Neutron kinetics plays an important role in reactor safety and analysis. The backward Euler method is the most widely used time integration method in the calculation of space-dependent nuclear reactor kinetics. Diagonally Implicit Runge-Kutta (DIRK) method owns high accuracy and excellent stability and it could be applied to the neutron kinetics for hexagonal-z geometry application. As solving the neutron kinetics equations is very time-consuming and the number of available cores continues to increase with parallel architectures evolving, parallel algorithms need to be designed to utilize the available resources effectively. However, it is difficult to parallel in time axis since the later moment is strongly dependent on the previous moment. In this paper, the Parareal method which is a time parallel method and implemented by MPI in the processor level is studied in the hexagonal-z geometry with the help of DIRK method. In order to make good use of the parallelism, a parallel strategy in the space direction is also used. In the coarse nodal method, many same operations are finished in the nodes and these operations could be parallel by OpenMP in the thread level since they are independent. Several transient cases are used to validate this method. The results show that the Parareal method gets a fast-convergent speed such as only 2∼3 iterations are needed to convergent. This space-time parallel method could reduce the cost time compared to the sequential method.

Flow elastic stability and vortex shedding were two important mechanisms of the flow induced vibration analysis. Due to the influence of manufacturing process, transportation and irradiation, the clamping action of grid on fuel rods may be invalid. Taking one fuel assemblies as an example, the effects of clamping failure on the natural frequencies, mode shapes, flow elastic stability and vortex shedding were studied. The results show that the effect of the rigid convex support failure on the natural frequency was directly related to the mode shape. The effect of the grid rigid convex failure near the nodes with larger amplitude on the natural frequency was obvious. The velocity of flow at the top and bottom of the fuel rods were larger and the size was comparable, this induced that the rigid convex failure of the top and bottom grids had a significant effect on the ratio of flow elastic stability and the vortex shedding.

Monolithic fuel is a fuel form that is considered for the conversion of high performance research reactors. In order to qualify this new fuel system, the fuel plates should meet the safety standards and perform well in reactor. The fuel system must maintain its mechanical integrity, sustain a geometric stability and should have stable and predictable irradiation behavior. The requirement to maintain mechanical integrity under normal operating conditions is primarily demonstrated by a successful testing of fuel plates up to the limiting conditions defined by the fuel performance envelope, including an adequate margin. Although large number of plates have been tested with satisfactory thermo-mechanical performance, post-irradiation examination of plates from previous RERTR-12 experiments have revealed that pillowing occurred in several plates, rendering performance of these plates unacceptable. To address such failures, efforts are underway to define the mechanisms responsible for the in-reactor pillowing, and suggest improvements to the fuel plate design and operational envelope. For this purpose, selected plates from previous experiments were simulated to understand the thermo-mechanical response of the plates to the fission density and thermally induced stresses. Simulation results were then comparatively evaluated with post-irradiation examinations of selected plates. The simulation results and experimental observations established a possible correlation between failure by plate pillowing, high porosity and a presence of tensile stress state. The study has implied that porosity leading to degradation of material properties, accompanied by a sufficiently large tensile stress state can lead to a pillowing-type failure at reactor shutdown. This paper presents these findings, discusses such failure modes, and the influence of fuel burn-up and power on the magnitude of the shutdown-induced tensile stresses.

Natural circulation with two-phase flow is a nonlinear dynamical systems, which can show a very complex and strange behavior under specific conditions. The application of stability analysis requires a large computational effort and is cumbersome in case of prediction the dynamical behavior by system codes alone. Therefore, model order reduction techniques are used to compensate this disadvantage by coupling with a bifurcation code such as MatCont. A reduced order model is derived by employing the POD-method to analyze the stability landscape of a low pressure natural circulation system representing passive safety systems such as the containment cooling condenser. The required full order model contains a classical natural circulation loop with a heated section and a riser. The two-phase region is modeled by a drift-flux mixture model. The reliability of the FOM is investigated by comparison with a reference model by the validated system code ATHLET.

