bluuebunny/crossref_metadata_2025 huggingface dataset:
The Numerical Simulation Research of Gas Entrainment Based on Three-Dimensional Model
Authors: Zhang Yilin , Huang Shanfang
Published: October 2017

Two kinds of three-dimensional model are built to simulate the gas entrainment process through a small break in the horizontal coolant pipe at the bottom of the stratified flow. The results were compared with the two-dimensional simulation results and the experimental data. In terms of the two-phase distribution, the simulation results agree well with the experimental data and show much superiority compared with the two-dimensional model. The results verify the reliability of model building, condition setting and calculating method qualitatively and quantitatively. In general, after gas entrainment, the average velocity over cross section increases obviously, but the mass flow rate decreases contrarily. This is because that void fraction meanwhile reduces the fluid density. In addition, it is found that the larger the void fraction of vapor is, the higher the average discharge velocity of the fracture cross-section fluid is. Besides, with the larger internal and external pressure difference, the gas volume fraction and the flow velocity in the break increase, resulting in the mass flow rate increasing along with them. However, since the critical height increases as well, the total loss amount of liquid in the stable effluent stage decreases, and the time before entrainment becomes shorter.
Neutronic Analysis of Accident-Tolerant Fuel Assemblies With Silicon Carbide Cladding in Pressurized Water Reactors
Authors: Tan Zhixiong , Cai Jiejin
Published: October 2017

After Fukushima Daiichi Nuclear Power Plant accident, alternative fuel-design to enhance tolerance for severe accident conditions becomes particularly important. Silicon carbide (SiC) cladding fuel assembly gain more safety margin as novel accident tolerant fuel. This paper focuses on the neutron properties of SiC cladding fuel assembly in pressurized water reactors. Annular fuel pellet was adopted in this paper. Two types of silicon carbide assemblies were evaluated via using lattice calculation code “dragon”. Type one was consisted of 0.057cm SiC cladding and conventional fuel. Type two was consisted of 0.089cm SiC cladding and BeO/UO2 fuel. Compared the results of SiC cladding fuel assembly neutronic parameters with conventional Zircaloy cladding fuel assembly, this paper analyzed the safety of neutronic parameters performance. Results demonstrate that assembly-level reactivity coefficient is kept negative, meanwhile, the numerical value got a relatively decrease. Other parameters are conformed to the design-limiting requirement. SiC kinds cladding show more flat power distribution. SiC cases also show the ability of reducing the enrichment of fuel pellets even though it has higher xenon concentration. These types of assembly have broadly agreement neutron performance with the conventional cladding fuel, which confirmed the acceptability of SiC cladding in the way of neutron physics analysis.
Preparation and Properties Investigation of Neutron Absorber Material Mo-Tb-Dy and Y-Tb-Dy Alloys
Authors: Lu Junqiang , Chen Lei , Qiu Congzhang , Yang Weiyan , Zhang Feifei , Yang Ruifang
Published: October 2017

Control rods with low worth absorber materials may provide a mechanical means of making relatively small adjustments in core reactivity. Mo-Tb-Dy and Y-Tb-Dy alloys were developed to obtain appropriate nuclear performance for low worth absorber material. The two alloys were prepared by powder metallurgy technology and vacuum melting technology individually. To clarify the effects of Mo and Y diluents, Tb-Dy was also prepared to be compared. The microstructures were analyzed by X-ray diffraction (XRD), Scanning Electron Microscope (SEM) and Transmission Electron Microscopy (TEM). The experiment results showed that homogeneous microstructures were obtained. Out-pile properties, including mechanical properties, thermal conductivities, thermal expansion, corrosion resistance properties and ion irradiation properties were measured and analyzed. Y-Tb-Dy has similar properties with Tb-Dy. With temperature increasing, yield strengths of Tb-Dy and Y-Tb-Dy decreases largely while Mo-Tb-Dy decreases slightly. Thermal conductivities of Mo-Tb-Dy were four times more than Tb-Dy and Y-Tb-Dy. Mo element significantly increases thermal conductivity. Tb-Dy and Y-Tb-Dy showed severe corrosion and became powders in 280°C/10MPa de-ionized water while Mo-Tb-Dy had very slow corrosion rate. All three alloys were irradiated at 400∼700°C for 25 displacement per atom (dpa). No voids was observed for Tb-Dy and Y-Tb-Dy. Void diameter increases and its density decreases with temperature increasing for Mo-Tb-Dy. Maximum irradiation swelling rate with 0.5% was observed at 500°C. Irradiation swelling significantly decreased with increasing irradiation temperature.
Study on the Technical Method of Level 2 Probabilistic Safety Assessment in Multiple External Hazards Compound Situations
Authors: Wei-dong Liu , Hong-wei Dai , Peng-yi Peng , Ze-feng Zhang , En-qi Wang
Published: October 2017

