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Abstract The grand challenge of “net-zero carbon” emission calls for technological breakthroughs in energy production. The traveling wave reactor (TWR) is designed to provide economical and safe nuclear power and solve imminent problems, including limited uranium resources and radiotoxicity burdens from back-end fuel reprocessing/disposal. However, qualification of fuels and materials for TWR remains challenging and it sets an “end of the road” mark on the route of R&D of this technology. In this article, a novel approach is proposed to maneuver reactor operations and utilize high-temperature transients to mitigate the challenges raised by envisioned TWR service environment. Annular U-50Zr fuel and oxidation dispersion strengthened (ODS) steels are proposed to be used instead of the current U-10Zr and HT-9 ferritic/martensitic steels. In addition, irradiation-accelerated transport of Mn and Cr to the cladding surface to form a protective oxide layer as a self-repairing mechanism was discovered and is believed capable of mitigating long-term corrosion. This work represents an attempt to disruptively overcome current technological limits in the TWR fuels. Impact statement After the Fukushima accident in 2011, the entire nuclear industry calls for a major technological breakthrough that addresses the following three fundamental issues: (1) Reducing spent nuclear fuel reprocessing demands, (2) reducing the probability of a severe accident, and (3) reducing the energy production cost per kilowatt-hour. An inherently safe and ultralong life fast neutron reactor fuel form can be such one stone that kills the three birds. In light of the recent development findings on U-50Zr fuels, we hereby propose a disruptive, conceptual metallic fuel design that can serve the following purposes at the same time: (1) Reaching ultrahigh burnup of above 40% FIMA, (2) possessing strong inherent safety features, and (3) extending current limits on fast neutron irradiation dose to be far beyond 200 dpa. We believe that this technology will be able to bring about revolutionary changes to the nuclear industry by significantly lowering the operational costs as well as improving the reactor system safety to a large extent. Graphical abstract

The phenomenon identification and ranking table (PIRT) is an important basis in the nuclear power plant (NPP) thermal-hydraulic analysis. This study focuses on the importance ranking of the input parameters when lacking the PIRT, and the target scenario is the small break loss of coolant accident (SBLOCA) in a pressurized water reactor (PWR) CPR1000. A total of 54 input parameters which might have influence on the figure of merit (FOM) were identified, and the sensitivity measure of each input on the FOM was calculated through an optimized moment-independent global sensitivity analysis method. The importance ranking orders of the parameters were transformed into the Savage scores, and the parameters were categorized based on the Savage scores. A parameter importance ranking table for the SBLOCA scenario of the CPR1000 reactor was obtained, and the influences of some important parameters at different break sizes and different accident stages were analyzed.

Best estimate (BE) codes are developed to carry out realistic safety analysis of the nuclear reactor, and generally hundreds of constitutive models are comprised in a BE code. Nevertheless, uncertainties of these constitutive models are often not properly handled in the best estimate plus uncertainty (BEPU) analysis. It is not sufficient to fully evaluate the uncertainties of different sorts of models with only one method. Thereby a structural method for uncertainty quantification (UQ) of constitutive models is proposed and relevant description is presented. Based on the method, constitutive models will be classified into two categories according to the characteristics, namely the independent model and the dependent model, and different methods will be adopted for different sorts of model. Several statistical methods for UQ of constitutive models such as the non-parametric curve estimation method, the Bayesian calibration method and the coverage calibration method are evaluated and the characteristics of these methods are discussed. In addition, several methods for the construction of surrogate model are utilized to reduce the computational cost, and a model selection technique is adopted to opt the optimal model among all alternative models. The large break loss of coolant accident (LBLOCA) experiment LOFT LP-02-6 is utilized to verify the proposed structural method, and the BEPU analysis of the LP-02-6 experiment is carried out. The results show that uncertainty intervals of the identified models obtained through the structural method are reasonable, uncertainties of the peak cladding temperature (PCT) as well as the accumulator injection time (AIT) are quantified, and the sensitivity analysis is carried out to evaluate the influence of different input parameters on the 1st PCT, 2nd PCT and the AIT.

A sub-channel analysis code, ATHAS-02, based on a two-fluid two-field model is developed in this study. The semi-implicit numerical solution scheme based on the staggered mesh system is adopted to get the finite difference equations. A pressure-velocity coupling algorithm is applied to iteratively solve the finite-difference equations. Meanwhile, preliminary validation studies are performed based on the in-tube subcooled boiling experiments, the GE 9-rod bundle experiments and the Studsvik bundle experiments. It is found that the predicted results agree well with the experimental data.

