China Experimental Fast Reactor Research Articles

Predicting the core thermal hydraulic parameters with a gated recurrent unit model based on the soft attention mechanism

Accurately predicting the thermal hydraulic parameters of a transient reactor core under different working conditions is the first step toward reactor safety. Mass flow rate and temperature are important parameters of core thermal hydraulics, which have often been modeled as time series prediction problems. This study aims to achieve accurate and continuous prediction of core thermal hydraulic parameters under instantaneous conditions, as well as test the feasibility of a newly constructed gated recurrent unit (GRU) model based on the soft attention mechanism for core parameter predictions. Herein, the China Experimental Fast Reactor (CEFR) is used as the research object, and CEFR 1/2 core was taken as subject to carry out continuous predictive analysis of thermal parameters under transient conditions., while the subchannel analysis code named SUBCHANFLOW is used to generate the time series of core thermal-hydraulic parameters. The GRU model is used to predict the mass flow and temperature time series of the core. The results show that compared to the adaptive radial basis function neural network, the GRU network model produces better prediction results. The average relative error for temperature is less than 0.5 % when the step size is 3, and the prediction effect is better within 15 s. The average relative error of mass flow rate is less than 5 % when the step size is 10, and the prediction effect is better in the subsequent 12 s. The GRU model not only shows a higher prediction accuracy, but also captures the trends of the dynamic time series, which is useful for maintaining reactor safety and preventing nuclear power plant accidents. Furthermore, it can provide long-term continuous predictions under transient reactor conditions, which is useful for engineering applications and improving reactor safety.

Challenges in high-fidelity thermal–hydraulic simulation of SFR cores: Insights and PACA-S4FR solutions

Numerical simulations in studying thermal–hydraulic behavior in sodium-cooled fast reactors (SFR) have gained attention for their safety and cost advantages. While low-fidelity numerical simulations commonly used in industry have the advantage of lower computational costs, they face challenges in accurately representing intricate flow phenomena. On the other hand, high-fidelity numerical simulations offer excellent characterization of turbulence but demand significant computational resources. The emergence of exascale supercomputers has provided opportunities and challenges for large-scale high-fidelity numerical simulations. In this paper, firstly, we review the research progress of thermal–hydraulic numerical simulations of SFR cores using both low-fidelity and high-fidelity methods. Secondly, we estimate the computational and storage requirements for high-fidelity numerical simulations of SFR core based on calculations performed on China Experimental Fast Reactor (CEFR). Subsequently, we analyze the challenges of conducting high-fidelity thermal–hydraulic simulations of the entire SFR core, including large-scale mesh generation, appropriate numerical discretization methods, efficient simulation acceleration techniques, coupled simulations, and the complexities of validation and verification processes. Finally, we present the research progress of PACA-S4FR, a high-fidelity thermal–hydraulic simulation software in the China Virtual Reactor (CVR1.0) project, along with the ongoing efforts to address challenges in high-fidelity simulations. This paper contributes to advancing the comprehension of SFR core thermal hydraulics, potentially aiding in safety and performance enhancements in SFR.

CEFR control rod drop transient simulation using RAST-F code system

This study aimed to verify and validate the transient simulation capability of the hybrid code system RAST-F for fast reactor analysis. For this purpose, control rod (CR) drop experiments involving eight separate CRs and six CR groups in the China Experimental Fast Reactor (CEFR) start-up tests were utilized to simulate the CR drop transient. The RAST-F numerical solution, including the neutron population, time-dependent reactivity, and CR worth, was compared against the measurement values obtained from two out-of-core detectors. Moreover, the time-dependent reactivity and CR worth from RAST-F were verified against the results obtained by the Monte Carlo code Serpent using continuous energy nuclear data. A code-to-code comparison between Serpent and RAST-F showed good agreement in terms of time-dependent reactivity and CR worth. The discrepancy was less than 160 pcm for reactivity and less than 110 pcm for CR worth. RAST-F solution was almost identical to the measurement data in terms of neutron population and reactivity. All the calculated CR worth results agreed with experimental results within two standard deviations of experimental uncertainty for all CRs and CR groups. This work demonstrates that the RAST-F code system can be a potential tool for analyzing time-dependent phenomena in fast reactors.