The article introduces the method of level 2 probabilistic safety assessment in multiple external hazards compound situations, and the application of probabilistic safety assessment technology in building a Level 2 probabilistic safety assessment model for a domestic nuclear power plant under such circumstances. This model can be used to quantitatively evaluate important risk indicators like the frequency of a large release of radionuclides, frequency of occurrence of each release category, in order to identify weaknesses of nuclear power plants in responding to risks. Corresponding risk insights are provided to enhance the safety of nuclear power plants.
Simulations of Interfacial Dynamics of Rising Bubble in Liquid LBE Alloy Based on Diffuse-Interface Method
Authors: Wang Chuntao , Cai Jiejin
Published: October 2017

This paper describes a novel numerical simulation method for simulating interfacial dynamics of a single bubble rising in liquid lead bismuth alloy. Numerical simulation of a single nitrogen bubble in liquid lead bismuth alloy from static to the full development of dynamic behavior based on diffuse-interface method was done to get the relationship of rising velocity and deformation of the bubble by time. Results of simulation of the flow characteristics of inert gas and liquid LBE alloy based on diffuse-interface method are consistent with the Grace emperical relationship graph, which verifies the feasibility and the accuracy of diffuse-interface method on simulation of rising bubble’s behavior in liquid LBE alloy. It is easy to find that with the increase of Eo, which is determined by bubble initial diameter when fulid is determined, the bubble deformation increases, and the bigger the bubble is, the easier it is to split the sub bubbles.
A New Method to Evaluate the Correlation Coefficients Between Nuclides in Uncertainty and Sensitivity Analysis
Authors: Jia Yi Xu , Xu Bo Ma , Jing Wen Shen , Jia Yi Liu , Yi Xue Chen
Published: October 2017

Uncertainty and sensitivity analysis is an essential component of nuclear engineering calculations. Uncertainties in the cross-section input data directly affect uncertainties in the results. The covariance values between different types of cross-sections are considered in the NJOY covariance library. However, the correlation coefficient between isotopes can depend on the specific problem. The correlation coefficient between 235U and 238U in a pressurized water reactor (PWR) might be different from that in a fast reactor. In this study, a new Monte Carlo-based method is proposed for calculating this effect. The correlation coefficients between different isotopes are calculated using a problem-dependent fraction parameter. The correlation coefficients between the capture cross-sections of 235U, 238U, 239Pu, and 241Pu are calculated. The same method can be extended to other reaction types. The correlation coefficients as a function of the isotopic atomic density uncertainty and the average one-group microscopic cross-section uncertainty are also studied. It is shown that the correlation coefficients vary very little with the uncertainty in the average one-group microscopic cross-section. The correlation coefficient of an isotope pair changes slightly over the course of a cycle because of atomic density and microscopic cross-section changes.
Small Punch Test on Mechanical Properties of Molybdenum Single Crystal Weld
Authors: Ma Yan , Wang Jianju , Lei Jinyun
Published: October 2017