As the development of the nuclear industry, passive technology turns out to be a remarkable characteristic in the design of advanced nuclear power plants. Since the 20th century, much effort has been given to the passive technology, and a number of evolutionary passive systems have been developed. The CPR1000 plant, which is one kind of mature pressurized water plants in China, is proposed to be improved with some passive systems to enhance safety. The proposed passive systems include: (1) the RMT (reactor makeup tank); (2) the ACC (accumulator); (3) the IRWST (in-containment refueling water storage tank); (4) the PEFS (passive emergency feed water system), which is installed on the secondary side of SGs; (5) the PDS (passive depressurization system). In modifying the passive safety system, it is found that the actuation conditions of the passive safety system and the accident assumptions bring great effects on the accidental mitigation. It is necessary to study the influence induced by them. In this paper, the transient phenomenon of SGTR is analyzed to show the process in the modification of the actuation conditions and the assumptions. Finally, the most proper assumptions and actuation conditions are determined. Through the sensitivity analysis, it is found that the RMTs’ and PEFSs’ behaviors bring a great impact on the accident transient. The strong mutual effect between each passive safety component in the SGTR is the reason of choosing this accident to analyze in this paper.

This paper presents an assessment of the heat transfer models under different heat transfer modes for water flow in helically coiled tubes based on the compiled datasets from the reviewed literatures. For single phase flow, most of the correlations of the heat transfer coefficient can fit well to the experiments. The correlations of Xin-Ebadian, Dravid and Kalb-Seader for laminar flow and those of Seban-McLaughlim, Mori-Nakayama, Xin-Ebadian, Hardik, Rogers-Mayhew, Mikaila-Poskas and El-Genk-Schriener for turbulent flow are recommended. For flow boiling heat transfer, Steiner-Taborek correlation could be utilized to predict the boiling heat transfer coefficients in helically coiled tubes for a relatively wide range of parameters. For dryout quality, the correlations of Hwang et al. and Santini et al. give relatively better predictions than others. However, more accurate correlations for flow boiling heat transfer coefficient and dryout quality need to be developed based on further investigations with wider parameter ranges in the future. The present work could provide references for the investigators for future uses of those correlations and for performing further investigations on the heat transfer characteristics of water flow in helically coiled tubes.

The performance of the droplets is vital of importance to the annular-mist flow and the heat transfer for the reflooding conditions. In this research, a two-fluid three-field model for the vertical upward flow is developed. The field equations of the vapor, continuous liquid phase and dispersed liquid droplet phase are established. The set of equations are solved by the semi-implicit difference algorithm based on the staggered grids. Wrutz's steam-water entrainment experiment is simulated to evaluate the performance of the selected closure relations. Becker's and Bennett's dryout and post-dryout heat transfer experiments are simulated to validated the interfacial and wall heat transfer models. Finally, the tubular reflooding experiments are simulated to study the effect of the droplets on the wall heat transfer for different injection conditions. The predicted results are compared to the experimental data and the predicted results by the additional two-fluid model. The results show that the predicted results by this two-fluid three-field model are more accurate than those by the additional two-fluid model.

The mainstream deterministic and probabilistic methods for safety analysis of the nuclear power plant (NPP) are the best estimate plus uncertainty (BEPU) analysis and probabilistic safety analysis (PSA), respectively. With the development of techniques, a more “realistic” simulation of the NPP has become a long-term goal to improve the economic efficiency. Nowadays, the combination of BEPU and PSA, also known as the risk-informed uncertainty method, is under development. In this paper, the cold leg double-ended break accident of a typical pressurized water reactor was selected to be the target scenario for a preliminary study of the combined method. A failure analysis of the emergency core cooling system based on PSA was carried out first, and the failure probability of each event sequence was determined. Then the conditional core damage probability of each event sequence was evaluated through BEPU analysis, and the core damage frequency of the reactor was finally determined.

In-tube condensation of pure vapor are widely adopted by passive safety systems of advanced nuclear reactors to remove the decay heat. Accurate calculation of the in-tube condensation heat transfer coefficient is important for evaluating the heat removal ability. A better understanding of the mechanism of pure vapor condensation heat transfer and its modelling are essential for the design and assessment of those passive safety systems. An analytical model for pure vapor condensation in a vertical tube is developed based on mass, momentum and energy conservation equations. The total axial pressure drop is calculated through the momentum equation of the vapor core assuming a uniform radial pressure distribution. An optimized transition area between the laminar and turbulent film flow is defined to avoid discontinuity of heat transfer coefficients. Based on the sensitivity analyses, Kay’s turbulent eddy diffusivity model and his turbulent Prandtl number model were recommended for the newly proposed model. The predicted results were compared with results of Shah (2016a,b), Chen et al. (1987) and Cavallini et al. (2002) correlations as well as Kuhn’s and KAIST experimental data. It was found that the proposed model predicts the heat transfer coefficient well with a mean absolute deviation of 11.99% for Kuhn’s data and 8.08% for KAIST’s data.

After Fukushima accident, the safety assessments within the full range of operation conditions for the current fleet of nuclear power plants in China are requested to be performed by the national nuclear safety administration. Accordingly, twelve typical cases were simulated to study the thermal-hydraulic behavior of loss of residual heat removal system for CPR1000 under cold shutdown operation with RELAP5/MOD3.3 code. The transient behaviors of system pressure, coolant temperature, water inventory, coolant boiling and core uncovering time, and other important parameters under 12 different plant configurations are predicted. The results indicate that the system pressure of cases with intact system increases continuously to the set value of 3.8 MPa for opening the relief valve in the RHR system without any mitigation action. The starting time of coolant boiling highly depends on the initial water inventory in the RPV as well as its behavior during the initial stage after losing the RHR system, while the starting time of core uncovering highly depends on the plant configurations. The results of the study can provide references for the plant safety improvement and guideline for an operator to strategize a mitigation action.