Uncertainty analysis of the China experimental fast reactor start-up tests based on refined calculation results

In this paper, the numerical calculations of China Experimental Fast Reactor (CEFR) start-up tests were performed by using SARAX code with the improved core model. The results of criticality, control rod worth, and reactivity effect were compared against the experimental data. The numerical results showed the SARAX code could be reliably used for CEFR. Moreover, the sensitivity of the keff value due to the change of input parameters was analyzed. At the same time, the uncertainty quantification of calculated keff value and control rod worth and detailed contribution to the total uncertainty were analyzed. For the reactor loaded with highly enriched UO2 fuel, the uncertainty of keff value could reach more than 2%. Besides, the capture reaction of 235U contributed the most with a value of 1.91%. The uncertainty analysis of the control rod worth showed that the total uncertainty depended on the control rod layout rather than the composition.

Transient three-dimensional numerical simulation on asymmetrical flow and thermal characteristics of CEFR sodium pool under one secondary pump shaft stuck accident

The China Experimental Fast Reactor (CEFR) primary circuit adopts a pool-type structure. The sticking of the main circulating pump shaft in the secondary circuit is an important design-based accident for CEFR, and it is necessary to investigate the three-dimensional thermal-hydraulic phenomena in the sodium pool under the accident. However, conducting relevant experiments directly is almost impossible due to the large dimensions and intricate flow paths within the reactor. In the present work, the “modular modeling and integrated coupling calculation” method is used to build the detailed model and calculate the primary circuit of CEFR in a 1:1 scale. The three-dimensional thermal-hydraulic phenomena inside the sodium pool under the accident condition of one secondary circuit pump shaft stuck are simulated. According to the calculation results, the backflow phenomenon of the fault loop pressure pipe and the intermediate heat exchanger (IHX) begins to appear around 50 s. An obvious thermal stratification phenomenon occurs in the hot pool at the beginning of the accident. As the accident progresses, the temperature difference between the hot and cold pools decreases gradually, and the thermal stratification of the hot pool gradually diminishes. The temperature difference between the inlet and outlet of the same IHX decreases from a maximum of 170 °Cto 2 °C. During the course of the accident, the peak temperatures calculated for the fuel assembly cladding and fuel pellet are 541.78 °C and 1040.17 °C, respectively, which is far below the designed temperature limit of 800 °C and 2800 °C, respectively. Moreover, comparative analysis is conducted on the simulation results of pump shaft stuck in the primary and secondary circuits. The results suggest that when the shaft of one secondary circuit pump is stuck, the average temperature of the sodium at the outlet of the core is higher and the temperature difference between the inlet of different loop pumps is greater, which places more substantial demands on the reactor component structure. It provides significant numerical reference for the safety design and assessment of CEFR.

Experimental study on design optimization of the core catcher for the CEFR

ABSTRACT In severe accidents, sodium-cooled fast reactors (SFRs) are more concerned with core disruptive accidents (CDAs), and in-vessel retention (IVR) is an important mitigation measure to ensure the integrity of the reactor vessel. The distribution and accumulation uniformity of the melt in the lower chamber seriously affects the residual heat discharge capability. The installation of a core catcher is an effective mitigation method, and its structure and arrangement play a key role in mitigating severe accidents. In this study, the experimental facility is designed based on the core catcher of the China Experimental Fast Reactor (CEFR). An experimental platform is established to simulate the falling and relocation of melts in the core catcher under the condition of CDAs. At the same time, a quantitative evaluation index is established to analyze the influence of the number of chimneys, the inclination angle of the chimney canopy, the central chimney height, and the spacing of chimneys on relocation of melts in the core catcher. Finally, some suggestions for correlation optimization design are given.

