bluuebunny/crossref_metadata_2025 huggingface dataset:
Effect of Al2O3 and SiC Content in Pack Cementation Powders on the Microstructure of SiC Coatings on HTR Graphite Spheres
Authors: Zhao Hongsheng , Zhou Ping , Li Ziqiang , Liu Xiaoxue , Zhang Kaihong , Wang Taowei , Wu Bin , Liu Bing
Published: October 2018

To improve the anti-oxidation ability of graphite, Al2O3 and SiC as additives, a uniform SiC coating was prepared on the surface of HTR graphite spheres by the pack cementation and sintering process. The microstructure of SiC coating was characterized by XRD, Raman spectroscopy, scanning electron microscopy, and the effect of Si/C ratio, Al2O3 and SiC content in the starting powder, sintering process conditions on the microstructure of SiC coating was also analyzed. The results show that the SiC coating on the surface of spherical graphite sample can be obtained with Si/C ratio higher than 3: 1, and the SiC coatings with different raw materials possess different microstructure and phase constitutes. The SiC coating is a kind of clear porous microstructure, while the main composition of the coating is β-SiC with some α-SiC and unreacted graphite. It is found that the content of Al2O3 has a significant influence on the penetration into the SiC coating, while the SiC content has a great influence on the microstructure between SiC coating and the substrate. When Al2O3 content is 10%, the bonding between the SiC coating and the substrate is better. When SiC content is 20%, the SiC coatings obtained have good densification and less lamellar accumulation of on the surface of the coatings. The SiC coatings sintered in Ar gas at 1750 °C have larger thickness and porous porosity microstructure, and the maximum thickness, about 400 ∼ 600 μm, can be achieved with the 2h of sintering time. The obtained SiC coatings sintered in vacuum at 1750 °C have a plate-like dense microstructure, in which the main composition of the coating is free Si, α-SiC and β-SiC, and the transitional area between SiC coating and the substrate is not obvious.
Modular Boiling Water Reactor Concept
Authors: Yi Liao , Cong Wang , Lei Chen
Published: October 2018

Modular boiling water reactor (MBWR) can be considered as a small sized economic simplified boiling water reactor (ESBWR). It has the advantage of easier fabrication, transportation and construction. In this paper, a 65MWe MBWR core was designed with natural circulation, passive safety features, high power density and an 18 months fuel cycle. The MBWR core consists of 104 fuel assemblies with 4.6 w/o U-235, the assemblies were divided into 3 batches based on the depletion level, the batches shuffled at the end of each cycle. The core converged to equilibrium after 8 fuel cycles. A steady-state equilibrium fuel cycle depletion analysis was performed over a 540 day cycle using the HELIOS and PARCS software. The control blades insertion patterns were chosen to minimize axial and radial power peaking and provide uniform burnup throughout the cycle. At the end of the equilibrium cycle, 16% of total control blade worth remained inserted and the average assembly burnup is 21.318 GWd/MTHM. Thermal hydraulic analysis was also performed to insure the core’s safety feature.
Characterization and Experimental Investigation for the Dynamic Performance of the Hydraulically Suspended Passive Shutdown System in China Sodium-Cooled Fast Reactor
Authors: Song Jian , Wu Yingwei , Tian Wenxi , Qiu Suizheng , Su Guanghui
Published: October 2018

In order to enhance the inherent safety of sodium-cooled fast reactors, innovative hydraulically suspended absorber rod (HSR) passive shut-down system have been proposed for China demonstration fast reactor. In this study, based on the functional and performance requirements, a full-scale experimental setup has been designed and fabricated for the analysis of the HSR as applied to the prototype reactor. The main characteristic of the test facility is the actuation of the mobile safety rod is triggered by coolant flow rate decrease in the primary loop below half the nominal value and then the rod inserts into the stationary sleeve by gravity. The objective is to investigate the dynamic performance of HSR and establish the laws of its movement at lowering the flow rate modeling the coastdown of primary circulating pump. A series of tests have been performed, including start-up, steady-state operation, loss of flow accident, sensitivity analysis and reliability test. This study also focused on the effect of various factors on scram time, the effect of pump coasting time, rod weight, gap between rod and guide tube, bypass holes, cone angle of rod, flow rate and fluid temperature are analyzed. The experimental results demonstrate the functionality and reliability of the HSR, which would lay foundation for further optimization design.
High Efficient Detritiation Catalysts for Fusion Safety
Authors: Wu Quanwen , Luo Wenhua , Meng Daqiao , Bao Jinchun , Ba Jingwen
Published: October 2018

Tritium is indispensable to the fusion reactor engineering, and it must be seriously defended because of its radioactivity and permeability. The method of catalytic oxidation and absorption is the most widely used process for tritium cleanup so far, in which detritiation catalyst is of great importance. The poor stability caused by the agglomeration of noble metal limits the life of detritiation catalysts. Here, Anti–Ostwald Ripening is used to prepare single-atom detritiation catalysts S-Pt/Ce0.7Zr0.3O2 for tritium (HT, DT and T2) oxidation. Single-atom dispersed Pt ensures the catalytic activity and decreased the economic cost. The strong metal-support interaction (SMSI) keeps Pt from aggregating, thus increases the working life of catalyst. And Pd based catalyst supported by a cation ordered κ-Ce2Zr2O8 is prepared for tritiated methane (CH4-xTx) oxidation. Tritiated methane is mostly oxidized by Pd/κ-Ce2Zr2O8 at about 450 °C, which is at least 50 °C lower than normal catalysts (such as Pd/Al2O3).
Post-Test CFD Analysis of Non-Uniformly Heated 19-Pin Fuel Bundle Cooled by HLM
Authors: Marinari R. , Di Piazza I. , Angelucci M. , Martelli D.
Published: October 2018

