Volume 86, Issue 1

Recently, the FNPP (Floating Nuclear Power Plant) has got more and more attention and rapid development due to very wide prospect application on remote areas or islands. In general, the IPWR (Integral Pressurized Water Reactor) is adopted to meet the requirements of the limited space, the nuclear safety and the maneuverability in marine. The IPWR could depend on natural circulation operation to remove the residual heat of core under accident or low load operation condition. Because the driving head is low, the natural circulation flow is likely to be influenced by rolling and inclined condition. To clarify the natural circulation flow characteristics of the core in FNPP rolling motion and inclined condition, based on the modified THEATRe code by adding the ocean motion module and spatial coordinate convert module, the main thermal-hydraulic parameters variation in rolling and inclined condition were obtained. The effect of inclined angle, rolling amplitude and period on the natural circulation flow were discussed. The natural circulation flow in the core fluctuates periodically with rolling motion. And the inclination and rolling will also cause the degree of steam superheat of OTSG secondary side fluctuate, which could impact on the stable operation of secondary side system.

In this research, the analysis methodology for Maanshan PWR nuclear power plant (NPP) ATWS (Anticipated Transient Without Scram) is established by using RADTRAD code. This methodology can evaluate the dose at Exclusion Area Boundary (EAB), Low Population Zone (LPZ), and control room for the Maanshan ATWS. In addition, there are four leakage pathways in the Maanshan ATWS that are containment leakage, containment low volume purge, steam generator leakage, and ESF leakage. Therefore, for these leakage pathways, the RADTRAD models are established. Finally, according to the analysis results, they are below the criteria of 10 CFR 100.11 for the EAB and LPZ and Standard Review Plan (SRP) for the control room. Additionally, the steam generator leakage gives the maximum dose in these leakage pathways.

This paper describes the computational fluid dynamics (CFD) methodology to simulate the boiling flow in a typical Pressurized Water Reactor (PWR) 5 ⨯ 5 rod bundle. The method includes the Eulerian-Eulerian two-fluid model coupled with the improved wall heat partitioning model. The NUPEC PWR Subchannel and Bundle Test (PSBT) International Benchmark are used for validation. The simulated surface averaged void fraction agree well with the experimental data, which indicate the promising application of the present method for modeling the boiling flow in the fuel rod bundle. The main emphasis of current research has been given to the analysis of the phase distribution around and downstream the spacer grid, the effect of the spacer grid structure, including the mixing vanes, the springs and the dimples on the void fraction distribution is investigated. The findings can contribute to a better understanding of three dimensional flow boiling characteristics and can be used to assist in optimizing the spacer grid.

The optimization measures in the physical protection system monitoring center of a nuclear power plant include the prioritization of alarm signals, optimization of sound and light alarm form, improvement of the layout of video monitor screen, security training, and strengthening of organizational management. Based on the fuzzy analytic hierarchy process, the influence of these factors on the probability of alert assessment and guard’s respond time in the EASI method are quantitatively analyzed. Making full use of the measures for prioritization of alarm signals can effectively promote the improvement of human-computer interaction efficiency. The degree of influence of the four factors (guarder’s status, decision strategy, guarder’s training and organization management) on guard’s decision-making is roughly the same.

To evaluate the valves used in the nuclear power plants are working properly under the required conditions, the performance and capacity test should be performed. In the test system, the accumulator was employed to control the large amount of high pressure and high temperature steam generated in the boiler precisely. In the accumulating process, the steam is often condensed. In order to prevent condensation, it is needed to install heaters and preheat the accumulator. However, if the size of the accumulator becomes large, the installation of the heater may not be easy. Therefore, when the test is conducted, the system was preheated by the latent heat generated from the phase change. Insufficient thermal insulation may cause temperature differences and it can cause mechanical problems in the accumulator structure. If insulation is sufficient, the temperature difference is indicated by the height. As the cooled condensate moves downwards, the condensate is discharged by the drain valve control and the temperature difference of the structure can be disappeared. The results of this paper can be applied to the conceptualization of equipment that uses latent heat and for the design of high-precision steam experimental devices or the design of high-capacity steam utilization systems.

This paper presents results of calculation studies of the viability of coated particles in the conditions of the reactor core on fast neutrons with sodium cooling, justifying the development of the concept of the reactor BN with microspherical fuel. Traditional rod fuel assemblies with pellet MOX fuel in the core of a fast sodium reactor are directly replaced by fuel assemblies with micro-spherical mixed (U,Pu)C-fuel. Due to the fact that the micro-spherical (U, Pu)C fuel has a developed heat removal surface and that the design solution for the fuel assembly with coated particles is horizontal cooling of the microspherical fuel, the core has additional possibilities of increasing inherent (passive) safety and improve the competitiveness of BN type of reactors. It is obvious from obtained results that the microspherical (U, Pu)C fuel is limited with the maximal burn-up depth of ∼11% of heavy atoms in conditions of the sodium-cooled fast reactor core at the conservative approach; it gives the possibility of reaching stated thermal-hydraulic and neutron-physical characteristics. Such a tolerant fuel makes it less likely that fission products will enter the primary circuit in case of accidents with loss of coolant and the introduction of positive reactivity, since the coating of microspherical fuel withstands higher temperatures than the steel shell of traditional rod-type fuel elements.