Feasibility Study to Byproduce Medical Radioisotopes in a Fusion Reactor.

Currently, international nuclear fission reactors producing medical isotopes face the problem of shutdown and maintenance, decommissioning, or dismantling, while the production capacity of domestic research reactors for medical radioisotopes is inadequate, and the supply capacity for medical radioisotopes faces major challenges in the future. Fusion reactors are characterized by high neutron energy, high flux density, and the absence of highly radioactive fission fragments. Additionally, compared to fission reactors, the reactivity of the fusion reactor core is not significantly affected by the target material. By building a preliminary model of the China Fusion Engineering Test Reactor (CFETR), a Monte Carlo simulation was performed for particle transport between different target materials at a fusion power of 2 GW. The yields (specific activity) of six medical radioisotopes (14C, 89Sr, 32P, 64Cu, 67Cu, and 99Mo) with various irradiation positions, different target materials, and different irradiation times were studied, and compared with those of other high-flux engineering test reactors (HFETR) and the China Experimental Fast Reactor (CEFR). The results show that this approach not only provides competitive medical isotope yield, but also contributes to the performance of the fusion reactor itself, e.g., tritium self-sustainability and shielding performance.

Numerical simulation of natural circulation characteristics under different DHX layout schemes in pool-type SFR during station blackout accident

Decay heat removal system (DHRS) is an essential system for pool-type sodium-cooled fast reactor (SFR) to ensure natural circulation and remove decay heat during station blackout (SBO) accident. Decay heat exchanger (DHX) is the key equipment in DHRS. In China Experimental Fast Reactor (CEFR), DHX layout scheme consisting of two DHXs in hot pool (DHX Layout 1) is adopted. Different DHX layout scheme is designed in the advanced pool-type SFR, for example DHXs are installed in both cold and hot pool (DHX Layout 2). It is necessary to analyze the influence of different DHX layout schemes. In this paper, two DHX layout schemes were proposed. Numerical simulation was carried out, of which method was preliminarily validated. The numerical results show that DHX Layout 2 can reach long-term cooling stage of natural convection earlier than DHX Layout 1 with strengthened flow rate, resulting in more uniform temperature distribution in the hot pool. There is a local asymmetrical temperature distribution in the cold pool for DHX Layout 2. These findings by numerical simulations provide important numerical reference for the subsequent DHRS design of the advanced pool-type SFR.

Verification and Validation of the SPL Module of the Deterministic Code AZNHEX through the Neutronics Benchmark of the CEFR Start-Up Tests

A new module for the AZtlan Nodal HEXagonal (AZNHEX) code, which is part of the AZTLAN Platform, was recently developed based on the Simplified Spherical Harmonics (SPL) scheme to deal with the challenges presented in small fast reactor cores, such as the China Experimental Fast Reactor (CEFR), with high leakage and significant scattering effects. For the verification and validation process, we generated nodal homogenized macroscopic cross-sections (XS) through a full heterogeneous core model using the stochastic code SERPENT and subsequently, these XS were employed in AZNHEX. To verify the SPL implementation, several mesh sensitivity exercises were performed demonstrating that the SPL module was implemented successfully. Furthermore, to validate the code with this new implementation, we modeled some exercises contained in the CEFR benchmark with AZNHEX and compared the results with the experimental data available. The final results show a great improvement compared with the original diffusion solver reducing the deviations significantly from experimental data. In conclusion, it is shown and discussed the relevance of improved numerical models (transport approximations instead of diffusion) for the deterministic calculations of small fast reactors.