In the context of the studies on GEN. IV/ADS nuclear systems, the correct evaluations of the temperature distribution in the fuel pin bundle is of central interest. In particular, the use of lead or lead-bismuth eutectic (LBE) as coolant for the new generation fast reactors is one of the most promising choices. Due to the high density and high conductivity of lead or LBE, a detailed analysis of the thermo-fluid dynamic behavior of the heavy liquid metal (HLM) inside the sub-channels of a fuel rod bundle is necessary in order to support the Front-End Engineering Design (FEED) of GEN. IV/ADS prototypes and demonstrators. In this frame, the synergy between numerical analysis by CFD and data coming from large experimental facilities seems to be crucial to assess the feasibility of the components.

At ENEA-Brasimone R.C., large experimental facilities exist to study HLM free, forced and mixed convection in loops and pools: e.g. NACIE-UP is a large scale LBE loop for mixed convection experiments. The MYRRHA-19 like Fuel Pin Bundle Simulator installed in the NACIE-UP facility allows to make non-uniform and dissymmetric tests with only a few pins heated. This technical feature of the FPS is very interesting for CFD validation and this kind of data tests in HLM fuel bundles are not so common in the literature.

In the present paper, a post-test validation is made by a detailed CFD model of the test section. Experimental data, statistically treated by the error propagation theory, are briefly presented and a preliminary comparison with CFD results using different models/turbulent Prandtl numbers are shown. Three monitored section at different levels are compared both for wall and bulk temperatures. This post-test comparison with this experimental configuration is unique and represents a further step towards the validation of the CFD models and methods in fuel bundle geometries cooled by HLM.
Geometry Survey on the Convex Shaped Core for Recriticality Prevention Against CDA in Sodium-Cooled Fast Reactor
Authors: Chitose Keiko , Tachi Yoshiaki , Wakabayashi Toshio , Takaki Naoyuki
Published: October 2018

In sodium-cooled fast reactors, the core is not arranged in its most reactive configuration. In this case, when the fuel melts to form a molten pool, the recriticality may occur by positive reactivity insertion due to core compaction. To prevent such recriticality, special devices of the fuel subassembly structure for discharging the molten fuel from the core region, have been investigated by the Japan Atomic Energy Agency (JAEA). On the other hand, the inherent feature of core geometry and the neutron characteristics may provide the similar effect to prevent such recriticality. The purpose of this study is to design the core specification its deformation in CDA causes negative feedback to subcritical condition, without any fuel discharge device.

The convex shaped core has the longer fuel length in the inner-core region and the shorter fuel in the outer-core region. Therefore, the core geometry as intact status has a lower neutron leakage effect. When the fuel melts in CDA, the core height is compacted and negative reactivity insertion is expected during molten pool formation. The convex shaped core is based on the large-scale cylindrical homogeneous core (3,600 MWth, 4.95m in core diameter, and 0.75m in core height). The calculation showed that the compaction of cylindrical core leads to a reactivity gain, whereas the convex shaped core results in negative reactivity effect.

In this geometry, both inner-core and outer-core are divided into two regions. Furthermore, we introduced the smaller diameter pin for inner-core and keep uniform Pu enrichment for all regions. The smaller diameter pins in high importance region are effective for flat-distribution. Through pin diameter survey, we confirmed the advantages of smaller diameter pin, such as reducing pressure loss of core coolant and decreasing the height of molten pool.
Performance Analysis of Generation IV Nuclear Reactor Power Plant Using CO2 and N2: Case Study of a Recuperated Brayton Gas Turbine Cycle
Authors: Osigwe Emmanuel O. , Gad-Briggs Arnold , Nikolaidis Theoklis , Pilidis Pericles , Sampath Suresh
Published: October 2018

With renewed interest in nuclear power to meet the world’s future energy demand, the Generation IV nuclear reactors are the next step in the deployment of nuclear power generation. However, for the potentials of these nuclear reactor designs to be fully realized, its suitability, when coupled with different configurations of closed-cycle gas turbine power conversion systems, have to be explored and performance compared for various possible working fluids over a range of operating pressures and temperatures. The purpose of this paper is to carry out performance analysis at the design and off-design conditions for a Generation IV nuclear-powered reactor in combination with a recuperated closed-cycle gas turbine and comparing the influence of carbon dioxide and nitrogen as working fluid in the cycle. This analysis is demonstrated in GTACYSS; a performance and preliminary design code developed by the authors for closed-cycle gas turbine simulations. The results obtained shows that the choice of working fluid controls the range of cycle operating pressures, temperatures and overall performance of the power plant due to the thermodynamic and heat properties of the fluids. The performance results favored the nitrogen working fluid over CO2 due to the behavior CO2 below its critical conditions.
The Quality Experience Feedback in Nuclear Fuel Manufacture
Authors: MengYao Tong , FangGang Li
Published: October 2018

