bluuebunny/crossref_metadata_2025_split huggingface dataset:
New Exploration on TMSR: Redesign of the TMSR Lattice
Authors: Zhao Jinkun , Si Shengyi , Chen Qichang , Bei Hua 
Published: October 2017

Molten Salt Reactor (MSR) has been recognized as one of the Next Generation Nuclear Power systems. Most MSR concepts are the variants evolved from the ORNL’s Molten-Salt Breeder Reactor (MSBR) which employs Molten-Salt as both fuel and coolant, and normally graphite is used as moderator. Many evaluations have revealed that such concepts have low breeding ratio and might present positive power coefficient. Facing these impediments, TMSR (Thorium Molten Salt Reactor) with redesigned lattice is proposed in this paper. Based on comprehensive investigation and screening, important lattice parameters including molten salt fuel composition, solid moderator material, lattice size, structure and lattice P/D ratio (lattice pitch to channel diameter) are redesigned.
In this paper, new composition of fuel salt without BeF2, which is also recommend for Molten Salt Fast Reactor (MSFR), is employed instead of LiF-BeF2-ThF4-UF4 adopted in the design of single fluid MSBR. The new fuel composition makes TMSR to benefit from the increased solubility for actinides (e.g. Th4, UF4). Moreover, due to the decent slowing-down power and neutron multiplication effect by (n,2n) reaction of beryllium, BeO is employed as moderator to improve neutron economy instead of graphite. To avoid corrosion on the one hand, Ceramic cladding (e.g. SiC) is introduced to separate the flowing liquid fuel and fixed solid moderator. More importantly, ceramic cladding is capable of maintaining a stable flow channel and supporting the core structure on the other hand. Concerning neutron spectrum, P/D ratio is an important parameter indicating the volume fraction of fuel in the lattice. In order to obtain a suitable spectrum for better breeding and safety features, lattice size and P/D ratio have been optimized for TMSR. Furthermore, since online reprocessing capability and refueling control are key parameters influencing depletion behavior which concerns the sustainability of the reactor system, these issues are also discussed in this paper.
Simulation of the redesigned TMSR system is performed to evaluate the outcomes of the lattice parameters optimization. SONG/TANG-MSR codes system is applied in the simulation, which is independently developed by Shanghai Nuclear Engineering Research & Design Institute (SNERDI). A traditional core model with LiF-BeF2-ThF4-UF4 fuel and graphite moderator is also evaluated by the codes for reference. Thanks to the optimized lattice parameters and as consequences of the redesigned lattice, TMSR has achieved a high breeding ratio close to 1.13. With a proper reprocessing and refueling strategy, the doubling time of TMSR can be shortened to about 15 years. Meanwhile a negative power coefficient is still maintained. Based on this lattice design, TMSR will have excellent performance on safety and sustainability.
Multi-Nuclide Separation Using Different Types of Zeolites
Authors: Mimura Hitoshi , Matsukura Minoru , Kurosaki Fumio , Kitagawa Tomoya , Kirishima Akira , Sato Nobuaki 
Published: October 2017

