive system is presented. Through the method, other relevant thermal-hydraulic parameters which are of interest in safety and reliability analyses of such NC-based systems can also be obtained. The case of the System-integrated Modular Advanced ReacTor (SMART) based on the method was demonstrated. Furthermore, the validity of the approach was proven using the test data and simulation results of a NC-based experimental facility. The method is simple and easy to apply to NC-based systems of similar configurations.

The paper investigates the lubrication performances of water film under the consideration of different lubrication models in the primary circuit loop system of nuclear power plant. Variations between lubrication performances and rotating speeds, eccentricity ratios and lubrication models are obtained. Research results show that rotating speeds and lubrication models have significant influences on the lubrication performances. There exist optimum eccentricity ratios at which the bearing load carrying capacity and friction characteristics can strike the balance point. Research conclusions have guiding significances on the structure designs and optimization for such bearings in the primary circuit loop system.

This paper describes a multi-scale thermal–hydraulic coupling system, which combines the capabilities of the open-source TrioCFD code and of the sub-channel code SubChanFlow (SCF) aiming to improve the description and prediction of the multi-scale thermal–hydraulic physical phenomenon inside the reactor vessel. This is a parallel coupling system based on the ICoCo (Interface for Code Coupling) concept, where each code is first wrapped by the ICoCo interface and then it is compiled to a shared library. A parallel C++ script was developed as the supervisor, which utilizes the two shared libraries and supervises the calculation of the coupled system. The MEDCoupling libraries developed by CEA provide a generic way to handle the mesh interpolation and field mapping between different domains of TrioCFD and SCF. Moreover, the explicit temporal coupling method and the domain-decomposition approach are adopted and presented in this paper. An academic problem is developed to test the multi-scale coupling code. The results indicate that the inter-code data exchange works well and that the coupled code TrioCFD/SCF can deal with various time-dependent boundary conditions.

This paper describes the pre-conceptual design for a high temperature (>1300 °C), ultra-small (1–10 MWe) modular reactor with a high efficiency (>50%) thermophotovoltaic (TPV) power-block. The integration of the TPV is attractive for increased efficiency over the heat cycles of traditional nuclear power plants (NPP) and increased inherent safety from the elimination of a working fluid. Allowing for heat removal through radiative and passive convective cooling, the design must operate at low power densities. A preliminary sampling of the design space was performed using coupled thermal and neutronic analysis on a simplified model. This study shows the viability of the design and reveals the favorability of using uranium nitride fuel that allow the core to operate safely at approximately 1–2 W/cm3 within the desired temperature limits. Although preliminary, the depletion and economic analyses reveal an operational time on the order of ten thousand days and approximately 43% savings in NPP capital costs.

This paper discusses one of the most important issues facing the regulatory body while performing the uncertainty analysis of the nuclear reactor parameter during accident conditions. This problem is the long computational time required by the statistical sampling methods to compute the uncertainty. We overcome this problem by introducing the Unscented Transform (UT) algorithm and singular value decomposition (SVD). Where both algorithms are combined (SVD/UT) to generate a set of sigma points, these sigma points are the representatives of whole probability distribution. The uncertainty quantification is performed during Loss of coolant accident in Pressurized Water Reactor (PWR), where the input variable of uncertainty is the coolant density reactivity. The SCALE 6.2 code is used for calculating the reactivity coefficients and the covariance matrix. The response variables are the peak cladding temperatures during the accident which are computed by ATHLET thermal-hydraulic code. The results obviously confirm the efficiency of the SVD/UT sampling in predicting the new mean values, and assure its ability to reduce the sampling size leading to a dramatic reduction of computational cost.

Safety by passive systems is a key design feature for new generation Nuclear Power Plants (NPPs). The Passive Containment Cooling System (PCCS) of the AP1000 NPP is a typical passive safety system, by which the heat produced in the containment is transferred to the environment through natural circulation and atmosphere is used as ultimate heat sink. Then, the climate conditions at the plant location influence the system performance. Monte Carlo (MC) simulation of random scenarios of embedding Thermal-Hydraulic (T-H) response of the passive system is commonly used for passive safety systems reliability assessment. However, T-H codes are usually computationally burdensome, in this paper, an effective way to overcome this issue and estimate passive safety system reliability is proposed and applied to the PCCS of the AP1000 NPP, showing that the T-H model runs can be reduced for efficient reliability assessment.

We developed a SPH equivalence technique in non-fundamental mode condition between a full-core model solved with the method of characteristics (MOC) in 2D and a simplified full-core diffusion model with two-group, finite-difference method over a pure Cartesian mesh. The MOC and diffusion calculations are performed with DRAGON5 and DONJON5 codes, respectively. An objective function is set as the root mean square (RMS) error (MOC-diffusion discrepancy) on absorption distribution and leakage rates defined over the macro-geometry in DONJON5. Three algorithms were developed to converge on the SPH factors in non-fundamental mode condition: a fixed point method, a pure Newton method for unconstrained optimization and a memory-limited Broyden-Fletcher-Goldfarb-Shanno (LBFGS) method. We investigated the benefit of all these three techniques on a series of LEU-COMP-THERM-008 BAW Core-XI loadings. We observe convergence success for all numerical techniques considered in this study.

The CABRI International Program (CIP) tests irradiated UO2 or MOX fuels submitted to Reactivity Initiated Accidents (RIA) representative power pulses in prototypical PWR thermal-hydraulic conditions. CIP is managed by The Institut de Radioprotection et de Sûreté Nucléaire (IRSN) within a OECD/NEA framework. Experiments are conducted in the CABRI reactor operated by CEA. For CIP, CABRI benefits from a new pressurized water loop. An important refurbishment program enhanced the facility safety and upgraded the experimental equipment such as the non-destructive examination bench IRIS and the Hodoscope on-line fuel motion monitoring system. Specific test devices with appropriate innovative instrumentation follow the test rod behavior during the transient. The successful first test in the pressurized water loop demonstrated the CABRI capability to perform fully instrumented RIA tests in PWR conditions and to provide highly valuable results for RIA phenomena modelling (boiling crisis, post-failure events), for code validation, and for assessing PWR safety criteria.