One dimensional (1D) numerical simulations have been carried out of heated parallel channel experiments using the linear frequency domain approach to investigate the effect of wall heat on the system’s stability for supercritical water in vertical heated channels. Oscillatory instability boundaries of the experimental cases with two different wall thicknesses and uniform power distribution are produced, assessed and discussed. These new numerical results with wall heat are compared with experimental data and previous numerical results without wall heat. There is improved agreement on the stability boundary predictions when the wall heat storage effect is included. However, the effect of wall heat storage was small for these experiments.

Korea Atomic Energy Research Institute (KAERI) has designed and constructed a test facility for reactor coolant pumps (RCPs). In RCP test facility (RCPTF), it is possible to perform a type test for a newly-developed RCP as well as a production test for the Advanced Power Reactor 1400 (APR1400) RCP before its installation in nuclear power plants. The thermal hydraulic and electric capability of the RCPTF covers up to 18.5 MPa, 343 °C, 11.7 m 3 /s, and 14.0 MW for the design pressure, temperature, flow rate, and maximum electric power, respectively. In 2013, a commissioning test was performed to verify its designed capability, followed by several modifications in the RCPTF including signal processing and control logic to enhance the verification and evaluation capability of the RCP performance. During the commission test period, the technical activity for RCP performance verification test and optimization of flow stability have also been performed. The developed techniques can be divided into four subjects; 1) Development of flow stabilization technique for high flow condition test facility, 2) Improvement on the accuracy of performance verification factor and development of evaluation method, 3) Evaluation of test facility vibration characteristics and optimization of measurement system, and 4) RCP coast-down data production. KAERI has completed a full set of technique developments, a prerequisite for the RCP performance test.

The numerical calculations with and without rolling motions were conducted to investigate the effects of ocean environment on flow and heat transfer of supercritical CO 2 . The AKN k -epsilon model was selected to model the turbulent flow and heat transfer of supercritical fluid. It concludes that the effect of rolling motion on the supercritical CO 2 in LPV region is much greater than that on conventional single-phase fluid. The rolling motion can cause the periodic oscillation of the local heat transfer and suppress the nonuniform heat transfer. The secondary flow is induced by both the rolling motion and buoyancy force. The heat transfer is enhanced gradually as the rolling period decreasing or rolling amplitude increasing, but the corresponding pressure drop varies more violently to affect the stability and controllability of the heat exchanger.

The vibration of pipeline system induced by pump is one of the most important factors affecting the safety and stability of nuclear power plant (NPP) operation, and is also the most prominent vibration and noise transmission channel in the plant area. Analysis results show that various support structures such as pipe clamps, hangers and brackets are the main transmission routes of pipeline vibration. However, it is usually difficult to obtain a satisfactory absorption effect by using traditional passive vibration isolation device, especially for a variety of low-frequency vibration. Besides, the effective isolation frequency is fixed which cannot adapt to the variation of the operation condition excitation source. Therefore, it is of great practical significance to propose a method and scheme which can effectively restrain the low-frequency vibration of pipeline system of NPP. In order to meet the above requirements, an electromagnetic semi-active vibration absorber is designed in this paper. Its electromagnetic stiffness can be adjusted by changing the excitation current, so as to achieve the purpose of changing the natural frequency online and improving the vibration attenuation effect. In order to verify the rationality of its electromagnetic structure, the magnetic field distribution of vibration absorber is simulated and analyzed. Results show that it is close to magnetic saturation inside without leakage. On this basis, a platform of simulated condensate water pipeline system for vibration experiment is set up, on which the frequency response together with time & frequency-domain vibration characteristics of typical support points are tested, and the stiffness, damping and other parameters of the vibration absorber are optimized according to the test results. Finally, the vibration of pipeline system is simulated and compared. The results show that under the condition of the optimal combination of parameters, the vibration spectrum of the pipeline is obviously reduced after the vibration absorber is used, which indicates a significant vibration control effect.