In this research, the effect of nanofluids including graphite and water on the heat transfer enhancement in heat Exchangers with both straight and rotational inner tube were examined. Nano-fluids including graphite and water were obtained at 0.01, 0.02 and 0.03 %volume concentrations by means of pure water as a base fluid. The aim of the study is to experimentally investigate and compare the influence of straight inner tube on the performance of a double tube heat exchanger at different flow rates, different capacity ratios (0.25, 0.5, 0.75, 1), different rotation rates (0 RPM, 100 RPM, 150 RPM, 200 RPM, 300 RPM, 500 RPM and 750 RPM) and by using nanofluid including graphite and water volume at values of 0.01, 0.02, 0.03%. In the experiments the flow is turbulent. In order to determine the heat transfer enhancement, the experimental datas were compared for pure water and nano-fluids. According to the results, Nusselt number, pressure loss, efficiency of heat exchanger were gauge. By the results achieved from experiments, correlations for Nusselt number have been reproduce and experimental data have been compared with ones in the literature by drawing graphics. The experiment that provides the best increase in heat transfer is the nanofluid including 0.02% volume concentrations of graphene/water at 500RPM speed.

In this study the radiation shielding properties of mortar samples were investigated. The samples were created by replacing heavyweight particles of iron, steel and chromium waste by calcareous sand in different volume percentages. Additionally, the effects of the physical properties of particles and samples on shielding properties is also discussed. In the scope of this work, the radiation shielding properties of mortar samples were measured by determining the values of mass attenuation coefficients by means of an experimental setup consisting of 133 Ba, 137 Cs and 60 Co standard point sources and high purity germanium (HPGe) detector. It’s concluded that the highest mass attenuation coefficient values among the heavy particles were achieved by chromium particles. However, all the high-density particles used in the study improved the radiation shielding properties of the mortars considerably compared to ordinary mortars.

The Segmented Gamma Scanner(SGS) technology is specially used for type identification and activity quantitative analysis of radioactive material in sealed containers, which mainly divided into the transmission measurement and the emission measurement. In actual measurement process, it cannot meet the need for rapid analysis because the efficiency calibration time takes up more than 60% of the whole detect time in the data analysis. Whereas most previous research have focused on theory or specific applications, this research groups aim to different aspects of the SGS technique and direct it at an audience with interests in the need for rapid analysis. The Monte Carlo simulation calculation models were established by the passive efficiency calibration method, and then the detection efficiency database was carried out based on the experimental verification. Lots of work was used to analysis the influence of crosstalk between layers and interpolation step. Furthermore, the database was implanted into the control and analysis system to complete the measurement experiments. The results show the measuring time of single waste drums is about 30 min, and the average relative deviation is less than 10%, so the problem that the time taken to calibrate the detection efficiency is too long has been solved effectively. The efficiency in the use of SGS technology has been increased.

Level density models have increasing importance to gain more in-depth into the nature of nuclear reactions. Many novel and advanced medical application use radioisotopes, which are produced with nuclear reactions. In this study, the effect of the level density parameters of the nucleus on the cross sections of neutron-fission reactions for 233 U, 235 U, 237 U, 239 U, 249 Cf, 251 Cf, 237 Pu and 247 Cm nuclei were investigated for up to 20 MeV neutrons. TALYS 1.8 software was used to calculate the cross-sections of neutron-fission reactions for different level density parameters. The calculations were compared with the EXFOR nuclear data library and the level density parameters, and the closest fit were searched. As outputs of the study, the effect of selection of level density parameter on cross section calculations was observed. The theoretically obtained data were compared with the experimental data taken from the literature. The results are presented graphically for better interpretation.

The methodology of analysing ionizing mattresses using the RESRAD-BUILD code is described. RESRAD-BUILD is a free computer program developed by Argonne US National laboratory. The RESRAD-BUILD code can analyze radiation exposures resulting from occupying a building contaminated with radioactive materials or housing contaminated furniture or equipment. In May 2018, the South Korean government pulled the mattresses of manufacturer Daijin Bed Co from its markets because these mattresses had unsafe radiation levels. After that, the AEC (Atomic Energy Council, Taiwan) had immediately conducted an inspection of ionizing mattresses of Taiwan and found that the radiation level of mattresses of two domestic brands exceed the legal standards. For assessing the dose of the ionizing mattress, it was necessary to establish a methodology for analyzing these in case of nuclear industry unusual products. This was done by using the RESRAD-BUILD code and the data are presented in this manuscript. The analysis results of RESRAD-BUILD were compared with the assessment result to the dose announced by the AEC. The RESRAD-BUILD result which the air exchange rate is 10 –2 /h is similar with the AEC data. In addition, the results present that the air exchange rate is very significant for calculating the dose.