Neutronics simulation of China Experimental Fast Reactor start-up tests using FARCOB and ERANOS 2.1 code systems

The neutronics analysis of selected start-up tests conducted from 2010 to 2011 in China Experimental Fast Reactor (CEFR) is carried out using in-house FARCOB and European ERANOS-2.1 code systems as a part of IAEA coordinated research project (CRP) on “Neutronics Benchmark of CEFR Start-Up Tests”. This exercise has the main objective of validation and verification of physical models, neutronics simulation methodology including cross-section libraries that is used for fast reactor core simulations. It is challenging to simulate such a small compact core of enriched uranium with high neutron leakage having stainless steel radial reflector by using 3-D diffusion theory codes. This benchmark analysis used a new simplified approach such as a 3-step method and experimental method for simulating temperature and sodium void reactivity coefficients. Net criticality, control rod worth, temperature and sodium void reactivity coefficients and swap reactivity could be predicted well within their measured error values. This exercise provided an additional validation of the FARCOB system against small sodium-cooled fast reactor (SFR) cores, and it can be confidently used for the design and analysis of small experimental reactor cores.

Preliminary verification of multi-group cross-sections generation and locally heterogeneous transport calculation using OpenMC with CEFR start-up tests benchmark

The multi-group (MG) Monte-Carlo (MC) scheme is a hybrid method based on MC solvers and deterministic strategies. This paper focuses on the study of MG schemes for the analysis of fast reactors. The MG scheme in this paper is composed of two steps. In the cross-section generation step, the OpenMC continuous-energy (CE) mode is used with 2D or 3D high-fidelity cluster geometry. In the core calculation step, both homogeneous and locally heterogeneous core models are computed using OpenMC in multi-group mode. The MG schemes are applied and validated in the China Experimental Fast Reactor (CEFR) start-up tests and compared to the CE scheme. The results show an overprediction of core criticality. The locally heterogeneous scheme improves significantly the accuracy in predicting control rod worth. Moreover, synthesizing the results from deterministic code, the locally heterogeneous scheme exhibits a stable performance across different codes, core sizes, and control rod designs. The locally heterogeneous core model with cross-sections generated from the 3D cluster has a good ability in the prediction of feedback coefficients including the sodium void reactivity, temperature reactivity, and subassembly swap reactivity. The computation speed of the locally heterogeneous MG scheme is almost the same as the homogenous MG scheme and its core calculation uses only 7.2% computation time of the reference CE scheme.

Investigation of the Effect of Rod Worth Uncertainty on the Reactivity Measurements of CEFR Start-Up Tests via McCARD Monte Carlo Calculations

In the reactivity measurements of the China Experimental Fast Reactor (CEFR) start-up tests, five independent control rods were moved to compensate for the reactivity insertion. Since the recorded control rod positions were converted to reactivity via S-curves (rod worth curves), any uncertainty in the S-curves can propagate to all reactivity measurements. In this work, we rigorously derived the uncertainty of the reactivity in terms of the statistical uncertainty of the S-curves with Monte Carlo (MC) simulations. Additionally, the average error of the control rod worth from the MC calculation was estimated from experiments and embedded into the uncertainty formulation. The formulation shows that the error of the reactivity is highly correlated to the changes in the control rod position during the reactivity measurements. McCARD MC simulations were then conducted for the CEFR start-up tests, and the calculated reactivity and uncertainty were compared with the measurements. The main error factor of each reactivity calculation was figured out by quantifying the uncertainty components. With the uncertainty formulation, the calculation results showed a better agreement compared with the measurements, as the relative errors were observed mostly within 2σ of the uncertainty.

Prevalence and factors associated with sarcopenia in a cohort of patients with chronic pancreatitis

The neutronics analysis of selected start-up tests conducted from 2010 to 2011 in China Experimental Fast Reactor (CEFR) is carried out using in-house FARCOB and European ERANOS-2.1 code systems as a part of IAEA coordinated research project (CRP) on “Neutronics Benchmark of CEFR Start-Up Tests”. This exercise has the main objective of validation and verification of physical models, neutronics simulation methodology including cross-section libraries that is used for fast reactor core simulations. It is challenging to simulate such a small compact core of enriched uranium with high neutron leakage having stainless steel radial reflector by using 3-D diffusion theory codes. This benchmark analysis used a new simplified approach such as a 3-step method and experimental method for simulating temperature and sodium void reactivity coefficients. Net criticality, control rod worth, temperature and sodium void reactivity coefficients and swap reactivity could be predicted well within their measured error values. This exercise provided an additional validation of the FARCOB system against small sodium-cooled fast reactor (SFR) cores, and it can be confidently used for the design and analysis of small experimental reactor cores.

