bluuebunny/crossref_metadata_2025 huggingface dataset:
Structural Integrity Research for Reactor Pressure Vessel Under In-Vessel Retention of a Core Melt
Authors: Gao Yongjian , He Yinbiao , Cao Ming , Li Yuebing , Bao Shiyi , Gao Zengliang
Published: October 2016

In-Vessel Retention (IVR) is one of the most important severe accident mitigation strategies of the third generation passive Nuclear Power Plants (NPP). It is intended to demonstrate that in the case of a core melt, the structural integrity of the Reactor Pressure Vessel (RPV) is assured such that there is no leakage of radioactive debris from the RPV. This paper studied the IVR issue using Finite Element Analyses (FEA). Firstly, the tension and creep testing for the SA-508 Gr.3 Cl.1 material in the temperature range of 25°C to 1000°C were performed. Secondly, a FEA model of the RPV lower head was built. Based on the assumption of ideally elastic-plastic material properties derived from the tension testing data, limit analyses were performed under both the thermal and the thermal plus pressure loading conditions where the load bearing capacity was investigated by tracking the propagation of plastic region as a function of pressure increment. Finally, the ideal elastic-plastic material properties incorporating the creep effect are developed from the 100hr isochronous stress-strain curves, limit analyses are carried out as the second step above. The allowable pressures at 0 hr and 100 hr are obtained. This research provides an alternative approach for the structural integrity evaluation for RPV under IVR condition.
New Structure Design for SCWR Internal Components
Authors: Zhou Qin , Song Yu , Wang Xiaoxin , Li Xiaotian
Published: October 2016

Structure design of supercritical light water cooled reactor (SCWR) faced some problem such as materials selection for their special application environment. SCWR vessel internal component materials not only service with irradiation damage, but also various temperatures. The temperature of reactor coolant inlet is 280 °C while outlet is 550 °C, in addition, the thermal expansion of internal components materials and reactor pressure vessel (RPV) components are different. Thus, reducing thermal stress in SCWR internal components and keeping inlet/outlet pipe seal are contradictory. The authors designed a SCWR internal components structure with combined application of 316Ti and N10242. The mechanical properties and irradiation performances of these materials were proved to be available, and finite elements calculation suggests that this new structure can evidently reduce the thermal stress of SCWR internal components.
Benchmarking the NEM Real-Time Core Model for VVER-1000 Simulator Application: Asymmetric Core
Authors: Georgieva Emiliya , Dinkov Yavor , Ivanov Kostadin , Stieglitz Robert
Published: October 2016

A real-time version of the Nodal Expansion Method (NEM) code is developed and implemented into Kozloduy 6 full-scope replica control room simulator. Combined with an enhanced thermal-hydraulics and I&C models the whole package is a high-fidelity simulation tool for operator training and various other applications. The fidelity and accuracy of simulation with emphasis on reactor core model is illustrated through comparison with plant-specific data. The transient of ‘Switching-off of One of the Four Operating Main Circulation Pumps at Nominal Reactor Power’ as described in OECD/NEA Kalinin 3 Coolant Transient Benchmark is an example of an asymmetric core scenario with a range of parameter changes.

Simulation results concerning fuel assembly power and axial power distribution during the transient are compared with records from Kalinin 3 in-core monitoring system.

Main operating parameters of nuclear steam supply system of a VVER-1000/V320 series units vary to a considerable degree. While Kalinin 3 benchmark specification contains very good description of the transient, as well as record of many parameters of the unit, the document provides only superfluous description of the reference unit. In such a case, an approach based on a ‘generic’ V320 model by default introduces deviations which are difficult to quantify. There are several examples which warrant discussion.

One example is core coolant flow and pressure loss during the transient. Pump head and pressure loss across reactor vessel are measured and recorded and in-core monitoring system provides estimation of core coolant flow, which is quite high in comparison with some other V320 units (e.g. by about 5 % larger). Without more detailed pressure loss data across the main circulation loop and specific pump characteristics, however, one can only guess how much simulation is off the mark. Another detail of the same problem is coolant flow through a specific fuel assembly. The presence of a fuel assembly of different design (TVS-M type) surrounded by TVSA type fuel assemblies shall be thoroughly considered, because secondary sources indicate significant differences in fuel assembly pressure loss coefficients between the two types. Coolant flow affects coolant (and fuel) temperature profile and thus neutron cross-sections.