Due to its good creep strength, molybdenum (Mo) single crystal is used for space thermionic reactor fuel element. However, stray grain formation during Mo single crystal welding leads to the degradation of mechanical properties. In this paper, Mo single crystal is welded by electron beam welding and the joint properties are evaluated by means of small punch test (SPT). The load-displacement curves of Mo single crystal matrix and fusion area are measured. The experimental data is fit to establish the relationship between yield load and standard test yield strength. The results show that the yield strength of the fusion is much lower than the other material. The micro hardness of the both materials is almost equal. Simultaneously the displacement value of the fusion area is the smallest which means the joint brittleness increases. SEM observation of rupture fracture demonstrates the embrittlement of the weld.
Numerical Simulation of Fluid Mixing and Stratification in Large Space
Authors: Cao Xingdi , Niu Fenglei , Wang Bin , Su Xiaowei
Published: October 2017

For AP1000 reactor, passive containment cooling system is a vital way to release heat to the environment, so an accurate prediction of distribution of temperature or density in large layered space plays an important role on the reactor optimization design and safety analysis. This paper investigates in comparing the results of different kinds of models in FLUENT with the experiment results in order to find out a more effective model and more suitable mesh number to simulate the mixing and stratification phenomenon. When LOCA or MSLB occurs in containment, the radius, position, and angle of the break can affect the containment mixing and thermal stratification. So this paper also studies the influence of height of the break and angle of the break on stratification with Fluent, and makes comparative analysis with the experiment results.
Visualization Study of Bubble Sliding Characteristics in a Subcooled Flow Boiling Narrow Rectangular Channel Under Natural Circulation Condition
Authors: Yang Kuan , Cao Xiaxin , Yan Changqi , Yang Yongyong , Tian Chunping , Xu Jianjun
Published: October 2017

Heat transfer enhancement by the motion of the bubbles sliding along the heating surface are wildly reported by many researchers, thus it is of great importance to quantitatively investigate the characteristics of sliding bubbles. A visualization study of subcooled flow boiling of water in a vertical single face-heated narrow rectangular channel under a series of natural circulation working conditions was conducted. Pictures of the bubble sliding behaviors were captured by a high speed camera simultaneously with thermal data. A sequence of digital image processing algorithms were applied to the original picture to extract bubble shape and location information, which post-processing methods were adopted to obtain characteristic sliding parameters (including the distribution of the equivalent sliding bubble diameter and velocity, number density of the sliding bubbles).

It is found that bubbles can be able to nucleate and grow while sliding on the heating plate after the ONB point; the bubble number density, average bubble sliding velocity and the average sliding diameter continue to increase along the test section; heat transfer in the flow channel are significantly enhanced along flow direction with relatively low local void fraction. The average bubble sliding velocity near the inlet is significantly smaller than the sectional average velocity of single-phase fluid of the flow channel, and then it exceeded near the outlet of the test section. The average bubble sliding velocity and diameter increase with increasing heat flux and decreeing local subcooling degree. The equivalent diameter of sliding bubbles and the bubble sliding velocity approximately follow normal distribution. The distribution of the bubble diameter and velocity both cover a wild range. The standard deviations of the probability density function of the sliding bubble diameter and velocity increase with increasing heat flux and decreasing subcooling degree.
Numerical Investigation of Spacer Effects on Heat Transfer Enhancement in a Vertical Annular Channel at Supercritical Pressure
Authors: Pan Jinsong , Xiao Yao , Gu Hanyang
Published: October 2017

