ALBERT

Description: A new radioactive liquid waste cementation facility was under commissioning recently in the Institute of Nuclear and New Energy Technology of Tsinghua University, which is designed to simultaneously process multiple intermediate-level radioactive waste drums. Therefore, the multiple volume sources and the scattering effect becomes a key issue in its radiation protection. For this purpose, the Monte Carlo program FLUKA code and experimental measurement were both adopted. In the FLUKA simulation, five different scenarios were considered, i.e., one drum, two drums, four drums, six drums, and eight drums. For the multiple volume sources, the source subroutine code of FLUKA was rewritten to realize the sampling. The complex shielding also leads to a deep penetration problem; hence, the optimization algorithm and variance reduction techniques were adopted. During the measurement, two scenarios, outdoor and indoor, were carried out separately representing the dose field when only one drum is considered and when the scattering effect is considered. A comparison between the experiments and calculations shows very good agreement. From both of the Monte Carlo simulation and the experimental measurement, it can be drawn that, in the horizontal direction, with the increase of the drum number, the dose rate increases very little, while in the vertical direction, the increase of the dose rate is very obvious with the increase of the drum number. The complicated source term sampling methods, the optimization algorithm and variance reduction techniques, and the experimental verification can provide valuable references for the similar scattering problem in radiation protection and shielding design.

Description: Uncertainty analyses of fission product yields are indispensable in evaluating reactor burnup and decay heat calculation credibility. Compared with neutron cross section, there are fewer uncertainty analyses conducted and it has been a controversial topic by lack of properly estimated covariance matrix as well as adequate sampling method. Specifically, the conventional normal-based sampling method in sampling large uncertainty independent fission yields could inevitably generate nonphysical negative samples. Cutting off these samples would introduce bias into uncertainty results. Here, we evaluate thermal neutron-induced U-235 independent fission yields covariance matrix by the Bayesian updating method, and then we use lognormal-based sampling method to overcome the negative fission yields samples issue. Fission yields uncertainty contribution to effective multiplication factor and several fission products’ atomic densities at equilibrium core of pebble-bed HTGR are quantified and investigated. The results show that the lognormal-based sampling method could prevent generating negative yields samples and maintain the skewness of fission yields distribution. Compared with the zero cut-off normal-based sampling method, the lognormal-based sampling method evaluates the uncertainty of effective multiplication factor and atomic densities are larger. This implies that zero cut-off effect in the normal-based sampling method would underestimate the fission yields uncertainty contribution. Therefore, adopting the lognormal-based sampling method is crucial for providing reliable uncertainty analysis results in fission product yields uncertainty analysis.

Description: The experiments are carried out in a three-dimensional channel with a screw conveyor, which plays the role of granular drives for the granular flow system and determines the injection of granular in the test target section. The jam-to-dense transition of granular flow is studied with the different inclination angle. The results show that, with a fixed diameter of hopper orifice and initial filling position, there is a change from jam to dense when the inclination angle larger than 22°. Variation of the flow rate with elevated frequency of the screw conveyor is further studied. The flow pattern is changed from dilute to dense with increasing rotation frequency of the screw rod. When the rotation frequency is larger than 5 Hz, the flow is dense. The dynamic balance of the interface between dilute to dense granular is observed in the main target section. We further research the dynamic interface by measuring the highest and lowest location with time and also simulate the gravity flow rate and screw conveyor flow rate with EDEM. From the results, we find that the interface between dilute flow and dense flow is influenced by the combined action of crew conveyor flow and dense gravity flow.

Description: CAP1400 nuclear island structure is an advanced and novel nuclear power plant structure. In order to explore the seismic response characteristics of CAP1400 nuclear island structure on soft rock sites, a three-dimensional refined nonlinear seismic response analysis model was established for a soft rock foundation-nuclear island structure system using ABAQUS software. The influences of the input ground motion intensity and the frequency spectrum characteristics on the acceleration, relative displacement, and floor response spectrum, as well as the critical shear wave velocity of nonbedrock sites for CAP1400 nuclear island structure, were proposed. The results suggested that the increasing amplitude of the peak acceleration and relative displacement of nuclear island structure decreased as the soft rock site entered a nonlinear state, and the high-frequency components of the input ground motion became more abundant. Specifically, the earthquake response was the largest at the cooling water tank on the top of the shield building, which was the focus of the seismic research on nuclear island structure. Due to the influence of the ground motion frequency spectrum characteristics and the nonbedrock site effect, the peak acceleration, peak relative displacement, and acceleration response spectrum of the nuclear island structure showed different changing trends for the near-field and far-field ground motions. Based on the influence of the site shear wave velocity on the seismic response of nuclear island structure, it was recommended that the critical shear wave velocity of nonbedrock sites for CAP1400 nuclear island structure should be 1250 m/s, and the nuclear island structure-foundation dynamic interaction could be ignored at this time. The research conclusions could provide some technical support and theoretical basis for the construction and seismic performance research of CAP1400 and other nuclear power plants.

Description: This paper proposes to design security measures based on the radioactive material package as the basic unit. The principle of four-layer defense in depth is put forward. Based on the concept of self-security intelligence, combined with out-of-vehicle monitoring, in-vehicle monitoring, and Beidou positioning technology, a security system for transport of radioactive materials was designed. It realized the perception, early warning, delay, and alarm functions and greatly improved the security.