Yet another example, even more strongly affecting our ability to interpret simulation results is core power reconstruction provided by the in-core monitoring system of the unit. The SPND (Self-Powered Neutron Detector) current readings are subject of conversion by an algorithm based upon simulated spatial neutron flux distribution across the reactor core. While error estimation of the parameters in stationary conditions is available from secondary sources, there is no reliable estimation of error magnitude during the transient.
Thermo-Hydraulic Analysis of the Optimized Helium Cooled Solid Breeder Blanket for CFETR
Authors: Cui Shijie , Zhang Dalin , Chang Jie , Tian Wenxi , Qiu Suizheng , Su G. H.
Published: October 2016

China Fusion Engineering Test Reactor (CFETR) is under design recently, in which a conceptual structure of the helium-cooled solid breeder blanket is proposed as one of the candidate tritium breeding blankets. In this concept, three radial arranged U-shaped breeding zones are designed and optimized for higher Tritium Breeding Ratio (TBR) and structure simplification. This blanket uses the Li4SiO4 lithium ceramic pebbles as the breeder, while beryllium pebbles as the neutron multiplier. In this paper, the thermal and fluid dynamic analyses of the optimized typical outboard blanket module are performed by CFD method, where the nuclear heating rate is obtained from the preliminary neutronics calculations. The thermal hydraulic behaviors of the first wall (FW), the temperature distributions of submodule structure material, Li4SiO4 pebble bed and Beryllium pebble bed under normal and critical conditions are calculated, respectively. The results show that the temperature on the blanket module can be effectively cooled below allowable temperature limits of the materials, even if the FW is suffering the maximum surface heat flux, which verified the reasonability of the design of the blanket cooling scheme. In addition, several parametric sensitivity studies are conducted to investigate the influences of main parameters (e.g. coolant mass flow rate, inlet temperature, pebble bed thermal conductivity and fusion power) on the temperature distributions of the blanket components.
Research on Tensile Properties of Ni-Base Super Alloy GH3128
Authors: Zhao Xi , Zhou Yu , Yuan Kun
Published: October 2016

GH3128, which is one of the domestic Ni-base super alloy, shows good performances under high temperature. Preliminary research has shown that this alloy is promising as structural material for Intermediate Heat Exchangers (IHX), which are the key components of Very High Temperature Reactor (VHTR). In this paper, the tensile properties of GH3128 are tested by the uniaxial tension experiments, and then the fractographic evaluation is conducted by using the scanning electron microscope (SEM). In the paper, the variation tendency of tensile strengths of GH3128 as well as plasticity is described along with the variation of temperature. The results also show that the tension fracture mode will gradually transform from ductile fracture to cleavage-intergranular compound fracture with the increasing of temperature, and the transition of fracture mechanism corresponds well with the variation of alloy plasticity.
Natural Convection Heat Transfer From Vertical 5×5 Rod Bundles in Liquid Sodium
Authors: Hata Koichi , Fukuda Katsuya , Mizuuchi Tohru
Published: October 2016

Natural convection heat transfer from vertical 5×5 rod bundles in liquid sodium was numerically analyzed for two types of the bundle geometry (equilateral square and triangle arrays, ESA and ETA). The unsteady laminar three dimensional basic equations for natural convection heat transfer caused by a step heat flux were numerically solved until the solution reaches a steady-state. The PHOENICS code was used for the calculation considering the temperature dependence of thermophysical properties concerned. The 5×5 test rods for diameter (D = 7.6 mm), heated length (L = 200 mm) and L/d (= 26.32) were used in this work. The surface heat fluxes for each cylinder were equally given for a modified Rayleigh number, (Rf,L)ij and (Rf,L)5×5,S/D, ranging from 3.08 × 104 to 4.19 × 107 (q = 1 × 104∼7 × 106 W/m2) in liquid temperature (TL = 673.15 K). The values of S/D, which are ratios of the diameter of flow channel for bundle geometry to the rod diameter, for vertical 5×5 rod bundles were ranged from 1.8 to 6 on each bundle geometry. The spatial distribution of local and average Nusselt numbers, (Nuav)ij and (Nuav,B)5×5,S/D, on vertical rods of a bundle was clarified. The average value of Nusselt number, (Nuav)ij and (Nuav,B)5×5,S/D, for two types of the bundle geometry with various values of S/D were calculated to examine the effect of the bundle geometry, S/D, (Rf,L)ij and (Rf,L)5×5,S/D on heat transfer. The bundle geometry for the higher (Nuav,B)5×5,S/D value under the condition of S/D = constant was examined. The correlations for (Nuav,B)5×5,S/D for two types of bundle geometry above mentioned including the effects of (Rf,L)5×5,S/D and S/D were developed. The correlations can describe the theoretical values of (Nuav,B)5×5,S/D for two types of the bundle geometry for S/D ranging from 1.8 to 6 within −11.77 to 13.34 % difference.
3-D Numerical Simulation of the Vapor-Liquid Flow at the Shell Side of Shell-and-Tube Heat Exchangers
Authors: Xiao Qi , Yang Ning , Zhao Zhenxing , Dai Chunhui , Wu Jun , Bai Fan , Wang Junrong , Wei Zhiguo , Tao Mo
Published: October 2016