Development of selective adsorbents is very important subject for the effective multi-nuclide decontamination related to the severe accident of Fukushima Daiichi Nuclear power Station (Fukushima NPS). In this study, the adsorption properties for nine kinds of zeolites (Zeolite A, Zeolite X, Zeolite Y, Zeolite L, Modified Chabazite, Phillipsite, Erionite, Synthetic Mordenite, Natural Mordenite and Clinoptilolite) are evaluated in the presence of sodium salts, boric acid and seawater. The present study deals with (1) selective adsorption properties for single nuclide ions (Cs+, Sr2+, Eu3+, I−, UO22+, Am3+ and NpO2+), and (2) multi-nuclide adsorption properties of 26 elements (typical elements in Advanced Liquid Processing System (ALPS) in Fukushima NPP-1) for the above zeolites. The distribution coefficient (Kd, ml/g) and uptake (R, %) were estimated by batch method using NaI (Tl) scintillation counter, ICP-AES and AAS.
Zeolites with different crystal structures have the diversity of the adsorption selectivity for various radioactive nuclides. Chabazite, mordenite and clinoptilolite with lantern or tunnel structure were very effective for the adsorption of monovalent Cs+ ions even in real seawater. Zeolite A and X with three-dimensional cage structures were effective for the adsorption of divalent Sr2+ and Co2+ ions under the practical condition (30% diluted seawater). Zeolite L was effective for the adsorption of Eu3+ ions under the practical condition. As for I− adsorption, Ag-zeolites are found to be effective, and the uptake (%) of I− (NaI in pure water) for Ag-zeolites was estimated to be above 98% in pure water. As for actinoid adsorption, the distribution profile, Kdvs pH, had a maximum depending on the hydrolysis pH. Zeolite A, Zeolite L and Zeolite X showed an excellent adsorption property for UO22+, Am3+ and NpO2+, respectively.
Selective adsorption tendencies of different zeolites were evaluated for 26 elements referred to ALPS. Comparing the uptake results for different zeolites, the following tendency of adsorbability was observed. Mordenite had adsorption selectivity for monovalent alkali metal ions of Rb+ and Cs+. Zeolite A and X exhibited relatively high adsorption selectivity for divalent ions of Sr2+ and Co2+. Zeolite L had adsorption selectivity for trivalent lanthanide ions such as Ce3+ and Eu3+. These tendencies were the same as those without boric acid.
Thus, the zeolites with diverse adsorption selectivity are effective for the multi-nuclide decontamination of radioactive contaminated water.
Development and Application of Cobalt Adjuster Rod for 60Co Medical Radioactive Sources Production in CANDU-6 Reactor
Authors: Ding Jie , Dang Yu , Zhu Libing , Tang Chuntao , Li Hui , Qin Huimin 
Published: October 2017

60Co radioactive sources are widely used in industrial and medical field. In 2008, Cobalt adjuster rod assemblies were loaded in the core of CANDU-6 reactor in order to produce industrial radioactive sources. Based on successful production and application of 60Co industrial radioactive sources, the program of developing cobalt adjuster rod for 60Co medical radioactive sources production was launched.
Comparing with industrial radioactive sources, medical radioactive sources require higher activity ∼220Ci/g and smaller sources size, which cause challenges to cobalt adjuster rod design and manufacturing. In this paper, several types of cobalt pellets have been developed to match the medical radioactive sources size and the corresponding nuclear design, TH design & safety analysis, shielding design & safety analysis was also completed to prove that the design requirements could be satisfied.
By now the development of cobalt adjuster rod for 60Co medical radioactive sources production has finished and the first batch of cobalt adjuster rod assemblies have been successfully loaded in core. The first batch of 60Co medical radioactive sources is expected to be put into application in 2019.
The Verification of SONG Library
Authors: Bei Hua , Zhao Jinkun , Chen Qichang , Si Shengyi 
Published: October 2017

We have developed the 293-group cross-section library SONGLIB with complete reaction pathes, broad-spectrum solving ability and wide range isotopes, which is adopted in the lattice code SONG, the next generation reactor lattice calculation and analysis code. SONGLIB provides the multi-group data for the transport calculation, resonance calculation and burnup calculation of SONG. So the precision and reliability of data is very important. However, the size of the data is very large and there are still a lot of personal factors in determination of the processing parameters and methods. In order to evaluate the library, it’s necessary to carry out the test work comprehensively.
Amount of test work is then undertaken based on three kinds of reactors such as pressurized water reactor with UO2 fuel, sodium cooled fast reactor and new-type molten salt reactor with none-beryllium core. In order to test the adaptivity of the library, 39 cases are designed totally, changing the fuel composition, the absorber compositon, the burnup depth and the working temperature. Only the results of a few cases are given in details to shorten the length of this article. For comparison, the Monte Carlo code MCNP is chosen and the library source of evaluated nuclear data is ensured to be the same. The model that has been calculated is simple, the cell with “fuel-cladding-modulator” the three-layer structure and total reflection boundary for the aim of minimizing the differences between the codes. From the macroscopic angle, the results of infinite multiplication factor fit well with each other in total. From the microscopic angle, the absorption rate for actinides, fission products and absorbers, the fission rate for fissile isotopes and scattering rate for moderators are compared with that of MCNP whose output is normalized to the actual cell power. The difference between the reaction rates for most isotopes can be neglected. But there still exist some obvious differences for some isotopes whose resonance effect should have been considered, such as Fe56, ZrNat and W, which might affect the macoscopic results to some extent.
As a conclusion, the data of SONGLIB can be expected with high precision and reliability. And SONGLIB is adapted in solving various problems for reactors with different spectrum, depth of burnup, operation condition or fuel cycling. Nevertheless, there still exist some factors that may affect the final results during the process and use of the library, which should be paid attention to while taking a further step in optimizing and updating the library.
Comparison of Heat-Transfer Correlations Obtained in Supercritical-Water-Cooled Bundles
Authors: Xie Hanqing , Maloufi Hakim , Zopf Andrew , Anderson William , Langevin Christian , Pioro Igor 
Published: October 2017