Experience feedback refers to the timely information collection, transformation, analysis, processing and summary, when some good experience or occasional problems appeared during the manufacturing process. In the manufacturing process of nuclear fuel, CNNC JianZhong Nuclear Fuel Assembly Co., Ltd (CJNF) established a comprehensive experience feedback system, and consolidated experience feedback processing flow. Using classification and gradation to collect, filter, organize and use information. CJNF feeds back some quality problems in fuel assembly manufacture, through the analysis of causes and the implementation of measures to avoid the occurrence of similar problems. Meantime, feeding back and sharing good practical experience in manufacturing and management process, it is benefit to pass on experience and learn from each other. The experience feedback of CJNF is the solid foundation of quality management system’s operation and improvement.
Experimental Study of Aerosol Behavior During Pool Scrubbing: Part 1 — Visualization Measurement of Aerosol Particle in a Single Rising Bubble
Authors: Fujiwara Kota , Kikuchi Wataru , Nakamura Yuki , Saito Shimpei , Yuasa Tomohisa , Kaneko Akiko , Abe Yutaka
Published: October 2018

As pool scrubbing plays an important role in fission product (FP) decontamination, a reliable model is needed. Despite the needs, mechanism of FP transfer from air-water from the swarm is not explained exactly which means that the evaluation of physical model used in pre-existing model couldn’t be done enough.

Existing model for pool scrubbing is predicted in the MELCOR code. Inside the code, a simple model of bubbly jet divided in 3 regions is used: 1) Globule region where the gas including FP enter and collapse, 2) Swarm rise region where the bubble rises up stationary after the collapse is done and, 3) Entrainment region where the bubble pop out to the atmosphere. In each region, the decontamination factor (DF), the particle density ratio before and after each region, is calculated. On these region, flow and physical force inside the gas phase is predicted to be one of the driving force which cause the FP transfer.

Therefore, our study aims at the particle behavior on the gas phase. As to understand the physical phenomenon individually, the study focuses on flow behavior and inner flow of a single rising bubble. As an approach, comparison of bubble containing aerosol and no aerosol has been done for each parameter of size, aspect ratio, velocity. Compared with existing equations, the rising speed of clean bubble condition and aspect ratio of CSI condition agreed well to the MELCOR code [1]. On the other hand, many difference were also measured in other condition.

Application of parameters obtained from experiment were done against the MELCOR model. Calculation of velocity inside the oil droplet using the experimental parameters obtained from visualization measurement was done. The local gravitational sedimentation and centrifugal velocity took a higher value in clean bubble and OX50 condition compared to CSI condition. On the other hand, Brownian diffusion velocity had an opposite trend. PIV measurement were performed by a silicone oil to visualize the inner flow clearly and compared with the calculation. Seen from the results, the local diffusion velocity took a lower value compared to the calculation using the MELCOR model.
Fuel Cycle Economy of Accident Tolerant Fuel Assemblies
Authors: Duoting Xu , Tong Liu , Heng Huang
Published: October 2018

Taking the large commercial pressurized water reactor and its mature fuel assembly as reference, this paper has analyzed economic performance of two accident tolerant fuel (ATF) designs based on once-through fuel cycle. The results show that the fuel cycle costs of both AT F designs have grown due to application of BeO powder, which is expensive. In order to reach the same electric cost as that of the referred fuel assembly, burn-up of these two AT F designs should be enhanced to 51323MWd/tU and 52054MWd/tU respectively.
Blind Simulations of NACIE-UP Experimental Tests by STH Codes
Authors: Forgione Nicola , Angelucci Morena , Barone Gianluca , Polidori Massimiliano , Cervone Antonio , Di Piazza Ivan , Giannetti Fabio , Lorusso Pierdomenico , Hollands Thorsten , Papukchiev Angel
Published: October 2018

In the frame of the SESAME project, a benchmarking activity was proposed to validate the existing system thermal-hydraulics codes for Heavy Liquid Metal reactors. More specifically, blind simulations on three well-defined experiments were carried out on the NACIE-UP facility, using CATHARE by ENEA, ATHLET by GRS, RELAP5-3D by University of Roma and RELAP5/Mod3.3 by University of Pisa. The numerical models were calibrated in terms of system thermal losses and gas enhanced circulation by means of the outcomes from specific experimental preliminary tests.

The present discussion expose, compare and analyze the numerical results of some representative parameters (primary lead-bismuth eutectic (LBE) mass flow rate, temperatures and pressure) charaterizing the system behaviour in transiet scenarios in a “pre-test” blind numerical assessment.
Facilities and Experience on Impact Test of Packages for Radioactive Materials Transport
Authors: Guoqiang Li , Dajie Zhuang , Xuexin Wang , Dongyuan Meng , Jiangang Zhang , Hongchao Sun , Shutang Sun , Anping Ma
Published: October 2018

