The AMB System’s EMI Analysis of Backup Helium Circulator for HTR-10GT

With the application in HTR-10GT, the reliability and stability of the AMB system should be studied deeply. Especially EMI analysis on the switch power amplifier is needed to be done, since which one is the main interference source for AMB during the switch turn-on and turn-off. Based on it, a simplified and improved modeling method is listed by dividing the nonlinear transition into several stages, and the models of the voltage source and current source are built in the form of the piece-wise linear way. The conducted emission on the differential mode noise and the common mode is shown by simulation. The result could provide the theoretical basis on the designing the grounding, filter and isolation for the AMB system.

Study on Microwave Curing of Radioactive Waste

Processing of nuclear waste has become an important issue nowadays. A microwave waste disposal method for solid waste mainly containing Ni, Fe element and their oxide produced in the primary loop is investigated in this article. The powder waste, which is treated as absorber of microwave and energy source for curing, was doped in the curing agent that using SiO2 as a main component. In the experiment, using a household microwave oven (Microwave-frequency: 2.45GHz, Output power: 700W), thermal effects and non-thermal effects make samples heated to a high temperature quickly and then curing. In order to evaluate the effect parameters including different doping ingredient, heating period, pressure were changed. A pressure is applied to make a better curing effect. After curing, the measurement of the density of the cured body was made. The microstructure of the samples was observed with an optical microscope, and the cured body composition was analyzed by XRD and SEM. The results show that samples are sintered quickly with less than 15 minutes and the cured body has a good performance. It’s therefor that microwave technology can be an effective way to dispose microwave absorbing substances such as Fe, Ni-containing powder waste.

Selective Adsorption Properties and Stable Solidification of Cs by Insoluble Ferrocyanide Loaded Zeolites

In Fukushima NPP-1, large amounts of high-activity-level water (HALW) accumulated in the reactor, turbine building and the trench in the facility is treated by circulating injection cooling system. The development of highly functional adsorbents and stable solidification method contributes to the advancement of the decontamination system and environmental remediation. The present study deals with (1) preparation of insoluble ferrocyanide loaded zeolites, (2) selective uptake of Cs+ in seawater, (3) estimation of Cs immobilization ratio and stable solidification. Various kinds of Cs-selective composites loaded with insoluble ferrocyanides (CoFC, NiFC) into the zeolites (zeolite A (A51, A-51J), zeolite X (LSX), chabazite (modified chabazite) and natural mordenite (SA-5)) matrices have been prepared using successive impregnation/precipitation methods by Tohoku University. As for Cs+ adsorption, these composites had relatively large uptake (%) above 95%, distribution coefficients (Kd) above 103 cm3/g and excellent adsorption kinetics even in seawater. The immobilization ratio (%) of Cs for the CoFC saturated with Cs+ was estimated at different calcination temperatures up to 1,200°C in advance. The immobilization ratio was less than 0.1% above 1,000°C, indicating that the adsorbed Cs+ ions are completely volatilized and insoluble ferrocyanides had no immobilization ability for Cs. In contrast, the insoluble ferrocyanide-loaded zeolites had excellent Cs immobilization ability; in the case of insoluble ferrocyanide-loaded natural zeolites (NiFC-SA-5, CoFC-modified chabazite), the immobilization ratio was above 99% and 96% even after calcination at 1,000°C and 1,100°C, respectively, indicating that nearly all Cs ions are immobilized in the sintered solid form. On the other hand, the immobilization ratio for the insoluble ferrocyanide-loaded A and X zeolites (NiFC-A (A51, A51J), NiFC-X) tended to decrease with calcining temperature; for example, the immobilization ratio for NiFC-X at 1,000°C and 1,100°C was estimated to be 74.9% and 55.4%, respectively, and many spots concentrating Cs were observed on the surface. The difference in immobilization behavior between natural zeolites and synthetic ones is probably due to the phase transformation and surface morphology at higher temperature above 1,000°C. The stable solidification of insoluble ferrocyanides was thus accomplished by using the excellent Cs immobilization abilities of zeolite matrices (Cs trapping and self-sintering abilities).

Study for Application of Safety Soft Control and Information System in Highly-Integrated Control Rooms of CPR1000 NPP

For most of the newly built nuclear power plants, the computerized main control rooms (MCR) are adopted. The soft control, the typical feature of computerized Human-Interface System (HIS) in the computerized main control room and mediated by software rather than by direct physical connections, is comprised of safety and non-safety control interface which provides the operators with manual control for component-level, and allows both continuous control of plant process and discrete control of components in nuclear power plant. The safety soft control and information system (SSCIS) is used to give the safety commands to and check the immediate response of the safety process. This paper describes the application of the system design basis, functionality, communication, operation faceplate and system modes for SSCIS which is firstly introduced in CPR1000 nuclear power plant. The design criteria and basic design features of SSCIS is developed to be as the design basis of the design implementation. The ISG-04 ‘Highly-Integrated Control Rooms-Communications issues (HICRc)’ provides acceptable methods for addressing SSCIS communications in digital I&C system design. The NUREG0700 ‘Human-System Interface Design Review Guidelines’ is applied as reference for human factor engineering requirement in the SSCIS design. And the SSCIS design has also fully considered the possible customer usual practice.