In this paper, experiments on breakdown voltages of helium gas in a uniform field have been carried out at a separation from 0.25 to 3.02 mm. The pressure and temperature respectively vary from 1 to 7 MPa, 25 °C to 180 °C, which is the working condition for the main helium blower and its Active Magnetic Bearings (AMBs) in the High Temperature Reactor-Pebble-bed Module (HTR-PM). COMSOL Multiphysics and a self-programmed FD-FCT (Finite Difference-Flux Corrected Transport) code with a two-dimensional drift–diffusion plasma model have been used to simulate the breakdown process and the results agree well with the experimental data. This kind of theoretical and experimental work provides valuable references for helium insulation design of electric devices in HTR-PM and other helium gas-cooled reactors. It also helps understand the mechanisms in helium discharge at high temperature and pressure which is quite different from its low counterpart.

We use a novel nonintrusive adaptive Reduced Order Modeling method to build a reduced model for a molten salt reactor system. Our approach is based on Proper Orthogonal Decomposition combined with locally adaptive sparse grids. Our reduced model captures the effect of 27 model parameters on keff of the system and the spatial distribution of the neutron flux and salt temperature. The reduced model was tested on 1000 random points. The maximum error in multiplication factor was found to be less than 50 pcm and the maximum L2 error in the flux and temperature were less than 1%. Using 472 snapshots, the reduced model was able to simulate any point within the defined range faster than the high-fidelity model by a factor of 5×106. We then employ the reduced model for uncertainty and sensitivity analysis of the selected parameters on keff and the maximum temperature of the system.

The critical power limit for a natural circulation-cooled research reactor core is estimated by utilizing a system analysis code RELAP5/MOD3.3 and an in-house companion program RCPP. The variation range of thermal-hydraulic parameters is found from steady-state and transient simulation results. From literature, potential candidate correlations for critical power estimation were selected and the design limit critical heat flux ratio was evaluated for the chosen correlations based on error statistics. A computational fluid dynamic simulation was carried out to observe the power-equilibrium flow relationship and used to prepare the system code input. To ease a rigorous critical power estimation process, an in-house code was developed to automate batch run and output post-processing of system code. The critical power limit was found for each combination of major operation parameters. A critical power map was generated by merging each power limit value, and the effect of the correlation and core states were studied.

The accident occurred in Japan on 11 March 2011 has revived the interest for the analysis of severe accidents. The scarce and sometimes unreliable data concerning boundary conditions, effectiveness of accident management measures and equipment performance, pose a tough challenge in modelling the accident scenarios. Throughout an analysis of the challenges posed by the Fukushima Daiichi Unit 1 data recorded, this paper describes the major postulates proposed by CIEMAT concerning the equipment and component responses, the effectiveness of accident management actions and the MELCOR model applied. Among most influencing assumptions are those related to the reactor pressure vessel (RPV) leaking pathways and failure mode, the water flow rate entering the reactor, the potential leaking pathways and failure mode and location from the primary containment vessel (PCV) to the reactor building, the corium relocation from RPV to the cavity and its distribution in the PCV, the potential stratification of the suppression pool and the hypotheses made a priori concerning fission product release and transport. Based on the postulated scenario and model, a remarkable agreement of the thermal footprints in terms of RPV and PCV pressures during 500 h has been achieved, in which the RPV and PCV leaks/failures as well as venting played a determining role in the short run of the accident and water injection heavily conditioned the long one. As for the scarce data related to fission products (FP), a consistent agreement is found in the suppression chamber, but estimates in the Dry-Well are about an order of magnitude below measurements despite showing the observed trend. A number of factors might affect FP comparisons to data, from the approximate method to derive dose rates (measurements) from FP masses (MELCOR results) to the RPV and PCV postulated failures. Anyway, based on the data available the set of hypotheses and approximations made seem to make up a defensible scenario for Fukushima Daiichi Unit 1. The studies and results presented in this paper have been achieved under the frame of the OECD/BSAF projects through the CSN-CIEMAT collaboration agreement on severe accidents research.

To further understand and monitor the thermal–hydraulic (TH) status of large domain of pressurized water reactor (PWR) core, an applicable engineering approach with high efficiency and high spatial resolution is critical. Traditional engineering computational fluid dynamics (CFD) computation needs too many computing resources to effectively analyze large domain engineering application for PWR core. In this study, a distributed-parallel (DP) CFD scheme is presented. This scheme separates the large domain into some sub-domains in order to optimize computing time and resources. The design completely retains the complex structures and fine-scale CFD mesh. In addition, it also significantly reduces the computing resources and time, and a fine-scale, full-height CFD analysis can be done in hours for all assemblies. Moreover, important design requirements such as energy consumption ratio, relative computing domain, relative assembly number, and the ranking method of important locations are also designed to optimize the applications of DP scheme to satisfy different engineering demands. The performance of the proposed scheme is evaluated using the CFD computation of a representative region from all 121 assemblies in a PWR core. This research serves to advance the development of engineering CFD computation for PWR core.

The corrugated plate is a vital steam-water separation device in nuclear power plants. Research on steam-water separation mechanism and separation efficiency of corrugated plates has been a hot research direction. In this paper, the film lateral movement distance of the film reflux on the corrugated plate wall is measured by PLIF method. Based on experimental results, calculation equation of film lateral movement distance when the film reflux with good applicability is fitted. The degree of liquid film reflux is characterized by the lateral distance between the apex of the liquid film reflux profile and film mainstream. This parameter is defined as distance of liquid film reflux. Based on the law of conservation of mass, theoretical equation of distance of liquid film reflux is derived. Results show that theoretical results of distance of liquid film reflux agree well with experimental results in small and medium Reynolds number regions (Re < 2625).

A large amount of general steel waste is generated during decommissioning and dismantling of nuclear facilities. Very low contaminated radioactive waste, whose radioactivity is below clearance level, generated from the demolition process may be reused for general use. We examined the feasibility of the clearance verification system for uranium waste. The relative error of uranium determination was within 30 % for 1 g of uranium when measuring steel materials (angle bar, channel steel, pipe steel, square steel tube, fragments of steel tube).