Supercritical water-cooled reactor (SCWR) is one of the Gen-IV reactors, which shows higher economy and safety. Reactor core of a SCWR employs a tight lattice in order to efficiently remove heat from nuclear fuels. In the narrow sub-channels of a tight lattice reactor core, a spacer has been used as a turbulence generator and a space-keeper between the fuel rods. Meanwhile, the spacer has significant influences on the heat transfer of the fuel rods. In order to investigate the effects of simple spacers on heat transfer to upwardly flowing supercritical fluid in a vertical annular channel, an experiment is underway at the supercritical model fluid thermal hydraulics test facility (SMOTH) with supercritical R134A. The equivalent diameter of the annular channel is 6 mm. The outer metal circular tube of the annular channel test section is electrically-heated. The blockage ratio of the simple spacers ranges from 0.2 to 0.4. Based on the geometry parameters of the test section, preliminary numerical investigations were carried out for the effects of simple spacer on the local heat transfer performance of supercritical R134A using commercial CFD code FLUENT. Heat transfer characteristics in the spacer downstream were analyzed with respect to the variations of heat flux, mass flux, pressure, blockage ratio and local enthalpy. And the reason for the different heat transfer enhancement under different conditions is given preliminarily. Finally, existing empirical correlations were selected to be compared with the results of CFD numerical simulation. The applicability of conventional subcritical heat transfer enhancement correlations for spacer grids to supercritical fluid was discussed.
Effects of Various Parameters on Supercritical-Pressure Heat Transfer and Limits for Heat Transfer Correlations Obtained in Vertical Bare Tubes
Authors: Maloufi Hakim , Xie Hanqing , Zopf Andrew , Anderson William , Langevin Christian , Pioro Igor
Published: October 2017

Currently, there is a number of Generation-IV SuperCritical Water-cooled nuclear-Reactor (SCWR) concepts under development worldwide. These high temperature and pressure reactors will have significantly higher operating parameters compared to those of current water-cooled nuclear-power reactors (i.e., “steam” pressures of about 25 MPa and “steam” outlet temperatures up to 625 °C). Additionally, SCWRs will have a simplified flow circuit in which steam generators, steam dryers, steam separators, etc. will be eliminated, as the steam will be flowing directly to a steam turbine.

In support of developing SCWRs studies are being conducted on heat transfer at SuperCritical Pressures (SCPs). Currently, there are very few experimental datasets for heat transfer at SCPs in power-reactor fuel bundles to a coolant (water) available in open literature.

Therefore, for preliminary calculations, heat-transfer correlations developed with bare-tube data can be used as a conservative approach. Selected empirical heat-transfer correlations, based on experimentally obtained datasets, have been put forward to calculate Heat Transfer Coefficients (HTCs) in forced convective in various fluids, including water at SCPs.

The Mokry et al. correlation (2011) has shown a good fit for experimental data at supercritical conditions within a wide range of operating conditions in Normal and Improved Heat-Transfer (NHT and IHT) regimes. However, it is known that a Deteriorated Heat-Transfer (DHT) regime appears in bare tubes earlier than that in bundle flow geometries. Therefore, it is important to know if bare-tube heat-transfer correlations for SCW can predict HTCs at heat fluxes beyond those defined as starting of DHT regime in bare tubes.

The Mokry et al. (2011) correlation fits the best SCW experimental data for HTCs and inner wall temperature for bare tubes at SCPs within the NHT and IHT regimes. However, this correlation might have problems with convergence of iterations at heat fluxes above 1000 kW/m2.
The Application of Risk-Informed Decision-Making Approach in Optimization of Allowed Outage Time for Safety System in Nuclear Power Plant
Authors: Wu Licun , Ma Chao , Deng Wei
Published: October 2017

To facilitate online maintenance for High Pressure Safety Injection (HPSI) system in nuclear power plant, the optimization measure for Allowed Outage Time (AOT) is proposed in this paper, and risk-Informed decision-making approach is used to demonstrate its feasibility. According to the analysis results, the extension of AOT meets the requirements of relative guides and deterministic analysis, and the impact on risk of plant is acceptable. Based on the extension of AOT, it is feasible to carry out online maintenance for safety system and corresponding support systems, which can effectively reduce workload during shutdown.
A Study on Maintenance Rule Application in VVER NPP
Authors: Chao Ma , Wei Deng , Jin An
Published: October 2017