According to the requirements of Chinese GB11806 and IAEA’s SSR-6, the packages for radioactive materials transport shall be tested to demonstrate the ability to withstand transport accidents. A small facility and a large one for package impact test have been designed and constructed to conduct the free drop test I experiment required in the GB11806 and SSR-6. The target of the smaller facility is a reinforced concrete component with volume of 3800mm × 3550mm × 3500mm. It is riveted by steel plate marked Q460 with volume of 3500mm × 3250mm × 100mm at surface. Total mass of the target is about 130t. The hoisting facility is mobile crane. The facility is used for the impact test of package within 13t. The target of the larger facility is a reinforced concrete component with volume of 11000mm × 7350mm × 7000mm. Steel plate marked Q460 with a volume of 11000mm × 5380mm × 100mm was anchored at concrete component top. Total mass of the target is about 1400t. The maximum height of the lifting hook in tower is 17.1m. The larger facility is used for the impact test of package within 130t. The two facilities for package impact test are both equipped with acceleration and stress measuring system, image measuring instrument, and geometrical measuring system. These two facilities have been used to conduct a lot of impact test of small packages weighing less than 12t. Some experience in construction and experiment has been concluded.
Multiphysics Analysis of CMC Silicon Carbide and Zircaloy Cladding
Authors: Altahhan Muhammad , McFerran Noah , Morrell Jonathan , Avramova Maria
Published: October 2018

Nuclear fuel cladding is an integral part of nuclear reactors and choosing the proper material is imperative to the design of a reactor. In this paper, the neutronic properties and the fuel performance of a 17 × 17 Westinghouse Pressurized Water Reactor (PWR) assembly using ceramic matrix composites (CMC) Silicone Carbide (SiC) as a cladding material is investigated. The material analysis is compared against traditional Zircaloy-4 cladding used in a 17 × 17 Westinghouse PWR assembly. The codes used in the analysis are the Michigan Parallel Characteristics based Transport (MPACT) code coupled with CTF, the North Carolina State University version of the Coolant Boiling in Rod Arrays Two Fluids (COBRA-TF) code, and the fuel performance code BISON as well as the uncertainty analysis code DAKOTA. Additionally, annular geometry for the fuel pellet is modeled to assess its merit compared to ordinary CMC SiC or traditional Zircaloy-4 claddings. It is found that on the neutronics side, the CMC SiC shows lower achievable U-235 enrichments required to reach the same burnup and effective neutron multiplication factor as Zircalloy-4 claddings. These results are an advantage that can be seen in the economic cost analysis done and additionally from the reactor operation point of view. Also, it is found that the different criteria of safe operation of Westinghouse PWR assemblies like the plenum pressure, the fuel-cladding contact pressure, the peak fuel temperature, and the fission gas release criteria are all achieved with CMC SiC with some criteria having larger design margins than of the Zircaloy-4 cladding. Furthermore, a critical heat flux (CHF) study shows that CMC SiC has even larger thermal margins than the ordinary Zircaloy-4 cladding, resulting in a more profitable fuel cycle due to the greater amount of power that the fuel pins can be operated at. An uncertainty quantification for the CHF Ratio (CHFR) is also done to assess the largest magnitudes of importance that affect the CHFR calculated.
Preliminary Neutron Simulation of Ceramic Fast Reactor
Authors: Yan Xuesong , Zhang Xunchao , Zhang Yaling , Yang Lei , Duan Wenshan
Published: October 2018

In this paper, preliminary neutron physical properties of ceramic fast reactor (CFR) are simulated and analyzed. The CFR core consists of ceramic materials, including nuclear fuels, coolants, structural materials, reflective and absorption materials. These ceramics improve inherent safety levels substantially, increase breeding performance, and enhance the power-generation efficiency. The CFR has the potential to operate and breed more than 30 years. The performance of the CFR was simulated focusing on neutron-related effects. The parameters discussed contain fast neutron energy spectrum, the ideal effective multiplication-factor, nuclides mass changes, breeding performance, operation mode, etc. Furthermore, the strengths of the proposed reactor system are discussed. In the future nuclear energy system, CFR may be one of the existing alternative novel reactor type.
Subcooled Flow Boiling Inception and Heat Transfer of Water in a Circular Tube Under Pulsatile Flow
Authors: Yuan Hongsheng , Tan Sichao , Cheng Kun , Wu Xiaoli , Guo Chao , Zhong Mingjun
Published: October 2018

The flow rate can fluctuate in offshore nuclear power systems which are exposed to wind and waves, as well as in loops where flow instabilities occur, resulting in different thermal-hydraulic characteristics compared with that under steady flow. Among the thermal-hydraulic characteristics, onset of nucleate boiling (ONB) model determines whether the fluid is boiling, and boiling heat transfer is crucial to equipment performance and safety, both being key issues in subcooled flow boiling. Therefore, an experimental study was conducted to investigate how an imposed periodic flow oscillation affects the boiling inception and heat transfer of subcooled flow boiling of water in a vertical tube. The experiments were conducted under atmospheric pressure with the average flow rate ranging from 96kg/m2s to 287kg/m2s and heat flux ranging from 10kW/m2 to 197kW/m2. The relative pulsatile amplitude range is 0.1–0.3 and pulsatile period range is 10s-30s. Photographic images and thermal parameters such as temperatures and flow rate were recorded. The lack of nucleation site on the heated surface of the test section results in high wall superheat at ONB. The effects of pulsatile amplitude and period on superheat at boiling onset and average heat transfer were analyzed. The results show that the superheat at boiling inception is decreased when the average heat flux is lower than the heat flux at boiling inception of the corresponding steady flow, and the superheat at boiling onset is increased when the average heat flux is higher than the heat flux at boiling onset of the corresponding steady flow. The above effect of flow rate pulsation on superheat increases with increasing amplitude and decreasing period, and the mechanism can be explained by boiling nucleation theory. The lack of large active nucleation site also affects the boiling heat transfer. By comparing the contribution of nucleate boiling to heat transfer with the widely used Cooper’s pool boiling correlation, the subcooled flow boiling was found suppressed by convection. The average heat transfer of both the intermittent flow boiling and the single phase flow is influenced by flow oscillation.
Computation of Solid Melting With Wall Boiling Under External Reactor Vessel Cooling
Authors: Song Hyuk-Jin , Park Jong Woon
Published: October 2018