The process of energy transfer from heat structure to control volume is determined by the wall-to-fluid heat transfer package, which is crucial for nuclear reactor safety analysis codes. The current logic for selection of heat transfer modes of RELAP5/MOD3.3 code is too complex and may result in incorrect heat transfer mode judgment. Also, the narrow application scope of film boiling heat transfer correlations may result in large errors in film boiling region which is of paramount importance for the predicted peak clad temperatures during hypothetical LB-LOCAs in PWRs. In this study, a new heat transfer package has been developed and incorporated into the RELAP5/MOD3.3 code. Differing from the original package, the modified one consists of twelve heat transfer modes and proposes a new logic for selection of heat transfer modes. For each mode, the models in the existing safety analysis codes and the leading models in literature have been reviewed in order to determine the best model which can easily be applicable to the RELAP5/MOD3.3 code. Particularly (1) a new package of heat transfer correlations are produced; (2) a new logic for selection of film boiling and transition boiling heat transfer modes is proposed which use minimum film boiling temperature and critical heat flux temperature as distinguished points. The modified code has been validated by comparing the analysis results with available experimental data from tube post dryout experiments and loss-of-fluid test (LOFT) facility. The calculation results showed that the improved package could better predict the experimental phenomena with higher prediction accuracy.

Best estimate plus uncertainty methodology is an advanced technology for licensing application of the nuclear power plants, but some aspects can still be improved in the conventional input-driven methodologies. For instance, uncertainties of the closure models are not properly handled; local sensitivity analysis methods not suitable for complex nuclear systems are still widely utilized; the sensitivity analysis process needs to be performed again every time new parameters are identified. Optimization researches were carried out in the study. A structural method was proposed to quantify the uncertainties of the closure models, and a low-cost moment-independent global sensitivity analysis method was developed to improve the sensitivity analysis framework. By integrating the improvements into the conventional input-driven framework, an uncertainty method with model treatment and improved sensitivity analysis was obtained. Finally, the method was validated through a large break loss of coolant accident in a three-loop pressurized water reactor.

An upgraded version of ATHAS sub-channel code ATHAS/OE is developed for the investigation of the thermal hydraulic behavior of nuclear reactor core in ocean environment with consideration of heaving and rolling motion effect. The code is verified by another modified sub-channel code based on COBRA-IV and used to analyze the thermal hydraulic characteristics of a typical SMR under heaving and rolling motion condition. The calculation results show that the heaving and rolling motion affect the thermal hydraulic behavior of a reactor significantly.

A space power with high power density, light weight, low cost and high reliability is of crucial importance to future exploration of deep space. Space reactor is an excellent candidate because of its unique characteristics of high specific power, low cost, strong environment adaptability and so on. Among all types of space reactors, heat pipe cooled space reactor, which adopts the passive heat pipe (HP) as core cooling component, is considered as one of the most promising choices and is widely studied all over the world. This paper develops a transient analysis code (TAPIRS) for heat pipe cooled space reactor power system (HPS) based on point reactor kinetics model, lumped parameter core heat transfer model, combined HP model (self-diffusion model, flat-front startup model and network model), energy conversion model of Alkali Metal Thermal-to-Electric Conversion units (AMTEC), and HP radiator model. Three typical accidents, i.e., control drum failure, AMTEC failure and partial loss of the heat transfer area of radiator are then analyzed using TAPIRS. By comparing the simulation results of the models and steady state with those in the references, the rationality of the models and the solution method is validated. The results show the following. (1) After the failure of one set of control drums, the reactor power finally reaches a stable value after two local peaks under the temperature feedback. The fuel temperature rises rapidly, however it is still under safe limit. (2) The fuel temperature is below a safe limit under the AMTEC failure and partial loss of the heat transfer area of radiator. This demonstrates the rationality of the system design and the potential applicability of the TAPIRS code for the future engineering application of heat pipe cooled space reactor.

Previous studies showed that the subchannel codes need most CPU time to solve the matrix formed by the conservation equations. Traditional matrix solving method such as Gaussian elimination method and Gaussian-Seidel iteration method cannot meet the requirement of the computational efficiency. Therefore, a new algorithm for solving the block penta-diagonal matrix is designed based on Stone's incomplete LU (ILU) decomposition method. In the new algorithm, the original block penta-diagonal matrix will be decomposed into a block upper triangular matrix and a lower block triangular matrix as well as a nonzero small matrix. After that, the LU algorithm is applied to solve the matrix until the convergence. In order to compare the computational efficiency, the new designed algorithm is applied to the ATHAS code in this paper. The calculation results show that more than 80 % of the total CPU time can be saved with the new designed ILU algorithm for a 324-channel PWR assembly problem, compared with the original ATHAS code.