The boiling vapor-liquid flow at the shell side of shell-and-tube heat exchangers was simulated by Euler-Euler transient 3D method in this paper. The mass and heat transfers between the two-phase fluid and heated wall for the subcooled boiling phenomenon were described by the Rensselaer Polytechnic Institute model (RPI model), while the steam condensation within the subcooled liquid was described by the Lee model. Firstly, different turbulence and interfacial force models were evaluated by comparing with the experimental data of Bartolomej (1982). It was found that the turbulence models have minor influence on the temperature and vapor volume fraction distributions. As the bubble size in the subcooled boiling process is small (usually <1 mm), the velocity slip between the vapor bubbles and the liquid is not so important. The simulation results using different drag force models are similar, and the Tomiyama model offers relatively better predictions. The non-drag forces could not significantly improve the accuracy in our simulations. Then the gas-liquid boiling flow at the shell side of shell-and-tube heat exchangers was then simulated. It was found that the water temperature increases almost linearly near the inlet zone, and the increase speed was slowed down when the bulk temperature approached to the saturated point as the boiling process happened more frequently and consumed much heat. The heat exchangers with the triangle and square configurations have similar temperature and vapor distributions. Further analyses for those two kinds of tube configurations are needed.
Numerical Study of the Effects of Wall Force Upon Bubble Motion Under Static and Rolling Condition
Authors: Huang Ying , Gao Puzhen
Published: October 2016

Much research on bubble dynamics under rolling motion has been carried out in recent years, but the effect of the wall force on bubble motion behavior is far from satisfactory and still needs to be further investigated. In order to study the interaction between bubbles and near wall under rolling condition, the paper adopts the VOF method, one momentum equation coupled with surface tension mode and the function of UDF. The bubble motion behaviors in a vertical channel under static and rolling conditions were numerically simulated.

The impact of the wall force on single bubble motions were discussed. Plus, the effect of the wall force on bubbles when they are approaching the wall was analyzed in detail. The numerical results agree well with the experimental results from other researchers. The simulating results show that as bubbles gradually approach the wall, the wall force prevents them from further coming closer to the wall and pushes them in the opposite direction. Under the rolling condition, the wall force affects the bubble motion only when the wall is located in the positive bubbles’ rolling direction.
Single Phase Flow in the Randomly Packed Bed With Spheres: Simulation and Experimental Validation
Authors: Zhang Nan , Sun Zhongning , Ding Ming
Published: October 2016

A computational fluid dynamic (CFD) model for single phase flow in the three dimensional randomly packed bed with spherical particles has been developed and validated with experimental results. The flow characteristics within this complex geometry are very complicated. In order to obtain insight into the interior and local flow characteristics, Three-dimensional simulation is required. First, we constructed the randomly packed bed with spherical particle, using Discrete Element Method (DEM) based on the integration of Newton’s laws of motion. To validate the DEM simulations the global bed porosity and the radial porosity distribution were compared with empirical correlation from literature. Second, the complex geometrical properties of random packed bed make it difficult to produce a fine mesh. Herein, the bridge method for meshing the particle-particle and particle-wall contact points in the packed bed was applied. The contact zones are modified and then partitioned into several regular parts, so the structure gird was meshed. Finally, the simulation of water flow in the randomly packed bed with a tube-to-particle diameter ratio of 6.325 has been carried out by the commercial CFD code. A comparison with previously published correlations and experimental data shows that the relationship proposed by KTA agree well with the measured pressure drop. Furthermore the results of simulation for distribution of velocity in the bed were analyzed and discussed.
A Systematic Approach for Severe Accident Uncertainty Analysis
Authors: Hoseyni Seyed Mohsen , Pourgol-Mohammad Mohammad
Published: October 2016

Uncertainty exists in every modeling process especially in those areas with complexity of the calculations like severe accident (SA) code which cover a broad range of physical and chemical phenomena. A systematic framework is proposed here for effective uncertainty assessment of SA computations by efficient use of available data and information. Available methodologies are either input-based or output based. The proposed methodology takes the advantages of both approaches and introduces an integrated one which quantifies the uncertainty of code input parameters (parameter uncertainty), code internal structure (model uncertainty) and code outputs (output uncertainty). The proposed methodology is comprisd of a hybrid qualitative and quantitative approach for identification of uncertainty sources. Using a Bayesian ensemble of sensitivity measures, identified severe accident phenomena are ranked according to their effect on the figure of merit. The other feature of the proposed methodology is the consideration of the SA code structural uncertainties (generally known as model uncertainty) explicitly by treating internal sub-model uncertainties and by propagating such model uncertainties in the code calculations, including uncertainties about input parameters. The code output is further updated through additional Bayesian updating with available experimental data from the integrated test facilities. In this paper, the key elements are discussed for the uncertainty analysis methodology and its application is demonstrated on the LP-FP2 experiment of LOFT test facility.
Design Approach for Mitigation of Air Ingress in High Temperature Gas-Cooled Reactor
Authors: Sato Hiroyuki , Ohashi Hirofumi , Nakagawa Shigeaki
Published: October 2016

One important safety design consideration for high temperature gas-cooled reactor (HTGR) is air ingress following a rupture of the reactor pressure boundary such as primary piping. The air intrusion to the reactor core held at high temperature through the break will results in significant oxidation of graphite components and fuels. Such oxidation may leads to the weakening of core support structures as well as fuel element damage and subsequent fission product release.