The solid wall convection melting physics is computed with coupling external boiling physics based on the volume-of-fluid model in the STAR-CCM+ code. A three-dimensional rectangular computational sample domain is used with 5.0 cm height, 4.0 cm width and 1.0 cm depth. This domain is separated by carbon steel representing a reactor vessel section in contact with water channel representing the gap between the reactor vessel outer surface and the insulation. The flow rate of the bottom inlet flow in the water flow channel is 2 × 10−2 kg/sec with temperature at 92 °C. The concentrated heat flux of 2.1 MW/m2 obtained from a typical APR1400 reactor condition is imposed at the center of the left wall of the solid region. At the same time, subcooled water entering the right channel boils and the vapor film covers the surface which expedites melting process in the solid heated region. Important finding is that solid wall melt-through is not expected as far as for the present analysis conditions and there is a minimum water flow rate below which the rate of solid melting is saturated.
Study on Flow Induced Vibration Analysis and Evaluation for Heat Transfer Tube of Steam Generator in Two Phase Flow
Authors: Xuan Huang , Huan-huan Qi , Feng-chun Cai , Zhi-peng Feng , Shuai Liu , Qian Huang
Published: October 2018

The flow induced vibration of heat transfer tube is one of the main factors leading to the failure of heat transfer tube in steam generator, the flow induced vibration of steam generator should be analyzed and evaluated in the design of steam generation. The flow induced vibration of heat transfer tube in two phase flow is the difficult and important content in the analysis. In this paper, the finite element model of heat transfer tube is established and the modal analysis is carried out. Then in order to evaluate the influence of two phase flow in the secondary side and support boundary constraint, the analytical results are compared with the natural frequency of the heat transfer tube measured in the two phase flow test. On the basis of accurate simulation of the dynamic characteristics of heat transfer tube in two phase flow, the paper calculate the turbulent excitation response and the fluid-elastic instability ratio aiming at the main mechanism causing the flow induced vibration of heat transfer tube in two phase flow. The modified PSD of turbulent excitation is proposed on the foundation of root mean square displacement amplitude of heat transfer tube measured in two phase flow test. The calculation result of the amplitude of heat transfer tube with different void fraction can envelope the test result by using the modified PSD as input, and the safety margin is reasonable. We also verify whether the analysis conclusion of fluid-elastic instability is in agreement with the test. The paper studies on flow induced vibration analysis and evaluation for heat transfer tube of steam generator in two phase flow, modifies the analysis program of flow induced vibration on the basis of the test results, the investigation can be used at the risk prediction and evaluation of flow induced vibration of heat transfer tube in two phase flow, solve the technical difficulties of flow induced vibration analysis in two phase flow, provide the technical support for the flow induced vibration analysis of steam generator in Hualong one.
CFD Analysis on Wall Boiling Model During Subcooled Boiling in Vertical Narrow Rectangular Channel
Authors: Ren Tingting , Yan Changqi , Yan Meiyue , Yu Shengzhi
Published: October 2018

Two-fluid model is a common method to simulate the subcooled flow boiling heat transfer, in which the wall boiling model is mainly used for the partition of wall heat flux and the mass transfer between two phases on the wall. The model determines the amount of vapor phase and predicts the cross-sectional void fraction in the channel, nucleate site density and bubble departure diameter play an important role in the accurate prediction of wall boiling model. Eulerian two-fluid model coupled with Rensselaer Polytechnic Institute (RPI) wall boiling model is employed to simulate the heat transfer characteristics and boiling phenomena in vertical narrow rectangular channels by using FLUENT code. Based on the experimental data of subcooled boiling in vertical narrow rectangular channel, different combinations of nucleate site density and bubble departure diameter correlations are used to calculate under different conditions of heat flux and inlet subcooling. Comparing the calculated heat transfer coefficients along the vertical height with experimental results, it can be found that these two parameters have a significant effect on the subcooled boiling heat transfer in narrow rectangular channels. Different parameter combinations lead to differences in wall heat flux distribution, different heat flux and inlet subcooling also have different effects on these models, which eventually lead to different evaporative heat flux, thus affecting the prediction of void fraction.
Experimental Study of Quench Temperature During Reflood Phase
Authors: Wang Jinyu , Wang Jun , Zan Yuanfeng , Huang Yanping
Published: October 2018