Maintenance effectiveness is important for the safety and power production of Nuclear Power Plants (NPP). U.S. Nuclear Regulatory Commission (NRC) Maintenance Rule (10CFR50.65, MR) became effective in 1996, and it is mandatory for all the US plants to use Maintenance Rule in their daily maintenance activities. With the development and wide usage of Probabilistic Risk Assessment (PRA) technique in China, China regulator and utilities are trying to adopt MR in maintenance activities. Brief study on application of MR in some VVER-typed China nuclear plant is carried out and some main results are shown. All the application process and results will be useful for later official application of MR in China.
Development of Wireless Tomography System to Measure Solid-Liquid Flow in Rotating Body
Authors: Atagi Yuya , Akimoto Yuya , Nagae Fumiya , Okawa Kazuya , Iso Yoshiyuki , Ikeda Ryosuke , Matsumi Yuuki , Takei Masahiro
Published: October 2017

A wireless tomography system has been developed to measure the real-time behavior of solid-liquid two phase flow in the centrifuge for controlling a rotational velocity and particle supply rate.

The purpose of this study is development of real-time behavior of solid-liquid two phase flow measurement technique and image conductivity distribution in rotating body because a technique for measuring the behavior of solid-liquid two phase flow in the centrifuges has not existed yet. In this study, a device to measure wirelessly behavior of solid-liquid two phase flow in stationary body has developed as a preliminary to measure in rotating body.

A centrifugation technology for industry process should be improved to obtain more effective separation, shorten process time and save energy. These requirements are achieved by optimizing rotational velocity and particle supply rate. The real-time measurement of behavior of solid-liquid two phase flow for controlling the rotational velocity and the particle supply rate is essential. In other words, the real-time behavior of solid-liquid two phase flow measurement and the rotational velocity control become innovative technologies. Typical behavior of solid-liquid two phase measurement technics with easy handling in particles liquid two-phase flow uses process tomography because the process tomography is suitable for real-time measurement in a rotating body. Process tomography has high temporal resolution. This detector was used with the wireless because electrical cables are not available for centrifuges under high speed rotational condition. This wireless tomography system was used for a lab scale rotating machinery measurement experiment.

Consequently, we could wirelessly measure the behavior of solid-liquid two phase flow in stationary body in real time and get data to image the behavior of solid-liquid two phase flow. This result images the behavior of solid-liquid two phase flow. Observing it enable to control a rotational velocity and particle supply rate.

So this wireless tomography system can separate solid-liquid two phase flow efficiently.
Technology Development on Reactor Dismantling and Investigation of Contamination in FUGEN
Authors: Soejima Goro , Iwai Hiroki , Nakamura Yasuyuki , Hayashi Hirokazu , Kadowaki Haruhiko , Mizui Hiroyuki , Sano Kazuya
Published: October 2017

Advanced Thermal Reactor (ATR) FUGEN is the heavy water-moderated, boiling light water-cooled, pressure tube-type reactor. The commercial operation of FUGEN started on Mar. 1978 and terminated on Mar. 2003 and the decommissioning of FUGEN has been carried out since the decommissioning plan was approved in 2008.

In order to perform the decommissioning work such as dismantling and decontamination safely and reasonably, technology development for the decommissioning has been promoted actively.

This paper describes a part of technology development as follows.

(1) Technology development on reactor dismantling

The reactor of FUGEN is made of various materials such as stainless steel, carbon steel, zirconium alloy and aluminum which have relatively high activity concentration by operation for 25 years.

With consideration of these characteristics, the reactor will be dismantled under water remotely in order to shield the radiation and prevent dust from migrating from water to air generated by the cutting considering the usage of zirconium alloy which is likely to be oxidized. In addition, laser cutting method whose features are fast cutting speed and less secondary waste in cutting will be applied for reactor dismantling.