This paper intends to propose a practical solution to protect the reactor from severe oxidation against air ingress accidents without reliance on subsystems. Firstly, a change is made to the center reflector structure to minimize temperature difference during the accident condition in order to reduce buoyancy-driven natural circulation in the reactor. Secondly, a modified structure of the upper reflector is suggested to prevent massive air ingress against a rupture in standpipes. As a preliminary study, a numerical analysis is performed for a typical prismatic-type HTGR to study the effectiveness of the proposed design concept using simplified lumped element models. The analysis considers internal decay heat generation and transient conduction from inner to outer regions at the reactor core, cooling of vessel outer surface by radiation and natural convection, and natural circulation flow in reactor. The results showed that amount of air ingress into the reactor can be significantly reduced with practical changes to local structure in the reactor.
Preparation for a New Experimental Program Addressing Core-Material-Relocation Behavior During Severe Accident With BWR Design Conditions: Conduction of Preparatory Tests Applying Non-Transfer-Type Plasma Heating Technology
Authors: Abe Yuta , Sato Ikken , Ishimi Akihiro , Nakagiri Toshio , Nagae Yuji
Published: October 2016

A new experimental program using non-transfer type plasma heating is under consideration in JAEA to clarify the uncertainty on core-material relocation (CMR) behavior of BWR. In order to confirm the applicability of this new technology, authors performed preparatory plasma heating tests using small-scale test pieces (107 mm × 107 mm × 222 mmh). Based on these preliminary results, an excellent perspective in terms of applicability of the non-transfer plasma heating technology to the SA (Severe Accident) experimental study was obtained. Furthermore, JAEA is preparing for the next step intermediate-scale preparatory tests in 2016 using about 50 rods and a control blade that would not only confirm its technical applicability, but also some insights relevant to the issue on CMR itself.
Phased Array Ultrasonic Inspection Technique for Cast Austenitic Stainless Steel Parts of Nuclear Power Plants
Authors: Yamamoto Setsu , Semboshi Jun , Sugawara Azusa , Ochiai Makoto , Tsuchihashi Kentaro , Adachi Hiroyuki , Higuma Koji
Published: October 2016

For safety operation of nuclear power plants, soundness assurance of structures has been strongly required. In order to evaluate properties of inner defects at plant structures quantitatively, non-destructive inspection using ultrasonic testing (UT) has performed an important role for plant maintenances. At nuclear power plants, there are many structures made of cast austenitic stainless steel (e.g. casings, valve gages, pipes and so on). However, UT has not achieved enough accuracy measurement at cast stainless steels due to the noise from large grains. In order to overcome the problem, we have developed comprehensively analyzable phased array ultrasonic testing (PAUT) system.

We have been noticing that dependency of echo intensity from defect is different from grain noises when PAUT conditions (for example, ultrasonic incident angles and focal depths) were continuously changed. Analyzing the tendency of echoes from comprehensive PAUT conditions, defect echoes could be distinguished from the noises. Meanwhile, in order to minimize the inspection time on-site, we have developed the algorithms and the full matrix capture (FMC) data acquisition system. In this paper, the authors confirmed the detectability of the PAUT system applying cast austenitic stainless steel (316 stainless steel) specimens which have sand-blasted surface and 3 slits which made by electric discharge machining (EDM).
Numerical Investigation of Slug Flow in a Horizontal Pipe Using a Multi-Scale Two-Phase Approach to Incorporate Gas Entrainment Effects
Authors: Wenzel Stefan , Czapp Marek , Sattelmayer Thomas
Published: October 2016

Numerical as well as experimental investigations of the highly intermittent slug flow regime of a gas-liquid mixture in horizontal pipes are of particular interest for nuclear reactor safety in post loss-of-coolant accident (LOCA) situations. The strong variation of governing interfacial length scales, as they are characterizing the slug flow regime, pushes common numerical multi-phase approaches to their limits, since they are designed either for interface capturing or for modeling the sub-grid behavior of the dispersed mixture. In this work an enhanced hybrid two-phase flow solver is employed to investigate the global and local characteristics of adiabatic, horizontal slug flows in a water-air system. A dynamic switching algorithm for an interface capturing procedure is introduced to examine segregated and dispersed parts in the same flow domain. The inter-facial area transport equation (IATE) is used to detect dispersed flow regions as well as to determine variable bubble sizes and their distribution within the slug body. Experimental results of videometry measurements on a horizontal, 10 m long pipe with an inner diameter of 54 mm at atmospheric pressure and room temperature are compared with numerical results of the same geometry in terms of global characteristics such as slug frequency and onset position. Local properties, such as the interfacial area density in the slug body, are also examined. This study demonstrates the capability of a coupled multiscale approach based on the Euler-Euler two-fluid model (TFM) for the simulation of slug flow in horizontal pipes with a high amount of entrainment.
Water Level Control for Steam Generator in Nuclear Power Based on Port-Controlled Hamiltonian Method
Authors: Tao Mo , Yu Rui , Li Xianling , Ke Zhiwu , Zhao Zhenxing , Xiao Qi
Published: October 2016

This article designs the water level of steam generator control system applying PCH system modeling and control principle of Energy-Shaping.