One of the core problems of waste canisters layout in high-level radioactive waste repository is the evolution of the temperature field. On basis of the layered thermal analysis model for single waste canister, the expression of temperature increment at any location in surrounding rock in the repository was obtained by the superposition principle. The initially estimated value of the canister spacing (CS) was determined according to the temperature design criterion. Finally, the influence of relevant parameters on the canister surface temperature (CST) was analyzed. The results were drawn as follows: (a) Taking thermal conductivities of 2.4 and 2.8 W/(m × K) for the rock as examples, the appropriate CS is 12.2 and 13.5 m under the tunnel spacing of 40 m, respectively. (b) The greater the CS, the greater the thermal conductivity of bentonite and rock, the smaller the CST would be. (c) The thicker the buffer layer, the less the heat flux inside the canister would spread out.

To determine heat transfer regimes of the pre and post CHF, the SPACE code calculates the wall temperature from a nucleate boiling heat transfer model at the given CHF. It needs iterations and consumes a large amount of computing time. To reduce the calculation time, this paper introduces the application of a machine learning method. Big data of the wall temperature at CHF was built by using the subprogram constructed as is in the SPACE code. Based on that database, the neural network models were trained and two neural network models having different configurations were suggested. The developed neural network models were implemented in the SPACE code and test calculations were performed. The neural network applied SPACE code properly predicted the wall temperature at CHF. In test calculations, the calculation time was also investigated. All suggested neural network models highly enhanced the calculation speed corresponding to a maximum 86% time reduction.

The energy spectrum of the prompt-neutron flux at the Missouri University of Science and Technology Reactor (MSTR) was obtained using an activation method. Foils were irradiated at the bare rabbit tube (BRT) of the reactor core (which has a 120 W configuration). The neutron spectrum was determined using an unfolding method implemented in SAND-II code. The Monte Carlo N-particle (MCNP) model was used to calculate the spectrum at the bare rabbit tube, which was used as the initial guess for input into the SAND-II code. Various MCNP spectra with 620-, 143-, 89-, 50-, 22-, and 12-energy groups were used as the initial guess for input into SAND-II. Seventeen different foils were irradiated at 100 kW. The foil set covers energies from 0.025 eV to 13 MeV for a broad spectrum analysis. The photon counts for the activated foils were obtained using a high-purity germanium (HPGe) detector. The count results were input into the SAND-II code to predict the neutron flux spectrum for the MSTR. The spectra were collapsed into three groups: thermal, epithermal, and fast flux. With the 620-group initial guess, the thermal, epithermal, and fast neutron fluxes were 1.43×1012±2.82×1011n/cm2s, 4.51×1011±2.85×1010n/cm2s, and 5.38×1011±4.85×1009n/cm2s, respectively, giving a total flux 2.42×1012±3.02×1011n/cm2s. Disparities were noted in the distribution of the thermal and epithermal flux predictions, as the number of groups in the initial spectrum guess changed. The 59%/19% distribution of the thermal/epithermal flux, as predicted with the 620-group guess, is inconsistent with the 49%/26% distribution predicted with the 89- and 143-group guesses. The predictions based on the 89- and 143-group guesses are fairly consistent with the predictions obtained with the 22- and 50-group guesses. The 12-group initial guess resulted in the prediction of a fairly even distribution between the thermal (38%) and epithermal (36%) fluxes. Regardless of the number of groups in the initial guess, the SAND-II prediction is fairly consistent in the fast energy range. The fast neutron flux was found to range between 22% and 26%.

Monte Carlo (MC) shielding calculations often use weight windows (WWs) and biased sources formed from a deterministic estimate of the adjoint flux to improve the convergence rate of tallies. This requires a significant amount of computer memory, which can limit the memory available for high-resolution tally output. A new method is proposed for reducing these memory requirements by using singular value decomposition (SVD) in linear or logarithmic space to approximate the adjoint flux. This method’s performance is evaluated using the Shift and Denovo codes for streaming and diffusion base case problems, followed by problems using the Westinghouse AP1000 and the Joint European Torus. The log SVD reduced WW memory requirements by an order of magnitude in all cases without a significant performance penalty. Additionally, the linear SVD reduced biased source memory requirements by an order of magnitude, but further investigation is needed to account for observed limitations.

Steam separator package including swirl vane separator, gravity and dryer is crucial for guaranteeing steam humidity below 0.1% in SG. In this paper, a high-pressure flow system is designed to study the performance of a full-scale steam separator package at a rated pressure of 6 MPa and temperature of 275.6 °C. To obtain separation efficiency, moisture carryover and pressure drop of each stage, the Reynolds number steam and water are 1 × 106–6 × 106 and 8.27 × 105–3.2 × 106, corresponding to 25–145% load. Results indicate that the separation efficiency ranges from 97 to 99%, 0.5 to 2% and 0.2 to 0.75% for separator, gravity and dryer. The moisture carryover ranges from 2 to 16% and 1 to 4% for separator and gravity, and is below 0.1% at dryer outlet. The pressure drop ranges from 2 to 24 kPa and 0.3 to 3 kPa for separator and dryer. Empirical correlations for swirl vanes are proposed, which are responsible for more than 85% and 70% of total separation efficiency and pressure drop.

The passive residual heat removal system (PRHRS) is adopted by the HTR-10 and HTR-PM to remove the decay heat in the reactor core, so that the safety of the reactor can be guaranteed under accident conditions. A multi-scale CFD-system coupled code used to calculate the transient characteristics of the PRHRS of HTR-10 was developed in this study. With this coupled code, the transient characteristics of the PRHRS during the startup process were calculated. The results show that the computational results are in good agreement with the experimental results, which validates the accuracy of the coupled code. The ability of the coupled code to simultaneously calculate the local heat transfer process in the reactor cavity and the global heat removal characteristic of the PRHRS is also shown. Moreover, radiation heat transfer plays a very important role in the heat removal process of the PRHRS.