In this paper, preliminary safety studies on the 800 MWth accelerator-driven system (ADS) proposed by Xi'an Jiaotong university are presented. The system is a pool type facility coupling a proton accelerator with current in the range of 17–23 mA and a sub-critical core by means of a spallation target. The RELAP5/MOD3.3 code is selected as a base tool. In order to simulate the system, the point kinetics model is modified and the property of lead-bismuth is implemented to meet the requirement of ADS analysis. This paper focuses on the assessment of its response to the loss of flow events. The first part is originated from the failure of the pump and the second part derives from the significant flow blockage at a fuel assembly inlet. The reactivity insertion accidents are caused by the change of the proton beam current. The results show that the safety and criteria are satisfied and the system is tolerant to the loss of flow accidents and proton beam doubled accident and is sensitive to the external neutron changing.

Future exploration of deep space requires space power with high power density, light weight, low cost and high reliability. Space reactor is an excellent candidate with its unique characteristics of high specific power, low cost, strong environment adaptability and so on. Among all types of space reactors, heat pipe cooled space reactor, which adopts the passive heat pipe as core cooling component, is considered as one of the most promising choice and is widely studied all over the world. Startup characteristics of this type space reactor are an active topic. Previous studies mainly focused on the startup from high temperature rather than environmental temperature. In order to simulate the transient startup process from frozen state, a transient analysis code (TAPIRS) for heat pipe cooled space reactor power system (HPS) has been developed and applied to investigate the system transient performance during a startup from zero cold power to full power. The code integrates separately validated point reactor kinetics model, lumped parameter core heat transfer model, combined heat pipe (HP) model (self-diffusion model, flat-front startup model and network model), energy conversion model of alkali metal thermal-to-electric conversion units (AMTEC), and HP radiator model. By comparing the simulation results of the models and steady state with those in the references, the rationality of the models and the solution method is validated. It is found that by adjusting the control drum's rotational speed, the reactor can startup from subcritical state to full power state while the heat pipe and AMTEC from solid state to normal operational state. HPS can startup entirely depending on the nuclear power, and the maximum temperature of the heat pipe does not exceed 1250 K in the whole startup process. The maximum errors of the parameters between the simulation results of this paper and those in the literature at the full power condition are less than 5%. Under the accident of control drum failure with largest reactivity insertion, the fuel temperature can be controlled within the safety limits. These show that the reactor system has characteristics of no single-point failures, the self-stabilization capability under accident conditions.

Hydraulic load analysis is of great importance in the design of pressurized water reactor. It provides load conditions for structural mechanics analysis of primary loop system and ensures the reliability of primary loop pipeline and main equipment. Currently common hydraulic load analysis software mainly includes RELAP5 and WAHA, both of which adopt the classical two-fluid six-equation single pressure model and the first-order differentiation algorithm. The single pressure model is not stable, and its numerical results are prone to oscillation under high-order precision schemes. Meanwhile, the low-order algorithm is not accurate enough, which is not conducive to the simulation of pressure wave problems. However, the two-fluid two-pressure model is unconditionally adaptive. In order to improve the accuracy of the numerical algorithm and ensure the stability of the numerical value, a high-order semi-implicit algorithm for solving the transient term and convection term is developed based on the two-fluid two-pressure model. Finally, the high-order algorithm is applied to the hydraulic load problems such as shock tube, water hammer and blowout. The results show that the algorithm can effectively reduce the numerical diffusion and improve the prediction accuracy.

Helically coiled once-through steam generators being compact in size and most efficient for heat transfer have been widely used in many SMRs (Small and Medium-size Reactor). It is very imperative to study the hydraulic characteristics of the helically coiled once-through steam generator prior to their integration with SMRs. In this study a thermal hydraulics code that uses a two-fluid two-phase flow model with the consideration of thermal non-equilibrium is developed for thermal-hydraulic characteristics analyses of helically coiled steam generators. A pressure-velocity coupling algorithm is applied to solve equations and a set of heat transfer correlations and frictional pressure drop correlations are embedded into the code. Validation of the code is done using experimental data of pressure drop and heat transfer for helical coil tubes. The code is also tested for simulation of the transient as well as steady state thermal-hydraulic characteristics of OTSG (Once-Through Steam Generator) of IRIS (International Reactor Innovative and Secure) and results are compared with the design data and the RELAP5’s result, the simulation results are found in close agreement with these available results.