However, laser cutting method has no experiences to be applied to dismantlement of reactor facilities. Therefore, laser cutting for actual dismantled objects in air was demonstrated in controlled area in FUGEN using laser cutting system composed of articulated robot and laser cutting head. As a result, safety and applicability of laser cutting system was confirmed. From now on, primary cutting work in air, cutting demonstration with a relatively high dose rate and mock-up test in water for dismantling the actual reactor will be carried out.

(2) Technology development on investigation of contamination

It is necessary to evaluate radioactive inventory in the facilities accurately in order to reflect the evaluated data to dismantling plan appropriately. Therefore, the investigation of the contamination for the facilities has been carried out for safe and reasonable decommissioning work.

The in-situ simple investigation method for the contamination of inner pipes which is mostly dominated by Co-60 is started to develop using the portable NaI(Tl) spectrometer. This method complements conventional investigation method to take samples from the pipes and to analyze them by radiochemical method to figure out the contamination of the whole facility.
Simulation of Cross-Flow Vortex-Induced Vibration of Single Tube Based on Two-Way Fluid-Solid Interaction Method
Authors: Zengzeng Wang , Tao Lu , Bo Liu
Published: October 2017

The fatigue damage and lift force caused by vortex induced vibration occur very often in the core of the Pressurized Water Reactor (PWR) [1] It is extremely complex to illustrate the mechanism of vibration which induced by Cross-flow. With the spacer grids and wings, the flow direction which in axial direction at the inlet will change and create swirls, so there are many flow directions in the nuclear fuel component. Assumed the tube endure cross-flow only in this article to simplify the fluid model. Most researchers in this field often ignore the displacement of structure induced by the cross flow because the value is so small that not enough to change the fluid region. In truth conditions, the motion of the cylinder caused the wake oscillation and strengthen the vortex shedding, in turn, the vortex shedding will aggravate the vibration amplitude. According that, one way FSI (Fluid Solid Interaction) can’t capture the influence from the cylinder vibration. In this article, Two-way FSI method was executed to get the vibration in time history in order to get the random vibration induced by the cross flow more close to the actual project. Using Finite Volume Method to discrete the fluid control equation and finite element method to discrete structure control equation combined with moving mesh technology. An interface between the fluid region and the structure region was created to transfer the fluid force and the structure displacement. Coupling CFD code and CSD (Computational Solid Dynamics) code to solve the differential equation and obtain the displacement of the cylinder in time history. A Fast Fourier Transfer (FFT) has been done to get the vibration frequency. An Analysis of the vortex shedding frequency and vibration frequency to find the correlation between the vortex shedding and the vibration frequency has been done. A modal analysis for the cylinder without water has been done to get the natural frequency. Results shows the cylinder has different response to the vortex shedding at different position of the cylinder in the same condition. There are more works need to be done aim to get the vibration mechanism in tandem tube and parallel tube to get clearly mechanism of vortex induced vibration in nuclear fuel assembly. The research of the vortex induced vibration in this article is a key to get on the follow research in more tubes array in different methods.
Breeding Properties Study on High Power Thorium Molten Salt Reactor
Authors: Liu Yafen , Guo Rui , Cai Xiangzhou , Yan Rui , Zou Yang , Zhou Bo
Published: October 2017

The Molten Salt Reactor (MSR), the only one using liquid fuel in the six types ‘Generation IV’ reactors, is very different from reactors in operation now and has initiated very extensive interests all over the world. This paper is primarily aimed at investigating the breeding characteristics of high power (1000 MWe) Thorium Molten Salt Reactor (TMSR) based on the two-fluid Molten Salt Breeder Reactor (MSBR) with superior breeding performance. We explored the optimized structure to be a thorium based molten salt breeder reactor with different core conditions and different post-processing programs, and finally got the breeding ratio of 1.065 in our TMSR model. At last we analyzed the transient security of our optimized model with results show that the temperature coefficient of core is −3 pcm/K and a 2000 pcm reactivity insertion can be successfully absorbed by the core if the insertion time is more than or equal to 5 seconds and the core behaves safely.
Education and Exercise for Severe Accident Management at KANSAI
Authors: Kawahara Takumi , Shiozawa Tsugio , Nishioka Tsutomu , Shimominami Yukimoto , Onooka Hiroaki
Published: October 2017