Reflood process plays an important role in protecting fuel rods from damage and mitigating consequence during a hypothetical loss of coolant accident (LOCA). Study of quench temperature which is a symbol of quenching process is important for deep understanding of quench mechanism and development of quench models. However, there is no consensus on effects of different parameters such as inlet mass flux on quench temperature. More experimental studies on quench temperature are still needed. Based on an experimental study in an annular channel, the influences of inlet mass flux (20∼300kg/m2•s) and power (0∼1.5kW/m) on quench temperature have been investigated. It is shown that the effect of inlet mass flux on quench temperature is related to power. Quench temperature increases with increasing inlet mass flux when power is low enough but decreases with increasing mass flux when power is high.
Development and Basic Verification of Decay Heat Removal Analysis Code of Sodium-Cooled Fast Reactor
Authors: Song Ping , Zhang Dalin , Feng Tangtao , Wang Shibao , Zhang Yapei , Wang Mingjun , Qiu Suizheng , Su G. H.
Published: October 2018

As one of the generation IV reactors, pool-type Sodium-cooled Fast Reactors (SFRs) is attracting more and more attention. Loss of flow and heat sink accident is one of the most serious accidents for SFRs. Therefore, the decay heat removal capacity after emergency shutdown is of great importance. This paper has developed a one-dimensional code named Decay heat Removal Analysis Code for Sodium-cooled Fast Reactor (DRAC-SFR) to analyze the decay heat removal capacity. In the code, the decay heat removal system contains the primary loop, the intermediate loop and air circuit. The decay heat is removed out step by step with the above three loops. Many studies have been conducted on code verification. The international benchmark analysis of EBRII reactor is applied in the code verification. The calculation is compared with the experimental data and the results of DRAC-SFR agreed well with the experimental data. The comparison with the steady state of China Experimental Fast Reactor (CEFR) shows a good agreement with the design value. The errors of all the compared parameters are within 2%. What’s more, calculation is performed to analyze the characteristics of the decay heat removal capacity for CEFR. Thus, code verification shows that DRAC-SFR is proper to evaluate the decay heat removal capacity for SFRs and has the ability to provide references and technical supports for the design and optimization of the pool-type sodium-cooled fast reactor.
Assessment of Radiological Source Term Releases for Potential Severe Accident Scenarios in a PWR SFP
Authors: Ahn Kwang-Il , Shin Jae-Uk
Published: October 2018

The primary purpose of this study is to assess the release of source terms into the environment for representative spent fuel pool (SFP) severe accident scenarios in a reference pressurized water reactor (PWR). For this, two typical accident scenarios (loss-of-cooling and loss-of-pool-inventory accidents) and two different reactor operating modes (normal and refueling modes) are considered in the analysis. The secondary purpose of this study is to assess the impact of an emergency makeup water injection strategy, which is one of representative SFP severe accident mitigation (SAM) strategies being employed after the Fukushima accident, upon the release of the radiological source terms. A total of 16 cases, consisting of four base cases and three injection cases for each base case were simulated using the MELCOR1.8.6 SFP version. The, analysis results are given in terms of (a) the key thermal-hydraulic behaviors during an accident progression and (b) releases of radiological fission products (such as Cesium and Iodine) into the environment. In terms of a release of Cesium and Iodine into the environment, the present study show that the two cases subject to a loss of pool inventory (i.e., LOPI-N-03 and LOPI-R-00) lead to the worst results with the respective release fractions of 77.5% and 59.4%.
Experimental Investigation on Dependence of Decontamination Factor on Aerosol Number Concentration in Pool Scrubbing Under Normal Temperature and Pressure
Authors: Sun HaoMin , Machida Shinichi , Sibamoto Yasuteru , Okagaki Yuria , Yonomoto Taisuke
Published: October 2018

During a severe accident of a nuclear reactor, radioactive aerosols may be released from degraded nuclear fuels. Pool scrubbing is one of the efficient filters with a high aerosol removal efficiency, in other words a high decontamination factor (DF). Because of its high performance, many pool scrubbing experiments have been performed and several pool scrubbing models have been proposed. In the existing pool scrubbing experiments, an experimental condition of aerosol number concentration was seldom taken into account. It is probably because DF is assumed to be independent of aerosol number concentration, at least, in the concentration where aerosol coagulation is limited. The existing pool scrubbing models also follow this assumption.

In order to verify this assumption, we performed a pool scrubbing experiment with different aerosol number concentrations under the same boundary conditions. The test section is a transparent polycarbonate pipe with an inner diameter of 0.2 m. 0.5 μm SiO2 particles were used as aerosols. As a result, DF was increasing as decreasing the aerosol number concentration. In order to ensure a reliability of this result, three validation tests were performed with meticulous care. According to the results of these validation tests, it was indicated that DF dependence on the aerosol concentration was not because of our experimental system error including measurement instruments but a real phenomenon of the pool scrubbing.
Verification of Shielding Calculation Capability of RMC With H.B.Robinson-2 Pressure Vessel Benchmark
Authors: Rao Junjie , Shang Xiaotong , Wang Kan
Published: October 2018