The radon from uranium tailings spreads fast and has a wide range of pollution, which poses a potential radiation hazard to the environment and the public in downwind region. In this paper, the open and naked uranium tailings are selected as research object. By setting up multiple Gaussian plume models with single point source, the diffusion of radon in the uranium tailings is simulated with different atmospheric stability, average wind speed, height and downwind distance. The results show that the maximum radon concentration increases while the related downwind distance decreases as the atmospheric becoming stable. The higher wind speed does not affect the downwind distance where the maximum radon concentration occurs, but it decreases the maximum radon concentration. The concentration of radon in residential area decreases but the decreasing rate speeds up with height going up. The distribution of radon in vertical and horizontal direction tends to be homogeneous while the near-surface area concentration decreases rapidly as farther downwind distance.

Tritium is indispensable to the fusion reactor engineering, and it must be seriously defended because of its radioactivity and permeability. The method of catalytic oxidation and absorption is the most widely used process for tritium cleanup so far, in which detritiation catalyst is of great importance. The poor stability caused by the agglomeration of noble metal limits the life of detritiation catalysts. Here, Anti–Ostwald Ripening is used to prepare single-atom detritiation catalysts S-Pt/Ce 0.7 Zr 0.3 O 2 for tritium (HT, DT and T 2 ) oxidation. Single-atom dispersed Pt ensures the catalytic activity and decreased the economic cost. The strong metal-support interaction (SMSI) keeps Pt from aggregating, thus increases the working life of catalyst. And Pd based catalyst supported by a cation ordered κ-Ce 2 Zr 2 O 8 is prepared for tritiated methane (CH 4-x T x ) oxidation. Tritiated methane is mostly oxidized by Pd/κ-Ce 2 Zr 2 O 8 at about 450 °C, which is at least 50 °C lower than normal catalysts (such as Pd/Al 2 O 3 ).

Segmented Gamma Scanner (SGS) is a commonly used nondestructive testing (Non-Destructive Assay NDA) method. SGS uses radial rotation, axial segmentation, segmented scanning of the non-uniform sample of the uniform treatment, making it possible to accurately measure the radioactivity on each segment and is currently widely used in the field of nuclear material management. This paper introduces a self-designed SGS measuring device and uses this device to perform a large number of scanning experiments on different measurement objects in the laboratory. It also studied the detection performance, stability and SGS method of different objects the accuracy of the measurement results.

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.

Steam generator (SG) is one of the key equipment of nuclear power units. Because of the large range of its loads changing, the water level control of SG effectively is an essential secure guarantee of nuclear power plants. SG is a complex system, besides imbalance and non-minimum phase characteristic, it also has the properties of nonlinearity, time-varying and with small stability margin. There are many difficulties in water level control of SG. Of which false water level and varying parameters are the most severe problems. In this paper, first the water level features and the water level control principle of U-tube steam generator (UTSG) are introduced. Then mathematical model mechanism and both the static and dynamic characteristic of it water level are discussed. Finally various control methods are used for comparing the control effect. Intelligent control is a type of control strategy which imitates human intelligence behavior. It is mainly aimed at the controlled plant with complicate model parameters, or which model structure hard to describe accurately by mathematical method. Cloud Model theory is proposed by Academician Li Deyi based on the idea of artificial intelligence with uncertainty. This theory focus on analyzing the uncertainty of control plant, realizes the uncertain conversion between qualitative concept and quantitative numerical by combining ambiguity and randomness. In the field of control technology, ambiguity and randomness make it difficult to establishing precise mathematical model of control plant, and become a bottleneck during the research of improving stability, accuracy and quickness of control system. In this context, Cloud Model can be a good conversion between qualitative concept and quantitative numerical due to its ability of showing the uncertainty of qualitative concept which described by natural language. Under the action of external input, system control can be realized by inferencing according to the qualitative concept and uncertainty rules of Cloud Model. In this paper, the researched Cloud Model control system is based on normal distribution, because a large number of random events in nature and society obey or approximately obey normal distribution. The rate of convergence of Cloud Model control is evidently faster than PID. Moreover, the capability of Cloud Model control in tracking, adapting, anti-interference and overcoming large time lag are apparently superior when comparing with the control effect of PID.