The best estimate plus uncertainty (BEPU) methodology is developed to carry out realistic safety analysis of the nuclear power plant. However, the sensitivity analysis in the conventional methodologies often leads to inaccurate results when lacking of the phenomenon identification and ranking table (PIRT) or adopting the local sensitivity analysis methods. As the initial inputs determined in short of the PIRT may neglect some important parameters or identify some unimportant parameters, and the local methods are not appropriate for complex and non-linear systems. In this study, an optimized moment-independent global sensitivity analysis method with very low computational cost was utilized to improve the sensitivity analysis framework of the BEPU methodology. The reliability as well as the applicability of the method were assessed through some tests and a large break loss of coolant accident (LBLOCA) of the LOFT facility. The improved sensitivity analysis framework was then applied to the BEPU analysis of a typical three-loop pressurized water reactor CPR1000, and the reliability of the optimized moment-independent method was further confirmed by comparing the sensitivity analysis result of the method to the results of the BEMUSE project and the Spearman method. Besides, two new parameters were identified to have great influence on the PCT, and the two parameters should be taken into consideration in the uncertainty analysis of the BEPU methodology.

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.

The best estimate plus uncertainty (BEPU) is an advanced method which has been widely adopted in nuclear power plant safety analysis. In order to investigate the transient behaviors of the pressurized water reactor CPR1000 during the cold leg small break loss of coolant accident (SBLOCA) at different break sizes, BEPU analysis of the SBLOCA was carried out in this study, and the sizes of the break were 0.2%, 0.4%, 0.6%, 0.8% and 1.0% of the cross-sectional area of the cold leg, respectively. The minimum collapsed level (MCL) in the active core section was determined to be the figure of merit (FOM). The base case calculations, uncertainty propagation calculations and sensitivity analysis calculations were performed with the NUSOL-SYS code. Based on the results, the transient behaviors were analyzed, the lower 95/95 tolerance limits of the MCL at different break sizes were quantified, and the important input parameters were determined.

The Ransom water faucet problem has become one of the most important benchmark tests to study two-fluid/phase flow because it exists an analytical discontinuous solution. The water faucet problem is a gravity-driven wave problem and its analytical solution is derived from the liquid free fall motion. In this paper, the opposite gravity-driven wave problem named the reversed water faucet problem where the liquid admits a rising motion with reduced speed driven by gravity is studied. With assumptions of decoupled phasic pressures, approximate incompressible flow, no liquid pressure gradient, no phase change, no wall and interfacial drag, the analytical solutions of the gas volume fraction and liquid velocity distribution are derived. From the gas volume fraction analytical solution, there is a moving discontinuity which is a very important point for testing accuracy of the numerical scheme and its stability near discontinuities. Two-fluid seven-equation two-pressure model is of particular interest due to the nature of inherent well-posed advantage in all situations. What’s more, high-order accuracy schemes have attracted great increasing attention to overcome the challenge of serious numerical diffusion from 1st-order scheme for accurately simulating many nuclear thermal-hydraulics applications such as long term transient natural circulation problems. In this paper, the solution algorithms with high-order accuracy in space and time are developed for this well-posed two-fluid model and its robustness and accuracy are verified and assessed against the derived analytical solutions. The numerical results show that high-order schemes could prevent excessive numerical diffusion and are more accurate than first-order time and space schemes; the space high-order scheme could give more accurate numerical results than the time high-order scheme for discontinuous solutions.

The best estimate plus uncertainty (BEPU) method is recommended by IAEA for nuclear safety analysis. Most of the existing BEPU methodologies rely on the uncertainty propagation of input parameters, while uncertainties of the constitutive models in best estimate codes tend not to be valued or even neglected. A structural method is proposed in this paper to quantify the uncertainties of the constitutive models. Different constitutive models will be classified according to the characteristics and corresponding method could be utilized for each model based on the method. Specific uncertainty quantification (UQ) methods adopted in this paper include the non-parametric curve estimation method, inverse method and design of experiment (DOE) method, and a model selection technique is adopted to opt the optimal model among all alternative models. The structural method is applied to the uncertainty evaluation of LOFT LP-02-6 experiment. Important models are identified, uncertainties of these models are quantified and propagated to the peak cladding temperature (PCT) through code calculations. Uncertainty of the PCT is quantified and the result shows that the calculated values could well envelop the experimental value.

Due to their advantages, such as compactness and high efficiency in heat transfer, helically coiled heat exchangers have been widely used by different type of nuclear power plants, especially by small and medium size reactors (SMRs). In order to analyze the thermal-hydraulic characteristics of a helical coiled once through steam generator (OTSG) for a small integral pressurized water reactor, a computer code is developed in this paper. The code is based on two-fluid model. The constitutive correlations are recommended based on the assessments with the compiled databases from the reviewed literatures. NUSOL SG is validated and verified against heat transfer in helical coiled tubes, and the calculation results agree well with the experiment data. The present study could provide references for the investigators to perform further investigations on the thermal hydraulic characteristics of helical coiled OTSGs.