Validation of the KIKO3DMG neutronics code on the CEFR start-up tests

Selected start-up tests of the China Experimental Fast Reactor (CEFR) were calculated using the KIKO3DMG nodal neutronics code’s diffusion and SP3 solvers. The static cases of fuel loading and first criticality, control rod worth, void, subassembly swap, temperature reactivity measurements, as well as the dynamic cases of rod drop transients were investigated. A new reactivity curve smoothing method was implemented in the nodal code to eliminate the cusping effect due to the movement of the absorber. The excellent agreement between the KIKO3DMG calculation, the CEFR experimental measurement, and the reference Serpent Monte Carlo calculation results validates the KIKO3DMG code on a fast reactor for the first time.

A reaction rate measurement experimental program of China experimental fast reactor

The study of the nuclear reaction rate distribution in the MOX core of China Experimental Fast Reactor is a necessary condition for obtaining the operating license. From the characteristics of core structure, both the principle of experimental and experimental system are studies, emphasized on the scheme design such as the size of foils, the irradiation power and counting rate. Firstly, based on the Monte Carlo Code（MCNP）and NJOY codes, the distribution of reaction rates is obtained. MCNP is a general-purpose Monte Carlo N-Particle code that can be used for neutron, photon, electron, and so on. The NJOY nuclear data processing system is a comprehensive computer code package for producing to cross sections and related nuclear parameters from ENDF/B VIII.0 evaluated nuclear data. Secondly, the activity measurement of foils was obtained by HPGE, then the reaction rate is determined by data processing. It is worth mentioning that the experimental reactivity introduction has a negligible effect on the reactor core by MCNP. Finally, a set of activation experiment scheme that is suitable for CEFR MOX core is formed, which serves as a guide to carry out activation method experiments.

3D simulation of transient thermal-hydraulic characteristics in CEFR during emergency shutdown event

Safe shutdown behavior of China Experimental Fast Reactor (CEFR) in emergency status is one of the key points in its safety assessment. The CEFR design incorporates a pool-type structure. However, the flow and heat transfer characteristics in the pool are complex, which results in obvious three-dimensional (3D) effects. Therefore, obtaining the 3D flow paths and accurate temperature distribution characteristics using a one-dimensional system code is difficult. In the present work, 3D computational fluid dynamics simulations were performed on the pool-type sodium-cooled fast reactor CEFR in full-scale. The “decoupling modeling and coupling calculation” method was used for performing numerical simulation to comprehensively investigate the 3D thermal and flow characteristics under emergency shutdown conditions. The convection and the corresponding heat transfer effects were evaluated by investigating both inter-flow and intra-flow paths in the reactor core. FLUENT 18.0 results were compared and verified with system code SELTAC simulation results. The results revealed that the residual heat was mainly dissipated by the intra-subassembly flow under reactor emergency shutdown conditions. The maximum core outlet temperature did not exceed 543.4 °C under emergency shutdown event, which is lower than the safety limit. This phenomenon indicates that if the heat-removal capability is maintained in the main heat exchange system, CEFR shutdown protection is effective. The three-dimensional results of the present work provide important numerical references for CEFR safety analysis.