Helical coil once-through steam generators (HCOTSGs) have been widely used in the design of small modular reactor (SMR) during the past several decades. As the widely accepted system analysis code, RELAP5 underestimates the heat transfer capability and pressure drop of helical coil tube steam generators using the original geometrical parameters of HCOTSG, because the built-in thermal-hydraulic empirical correlations are only suitable for straight tubes. In this study, thermal-hydraulic models for helical coil tubes and tube bundles are implemented in RELAP5 while the original functions of the code are not affected. A new flag for helical component is proposed and heat transfer boundaries for tube side and shell side are developed. The separate-effect validations of friction factor and heat transfer in helical coil tube are carried out based on experimental data and code-to-code verification. Then, the original RELAP5 and developed RELAP5-HCOTSG are used to simulate helical coil tube steam generators of two SMRs. One is the integral reactor IRIS, and the other is MRX. The calculated results are compared to the design parameters. It shows that the HCOTSG module developed in this study can improve the capability of RELAP5 to predict the thermal-hydraulic characteristics in SMR once-through steam generators with helical coil tubes.

In this study, a novel assembly design is proposed for a fluoride salt-cooled high-temperature commercial-scale (FHCR) reactor. It employs tristructural isotropic (TRISO) fuel particles nestled within removable cylindrical beryllium carbide (Be2C) moderator blocks, which are further contained within prismatic graphite blocks. As the name implies, the FHCR is a preconception of a 3400 MW(t) commercial power reactor that uses FLiBe (LiF-BeF2) as the primary coolant of choice. This paper uses the SERPENT 2 code to conduct a parametric neutronics study on the two-dimensional lattice assembly design of the FHCR. The calculations examine the effects of various fuel enrichment levels, TRISO particle packing fractions, fuel compacts’ pitch sizes, and moderating materials on the cycle length, while also determining the neutron spectrum and various nuclide inventories. Finally, a preliminary core is modeled based on the study conducted on the assembly design. Based on the negative values of the fuel temperature coefficients (FTC), moderator temperature coefficients (MTC), and coolant temperature coefficients (CTC) obtained, the design is determined to be safe. This new assembly design is also able to achieve keff>1 for approximately 3.55 years, translating to a burn-up of 160.6 MWd/KgU.

The wire-wrapped assembly could enhance the safety and the economy of the reactor, however the complex structure brings some new problems for thermal-hydraulic analysis. In this paper, the hydraulic analysis of the 7-pin wire-wrapped assembly is carried out by computational fluid dynamics (CFD). The structural grid of the entire fluid domain is generated by the commercial grid software ICEM CFD and solved by CFX. Unlike most studies, this article explores the effect of assembly fuel flow by changing the shape of the fuel duct. Considering the influence of the different shapes of the fuel duct, the pressure drop, cross flow, sub-channel flow distribution are studied. Through this study, it can help to improve the flow environment of peripheral fuel rods and optimize the design of assembly.

Point kinetics equations (PKEs) is an significant model used to describe dynamic behaviors of neutron in nuclear reactors. How to cope with its stiffness problems is very important to solve PKEs accurately and efficiently. In this paper, a new semi-analytical algorithm based on Euler-Maclaurin Approximation (SAEMA) is developed to solve PKEs. SAEMA algorithm is applied and tested in different reactors with three typical reactivity insertion cases, including step, ramp and sinusoidal insertions. Firstly, the investigations on the computational stability are performed under different time step sizes. Secondly, the performances on computational accuracy are evaluated by comparing the results by SAEMA algorithm with the ones by analytical method and excellent CATS algorithm. Finally, by comparing the CPU time consumed by SAEMA algorithm with the physical time, the studies on the computational efficiency are also carried out. Just as results shows, SAEMA algorithm is reasonably an attractive way to solve the PKEs.

In this paper, a new approach aimed at the Fault Diagnosis with Full-scope Simulator based on the State Information Imaging (FDFSSII) in NPP is proposed. The FDFSSII approach first constructs a series of gray-image which presents the operating transient (included normal and fault condition) according to the real time monitoring data. Furthermore, the Machine Learning (ML) technology is employed to achieve image feature extraction and classification by analyzing and learning from massive amounts of historical and synthetic gray-image data – the image feature is extracted by the Kernel Principal Component Analysis (KPCA) and classified by the designed classifiers in different learning methods. Finally, diagnosis effect is evaluated by the F1 score. The simulation result shows that the FDFSSII approach has achieved good effect for the fault diagnosis in NPP. Meanwhile, it simplifies the process of nuclear reactor with the large monitoring data and provides useful support information to the operators.

In this paper coupled calculations with the CFD code ANSYS-CFX-19.0 and the Monte Carlo neutronics code MCNP6 were performed to analyze the heat transfer in supercritical water flowing through the typical fuel assembly of the high-performance light water reactor (HPLWR), in order to improve the characterization of the heat transfer phenomena in supercritical water under non-uniform axial heat flux distributions that is characteristic of this type of reactor. To check the capability of the CFX model to predict the thermal-hydraulic behavior of supercritical water, the computational results were compared with two experimental data. The Shitsmańs experiment in the presence of heat transfer deterioration (HTD) using four low-Re turbulence models (SST, k-ω, BSL-k-ω, and ω-Reynolds Stress) and the Wanǵs experiment in absence of HTD, using the low-Re-SST and the scalable-wall-function-SSG turbulence models. In the presence of the HTD phenomenon, results showed the high dependency of the wall temperature with the turbulence model and the turbulent Prandtl number selected. In the absence of HTD, both turbulence models studied adequately predicted the behavior of the wall temperature distribution. For the coupled neutronic/thermal-hydraulic analysis of the typical HPLWR fuel assembly, the low-Re SST turbulence model and the Prt = 1.5 were used. Different axial profiles of heat flux generated in the fuel rods were obtained for the different power values studied. For the analyzed conditions, the presence of HTD in the lower zone of the fuel assembly was observed. In addition, the results showed a strong non-uniformity of the circumferential surface cladding temperature distribution in the sub-channel located at the corner of the fuel assembly; a new curvature radius of the assembly box corner was proposed to obtain a well homogenized circumferential wall temperature distribution.