The effectiveness of heat transfer from the fuel rods to the moderator in a Canadian, pressure-tube type, Super-Critical Water-cooled Reactor (SCWR) has been assessed for the postulated loss-of-coolant accident (LOCA) coupled with the total loss of emergency core cooling (LOECC) event using the safety analysis code “SCTRAN”. A radiation heat-transfer model, including a two-dimensional heat conduction solution scheme, has been incorporated into the SCTRAN code for this assessment. The assessment result shows that the combined radiation heat transfer from the fuel rods to the fuel channel and natural convective heat transfer from the fuel rods to the steam are effective in removing decay heat for the LOCA/LOECC scenario. Maximum cladding temperatures of fuel rods in inner and outer rings of the fuel assembly in the highest-power channel are predicted at 1278 °C and 1192 °C respectively, which are lower than the melting point of the modified stainless-steel cladding. Sensitivity analyses of several analytical parameters on the predicted maximum cladding temperatures have been performed.

As compared to the two-fluid single-pressure model, the two-fluid seven-equation two-pressure model has been proved to be unconditionally well-posed in all situations, thus existing with a wide range of industrial applications. The classical 1st-order upwind scheme is widely used in existing nuclear system analysis codes such as RELAP5, CATHARE, and TRACE. However, the 1st-order upwind scheme possesses issues of serious numerical diffusion and high truncation error, thus giving rise to the challenge of accurately modeling many nuclear thermal-hydraulics problems such as long term transients. In this paper, a semi-implicit algorithm based on the finite volume method with staggered grids is developed to solve such advanced well-posed two-pressure model. To overcome the challenge from 1st-order upwind scheme, eight high-resolution total variation diminishing (TVD) schemes are implemented in such algorithm to improve spatial accuracy. Then the semi-implicit algorithm with high-resolution TVD schemes is validated on the water faucet test. The numerical results show that the high-resolution semi-implicit algorithm is robust in solving the two-pressure two-fluid two-phase flow model; Superbee scheme and Koren scheme give two highest levels of accuracy while Minmod scheme is the worst one among the eight TVD schemes.

The advanced accumulator is designed to improve the safety and reliability of CPR1000 by providing a small injection flow to keep the reactor core in flooded condition. Thus, the core still stays in a cooling state without the intervention of low pressure safety injection and the startup grace time of the low pressure safety injection pump can be greatly extended. A new model for the advanced accumulator, which is based on the basic conservation equations, is developed and incorporated into RELAP5/MOD 3.3. The simulation of the advanced accumulator can be carried out and results show that the behavior of the advanced accumulator satisfied its primary design target. There is a large flow in the advanced accumulator at the initial stage. When the accumulator water level is lower than the stand pipe, a vortex appears in the damper, which results in a large pressure drop and a small flow. And then the sensitivity analysis is performed and the major factors which affected the flow rate of the advanced accumulator were obtained, including the damper diameter, the initial volume ratio of the water and the nitrogen and the diameter ratio of the standpipe and the small pipe. Additionally, the primary coolant loop cold leg double-ended guillotine break LBLOCA in CPR1000 with advanced accumulator is analyzed. The results show that the criterion for maximum cladding temperature limit (1477 K) ( NRC, 1992 ) can be met ever with 200 s after the startup of the low pressure safety injection. From this point of view, passive advanced accumulator can strive a longer grace time for LPSI. Thus the reliability, safety and economy of the reactor system can be improved.

This paper describes the preliminary safety analysis of the Chinese Supercritical water cooled Reactor (CSR1000), which is proposed by Nuclear Power Institute of China (NPIC). The two-pass core design applied to CSR1000 decreases the fuel cladding temperature and flattens the power distribution of the core at normal operation condition. Each fuel assembly is made up of four sub-assemblies with downward-flow water rods, which is favorable to the core cooling during abnormal conditions due to the large water inventory of the water rods. Additionally, a passive safety system is proposed for CSR1000 to increase the safety reliability at abnormal conditions. In this paper, accidents of “pump seizure”, “loss of coolant flow accidents (LOFA)”, “core depressurization”, as well as some typical transients are analysed with code SCTRAN, which is a one-dimensional safety analysis code for SCWRs. The results indicate that the maximum cladding surface temperatures (MCST), which is the most important safety criterion, of the both passes in the mentioned incidents are all below the safety criterion by a large margin. The sensitivity analyses of the delay time of RCPs trip in “loss of offsite power” and the delay time of RMT actuation in “loss of coolant flowrate” were also included in this paper. The analyses have shown that the core design of CSR1000 is feasible and the proposed passive safety system is capable of mitigating the consequences of the selected abnormalities.

Rod bundle experiments with axially uniform and nonuniform heat fluxes are examined to explore the potential limitations of using uniform rod bundle CHF data for CHF correlation development of light water reactors with nonuniform axial power distribution (APD). The case of upstream burnout is presented as an example of unique phenomena associated with nonuniform rod bundle CHF experiments. It is a result from combined effect of axial nonuniform power shape and different interchannel mixing mechanisms. In addition, several key parameters are investigated with respect to their potential impacts on the thermal-hydraulic behaviors between rod bundles with uniform and nonuniform APDs. This type of misrepresentation cannot be amended or compensated through the use of correction factors due to the lack of critical information in the uniform rod bundle CHF testing as well as the fundamental difference in the underlining driving mechanisms. Other potential issues involved with the use of uniform rod bundle CHF data for nonuniform APD system applications also present strong evidence concerning the limitations and inadequacy of using uniform rod bundle CHF data for the correlation, prediction, and design limit calculation for safety analysis.