Research on the method of predicting CEFR core thermal hydraulic parameters based on adaptive radial basis function neural network

Alterations in thermal hydraulic parameters directly affect the safety of reactors. Accurately predicting the trends of key thermal hydraulic parameters under various working conditions can greatly improve reactor safety, thereby effectively preventing the occurrence of nuclear power plant accidents. The thermal hydraulic characteristic parameters in the reactor are affected by many factors, in order to preliminarily study whose forecasting methods and determine the feasibility of neural network forecasting, the China Experimental Fast Reactor (CEFR) is selected as the research target in this study, and the maximum surface temperature of fuel rod sheath and mass flow rate are used as predictive variables. After data samples are generated through the reactor sub channel analysis code (named SUBCHANFLOW), two widely used adaptive neural networks are used to perform the thermal hydraulic parameter forecast analysis of CEFR fuel assembly under steady-state conditions. The 1/2 core model of CEFR is used to perform a single-step predictive analysis of thermal hydraulic parameters under transient conditions. The results show that the adaptive radial basis function (RBF) neural network exhibits a better fitting ability and higher forecasting accuracy than that of the adaptive back propagation neural network, and the maximum error under steady-state conditions is 0.5%. Under transient conditions, poor forecasting accuracy is observed for some local points; however, the adaptive RBF neural network is generally excellent at predicting temperature and mass flow. The mean relative error of temperature does not exceed 1%, and the mean relative error of flow does not exceed 6.5%. The proposed RBF neural network model can provide real-time forecasting in a short time under unstable flow conditions, and its forecasting results have a certain reference value.

Research on decay heat removal effectiveness in pool-type fast reactor CEFR based on system code SAC-DRACS coupled with CFD software

In station blackout(SBO) accident, the establishment and maintenance of natural circulation in the China Experimental Fast Reactor(CEFR) mainly relies on the decay heat removal system(DHRS). However, due to the complex structure of the reactor, it’s very difficult to model the full-scale CEFR. In the present work, DHRS and the primary circuit system of the CEFR are modeled by the System Analysis Code for Direct Reactor Auxiliary Cooling System (SAC-DRACS) code and three-dimensional CFD software FLUENT, respectively. Then the long-term transient simulation of 4000 s is calculated by coupling SAC-DRACS and CFD codes. The overall temperature stratification and local core temperature variations in the pool caused by DHRS are analyzed. Meanwhile, the residual heat removal capacity of DHRS is evaluated, the average core outlet temperature does not exceed 500.3 °C under the action of DHRS. This work provides an extension of the research method for the pool-type fast reactor safety analysis.

Verification of OpenMC for fast reactor physics analysis with China experimental fast reactor start-up tests

High-fidelity nuclear data libraries and neutronics simulation tools are essential for the development of fast reactors. The IAEA coordinated research project on “Neutronics Benchmark of CEFR Start-Up Tests” offers valuable data for the qualification of nuclear data libraries and neutronics codes. This paper focuses on the verification and validation of the CEFR start-up modelling using OpenMC Monte-Carlo code against the experimental measurements. The OpenMC simulation results agree well with the measurements in criticality, control rod worth, sodium void reactivity, temperature reactivity, subassembly swap reactivity, and reaction distribution. In feedback coefficient evaluations, an additional state method shows high consistency with lower uncertainty. Among 122 relative errors in the benchmark of the distribution of nuclear reaction, 104 errors are less than 10% and 84 errors are less than 5%. The results demonstrate the high reliability of OpenMC for its application in fast reactor simulations. In the companion paper, the influence of cross-section libraries is investigated using neutronics modelling in this paper.

Influence of nuclear data library on neutronics benchmark of China experimental fast reactor start-up tests

Nuclear data is the basis of reactor physics analysis. This paper aim at studying the influence of major evaluated nuclear data libraries, CENDL-3.2, ENDF/B-VIII.0, JEFF-3.3, and JENDL-4.0u, on the neutronics modelling of CEFR start-up tests. Results show these four libraries have a good performance and consistency in the modelling CEFR start-up tests. The JEFF-3.3 results exhibit only an 8 pcm keff difference with the measurement. The difference in criticality is decomposed by nuclide, which shows the large overestimation of CENDL-3.2 is mainly from the cross-section of 52Cr. Except for few cases, the calculation results are within 1σ of measurement uncertainty in control rod worth, sodium void reactivity, temperature reactivity, and subassembly swap reactivity. In the evaluation of axial and radial reaction distribution, there are about 65% of relative errors that are less than 5% and 82% of relative errors that are less than 10%.