In case of a severe accident, all the staff members of a nuclear power plant (NPP), members of the Emergency Response Organization (ERO) on site as well as operators in the main control room (MCR) are required to take necessary actions to mitigate the consequential effects of the accident. Therefore, Nuclear Engineering Ltd (NEL) has been implementing education and exercise for severe accident (SA) management both at nuclear power plants and Nuclear Power Division of Kansai Electric Power Company (KANSAI) since FY 2014. For the education of commanders who take a lead at the ERO in case of an accident, table top exercise is provided by using simulators developed by NEL, including functions to respond to a SA involving core melt. Continuous implementation of this education and exercise program is expected to enhance KANSAI’s severe accident management ability and their voluntary safety improvement activities in the future.
Probabilistic Safety Assessment for CAP Nuclear Power Plant Spent Fuel Pool
Authors: Yiquan Xu , Yucheng Zhuo , Yajun Yang , Hao Fu
Published: October 2017

For china advanced passive (CAP) pressurized water reactor (PWR) nuclear power plant, spent fuel damage risk potentially induced by internal events is assessed. Spent fuel damage frequency (FDF) is quantified by event tree and fault tree model using probabilistic safety assessment (PSA) software RiskSpectrum. For different operation conditions total FDF is 2.05×10−9 per reactor-year which is much lower than core damage frequency (about 2.41×10−7 per reactor-year). By assumption of completely radioactivity releasing, large radioactive release frequency caused by FDF is one order of magnitude lower than that caused by core damage (about 2.38×10−8 per reactor-year). Since the multiple prevention and mitigation measures in CAP PWR nuclear power plant responding to spent fuel pool accidents, the risk of spent fuel pool is much lower than that of reactor core, and the safety goals of nuclear safety guide can be satisfied.
Investigation on Gas Dispersion Inside the Large-Scale Containment by CFD Simulation With Code_Saturne for Experimental Scenario Definition
Authors: Song Yuting , Yun Xiangyu , Chen Tian , Zhang Huiyong , Min Jiesheng , Chen Guofei
Published: October 2017

During accidental situations, large quantities of hydrogen are generated and released due to metal-water reaction. Eventual stratification formed in the top region of containment and locally high concentration threatens the integrity of nuclear power plants. The stratification of hydrogen may be eroded by several measures, such as air jet, spray, ventilation. So far, several activities have been carried out on the hydrogen stratification’s break-up study. The well-known OECD/SETH-2 project used PANDA and MISTRA test facilities to study this phenomenon with variable test conditions. These test cases are also wildly used to validate the computational fluid dynamics (CFD) codes.

The large-scale containment test facility – COntainment Thermal-hydraulics and Hydrogen Distribution (COTHYD) was designed and constructed at CGN to study the containment thermal hydraulics behavior and hydrogen risk in PWR during severe accidents. A series of tests including helium and air stratification, helium and air stratification eroded by air jet, as well as steam dispersion and condensation are planned to be performed on this test facility. The objective is to set up high quality database for code validation and physical phenomena research.

During the test cases preparation, the relevant tests carried out since 2000 in MISTRA, TOSQAN, THAI, PANDA were collected to give the reference for the cases design in COTHYD. For the tests of stratification’s break-up on COTHYD, helium is discharged from top and it accumulates at the top region of containment. On the preparation stage, Code_Saturne was used to define the test scenario and predict helium distribution. Both dead volume and open volume (with ventilation) are modeled to investigate helium stratification formation and helium-rich layer concentration evolution. These results will be used as test matrix configurations. Code_Saturne, EDF in-house open-source software, has been used in the simulation of hydrogen dispersion. In 2015, B. Hou presented their work on the simulation of the break-up phenomenon of helium stratification by air (ST1_7 and ST1_10 test cases of OECD/SETH-2 project) (Hou et al., 2015) [1]. Simulation results demonstrated that Code_Saturne can well predict and simulate this phenomenon. This software is, as a consequence, used as the case design and validation tools in the pre/post experiment steps.
Prospect Research on Application of Mobile Liquid Radwaste Treatment Unit in Nuclear Power Plant
Authors: Zhu Laiye , Zhu Maogang , He Yanhong , Wang Xin , Dong Wenshu
Published: October 2017