The rod drop experiment is an important transient in reactor operation. This study focuses on a comparison between experimental reactor physics data and respective calculation. The rod drop experiment was performed at the VR-1 reactor. A full core 3D model was used to calculate this transient using Serpent2. Additionally, a point kinetic solution is presented. The Serpent results were compared with the experiment and with the point kinetic calculation. Two different nuclear data libraries (ENDF/B-VIII.0, JEFF3.3) were used in the data sensitivity analysis. The Serpent 3D kinetics results were almost identical to the point kinetic equation solution. Experimental data differed in the rod worth. This study shows that the new Serpent dynamic toolkit provides an accurate description of the reactor behavior during this transient. The 3D calculation proved that the detector position during the transient has a direct impact on the measured control rod worth.

Extended Kalman filter (EKF) and cubature Kalman filter (CKF) are proposed to estimate the state parameters of an external neutron source driven sub-critical reactor, including power level, reactivity, external neutron source, six-groups of delayed neutrons precursor densities, equivalent fuel temperature, average coolant temperature and nuclear densities of iodine, xenon, promethium, samarium nuclides. Parameter settings and matters needed attention in EKF and CKF are also analyzed, especially the relationship between model prediction covariance matrix and measurement covariance matrix. In order to effectively identify the maneuvering of the external neutron source and reactivity on the uncertainty of the prediction model, two adaptive algorithms are proposed to adjust the covariance matrix of the prediction model online. The results show that these two adaptive algorithms can effectively detect various maneuvering such as neutron source variation and reactivity insertion, and realize the optimal estimation of the reactor state using EKF method. However, CKF has divergence and non-convergence. EKF achieves good results in all parameters estimation.

Flapper Spurious opening during reactor operation is one of the most perilous accidents for pool-type research reactors. In TRR during such event, if none of SCRAMs could shut-down reactor, partially fuel elements burn-out is occurred. To avoid that, a core cooling condition monitoring system is developed. There are also other PIEs, i.e. fuel element coolant inlet blockage or unintended fuel withdrawal, which can be detected by core DP. In flapper opening scenario, the experiments show core DP response is very tangible. About two other scenarios, results indicate core DP is capable to detect the abnormality in the core flow, but maybe some other supplementary data and signal are required to distinguish the accident. Preliminary data from the system exploitation during reactor operation are analyzed to take into account the uncertainties of various normal operation conditions and determine a minimum allowed working window for the system safety signal.

The neutronics calculation of the PB-FHR with the fraction ranging from 5% to 30% are carried out by the CLS method and ERM method. It can be observed that the infinite multiplication factor obtained with the CLS method are consistently smaller than those obtained with the ERM method when the fraction is less than 15%. Moreover, the differences of the two methods basically increases with the increasing burnup level when the fraction is less than 15%. For the fraction up to 30%, the CLS method shows significant bias of 520 pcm with the explicit method. After the CLS method with a modified packing fraction correction is adopted, the largest difference drops to 220 pcm when the fraction is 30%.

In the present paper, CFD simulations related to the operating conditions considered during the experimental campaign on CIRCE-HERO facility are presented, with the aim of investigating the observed temperature stratification phenomena. Calculations are performed using the commercial codes STAR-CCM+ and ANSYS Fluent adopting a RANS approach; the numerical results and the experimental data are compared. Four distinct experimental tests are investigated also performing sensitivity analyses regarding the boundary conditions. In particular, assumptions concerning the heat losses distribution and the shape of the pool inlet were taken into account. The numerical results provide support for further understanding of the involved phenomena, suggesting the possible causes of the thermal stratification observed experimentally inside the pool. Similar trends for the predicted and experimental data were obtained and – even from a quantitative point of view - the observed discrepancies can be considered acceptable, assuming the uncertainties in the experimental boundary conditions and measurement.

A new design integrating heat-pipes into a nuclear cooling system is presented. The heat-pipes are presented as the primary mode of heat transfer. Analyzing the prevailing limits to determine suitability for predicting the performance of the heat-pipe concept. When the limits are determined for the design integration, steady-state behavior needs to be quantified. A model that accounts for 3D behavior is evaluated for use. Using the limits to evaluate the design integration, with sodium, the operating regime would still remain below the predicted limits. With potassium the operating regime would exceed the capillary limit. This is caused by the increase in pressure drop. With the 3D model, a validation shows that conduction can give very good results for both transient and steady-state behavior for sodium. It shows that water has poor transient prediction but accurately predicts the steady-state behavior. Both solutions were close to the reported experimental results for steady-state.

CsI and CdI2 aerosol decomposition rate under irradiation has been quantified at 80 °C and 120 °C in presence of humidity and on different substrate (stainless steel, quartz and Epoxy paint). A model has been developed for the ASTEC-SOPHAEROS code to reproduce the data and help the identification of the gaps remaining in the understanding of iodine volatility in a severe accident of a Nuclear Power Plant (NPP). The current model applied to model the gaseous iodine behaviour in the containment of PHEBUS-FP tests does not fit with the experimental data probably because the nuclear aerosol reaching the containment are much more complex than pure CsI aerosols. It has been clearly shown than the radiolytic oxidation of metallic iodide aerosols into molecular iodine can significantly impact the source term evaluation even if additional experimental data area required to cover the variety and complexity of nuclear iodide aerosols.

In this paper a probabilistic methodology based on core nodalization is proposed to estimate the core power in the presence of xenon oscillation. A time-dependent Monte Carlo neutron transport code named MCSP-NOD is developed for dynamic analysis in arbitrary 3D geometries to simulate xenon oscillations as well as sub-critical condition with feedbacks. The new code is based on the approach adopted in MCNP-NOD which was previously introduced as a tool for core transient analysis using the MCNPX platform. As before, the core is divided into nodes of arbitrary dimensions, and all terms of the transport equation e.g. interaction rates, leakage ratio are estimated using the MC techniques. However, as a new option the concentration of iodine and xenon are also estimated which enables us to predict the oscillatory behavior of reactor power following poison oscillation. These quantities are then employed within the time-dependent neutron transport equation for each node independently to compute the neutron population. Simulations prove the robustness of the method.