The Failure Analysis and Processing of Digital Reactor Protection System

The Qinshan Nuclear Power Plant phase 1 unit (QNPP-1) has a power rating of 320 MWe generated by a pressurized water reactor that was designed and constructed by China National Nuclear Corporation (CNNC). The TELEPERM XS I&C system (TXS) is to be implemented to transform analog reactor protection system (RPS) in QNPP-1. The paper mainly describes the function, structure and characteristic of RPS in QNPP-1. It focuses on the outstanding features of digital I&C, such as strong online self-test capability, the degradation of the voting logic processing, interface improvements and CPU security. There are some typical failures during the operation of reactor protection system in QNPP-1. The way to analyze and process the failures is different from analog I&C. The paper summarizes typical failures of the digital RPS in the following types: CPU failure, communication failure, power failure, Input and output (IO) failure. It discusses the cause, risk and mainly processing points of typical failure, especially CPU and communication failures of the digital RPS. It is helpful for the maintenance of the system. The paper covers measures to improve the reliability of related components which has been put forward effective in Digital reactor protection system in QNPP-1. It will be valuable in nuclear community to improve the reliability of important components of nuclear power plants.

Updated Design and Development Route for CH HCCB TBM and its Mockup

Chinese helium-cooled ceramics breeder test blanket module (CH HCCB TBM) is determined to be tested in ITER machine to get data for fusion reactor design and development in future. Chinese TBM is designed to occupy half of port C with 484mm in torroidal and 1660mm in poloidal. Radial length is 675mm. TBM is composed of box, 12 submodules and independent backplate. Box formed by first wall, grids and caps have 12 caivities to hold submodules. Box and submodules are supported by backplate by welding. Backplate distribute helium with flow rate 1.36kg/s to cool first wall and then part of it go out of TBM by bypass. The rest 0.77kg/s go on to cool caps and girds first and then cool submodules. Submodules with dimensions 250mm×202mm×318mm have independent cooling and purging systems connected to backplate manifold systems. In a submodule, two U-shaped structures hold breeding material Li4SiO4 pebbles. Out of the structure filled beryllium pebbles. Neutronics results show that tritium production is ∼64mg/FPD. Maximum temperature 538°C of structure material occurs in the front of first wall with surface heat flux 0.5MW/m2. Maximum total stress at first wall is 471MPa at 394°C; that in submodules is 426MPa at 400°C; that in backplate is 526MPa at 410°C, In order to explore development technologies for the TBM, a mockup with dimensions 484mm (torroidal)×592mm (poloidal)×675mm (radial) has been designed. The mockup with similar structure ignores bypass and purge gas systems. In the mockup, there’s only one submodule and the other three are replaced by submodule replacements. By discussions and investigations, development route has been decided and the mockup is being fabricated.

Mechanical Performances of ITER In-Vessel Coil’s Conductor

The design and R&D for ITER In-Vessel Coils (IVCs) is being deployed. The concerned issue of “Edge Localized Modes” (ELMs) and “Vertical Stabilization” (VS) of the ITER plasma can be addressed by the implemented IVCs. The ELM and VS coils will be installed in the vessel just behind the blanket shield modules to reach the requirement of keeping strong coupling with the plasma. The 59mm Stainless Steel Jacketed Mineral Insulated Conductor (SSMIC) using MgO as the insulation is being designed for the IVCs to resist the special challenges, including the nuclear radiation, high temperature, electromagnetic and thermal fatigue. It is necessary to take the mechanical performances of the SSMIC and the feasibility of fabrication techniques into consideration of the R&D program. The mechanical performances of the SSMIC close to the actual work conditions, including the three point bend modulus, three point bend cyclical performance and the cyclical performance with a U-bend sample of the SSMIC prototypes have been investigated and the results are presented in this paper.

Research on Project Management of DCS Development for First AP1000 Unit

Sanmen Nuclear Power Plant is constructing the first AP1000 unit in the world, and DCS system is adopted as the non-safety control system. This paper reviews the experience of Sanmen project, and analyzes the necessity of nuclear power plant Owner’s management for developing the DCS system, as well as how to manage the development effectively.

Research on Feature Recognition of Nuclear Power Equipment Based on the Optimal Wavelet Basis

Touching upon that the crack fault of the rotor may occur after the reactor coolant pump (RCP) has operated a long time, the fault feature can be identified effectively by the method of the wavelet analysis. In this research, based on the simulation signal of crack fault and the method of discrete wavelet transform (DWT), the cross-correlation coefficients between the fault signal and the different wavelet basis which are selected from the wavelet basis library can be computed. After confirming the maximum of the cross-correlation coefficients, the optimal wavelet basis applied to the fault signal of the cracked rotor will be found. And the main frequency component of the fault feature is recognized by use of the wavelet packet transform (WPT) based on the optimal wavelet basis. The results of simulation illustrate that the wavelet basis selected by the maximum cross-correlation coefficients can become the optimal wavelet basis, and the fault feature of the cracked rotor can be recognized effectively.

The Application of AFAL Methodology in Substantiation of Instrument Calibration Intervals Extension in Nuclear Power Plant

This is an Introduction about the AFAL methodology of study the instrument drift characteristics, and the application of AFAL in substantiation of instrument calibration interval extension in nuclear power plant. AFAL methodology main aspects include collecting historical instrument calibration data of nuclear power plant, calculating statistics values such as: sample number, mean, median, standard deviation, then work out the drift tolerance interval of the instrument. Based on analysis of these statistical calculated values, will understand the instrument drift performance. This article also discusses technical issues associated with the application of AFAL and how to solve them, such as: grouping instrument, sample sizes, outliers detecting and processing, high-confidence deduction etc. Through the study of the instrument drift characteristics, evaluate the performance of instrument, determine the calibration interval can prolong properly. The application practices of AFAL methodology show, extending instrument calibration interval can support nuclear power plant to achieve the goal of prolonging the fuel cycle, under the nuclear safety precondition. The nuclear power plant can improve the capacity factor of the unit and economic performance.