The basic principles of port-controlled Hamiltonian (PCH) system modeling and control principle of Energy-Shaping are provided, including passivity, dissipation and so forth. Meanwhile, port-controlled Hamiltonian (PCH) system, energy-shaping principle, matching method and its corresponding realization and stability analysis are presented. The water level of steam generator is accomplished; the controlled object is mathematically modeled with principle modeling method. Then the mathematic model is translated into PCH model. the proper Hamiltonian function is finely selected. The water level equilibrium point of steam generator system is respectively determined; the PCH controller based on energy-shaping method is ultimately obtained.

The proposed modeling and control methods provide a potential idea for the other control methods applied method on steam generator system. Moreover, the proposed methods can be put into industrious process control.
Impact of Deviations Handling on Outages
Authors: Matahri Naoëlle
Published: October 2016

Based on the information provided by the operators, IRSN experts select and analyze different deviations presenting a possible generic nature which could affect the safety of power plants. Some of these deviations result in non-compliance (NC) with the safety requirements.

To maintain an acceptable level of safety, an operator has to implement corrective measures for any situation of non-compliance with safety requirements. IRSN, the Technical Support Organization of the French Nuclear Authority, analyzes the different deviations to assess the impacts on the concerned NPPs safety. Based on the impact on safety, measures should be applied immediately or during the next outages, on a reactor or on several of them. The permanent corrective measures schedule is defined taking into account the “NC” safety impact. However, for some of the “NCs”, it can be difficult to define and implement swift permanent corrective measures, especially when the lack of compliance affects several similar units and requires a design change.

This paper explains the French approach of deviations treatment and specifically the relationship between the Nuclear Safety Authority, the Technical Support Organization, IRSN and the Licensee, EDF during an outage.
Use of Algebraic Heat Flux Models to Improve Heat Transfer Predictions for Supercritical Pressure Fluids
Authors: Papp Vera , Pucciarelli Andrea , Sharabi Medhat , Ambrosini Walter
Published: October 2016

This work proposes simulations of heat transfer under supercritical pressure conditions showing improvements with respect to previous works. This is obtained by the introduction of the Algebraic Heat Flux Model (AHFM) for evaluating the turbulent heat flux in turbulence production terms, using the in-house code THEMAT and the STAR-CCM+ code. The first code makes use of the AHFM also in the energy balance equations, while for the commercial code simplifying assumptions are considered in the implementations.

Custom sets of parameters for every condition of inlet temperature and internal diameter are tuned in some cases, driven by the opinion that a single set of parameters cannot be suitable in every flow conditions, considering the complexity of the variables that concur in the heat transfer deterioration phenomenon. The AHFM model gives promising results with new sets of parameters in order to model the deterioration and the recovery phases because of its term related to the variance of temperature.
Accurate Simulation of HTGR Steam Generator for Pressurized Loss of Forced Cooling Accident
Authors: Lu Jianan , Guo Jiong , Li Fu
Published: October 2016

Nuclear power plant is a large-scale complicated system, which includes reactor core, steam generator, turbine and other important components. These components are tightly coupled with each other. Among these components, steam generator is the key link of primary circuit system and secondary circuit system. Heat transfers from primary side to secondary side. Once-through steam generator applied in high temperature gas cooled reactor (HTGR) has the properties of small heat capacity and rapid response speed. In HTGR system, the steam generator should match with the properties of reactor core such as large heat capacity and the slow response. Therefore, accurately simulating the steam generator is a complex task and has a great impact on the coupling property of a reactor system. To address this issue, effects of boundary conditions on the output water quality are analyzed and time integration schemes of backward differentiation formula (BDF) are implemented to HTGR steam generator simulation code BLAST in this work. The introduced BDF is a higher-order approximation to a transient term. It can reduce the numerical error from an explicit time integration scheme. The modified code is numerical tested in a noteworthy HTGR accident operation condition: Pressurized Loss Of Forced Cooling (PLOFC) accident. The performance of HTGR steam generator in the accident is analyzed. The accuracy of the improved algorithm is compared with the original BLAST code. Result shows the safety characteristics of steam generator in PLOFC accident and indicates that the numerical accuracy is significantly improved for both helium and water sides by BDF. For the consideration of accuracy and stability, BDF2 is chosen in the modified BLAST code.
Study on Current Status and Future Developments in Nuclear-Power Industry of Ukraine
Authors: Zvorykin Alexander , Pioro Igor , Panchal Raj
Published: October 2016

Nuclear power in Ukraine is the most important source of electricity generation. Currently, Nuclear Power Plants (NPPs) generate 45.5% of the total electricity in the country followed with coal generation – 38%, gas generation 9.6% and the rest is based on renewable sources, mainly on hydro power plants – 5.9%. Nuclear-power industry is based on 4 NPPs including the largest one in Europe – Zaporizhzhya NPP with about 6,000 MWel gross installed capacity.

These NPPs are equipped with 13 VVER-1000 and 2 VVER-440 Russian-design Pressurized Water Reactors (PWRs) with the total gross installed capacity of 13,800 MWel. Layout of these NPPs, thermodynamic diagram and thermal efficiencies are provided. Thermal efficiencies have been calculated with the IAEA Desalination Thermodynamic Optimization Programme DE-TOP and compared to the actual ones.