The potential reasons for the corium spreading difference over inert ceramic and reactive concrete channels of VULCANO VE-U7 experiment are investigated using Moving Particle Semi-implicit (MPS) method. A new thermal contact resistance model has been developed for MPS so that influence of the subscale heat transfer between the melt/crust and the substrate on spreading can be considered. The results indicate that the spreading difference is not much influenced by heat loss of the melt to different substrates, but more likely due to gas bubbles in the concrete channel. The most likely responsible gas bubble effect could not be well identified with the single channel analysis, because it could not consider the inflow mass interactions at the stabilization pool. The double-channel analysis with such consideration indicated enhancement of the effective thermal conductivity of the melt as the key influence of the gas bubbles that led to the difference.

Deformation and contact force of the fuel assembly due to thermal gradient are of great concerns to the integrity design and safe operation of the core in the China Fast Reactor (CFR). Up to now, only the 2D deflection measurement of the sub-assemblies has been investigated, and almost no attention is paid to the contact force measurement between the sub-assemblies with a small clearance. In this paper, the measurement method and system are developed for measuring the deformation and contact force of the fuel assembly under thermal gradient with the test fuel assembly simplified based on the stiffness equivalent strategy. In addition, the 3D deformation and contact force are measured through the non-contact industrial photogrammetry system and the thin film pressure sensor or strain gauge, respectively. Free thermal bowing test for a single assembly is performed at first for verifying the reliability of the 3D deformation measurement by comparing with the results of finite element (FE) simulation. Afterwards, the 3D deformation and the contact force are measured for the single assembly restrained thermal bowing test. The accuracy of the proposed contact force measurement is investigated through comparison with the method of spoke-type force sensors. The developed measurement method and system can provide experimental basis for safety design of the CFR due to its potential for comprehensive deformation and contact force measurements.

Suppression pools are an essential passive system for source term attenuation in boiling water reactors during severe accidents, particularly during Station BlackOut (SBO) sequences, as it happened in Fukushima. This paper investigates how uncertain predictions of suppression pools decontamination can be. Based on MELCOR 2.1 calculations of Fukushima Unit 1, a stand-alone version of SPARC-90 (Suppression Pool Aerosol Removal Code) has been used in combination with DAKOTA-6.4, to propagate the uncertainties in the input deck variables affecting the Decontamination Factor (DF). The results indicate that DF uncertainties may spread around two orders of magnitude and the uncertainty margin stays roughly constant over time. In addition, a sensitivity analysis based on the Pearson and Spearman correlation coefficients has been carried out and pointed that uncertainties associated to particle inertia (i.e., particle density and size) and in-pool phase change (i.e., non-condensible gas fraction in the carrier gas) dominate the uncertainties found in the DF for this specific scenario.

The differential operator method (DOM) has been developed to perform sensitivity analyses of generalized responses in the form of reaction rate ratios. In this method, the memory consumption required to store the source perturbation effect will become prohibitively large with a large number of particle histories. This work introduces the superhistory-based differential operator method (SH-DOM) to reduce the memory usage. In the superhistory algorithm, the source perturbation effect is estimated by following the source particle and its progenies over super-generations within a single particle history, which significantly reduces the memory usage. The new method is verified via the Jezebel, Flattop and the UAM TMI PWR pin cell benchmark problems calculated by the collision history-based method and the SH-DOM. Results show that the energy integrated sensitivity coefficients given by the present method agree within 5% with those of the collision history-based method and the SH-DOM can effectively reduce the memory consumption.

Comprehensive knowledge of neutron spectrum is significant in reactor design. Online wide-range neutron spectrum unfolding technology still requires improvement in accuracy and efficiency. In the work, a “two-step” neutron spectrum unfolding method based on artificial neural network (ANN) was developed to unfold spectrum with wide energy range. First, a default spectrum was reconstructed by using the ANN model which had been trained with a large amount of neutron spectra generated from Monte Carlo transport calculation. Second, the default spectrum was optimized by using iteration algorithm. The two-step method was verified with a thermal neutron reactor VENUS-3 and a fast neutron reactor BN-600. Comparison of mean square error (MSE) between this method and the traditional unfolding method showed reduction of 83.4% and 85.6% on VENUS-3 and BN-600 respectively, and average relative deviation (ARD) reduction of 89.3% and 86.1% respectively. Also, comparison of spectrum quality (Qs) showed reduction of 83.4% and 86.0% respectively for the two cases. This work demonstrated that the developed two-step method could obtain the better accuracy than traditional method.

A new method for detecting a gas–water interface in a circular pipe is proposed. In the method, ultrasonic signals are used for non-intrusive measurement and three types of signal analyses are conducted: time-of-flight (TOF), local amplitude, and global amplitude analyses. Horizontal, 45° inclined, and vertical pipe configurations were used to verify the applicability of the proposed detection method. In the case of a horizontal pipe with an acoustic beam directed perpendicular to the water surface, TOF and amplitude analyses detect the water level. In the cases of a horizontal pipe with an acoustic beam directed parallel to the water surface, a 45° inclined pipe, and a vertical pipe, when the pipes were filled with water, TOF analysis was not applicable and amplitude analysis detects the water level. In conclusion, the gas–liquid interface in circular pipes could be analyzed qualitatively and quantitatively through the proposed non-intrusive acoustic method.

In order to give a clear picture of key parameters uncertainties introduced from fuel pebble model in HTR and its effect on the uncertainty propagation, herein different methods of sampling the position of a random dispersed TRISO particle in the fuel region were studied. And total ten different fuel pebble models including high-fidelity models with random dispersed fuel particles and homogenized models were built to quantify the uncertainty of eigenvalue introduced by fuel pebble models. Meanwhile, the effect of fuel pebble model on the uncertainty propagation of nuclear data was also investigated. The numerical results indicate that the fuel pebble models introduce a great model uncertainty to the calculated multiplication factor and also have a significant effect on the uncertainty propagation of nuclear data. However, the uncertainty of the multiplication factors due to the random distribution of TRISO particles is relatively small compared with the uncertainty propagated from nuclear data.