SuperCritical Water-cooled Reactor (SCWR) as one of the six Generation-IV nuclear-power-reactor concepts will have increased thermal efficiency compared to that of current Nuclear Power Plants (NPPs) equipped with water-cooled reactors by operating the reactor coolant at supercritical conditions: Coolant pressure of about 25 MPa, inlet temperatures between 300–350°C, and outlet temperatures between 550–625°C. The major flow geometry inside the reactor core is the bundle flow geometry. For safe and efficient operation of an SCWR heat transfer coefficients should be calculated with minimum uncertainties.
Unfortunately, the vast majority of experimental datasets were obtained in vertical bare tubes cooled with SCW. Experiments in a bundle flow geometry are even more complicated and expensive compared to that in bare tubes. Due to this very few experiments have been performed in bundles.
According to the abovementioned, the vast majority of heat-transfer correlations are based on bare-tube data, and only one currently known correlation is based on a 7-element bundle cooled with SCW (the so-called, Dyadyakin and Popov correlation (1977)). Rods in this bundle are equipped with four helical ribs to enhance the heat transfer. However, the authors have not provided any dataset(s) associated with this bundle and correlation.
In the current paper a number of bare-tube heat-transfer correlations obtained in SCW and the Dyadyakin and Popov correlation were compared with two datasets obtained in an annular channel with the heated central rod and 3-element bundle. The central rod in this annular channel and rods in the 3-element bundle have the same heated length as those in the 7-element bundle tested by Dyadyakin and Popov in 1977, and are also equipped with four helical ribs.
The comparison showed that the Jackson correlation (2002) is the most accurate one in predicting Heat-Transfer-Coefficient (HTC) profiles in the annular channel at normal heat-transfer regime. The Dittus and Boelter correlation (1930) is the most accurate in predicting HTC profiles in the 3-element bundle at normal heat-transfer regime. No one correlation is capable to follow closely HTC profiles at the deteriorated heat-transfer regimes in both flow geometries. Aloo, it should be mentioned that bare-tube heat-transfer correlations, which have thermophysical properties based on bulk-fluid and wall temperatures, might have problems with convergence at high heat fluxes, i.e., above the heat flux at which the deteriorated heat-transfer regime starts in bare tubes.
Study on Social Stability Risk Analysis for Nuclear Power Projects
Authors: Rongxu Zhu , Feng Zhao , Haifeng Wang 
Published: October 2017

In this paper, the social stability risk analysis of nuclear power projects is carried out, and the working process of social stability risk analysis and risk evaluation index system of social stability is established. The social stability risk analysis of nuclear power plant in Guangdong province was taken as an example to introduce the application in this paper, at the same time for the current nuclear power project should focus on risk points to put forward risk mitigation measures and recommendations.
This study can provide theoretical support and technical support for the social stability risk analysis of nuclear power projects, and provide decision support for the construction of nuclear power projects.
The Impacts of Scale Effect on Flow Patterns in a Supersonic Steam Injector
Authors: Fujishiro Masaya , Abe Yutaka , Kaneko Akiko 
Published: October 2017