Description: With the wide application of sea-based reactors, the impact of ocean conditions on the safety performance of reactors has gradually attracted attention. In this paper, by establishing the thermal hydraulic transient analysis model and the critical heat flux (CHF) model of natural circulation system, the CHF characteristics in the rectangular channel of natural self-feedback conditions under ocean conditions are studied. The results show that the additional acceleration field generated by ocean conditions will affect the thermal hydraulic parameters of the natural circulation system, that is, the external macroscopic thermal hydraulic field. On the other hand, the boiling crisis mechanism will be affected, that is, the force on the bubble and the thickness of the liquid film. Within the parameters of the study, ocean conditions have a great impact on CHF of natural circulation, and the maximum degradation of CHF is about 45%. The obtained analysis results are significant to the improvement of design and safety operation of the reactor system.

Description: The luminescence of Kr-Xe, Ar-Kr, and Ar-Xe mixtures was studied in the spectral range 300–970 nm when excited by 6Li (n, α)3 H nuclear reaction products in the core of a nuclear reactor. Lithium was deposited on walls of experimental cell in the form of a capillary-porous structure, which made it possible to measure up to a temperature of 730 K. The temperature dependence of the radiation intensity of noble gas atoms, alkali metals, and heteronuclear ionic noble gas molecules was studied. Also, as in the case of single-component gases, the appearance of lithium lines and impurities of sodium and potassium is associated with vaporization during the release of nuclear reaction products from the lithium layer. The excitation of lithium atoms occurs mainly as a result of the Penning process of lithium atoms on noble gas atoms in the 1s states and subsequent ion-molecular reactions. Simultaneous radiation at transitions of atoms of noble gases and lithium, heteronuclear ion molecules of noble gases allows us to increase the efficiency of direct conversion of nuclear energy into light.

Description: In order to enhance the ability of severe accident mitigation for Pressurised Water Reactor (PWR), different kinds of severe accident mitigation strategies have been proposed. In-Vessel Retention (IVR) is one of the important severe accident management means by External Reactor Vessel Cooling. Reactor cavity would be submerged to cool the molten corium when a severe accident happens. The success criterion of IVR strategy is that the heat flux which transfers from the corium pool must be lower than the local critical heat flux (CHF) of the reactor pressure vessel (RPV) outside wall and the residual thickness of the RPV wall can maintain the integrity. The residual thickness of RPV is determined by the heat flux transfer from the corium pool and the cooling capability of outer wall of the RPV. There are various factors which would influence the CHF and the cooling capability of outer wall of the RPV. In order to verify the optimized design which is beneficial to the heat transfer and the natural circulation outside the actual reactor vessel, a large-scale Reactor Vessel External Cooling Test (REVECT) facility has been built. A large number of sensitivity tests were carried out, to study how these sensitivity factors affect CHF value and natural circulation. Based on the test results, the structure of the test section flow channel has an obvious effect on the CHF distribution. The flow channel optimized can effectively enhance the CHF value, especially to enhance the CHF value near the “heat focus” region of the molten pool. The water level in the reactor pit has also a great impact on the natural circulation flow. Although natural circulation can be maintained with a low water level, it will lead to a decrease of the cooling capacity. Meanwhile, some noteworthy test phenomena have been found, which are also essential for the design of the reactor pit flooding system.

Description: Although many types of simulated radionuclides have been widely used as a substitute for actual nuclear waste in the studies of nuclear waste solidification, the understanding of the applicability and validity of simulated radionuclides is still insufficient. In particular, the selection and use of simulated radionuclides, which can play a decisive role in the accuracy of the experimental results, still lack unified or integrated references. This paper provides a critical review on the selection, experimental methods, and applicability of the most commonly studied simulated radionuclides, followed by a careful discussion and recommendation of simulated radionuclides suitable for different solidified bodies. The main factors (e.g., temperature, pH, and atmosphere) affecting the choice of simulated radionuclides were analyzed in detail. This work helps to integrate the selection and use of simulated radionuclides, and it will be beneficial for improving the effectiveness of nuclide solidification research.

Description: The thermal hydraulic and neutronics coupling analysis is an important part of the high-fidelity simulation for nuclear reactor core. In this paper, a thermal hydraulic and neutronics coupling method was proposed for the plate type fuel reactor core based on the Fluent and Monte Carlo code. The coupling interface module was developed using the User Defined Function (UDF) in Fluent. The three-dimensional thermal hydraulic model and reactor core physics model were established using Fluent and Monte Carlo code for a typical plate type fuel assembly, respectively. Then, the thermal hydraulic and neutronics coupling analysis was performed using the developed coupling code. The simulation results with coupling and noncoupling analysis methods were compared to demonstrate the feasibility of coupling code, and it shows that the accuracy of the proposed coupling method is higher than that of the traditional method. Finally, the fuel assembly blockage accident was studied based on the coupling code. Under the inlet 30% blocked conditions, the maximum coolant temperature would increase around 20°C, while the maximum fuel temperature rises about 30°C. The developed coupling method provides an effective way for the plate type fuel reactor core high-fidelity analysis.