In nuclear power plant system, pump is the key equipment to maintain the flow of the primary loop coolant and the secondary loop heat transfer fluid. The main coolant pump and the feed water pump are in long-term operation status. Bearings are the key components to ensure stable operation of the pump, and which could be damaged in abnormal conditions. Once the failure occurred in the bearings, pumps would exhibit periodic vibration, which might cause the flow pulsations of coolant and heat transfer fluid; gradually, these situations could reduce the control accuracy and the stability of pump. Therefore, the detection and diagnosis of pump bearings are significant to improve the safety and stability of reactor system. We proposed an approach combined with signal processing and machine learning to extract the signal features and recognize the signal samples automatically. The proposed approach consists of three main steps: firstly, empirical mode decomposition (EMD) is applied to decompose the signals into several intrinsic mode functions (IMFs) which are corresponding to the different components of the original signals; secondly, calculating the correlation coefficient between each IMF and the original signal, the correlation coefficient sequence imply the components distribution of the signal which can be applied to recognize the signal samples; finally, extracting a part of correlation coefficient sequences to train the support vector machine (SVM), and then an classifier can be obtained and use to recognize the other signal samples automatically. Experimental results show that this method can effectively detect the pump bearing operating conditions and failures, and can provide a reference for the safe and stable operation of reactor pumps.

The advanced Mechanical Shim (MSHIM) core control strategy employs two separate and independent control rod banks, namely the MSHIM control banks (M-banks) and axial offset (AO) control bank (AO-bank), for automatic reactivity/temperature and axial power distribution control respectively. The M-banks and AO-bank are independently controlled by two closed-loop controllers called the coolant average temperature ( T avg ) controller and AO controller. Since the movement of M-banks and AO-bank can both affect the T avg and AO, the T avg controller is coupled with the AO controller. In order to avoid the interference between the two controllers, the MSHIM control system adopts an interlock design between them to avoid the simultaneous movement of the M-banks and AO-bank and make sure the priority of the M-bank movement. This design can enhance the stability of the MSHIM control system. However, the control performance is degraded at the same time. In the present study, the feedforward compensation decoupling method and multimodel approach are used to eliminate the coupling effect between the two controllers in the MSHIM control system during a wide range of power maneuvers. A multiple feedforward compensation system is designed with integration of feedforward compensators for the T avg and AO controllers at five power levels using the multimodel approach. By implementing it in the MSHIM control system, the interlock between the M-banks and AO-bank can be released to realize the independent and decoupled control between T avg and AO. The effectiveness of the decoupled MSHIM control system is verified by comparing its control performance with that of the original MSHIM control system during typical load change transients of the AP1000 reactor. The obtained results show that superior and decoupled control of T avg and AO can be achieved with the proposed decoupled MSHIM control system.

This paper introduces a method to control the calculation progress of the RELAP5 codes and integrate them with other codes by interacting boundary data (such as general tables, TDVs and TDJs) at each step. This work is basically finished with the support of the RINSIM simulation platform. The paper gives a brief introduction on RINSIM that how it controls the codes progress, sends the codes control commands, shares the values of different codes’ common blocks or modules. However, the work can’t be done by just using the RINSIM, it also needs to modify RELAP5 codes. With the codes’ modification of commands responding, data reading/writing interface, data interacting interface, time step control and so on, we can build interface subroutines to integrate codes onto the RINSIM. At the end, the paper gives out the result of a transient calculation with an advanced PWR model. Compared to some old integration method, the new method has far more strong stability. And the result shows that the integration progress of the code does not obviously affect the calculation accuracy, but definitely extends the application fields because of the multiple functions supplied by the RINSIM.