Best estimate plus uncertainty (BEPU) analysis has been widely adopted for safety evaluation of the nuclear reactor, and the sensitivity analysis is an important part of the BEPU methodology. Local sensitivity analysis methods are still widely utilized in the methodology, which is not suitable for complex non-linear nuclear systems. The global sensitivity analysis is essential but the biggest problem is that the computational cost can hardly be accepted. In this study, the moment-independent global method was adopted and optimized, and a low-cost method was obtained and assessed. The sensitivity analysis of a large break loss of coolant accident (LBLOCA) of the LOFT facility was carried out by using the method. Results show that the optimized method can well evaluate the sensitivity indices with very low cost and relatively high accuracy, and the result obtained through the optimized method is much more reliable than that of the local sensitivity analysis method.

This paper presents the verification and validation elements of the UNIST in-house Monte Carlo code, MCS, for the multi-cycle and multi-physics analyses of high-fidelity, large-scale commercial pressurized water reactors (PWRs). Analysis on the neutronic performance with thermal/hydraulic (T/H) feedback is the key to detecting the complex behavior of an operating nuclear power reactor. The MCS solutions with T/H feedback, TH1D, of the consortium for advanced simulation of light water reactors (CASL) virtual environment for reactor applications (VERA) core physics benchmark progression problems 6 and 7 showed excellent agreement in eigenvalues, temperature and power profiles with the MC21/COBRA-IE, MC21/CTF and VERA-CS solutions for the single assembly and whole core of Watts Bar Nuclear 1 (WBN1) Cycle 1 under beginning-of-cycle and hot-full-power condition. Furthermore, the core depletion analysis is one of the most compelling advances for reactor analysis. Therefore, this work is significantly focused on the nuclide depletion simulation coupled with TH1D of the first two cycles of WBN1 to address the VERA core physics benchmark problems 9 and 10. MCS is one of the few Monte Carlo codes that have the capability of depletion calculation for both WBN1 Cycles 1 and 2. The accuracy of MCS simulation of WBN1 Cycle 1 is within 40 ppm and 30 ppm in critical boron concentration (CBC) for all burnup points, compared with the measured data and VERA-CS solutions, respectively. To demonstrate the multi-cycle refueling capability of MCS, the WBN1 Cycle 2 is simulated and compared with the solutions of VERA-CS only, because of the lack of available measured data. MCS shows excellent agreement compared with VERA-CS within 30 ppm in CBC, and the average bias for the entire Cycle 2 is approximately 20 ppm. These results provide confidence in MCS’s capability in high-fidelity, multi-cycle calculations of the practical PWR core.

It is important to ascertain the mechanisms underlying the eutectic reaction between different reactor materials that might be encountered during a Core Disruptive Accident (CDA) of Sodium-cooled Fast Reactors (SFR) since such reactions will affect the accurate progression of a severe accident. In this study, motivated to understand the characteristics of the probable eutectic reaction between fuel debris and the lower head of reactor vessel, a series of simulated experiments has been conducted at the Sun Yat-sen University using a couple of rather lower-eutectic-point (456 K) materials (namely Sn particles and Pb pellet). The experiments were carried out in a self-designed experimental system, which mainly consists of a sample holder and a visible resistance furnace. To acquire a relatively comprehensive understanding, a variety of experimental parameters such as the reaction temperature (463–483 K), contact pressure (0.4–1.2 MPa), Pb pellet diameter (10–25 mm) along with the diameter (0.3–3 mm) and geometry (spherical, cylindrical and droplet-shaped) of the Sn particles have been taken. Through detailed analyses, it is found that the reaction temperature and contact pressure can have noticeable positive impact on the reaction rate. As for the size of Pb pellet and Sn particles, with increasing the diameter ratio of Sn particles to Pb pellet, a non-monotonous effect is observed due to the competing role between the contact area and contact pressure. An evident influence of Sn particle geometry on reaction rate has been verified in accordance with the variation of particle-bed porosity. The analyses in this work also suggest that the reaction rate in previous experiments using block-block samples is generally larger than present experiments using particle-pellet ones, especially at a higher temperature. Knowledge and evidence obtained from this work will be utilized for the design of future high-temperature experiments using actual reactor materials as well as for the improved validations of eutectic-reaction-related models incorporated in fast reactor severe accident codes.

A new Monte Carlo (MC) neutron/photon transport code, called MCS, has been developed at Ulsan National Institute of Science and Technology (UNIST) with the aim of performing the high-fidelity multi-physics simulation of large-scale power reactors, especially pressurized water reactors (PWR). The high-fidelity multi-physics analysis of large-scale PWR is a challenging problem due to two aspects, the first being the difficulty of implementing various state of the art techniques into a single code system, and the other making it feasible to run such simulations on practical computing machines within reasonable amount of memory usage and computing time. In this paper, features implemented into MCS for large-scale PWR simulations are described including but not limited to depletion, thermal/hydraulics coupling, fuel performance coupling, equilibrium xenon, on-the-fly neutron cross-section Doppler broadening, and critical boron search. The efficient memory usage for burnup simulation and the high performance of MCS through various algorithms and optimizations (parallel fission bank, hash indexing) are illustrated on Monte Carlo performance benchmarks. Finally, the large-scale PWR analysis capability is fully demonstrated with BEAVRS Cycles 1 & 2 calculations.

Boiling heat transfer over tube bundles has been extensively applied to various industries with a high demand for efficient heat transfer. This work presents a review of recently published studies on pre-CHF boiling heat transfer across plain and enhanced tube bundles. Bundle effect and heat transfer enhancement by modified heating surfaces under various operating and geometric parameters are analyzed. Flow regime maps for boiling two-phase flow in horizontal and vertical bundles are critically described separately. The local boiling heat transfer performance is affected by the non-uniform heat flux distribution in a bundle. A decreasing heat flux distribution along the bundle height can enhance the bundle effect. The effect of the pitch to diameter ratio on bundle effect also depends on the heat flux distribution. Significant influences of the bundle inclination angle and elevation angle on the boiling heat transfer were observed by researchers. Complex bundle effect was found in special shape bundles, such as V-shape, C-shape, and U-shape bundles, which suggests applying different HTC correlations to different regions in a bundle. Moreover, the bundle boiling behaviors under sub-atmospheric and sub-critical pressures have been examined. The heat transfer performance in tube bundles with enhanced surfaces is significantly impacted by the surface characteristics and the imposed heat flux. Bundle effect is still prominent, and the surface enhancement reduces along the bundle height. A mixed bundle with enhanced tubes only in the lower part can achieve the same heat transfer performance as a fully enhanced bundle.