From the viewpoint of an importance of safety, the nuclear power plant should be managed to prepare severe accidents. The performance of safety dropped by an accident is strongly to be minimized during the situation of station blackout. The installation of a steam injector (SI) into the nuclear power plant has long been expected.
In the SI, the steam condenses due to the direct contact at the surface of water jet, resulting in the force attracting water. The force drives the circulation of an amount of coolant water. SI also works as a reactor condenser thanks to its high efficient performance during the condensation. Because any external forces to circulate water and steam are not required, SI can be operated without the electric powers. The structure of SI is similar to a convergent-divergent nozzle. After the flow acceleration at a throat, the discharged pressure is expected to exceed the inlet pressure. Owing to its quite simple structure, the reduced cost of installation and maintenance is also expected.
The following previous studies for four cases of throat diameter clarified two-phase flow structures and heat transfer characteristics in water jet and performance of SI: (i) Narabayashi et al. (2000) examined for 5.5 and 6.5 mm in diameter; (ii) Osakabe et al. (2004) for 3.4 mm; (iii) Koizumi et al. (2006) for 4 mm; (iv) Abe et al. (2014) for 4, 6.5, and 8 mm. Although these clarified the operative state which formed a water jet, operative condition was not elucidated. Furthermore, the scale effect for various diameters of SI has not been discussed in detail.
The aim of this study is to clarify scale effect of a test section on operating criteria and performance. Experiment was performed to clarify the scale effect by using three types of throat diameters: 4, 6.5, and 8 mm.
As a result, three formations of a water jet were observed: (i) formation, (ii) incomplete formation, and (iii) no formation. We proposed a classification which enables us to categorize complex flow patterns into five regimes. We clarified the operating criteria of them by comparing water flow rate with steam flow rate. SI did not form a water jet on the condition with low steam flow rate. The suppling water was stopped, and only steam was supplied to the test section for the condition that steam latent heat was larger than subcooled water enthalpy.
Comparison of Test and Analytical Results on CAP1400 Reactor Flow Distributor Designs
Authors: Wei Zhang , Ming Zhang , Hao Yu , Qing Yu , Shaoxuan Lin 
Published: October 2017

The CAP1400 reactor internal is going to use a new component termed the “Even Flow Distributor (EFD)”, instead of the existing flow skirt (FS) design, to help distribute the incoming flow more evenly to the fuel assemblies. To verify the effect of the EFD, a scale model of the reactor and internals was built and hydraulic tests of both the EFD and the FS configurations were conducted. In addition, numerical simulations of the flow fields, using CFD, of both designs were also carried out.
From the scale model test results, the overall flow distribution of EFD is better than that of the FS. The core inlet flow distribution taken from the CFD results is slightly better than that from the hydraulic test. The differences between CFD result and test results are less than 3 percent for the most of fuel assemblies, and about 5 percent for a few assemblies. Based on this study, it is concluded that the EFD is a very effective means of controlling core inlet flow distribution in a CAP1400 reactor.
Verification of FPGA Based NPP I&C Systems Considering Multiple Faults: Technique and Automation Tool
Authors: Yasko Alexander , Babeshko Eugene , Kharchenko Vyacheslav 
Published: October 2017

Instrumentation and Control (I&C) systems for Nuclear Power Plants (NPP) are exceedingly complicated electronic solutions that include thousands of different components such as microcontrollers, Field-Programmable Gate Arrays (FPGAs), integrated circuits etc. Deployment of such safety-critical systems cannot be performed without complex safety and reliability assessment, verification and validation (V&V) activities that are addressed to exposing of overlooked faults. The examples of such activities are Fault Tree Analysis (FTA), Failure Modes and Effects Analysis (FMEA), Fault Injection Testing (FIT). Due to complexity of NPP I&C systems in most cases the process of assessment is very time consuming and the results mostly depend on experts’ qualification.
Traditional safety and reliability assessment methods are being constantly modified and enhanced so as to comply with increasing demands of national and international standards and guidance, as well as to be applied for I&C systems that contain number of complex components like FPGA.
Although much work related to analysis of FPGA-based systems has been performed, there is a lack of detailed technique for FPGA-based I&C systems failure identification that considers probability of several faults at the same time (multi-faults), development of preventive strategies for controlling or reducing of the risk related to such failures, as well as automation of this technique so as to make it utilizable for real NPP industry tasks.
FIT as verification for Failure Modes, Effects and Diagnostics Analysis (FMEDA) was used during Safety Integrity Level 3 (SIL3) certification process of RadICS NPP I&C platform, while the parts of proposed technique were used as internal verification and validation activities applied on several modules of the platform.
The Research and Comparison of Thermal-Hydraulic Test Facilities for Pressurized Water Reactors (PWRs)
Authors: Lu Yuan , Xiao Changzhi 
Published: October 2017