Since the Fukushima accident in 2011, more and more attention has been paid to nuclear reactor safety. A number of evolutionary passive systems have been developed to enhance the inherent safety of reactors. This paper presents a passive safety system applied on CPR1000, which is a traditional generation II+ reactor. The passive components selected are as follows: (1) the reactor makeup tanks (RMTs); (2) the advanced accumulators (A-ACCs); (3) the passive emergency feedwater system (PEFS); (4) the passive depressurization system (PDS); (5) the in-containment refueling water storage tank (IRWST). The model of the coolant system and the passive systems was established by utilizing a system code (RELAP5/MOD3.3). The SBLOCA (small-break loss of coolant) was analyzed to test the passive safety systems. When the SBLOCA occurred, the RMTs were initiated. The water in the RMTs was then injected into the pressure vessel. The RMTs’ low water level triggered the PDS, which depressurized the coolant system drastically. As the pressure of the coolant system decreased, the A-ACCs and the IRWST were put to work to prevent the uncovering of the core. The results show that, after the small-break loss-of-coolant accident, the passive systems can prevent uncovering of the core and guarantee the safety of the plant.

Design analysis is one of the main difficulties during the research and design of SCWRs. Currently, the development of safety analysis code for SCWR is still in its infancy all around the world, and very few computer codes could carry out the trans-critical calculations where significant changes in water properties would take place. In this paper, a safety analysis code SCTRAN for SCWRs has been developed based on code RETRAN-02, the best estimate code used for safety analysis of light water reactors. The ability of SCTRAN code to simulate transients where both supercritical and subcritical regimes are encountered has been verified by comparing with APROS and RELAP5-3D codes. Furthermore, the LOFA and LOCA transients for the CGNPC SCWR design were analyzed with SCTRAN code. The characteristics and performance of the passive safety systems applied to CGNPC SCWR were evaluated. The results show that: (1) The SCTRAN computer code developed in this study is capable to perform design analysis for SCWRs; (2) During LOFA and LOCA accidents in a CGNPC SCWR, the passive safety systems would significantly mitigate the consequences of these transients and enhance the inherent safety.

This paper proposes a design of passive cooling system for CPR1000 spent fuel pool (SFP). Our design can effectively manage the SFP temperature not to exceed 80 °C. Then the transient analysis of the CPR1000 SFP with designed passive cooling system is carried out in station blackout (SBO) accident by the best-estimate thermal-hydraulic system code RELAP5. The simulation results show that to maintain the temperature of CPR1000 SFP under 80 °C, the numbers of the SFP and air cooling heat exchangers tubes are 6627 and 19 086, respectively. The height difference between the bottom of the air cooling heat exchanger and the top of the SFP heat exchanger is 3.8 m. The number of SFP heat exchanger tubes decreases as the height difference increases, while the number of the air cooling heat exchanger tubes increases. The transient analysis results show that after the SBO accident, a stable natural cooling circulation is established. The surface temperature of CPR1000 SFP increases continually until 80 °C, which indicates that the design of the passive air cooling system for CPR1000 SFP is capable of removing the decay heat to maintain the temperature of the SFP around 80 °C after losing the heat sink.

This paper presents CFD analyses in heat unsymmetric subchannels and heat symmetric seven-rod bundle geometries of a Super Fast Reactor (Super FR) fuel assembly using STAR-CD. The purpose of CFD analyses in heat unsymmetric subchannels is to evaluate the effect of the power differences on the heat transfer in subchannels of the Super Fast Reactor. For heat symmetric seven-rod bundles, the effects of the gap clearance between the fuel rod and the assembly wall and the displacement of the fuel rod on the circumferential temperature distributions and Maximum Cladding Surface Temperature (MCST) are analyzed. The results show that larger power difference between fuel rods gives larger circumferential temperature difference of the hottest fuel rods. Considering cross flow between edge and ordinary subchannels, 1 mm gap between the fuel rod and the assembly wall is better for small MCST although the circumferential temperature difference in edge subchannel is large. MCST increases exponentially with the displacement. The relative error of displacement should be less than 1% if the allowable increment of MCST due to displacement is less than 6 °C.

This paper presents CFD analyses of heat transfer in subchannels of a Super Fast Reactor fuel assembly. Analyses are concentrated on the circumferential temperature distribution on the cladding outer surface because the Maximum Cladding Surface Temperature (MCST) has been a crucial design parameter to evaluate fuel cladding integrity of the Super Fast Reactor. Speziale non-linear high Re k–ɛ model, which can reproduce the anisotropic turbulence flow in non-circular flow channels, with two-layer near-wall treatment is adopted. The results show that heat conduction in the cladding should be considered in the CFD analyses. Larger circumferential temperature gradient occurs on the cladding surface in the edge and corner subchannels than that in the ordinary subchannel because of their special geometries causing larger heterogeneity of mass flow rate distribution inside the subchannels. Improved subchannel configurations to reduce the circumferential temperature gradient are proposed. This study will be a good guideline to the future core design improvement.