Two of these reactors have been built and put into operation in 70-s, ten in 80-s, one in 90-s and just two in 2004. Therefore, based on an analysis of the world power reactors in terms of their maximum years of operation (currently, the oldest reactors are 45-year old) several projections have been made for future of the nuclear-power industry in Ukraine. Unfortunately, all these projections are quite pessimistic.

There is a possibility that around 2030–2035 the vast majority of the Ukrainian reactors will be shut down, and Ukraine can be left without the basic and vital source of electricity generation. Also, current problems of Ukrainian NPPs are: 1) lower capacity factors (around 80%) compared to those in other countries (∼90%); 2) uncertainties with nuclear-fuel supply due to political situation; and 3) service and repairs of relatively old reactors.
Measurements of Fine Structures in the Lead-Bismuth Eutectic Alloy Melts by Differential Scanning Calorimetry
Authors: Ye Changqing , Li Qiulin , Wu Ping , Tang Guoyi , Liu Wei
Published: October 2016

The implementation of the Generation IV nuclear reactors and a spallation target of the accelerated driven system (ADS) concerning to the use of liquid lead-bismuth eutectic (LBE) alloy. The liquid LBE alloy should be fully characterized and especially its physical properties should be completely known to make sure the nuclear safety. Differential scanning calorimetry (DSC) experiments were employed on LBE alloy at a temperature range of room temperature (RT) to 500°C to detect the structure phase transition and obtain the thermal effect of LBE alloy.

The results of DSC curves showed that there existed two remarkable thermal signal events at the melting points zones of Pb and Bi during the melting process tested with a scan rate of 2°C/min. Even though the scan rate was increased to 5°C/min, the DSC signal still exhibited the changes of curve slop at the element melting point zone. Just this interesting DSC thermal signal change phenomenon around 300°C contributed to the relationship with the explanation on the severe embrittlement of T91 steel induced by liquid LBE alloy. The results suggest that there existed effective critical size and form of chemical clusters in liquid LBE alloy worked actively on the embrittlement of T91 Steel.
Development of Error Reduction Methods in Aerosol Measurement for Pool Scrubbing Experiment
Authors: Sun HaoMin , Sibamoto Yasuteru , Okagaki Yuria , Yonomoto Taisuke
Published: October 2016

Pool scrubbing is one of the effective mechanisms to filter out radioactive aerosols in a severe accident of a nuclear reactor. A lot of work has been done on the pool scrubbing models and experiments. However, large discrepancies still exist between the simulation and experimental results. To improve the pool scrubbing model, an accurate decontamination factor (DF) evaluation by an aerosol measurement and a detailed two-phase flow structure measurement is inevitable. A pool scrubbing experimental apparatus was constructed by the thermohydraulic safety research group in Japan Atomic Energy Agency. The test section is a transparent pipe with the inner diameter of 0.2 m and the length of about 4.5 m. The aerosol laden air flow was injected upwardly into the pool water. The aerosol particle diameter distribution was measured by a light scattering aerosol spectrometer. White polydisperse BaSO4 particles were used as the aerosol test particles. In the first step, we focused on investigating and reducing the error of DF experimentally. Several problems resulting in the error and their solutions for the error reduction were summarized in this paper. Based on the error reduction methods, the DFs of pool scrubbing were measured in two water submergences. The results showed that the DFs for the aerosol with small diameter were independent of the injecting air velocity in the submergence of 0.3 m. In addition, it was found that the DFs increased with increasing the air flow rate in the submergence of 2.9 m. It was presumed that the increase of DF was dominated by the increase of bubble surface area and/or turbulence intensity with the air flow rate increase, while the effect of the reduced bubble traveling time in the water, which may reduce the DF, was smaller than the increasing effect.
Effects of Applying the Implicit Particle Fuel Model for Pebble-Bed Reactors
Authors: Li Zhifeng , Cao Liangzhi , Wu Hongchun , Wan Chenghui , Hu Tianliang
Published: October 2016

In the pebble-bed high temperature gas-cooled reactor, there exist randomly located TRISO coated fuel particles in the pebbles and randomly located pebbles in the core, which is known as the double stochastic heterogeneity. In the previous research, the regular lattice pattern was used to approximately simulate the pebble unit cells because the difficulties in modeling the randomly located TRISO geometric. This work aimed at to quantify the stochastic effect of high-temperature gas cooled pebble-bed reactor unit cells, and in view of the strong ability to carry out the accurate simulation of random media, the implicit particle fuel model of Monte Carlo method is applied to analyze to the difference between regular distribution and random distribution. Infinite multiplication factors of the pebble-bed reactor unite cells were calculated by the implicit particle fuel model and simple cube regular lattice pattern at different TRISO packing factor from 0.5%–50%. The results showed that the simple cube regular lattice pattern underestimates the infinite multiplication factors for most packing fractions, but overrates the infinite multiplication factors when the packing fraction is very low.
Analyses of Severe Accidents in GFR Demonstrator ALLEGRO With MELCOR 2.1
Authors: Vácha Petr , Bělovský Ladislav
Published: October 2016

The helium-cooled Gas Fast Reactor (GFR) is one of the six reactor concepts selected for further development in the frame of the Generation IV International Forum (GIF). Since no gas cooled fast reactor has ever been built, a small demonstration reactor is necessary on the road towards the full-scale GFR reactor. A concept of this demonstrator is called ALLEGRO.