Recently, nuclear power safety draws more attention after Fukushima nuclear accident, for which it is essential to construct a large number of test facilities simulating possible issues occurred in the reactor. The thermal-hydraulic test facility is extensively used to simulate thermal-hydraulic response during a loss of coolant accident (LOCA) or an operational transient which can minimize the nuclear safety accidents. This paper focus on the research of thermal-hydraulic test facilities of PWRs in different countries. All of facilities were designed by scaling analysis method. Meanwhile, a wide range of data comprising of power data, pressure data, volume data, configuration and a series of nuclear safety test data is compared in over ten test facilities. Based on above comparable data and relevant research, the main conclusions are as follows:
Scaling analysis lays a solid foundation for the design and construction of scale-down nuclear reactor thermal-hydraulic test facilities. This would provide the reference for choosing scaling concepts in the reactor integral test facility.
Characterization of FUTURIX-FTA Irradiated Nuclear Fuel Samples
Authors: Andrello Concettina , Freis Daniel , Lo Frano Rosa , Papaioannou Dimitri , Delage Fabienne 
Published: October 2017

The amount of spent fuel and high-level waste already available, and which will be produced by the future NPPs operation, calls for the evaluation of any possible technological solution that could minimize the burden of their disposal: reduction of Minor Actinide (MA) content, in addition to the radiotoxicity and radioactivity, and of the generated thermal load (decay heat). In this context, R&D efforts currently focus on the development of methodologies and technical solutions for Partitioning and Transmutation.
MAs and long-lived fission products are in fact the main contributors to the long-term radiotoxicity of spent nuclear fuel, and their transmutation to short-lived fission products, in fast spectrum nuclear reactors, in transmuters or in Accelerator Driven Systems (ADS), by neutron irradiation of dedicated fuels/targets, is a promising and widely investigated option.
In order to provide substantial input for the safety assessment of innovative nuclear fuels dedicated to MA transmutation, several irradiation tests are being carried out. In some options, the investigated fuels/targets are uranium free, or of low uranium content, to improve the transmutation performance and contain high concentrations of MA and plutonium compounds.
Two molybdenum based CER-MET fuels, called ITU-5 & ITU-6, were prepared at JRC Karlsruhe for the irradiation experiment FUTURIX-FTA (FUel for Transmutation of transURanium elements in phenIX/Fortes Teneurs en Actinide). The experiment performed from 2007 to 2009 in the Phénix reactor, France, in cooperation with CEA. The experiment ended after 235 equivalent full power days (EFPD) at a Linear Heat Rate of circa 130 W/cm and reached burn-ups of 18 %FIHMA and 13 %FIHMA, respectively. Afterwards, the pins were transported to the Hot Cells of JRC Karlsruhe for Post Irradiation Examination.
After a short summary describing the fuel preparation and irradiation conditions of the FUTURIX FTA irradiation experiment, the paper will give an overview on the current status and further planning of the Post Irradiation Examinations of ITU-5 & ITU-6 at JRC Karlsruhe. Finally, the results of the characterisations will be discussed and conclusions on the irradiation performance will be drawn. The results of this experiment will help to increase the knowledge and understanding of the irradiation behaviour of metal based transmutation targets and the qualification and validation of models developed to predict fuel safety performance.