This article presents a theoretical investigation on the steady-state natural circulation characteristics of a new type of pressurized water reactor. Through numerically solving the one-dimensional steady-state single-phase conservative equations for the primary circuit and the steady-state two-phase drift-flux conservative equations for the secondary side of the steam generator, the natural circulation characteristics were studied. On the basis of the preliminary calculation analysis, it was found that natural circulation mass flow rate was proportional to the exponential function of the power and that the value of the exponent is related to the operating conditions of the secondary side of the steam generator. The higher the outlet pressure of the secondary side of the steam generator, the higher the primary natural circulation mass flow rate. The larger height difference between the core center and the steam generator center is favorable for the heat removal capacity of the natural circulation.

This paper presents a separated flow model of annular upward flow in a vertical narrow annular channel with bilateral heating. The theoretical model is based on fundamental conservation principles: the mass, momentum and energy conservation equations of liquid films and the momentum conservation equation of vapor core. Through numerically solving the equations, liquid film thickness, radial velocity and temperature distribution in liquid films, heat transfer coefficient of inner and outer tubes and axial pressure gradient are obtained. The predicted results of heat transfer coefficients and axial pressure gradient are compared with the experimental data and good agreements between them are found.

Large break LOCA (LBLOCA) is one of the limit design basic accidents in nuclear power plant. The large flow water in the advanced accumulator is injected into primary loop in early short time. When the vessel pressure drops and reactor core is re-flooded, the advanced accumulator provides a small injection flow to keep the reactor core in flooded condition. Thus, the startup grace time of the low pressure safety injection pump is extended, and the core still stays in a long-term cooling state. By deducing the original accumulator model in RELAP5 accident analysis code, a new model combining the advanced and the traditional accumulator is obtained and coupled into RELAP5/ MOD 3.3. Simulation results show that there is a large flow in the advanced accumulator at the initial stage. When the accumulator water level is lower than the stand pipe, a vortex appears in the damper, resulting in a large pressure drop and small flow. The phenomenon meets the demand of the advanced accumulator design and the simulation of the advanced accumulator is accomplished successfully. Based on this, the primary coolant loop cold leg double-ended guillotine break LBLOCA in CPR1000 is analyzed with the modified RELAP5 code. When the double ended cold leg guillotine accident with 200s delayed startup of the low pressure safety injection occurs, maximum cladding temperature in the core with traditional accumulator is 1860K which seriously exceeded the safety temperature (1477K)[1] prescribed limits while the maximum cladding temperature with advanced accumulator has the security temperature-1277K. From this point of view, adopting passive advanced accumulator can strive a longer grace time for LPSI. Thus the reliability, security and economy of reactor system were improved.Copyright © 2013 by ASME

According to the calculated axial and radial power distribution of reactor core, a numerical analysis of thermal-hydraulic behavior of coolant in supercritical water-cooled reactor of CGNPC (CGN-SCWR) was performed by using sub-channel analysis software ATHAS. The coolant flow in CGN-SCWR reactor core were in two path mode. In order to ensure the maximum cladding temperature (MCT for short) reached the requirements throughout the entire life cycle, the sub-channel analysis of the MCT in the BOC, MOC and EOC had been carried out. The temperature and density properties of coolant and moderator in the fuel assembly along the axial and radial direction were investigated. The MCT of the fuel assembly was obtained, and the selected typical sub-channels were analyzed in detail. The results show that the MCT, coolant outlet temperature and other parameters in CGN-SCWR reactor core meet the thermal-hydraulic design requirements through the whole life time.Copyright © 2013 by ASME

The spent fuel pool (SFP) is mainly used for cooling spent fuel assemblies (SFAs) discharged from the reactor core. Besides, it can also shield the radiation. The decay heat can be removed through normal operation cooling system, otherwise it can only rely on the natural circulation in the pool when the coolant pump loses power or the heat exchanger fails. Thus the pool water temperature will continue to rise until it begins to boil. During this period, if no active cooling measures are triggered, the water level will gradually decrease as water boiling. Once the water level drops to the top of the fuel assemblies, the fuels begin to be exposed in the environment. In this paper, the CPR1000 spent fuel pool was chosen as the analysis object and the best estimate system thermal hydraulic code RELAP5 was employed to investigate the process in spent fuel pool in case of loss of heat sink. The results of calculations show that when losing two sets of cooling line, the increase in water temperature in the pool from 55 °C up to 100 °C takes approximately 9.1 h, the evaporation of water volume above the SFAs takes approximately 75.4 additional hours; while in case of losing one set of cooling line, the water temperature of the pool surface reaches 76.6 °C, at which the pool water would not going to boil under the atmospheric environment condition. The results of performed analysis are useful for safety analysis and storage of the SFAs, and can be used to provide a reference for spent fuel pool engineering design.