The nuclear heating reactor (NHR) is a typical integral pressurized water reactor (iPWR) developed by the institute of nuclear and new energy technology (INET) of Tsinghua University, which has the safety advanced features such as the primary circuit integral arrangement, full-range natural circulation, self-pressurization. Power-level control is crucial for the operational stability and efficiency of the NHR, and the dynamic modeling is a basis for control system design and verification. From the conservation laws of mass, energy and momentum, a lumped-parameter dynamical model is proposed for the nuclear steam supply system (NSSS) based on the 200MW th nuclear heating reactor II (NHR200-II). The steady-state model validation is given by the comparing the parameter values of this model and that for plant design. Then, both the open-loop responses under the disturbances of reactivity and coolant flowrates as well as the closed-loop responses under the case of power ramp are given, where the rationality of the responses are analyzed from the viewpoint of plant physics and thermal-hydraulics. This model can be utilized for not only the control system design but also the development of a real-time simulator for the hardware-in-loop control system verification.

In order to study the effect of rolling motion on flow instability of parallel rectangular channels of natural circulation, the natural circulation reactor simulation system is used for physical prototype. And theory analysis model of parallel rectangular channels of natural circulation system under rolling motion is established and coded by Fortran. The results of the program are verified to the experiments, and the results are in good agreement. The flow instability boundaries of different pressure under static and rolling motion are calculated respectively. The results show that: 1) under static condition, with the increase of the pressure, the instability boundary line changes, and the system becomes more stable; 2) under rolling conditions, the heating power of instability boundary decreases comparing to the stable conditions. The instability occurs earlier; 3) the stability of the system decreases with the increasing of rolling amplitude and frequency.

In order to maintain the reactor core integrity in case of severe accidents, multi-layered installation of backup safety systems with the passive feature is crucial, as was evident from the Fukushima Daiichi accident. The supersonic steam injector (SI), also known as a passive jet pump driven by high-pressure steam and water, is capable of operating without external power supplies nor rotating machinery. Along with the Reactor Core Isolation Cooling (RCIC) system, the SI can possibly be considered as one of the candidates as a backup safety system for current and next-generation reactors. Although the fundamental experiments on SI have been carried out for past few decades, its operation characteristics are still not well understood. Especially, its stable operation condition as a passive jet pump hasn’t been investigated thoroughly. Our study shows that the SI’s operating range is highly dependent on inlet water flow rate and steam pressure. In addition, due to the direct contact condensation between steam and water jet within SI unit, the water jet stability may influence SI’s operation. In order to deepen understanding of SI’s operation characteristics, this study aims to investigate on the inlet water temperature effect, which has a direct influence on steam-water condensation as well as the water jet stability within the mixing nozzle section of SI unit.

Based on the external force method, the present study integrates the nonlinear dynamic model developed previously by the authors with vertical seismic accelerations to investigate the seismic-induced effect on the single nuclear-coupled boiling channel natural circulation loop. The natural frequencies of the states in the stable region are widely explored through nonlinear analysis. The results indicate that the natural frequency of initial state tends to increase as the increase in the phase change number (operating power) or as the decrease in the subcooling number (inlet subcooling). As supposing that a real seismic acceleration is directly imposed on the system, some parametric effects on the seismic-induced oscillations are performed in the present natural circulation loop system. The seismic-induced oscillations are found to be consistent with the combined results of the inherent system stability characteristics and the impact of external seismic acceleration. The system with a larger outlet loss coefficient of the heated channel, or a longer heated channel, would destabilize the seismic-induced oscillations, while the inlet loss coefficient of the heated channel has a stable effect on the system. A much stronger resonance oscillation could be induced by the increase in the core inlet subcooling. Notably, the enlargement and contraction of the heated channel diameter would move the natural circulation system closer to type-I and type-II instability regions, respectively. These both generate unstable effects on the seismic-induced oscillations due to the inherent stability characteristics of the initial states.