RMC is a 3-D continuous energy Monte Carlo code developed by REAL team in Tsinghua University, China. Besides the capability of fuel cycle burnup calculation, hybrid MPI/OpenMP parallelism strategy, sensitivity and uncertainty analysis, N-TH coupling calculation, shielding calculation methods including general source description, regional importance method, weight window method and source biasing method have been also developed for deep penetration problems. H.B.Robinson-2 Pressure Vessel Benchmark (HBR-2 benchmark) is used for the qualification of pressure vessel neutron flux calculation methods and shielding calculations based on this model have been performed by Monte Carlo codes such as SCALE, MCNPX and deterministic transport code DORT. In this work, the verification calculation of shielding calculation capability of RMC is conducted based on HBR-2 benchmark. The total calculation consists of two stages. Criticality calculation is performed first to obtain the fission neutron distribution in the reactor core assemblies. Then the fission neutron distribution is regarded as the initial neutron source in the following fixed source calculation. Variance reduction techniques such as source biasing and regional importance methods are combined together to be able to reduce the variance of the neutron flux in regions within and outside the pressure vessel including the downcomer and cavity regions. The preliminary calculation results show good agreement with MCNP and the shielding calculation of RMC is justified and applicable for deep penetration problems.
Detailed Experimental and Analytical Study on Long Two-Phase Closed Thermosiphons Related to Passive Spent-Fuel Pool Cooling
Authors: Graß Claudia , Krüssenberg Anne , Kulenovic Rudi , Weyermann Fabian , Starflinger Jörg , Schaffrath Andreas
Published: October 2018

New concepts are currently being discussed for passive residual-heat removal with heat pipes from spent-fuel pools and wet-storage facilities. Because of their high heat-transport capability and their simple design, two-phase closed thermosiphons have a great potential to satisfy the demands of a reliable and independent passive heat removal. The geometry of spent-fuel pools and the potential incorporation into existing plants requires thermosiphons of at least 10 m in length including bends. Such thermosiphons are neither available nor have they been investigated yet. Therefore, experimental and numerical investigations are being carried out.

At IKE the basic operational behavior of 10-m-long thermosiphons with water — as working fluid — are being experimentally investigated. Measurements for different pipe diameters (32 mm and 45 mm) are performed at various heat sink temperatures (10 °C, 20 °C and 30 °C), heat inputs (1000 W to 4000 W), and filling ratios (50%, 70% and 100%).

GRS is developing codes, such as AC2, in order to simulate all relevant phenomena within a nuclear power plant during normal operation, incidents, accidents, and severe accidents. Regarding passive residual-heat removal with thermosiphons, the models of AC2 are being improved to properly simulate the thermos-hydraulics of this heat transfer process. Starting with the module ATHLET (Analysis of Thermal-hydraulic of Leaks and Transients), the applicability of its existing models is checked for modeling long thermosiphons and calculating their operational behavior. The main model improvements are being validated against the new experiments of IKE.
Electrochemical Measurement of Radio-Activated Metal Under High Temperature Condition
Authors: Taguchi Ryota , Ihara Tomonori , Hazuku Tatsuya , Takamasa Tomoji , Kano Sho , Abe Hiroaki
Published: October 2018

Fundamental knowledge of electrochemical characteristics of metals under high temperature and radiation conditions is important in development of reactor core materials for light water reactors including a supercritical water-cooled reactor (SCWR). In the past studies, however, the majority of electrochemical measurements of the metals have been conducted at temperatures less than 290 °C and under a non-radiation condition. On the other hand, the radiation induced surface activation (RISA) which enhances of wettability and anticorrosive effect on metal surfaces was first revealed by authors in 1999. This study is aimed to confirm the RISA effect experimentally under high temperature conditions. The electrochemical characteristics measurement of a specimen immersed in pure water in an autoclave were performed at temperatures up to 300 °C. A normal 304 stainless-steel and a radioactivated 304 stainless-steel were used as the specimen. The radioactivated specimen was made by 140 MeV irradiation of C4+ ion beam at a cyclotron accelerator of Tohoku University in Japan. The experimental results showed that the corrosion potential of specimen was lowered due to radioactivation.
A PSA Case Study: Promoting the Reliability of the CRSS on the CMRR Through the ATWS Mitigation System
Authors: Yu Heng , Wang Guan-bo , Qian Da-zhi , Guo Yu-chuan , Hu Bo
Published: October 2018

An increasing number of PSA programs concerning research reactors have been launched across the world. As with many other reactors, the CMRR (China Mianyang Research Reactor), a typical pool-type research reactor, regards the control rod shutdown system (CRSS) as its primary shutdown system which enables the reactor subcritical by dropping control rods into the core after a specific initiating event is detected. As a result, the CRSS is an essential ingredient of engineered safety features. It is necessary to enhance the reliability of the CRSS, ensuring the reactor can be successfully shut down when the ATWS — the anticipated transients without scram occurs. Therefore, additional facilities should be designed to cope with the extremely severe circumstance. Accordingly, the purpose of this paper is to evaluate the promotion of the CMRR’s safety degree and the reliability of its CRSS from the PSA’s perspective with an ATWS mitigation system installed. Results indicate that, by introducing the ATWS mitigation system, the failure probability of the CRSS can decrease from 1.52e−05 per demand to 3.35e−06 per demand, while the aggregate CDF (core damage frequency) induced by all IE (initiating event) groups, is able to decrease to a relatively low value 1.17e−05/y from its previous value 3.11e−06/y. It is apparent that the reliability of the CRSS as well as the safety degree of the overall reactor can be enhanced effectively by adding the ATWS mitigation system to the elementary design of the normal CRSS.
Characteristic Tests on Transition Core of HTR-10
Authors: Wei Liqiang , Ding Dongmei , Liu Ling , Wang Yucheng , Chen Xiaoming , Xie Feng
Published: October 2018