The French Commissariat à l’énergie atomique et aux énergies alternatives (CEA) developed between 2001–2009 a pre-conceptual design of both the full-scale GFR called GFR2400 and the small demonstration unit called ALLEGRO (75 MWt). Since 2013 ALLEGRO has been under development by several partners from Czech Republic, France, Hungary, Poland and Slovakia.

No severe accident study of ALLEGRO using a dedicated computer code has been published so far. This paper is the first attempt to perform computer simulations of the ALLEGRO CEA 2009 concept, using MELCOR version 2.1.

A model of the ALLEGRO CEA 2009 concept has been developed with the aim to perform safety analyses; to confirm that MELCOR can be used for such a study, to investigate what scenarios lead to a severe accident and to study in detail the progression of the severe accident during the in-vessel phase.

Several pressurized and depressurized protected scenarios were investigated; four of them are presented in this paper. It was observed that even long-lasting station blackout (SBO) without further failures of the passive safety systems does not lead to a severe accident as long as there is enough water in the decay heat removal (DHR) system. Loss of coolant (LOCA) transients with DHR system in the forced-convection mode can lead to peak cladding temperatures causing limited core damage in the early phase of the accidents, but without further development into core meltdown. On the other hand, LOCA combined with SBO leads to excessive core melting in orders of minutes, which represents a weak point of ALLEGRO 2009 concept. Recommendations were formulated for the further development of the ALLEGRO concept.
Radiation Shielding Design of High Temperature Reactor Pebble-Bed Module (HTR-PM)
Authors: Sun Sida , Fang Sheng , Li Hong
Published: October 2016

Radiation safety is an important concern in the design and licensing of the 200MWe High Temperature Reactor Pebble-bed Module (HTR-PM) demonstration power plant in China. To meet the requirement of the regulatory, various radiation protection strategies and methods are applied in the design process of systems and components of HTR-PM. In this study, the radiation shielding design of HTR-PM is reviewed, which includes the radiation source analysis, in-house dose calculation tool, shielding and dose reduction methods used for primary systems. The underlying conservative assumption is also discussed for correctly evaluating the dose calculation result. This summary provides a relatively systematic review of the radiation shielding methods in the design phase of HTR-PM, which may provide useful information and experiences for the radiation shielding design of future pebble-bed reactors.
Effects of Temperature on Deposition Rate of Supersaturated Silicic Acid on Ca-Type Bentonite
Authors: Sasagawa Tsuyoshi , Chida Taiji , Niibori Yuichi
Published: October 2016

For the disposal system of high-level radioactive waste in Japan, Na-type bentonite is used as one of backfilling and buffer materials for preventing the migration of groundwater and radionuclide. However, the alteration to Ca-type bentonite will cause the degradation of the barrier performance. On the other hand, silicate minerals around the repository dissolve under the high alkaline condition of groundwater (about pH 13) altered by alkaline components leaching from cementitious materials used for the construction of the repository. Such high-concentration silicic acid becomes supersaturated with the decrease in pH by mixing with natural downstream groundwater (pH 8) because of the change in the solubility of silicic acid. So far, the authors have examined the deposition rates of supersaturated silicic acid on Ca-type bentonite under the condition of room temperature, showing the clogging effect of flow-paths with the deposition. However, the dynamic behaviors of silicic acid are much sensitive to temperature change. Therefore, the present study focuses on the effect of temperature on the deposition rate of silicic acid on Ca-type bentonite. As a result, in the range up to 323 K, the deposition of supersaturated silicic acid on Ca-type bentonite was promoted with the increase in temperature. This suggests that the deposition of silicic acid will clog the flow-paths in Ca-type bentonite in this temperature range.
Experimental Study on AlN/H2O and Al2O3/H2O Nanofluid Flow Boiling Heat Transfer and its Influence Factors in a Vertical Tube
Authors: Wang Y. , Deng K. H. , Liu B. , Wu J. M. , Su G. H.
Published: October 2016

In this work, AlN/H2O and Al2O3/H2O nanofluid was prepared by an ultrasonic oscillation. Moreover, saturated flow boiling heat transfer in a vertical tube is experimentally investigated by nanofluid, with 0.1% volume concentration and 30nm diameter for AlN and 20nm diameter for γAl2O3. Several factors are under consideration, including heat flux on the heating surface (48–289kW·m−2), pressure (0.2–0.8Mpa) and mass flow rate (350–1100 kg·m−2·s−1). The results show that the saturated flow boiling heat transfer of AlN/H2O nanofluid is improved mostly about 64% compared with deionized water and Al2O3/H2O nanofluid is improved mostly about 61% compared with deionized water, and the average Nusselt number enhancement rates of nanofluid compared with deionized water are 23% for Al2O3/H2O nanofluid and 31% for Al2O3/H2O nanofluid in the range of this work. Furthermore, the heat transfer capacity of nanofluid increases with increasing heat flux on the heating surface and pressure. It is proved that nanoparticle deposited on the heating surface by SEM observations and TEM observations for nanoparticle confirm that nanoparticle does not change obviously after boiling. In addition, the enhancement rate of nanofluid saturated flow boiling heat transfer capacity increases with increasing pressure, but the influence of mass flow rate is negligible. In this study, as the thermal conductivity of AlN is greater than that of Al2O3, the heat transfer capacity of AlN/H2O nanofluid is greater than that of Al2O3/H2O nanofluid under the same conditions.
Analysis on Safety System Operating Characteristic of SCWR
Authors: Liu Liang , Zhou Tao , Chen Jie , Muhammad Ali Shahzad , Chen Juan , Xia Bangyang
Published: October 2016