In this paper, the close process of contra-motion check valve in countercurrent pressure difference is studied in nuclear power plant experimental circuit. The transient mathematical model is established. Through the dynamic grid technique, transient flow field in different conditions of valve closing are calculated, and dynamic properties of valve head closing process is analyzed. Through the CFD transient analysis, inherent damping principle and inhibition of water hammer principle are revealed. By writing water hammer analysis, the inhibitation ability of water hammer in typical six conditions is analyzed. At last the performance verification experiments show that contra-motion check valve can effectively inhibit water hammer in the process of parallel double pump switching transition.

To provide a passive cooling system for the reactor containment, Passive Containment Cooling System (PCCS) was adopted in the design of i-POWER nuclear power plant. This study focused on validation tests for condensation heat transfer of the PCCS heat exchanger using the CLASSIC facility. The tests include investigation of the condensation heat transfer in prototypic single tube and bundle tubes. From the single tube experiments, condensation heat transfer model was proposed to reflect the PCCS heat exchanger tube geometry. Experimental results in the bundle tube show consistent trend compared to the proposed heat transfer model from the single tube test. The local condensation heat transfer coefficient of inside tubes was smaller than the average value due to a shadow effect by a larger mass fraction of non-condensable gas, so that design of the PCCS should take into account degradation of the condensation heat removal in the bundle geometry.

UO2 pellets containing 5 vol% diamond particles were irradiated in the Advanced Test Reactor at an LHGR of up to 310 W/cm to the burnup of 7.2 GWd/tHM. Fission gas release, measured during post-irradiation examination, was determined to be 1.08%. Fission gas release modeling performed using BISON fuel performance code showed that undoped UO2 fuel irradiated under the identical conditions, would have had fission gas release of 9.09%. Noting that diamond has the highest thermal conductivity of any known material, these results suggest that doping with a good thermal conductor is an effective means to reduce fission gas release in UO2 fuels.

The cause of inaccurate prediction of system dependability is the uncertainty associated with the modeling, which may lead to overestimation or underestimation of dependability metrics. The prediction of system abnormality or abnormal component of the system under uncertainty becomes a challenging issue and essential for control and safety-critical systems. To efficiently detect the abnormal component's causes of whole system failure, this paper proposes a Piping Possibilistic Timed Coloured Petri Net (PPTCPN) to model the uncertainty. The effectiveness of the proposed mechanism has been validated on 5 different safety critical and control systems of Nuclear Power Plant, and shown on a Digital Feedwater Control System (DFWCS) in this paper.

The stability analysis of a natural circulation integrated self-pressurized water reactor is investigated by the Lyapunov approach. The analysis of the pressurized water reactors (PWRs), particularly the integrated self-pressurized water reactors, is essential in keeping the neutronic and thermal-hydraulic parameters of the system stable. An appropriate nonlinear dynamic model is introduced based on conservation of mass, momentum and energy which is then linearized, in a state-space model. The Lyapunov approach and Routh-Hurwitz criterion are applied to assess the stability of this linearized system over its entire power range. The analysis is done for the primary coolant circuit in the RPV by assuming the steam dome pressure as the fixed parameter. It is found that the system remains stable over its entire power range. The influence of different geometrical features is studied at nominal conditions. It has been found that by reducing the chimney height results in a decrease in the coolant flow rate and a downward motion of the onset of flashing while the average core coolant temperature rises. For lower values of friction losses coefficient, the coolant flow rate increases, and the onset of flashing moves upward and the average core coolant temperature decreases. A change in independent parameters, which are effective in generating the natural circulation of the coolant, can influence the inherent safety of the system: An increase in reactor power and chimney height and a decrease in friction losses coefficient improve the system inherent safety. Two input functions as extra reactivity for increasing and decreasing the power from the nominal state are implemented into the dynamic model and the model response is therefore assessed by considering the lack of two phase flow entry to the core restriction. The boundary of the system stability lies in the range 32–107 MWt and using the system outside this range the system pressure needs to be controlled through spray and heater systems. The obtained results are based on the initial information, data and primary design of these types of reactors. Determining a more accurate boundary requires more detailed design and assessment together with experimental test facilities with respect to other restricting parameters of the system. The results presented in this paper can be implemented in further research on this type of reactors, particularly for nonlinear stability analysis and finding nonlinear Lyapunov function.

This paper summarizes the results of experimental measurements recently made on heavy water moderated plutonium-thorium (Pu-Th) mixed-oxide fuel bundles in the ZED-2 critical facility of Canadian Nuclear Laboratories. Pu-Th fuel in pressure-tube Heavy water reactors can help extend nuclear energy resources by utilizing plutonium from spent fuel, and extracting the energy potential of thorium. The ZED-2 experiment was a series of substitution measurements, with (Pu,Th)O2 test fuel placed into vertical channels filled with air or heavy water in a reference lattice of very low-enriched uranium (LEU) fuel. Sensitivity to the test fuel was optimized by placement into positions of higher power. MCNP5 simulations of the fuel substitution series are within uncertainty of critical moderator height measurements. The precision of the substitution measurements can help support the validation of reactor physics codes for evaluation of reactivity coefficients in PT-HWRs with fuel that contains a mixture of thorium and plutonium oxides.

Design improvement of two-phase natural circulation systems with thermoelectric generator (TEG) is proposed in this paper. In the systems designed, the TEG produces electricity with heat energy of the systems. The electricity is further supplied to a pump and a group of agitators. In this way, circulation flow in the systems and external heat transfer of heat exchangers can be enhanced. Feasibility and effectiveness of the designed systems are analyzed through numerical simulation. The results indicate that by regulating the allocation of electricity between different loads, enough inlet subcooling of the pump can be ensured. Thus, the potential cavitation can be avoided. Comparative analysis of the systems performance with that of the natural circulation systems are carried out. Analysis results indicate that circulation flow and heat transfer are markedly enhanced with the systems designed.