After a long-term shutdown, the 10MW high temperature gas-cooled test reactor (HTR-10) was restarted, and the operation & safety characteristics of the HTR-10 transition core are tested and verified. A series of the characteristic tests have been implemented, such as the value calibrating test of the control rod and boron absorber ball, the disturbance characteristic of helium circulator, the start-stop characteristic and the stable power operation characteristic, which indicated the characteristics of the reactor transition core meet the design and safety requirements.
External Hazard Coinciding With Small Break LOCA Thermohydraulic Calculation With System Code ATHLET
Authors: Kosowski Kai , Seidl Marcus
Published: October 2018

The safety behaviors of a NPP after an external hazard-initiated event, as well as after a small break LOCA are already well known as part of the analyses done for standard licensing. In this paper, a combination of both event categories is investigated by means of the thermohydraulic system code ATHLET (analysis of the thermohydraulics of leaks and transients) developed by the Gesellschaft für Anlagen- und Reaktorsicherheit (GRS, Germany). The scenario assumes an external event with a loss of coolant accident caused by induced vibrations on a small pipe attached to the primary circuit, although all pipes are designed to withstand the loads created by such an external event. The coincidence of both events leads to a beyond-design basis consideration. Furthermore, in the context of both robustness and enveloping analyses, the unavailability of both the internal power supply via the external grid connection and the emergency diesel power supply due to the external impact is postulated. For emergency cooling, the NPP in the scenario considered has only access to the passive accumulators and to systems supplied via the safeguard emergency diesel engines (2nd quartet of emergency diesel engines) — a design feature of the Siemens KWU type PWRs —, which are housed in the bunkered emergency feed building and operate with lower voltage levels.

The scenario is exemplarily modeled and simulated with ATHLET. The manner of the external event itself is not in focus, but rather the thermohydraulic behavior of the NPP is considered in the reported analysis. Beside the model assumptions and the boundary conditions, the accident sequence is explained in detail. It turns out that the remaining systems for emergency cooling are able to handle the LOCA under such demanding boundary conditions. Most importantly, the primary pressure can be lowered below the zero flow pump head of the pool cooling pumps. The long-term cooling can thereby be ensured. Furthermore, the heat removal out of the core is always sufficient.

Overall it can be concluded that all safety protection objectives have been complied for this beyond-design basis scenario.
Numerical Simulation of Corium Jet Breakup During Fuel-Coolant Interaction Based on the ALISA Test Performed at KROTOS Test Facility
Authors: Shen Pei , Zhou Wenzhong
Published: October 2018

Steam explosion is one of the consequences of fuel-coolant interactions in a severe accident. Melt jet fragmentation, which is the key phenomenon during steam explosion, has not been clarified sufficiently which prevents the precise prediction of steam explosion. The focus of this paper is on the numerical simulation of the melt jet behavior falling into a coolant pool in order to get a qualitative and quantitative understanding of initial premixing stage of fuel-coolant interaction. The objective of our first phase is the simulation of the fragmentation process and the estimation of the jet breakup length. A commercial CFD code COMSOL is used for the 2D numerical analysis employing the phase field method. The simulation condition is similar to our steam explosion test supported by the ALISA (Access to Large Infrastructure for Severe Accidents) project between European Union and China, and carried out in the KROTOS test facility at CEA, France. The simulation result is in relatively good agreement with the experimental data. Then the effect of the initial jet velocity, the jet diameter and the instability theory are presented. The preliminary data of melt jet fragmentation is helpful to understand the premixing stage of the fuel-coolant interaction.
Design and Distortion Analysis of Thermal-Hydraulics Test Facility for the Fuel Transfer Tube
Authors: Xi-dao Mao , Yang Liu , Hai-jun Jia , Qiang Guo
Published: October 2018

It is important to study the thermal-hydraulic safety issues of the fuel assembly during the transportation in the fuel transfer tube. The flow and heat transfer mechanism is complicated when cooling the fuel assembly by natural circulation in the transfer tube. After analyzing the flow field in the transfer tube, the position of highest surface temperature is determined. The test facility was designed by the sampling the high temperature zone. A set of similarity criterion was developed for distortion analysis of the test facility. The analysis results show that the test facility is conservative to evaluate the thermal-hydraulic safety of the assembly.
Study on Factors Influencing the Diffusion and Migration of Radionuclide Offshore in Nuclear Power Plant Accidents
Authors: Zichao Li , Tao Zhou , Xuemeng Qin , Haider Amir , Bing Li , Juan Chen
Published: October 2018

Studying the diffusion and migration mechanism of radionuclide in an offshore environment under accidental conditions of nuclear power stations is of great significance to emergency decision-making and nuclear accident mitigation. The migration mechanism of radionuclide in the ocean was studied. For the short-term situation, a multi-factor coupled model was established and the effects of turbulent diffusion, decay and adsorption sedimentation on the radionuclide concentration were analyzed. The results show that the radionuclide concentration decreases with turbulent diffusion. In the short-term, the decay of radionuclide has no effect on the concentration of 137Cs, but has a significant effect on the concentration of 131I. The maximum adsorption sedimentation rate of suspended matter to radionuclide is about 1%, which can be neglected in the short-term simulation and needs to be considered in the long-term simulation. The model can better simulate the variation of radionuclide concentration.