In this paper, operating characteristics of the safety system of Chinese Supercritical Water-cooled Reactor (CSR1000) is described. Selecting CSR1000 as the focus of research, and it’s active and passive safety systems are analyzed in turn. A comparison is given between these two types of safety systems. Henceforth, the features of the safety control systems of CSR1000 are obtained. The results show that for the active systems, the control speed of the pressure control system is the fastest and that of the power control system is the slowest. It is observed that the active control system exhibits simple harmonic oscillation. On the other hand, the control feature of passive control system is stable. In addition, coupling the safety systems can ensure the safety of CSR1000 in the event of a loss of flow accident (LOFA).
Activities of the Safety and Operation Project for the International Research and Development of the Sodium-Cooled Fast Reactor in the Generation IV International Forum
Authors: Sakai Takaaki , Ren Lixia , Tsige-Tamirat Haileyesus , Vasile Alfredo , Kang Seok-Hun , Ashurko Yury , Fanning Thomas
Published: October 2016

The Generation IV (GEN-IV) international forum is a framework for international cooperation in research and development (R&D) for the next generation of nuclear energy systems. Concerning the sodium-cooled fast reactor (SFR) system, there are five cooperation projects for R&D. The SFR Safety and Operation (SO) project addresses the area of the safety technology and the reactor operation technology developments. The aim of the SO project includes (1) analyses and experiments that support establishing safety approaches and validating performance of specific safety features, (2) development and verification of computational tools and validation of models employed in safety assessment and facility licensing, and (3) acquisition of reactor operation technology, as determined largely from experience and testing in operating SFR plants. The tasks in the SO topics are categorized into the following three work packages (WP): WP-SO-1 “Methods, Models and codes” is devoted to the development of tools for the evaluation of safety, WP-SO-2 “Experimental Programs and Operational Experiences” includes the operation, maintenance and testing experiences in experimental facilities and SFRs (e.g., Monju, Phenix, BN-600 and CEFR), and WP-SO-3 “Studies of Innovative Design and Safety Systems” relates to safety technologies for GEN-IV reactors such as passive safety systems. In this paper, recent activities in the SO project are described.
Experimental Study on Decontamination Factor of Methyl Iodide by Bubbling
Authors: Wen Jiming , Sun Zhongning , Gu Haifeng , Zhou Yanmin
Published: October 2016

Once a serious event occurs, plenty of gas should be discharged into environment, so it is in urgent need to filter radioactive material before the ventalition. Bubble columns which belong to a kind of wet filters have a potent ability of filtering methyl iodide. The efficency of filtration depends on interface area between liquid and gas, liquid mass transfer rate and retention time of bubbles. Furthermore the decontamination factor of methyl iodide increases with temperature, pressure and liquid level, but decreases with superficial velocity. The experimental condition in this paper is: maximum pressure 0.5 MPa and maximum temperature 150°C, superficial speed 0.2m/s, gas composed of air and steam, liquid level 0.1–2m. The relationship between DF and experiment condition is deep discussed.
The Early Depressurization Method for the Vapor Generation at the Top of Reactor Vessel
Authors: Tsuruoka Hokuto , Miyata Shinya , Maruyama Manabu , Kanamori Daisuke
Published: October 2016

Mitsubishi Heavy Industries (MHI) is developing improved reactor cooling process in accident conditions, by using steam generator (SG) on accident in order to enhance safety of operating plants.

In case of station black out (SBO), leak from the reactor coolant pump (RCP) seal should be terminated by depressurizing the primary system. In this process, operators use the main steam relief valves (MSRVs) to reduce primary coolant temperature and pressure. However, coolant temperature in vessel head and upper plenum tends to remain higher than reactor temperature, and the coolant in these regions eventually starts boiling because of depressurization. This phenomenon inhibits smooth decrease in primary coolant temperature and pressure by forming large void region in the primary system. MHI is investigating improved process to achieve smooth decrease of primary temperature and pressure. In the process, condensation of the void generated should be appropriately achieved.

In this study, experimental tests on void condensation in the vessel head were performed using Large Scale Test Facility (LSTF). In addition, applicability of M-RELAP5, a plant transient computer code used in safety analysis for Japanese PWRs, to actual plant behavior modeling was confirmed, in terms of void generation/condensation in vessel head in accidental condition.