With the introduction of third generation of nuclear power AP1000, Westinghouse uses the mobile device (a mobile wastewater treatment device 6 units shared) radioactive waste system design concepts. This design not only simplifies the process of nuclear island waste system; saves equipment layout space; improves equipment utilization; while increases the use of new technologies lifetime of the plant and the possibility of flexibility. This paper introduces the first AP1000 unit (Sanmen, Zhejiang Province) by using the advanced mobile device technology and application of wastewater treatment under the condition of the primary coolant source level. At the same time, the paper also discusses the periodic system inspection and the strategy of maintenance. In addition, the paper further expands the application direction of the mobile waste processing aspects, such as: decommissioning of nuclear facilities; enhancing the facility decommissioning radioactive liquid waste purification capability. Another example: After the Fukushima accident, people pay more attention to accident-mitigation-design and hope to accelerate the development of emergency radioactive liquid waste processing devices. Thus, in addition to strengthening the nuclear power plant inherent defense in depth and resistance emergency capability, mobile waste treatment device or combination device special regional settings can be made to improve and enhance the ability to get more diversified emergency response.
Experimental Study on Critical Heat Flux of In-Vessel Retention for Large Nuclear Reactor
Authors: Sun Canhui , Chen Wei , Geng Jun
Published: October 2017

During a rare accident involving severe core damage in a nuclear power plant, if the molten core material can be contained within the boundary of the reactor vessel, the severity of the accident is expected to be greatly reduced. Therefore, the severe accident management strategy based on in-vessel retention (IVR) of molten core debris is highly desirable, and has been adopted by advanced reactor designs such as AP1000, APR1400 and CAP1400.

Previous studies assumed a fully developed molten pool, where molten metal and molten oxides separated, and liquid metal form a layer on the top of the oxidic pool. The limiting failure mechanism of the lower head is related to occurrence of boiling crisis on the outer surface of the vessel. The boiling crisis would occur, when the local heat flux through the vessel wall exceeds the critical heat flux value. The evaluation of thermal regime is then done by comparing the heat flux distribution from the molten corium pool and the CHF variation on the outer wall in order to define the available margins to CHF. When a larger power nuclear reactor is designed, the margin of the IVR could be decreased. Therefore, the study on critical heat flux enhancement for lager nuclear reactor is necessary.

In this study, an experimental facility is conducted to analysis the critical heat flux mechanism and enhancement for lager power nuclear reactor. The facility includes two loops, one is primary loop, and the other is auxiliary loop which could import injection flow in the primary loop. The test section is an inclined downward heated rectangular channel, and the angle is variable. A series of critical heat flux experiments are conducted in this facility. The results are compared with previous experiment results and calculation correlation in the large nuclear reactor operating conditions.
Investigation of Diameter Effect on Critical Flow
Authors: Minfu Zhao , Yufeng Lv , Dongxu Zhang , Yuzhou Chen , Keming Bi
Published: October 2017

Discrepancy has long existed about nozzle diameter effect on critical flow rate in two-phase critical flow analysis. The issue is of great importance because it involves whether the scaling test can accurately simulate the actual process during loss of coolant accident. A series of transient critical flow experiments has been performed in China Institute of Atomic Energy (CIAE), which aims to study the nozzle diameter effect. The diameter of adopted test nozzle is 5mm, 10mm and 15mm, respectively. The experiment result shows that the discharge mass flow rate decreases as nozzle diameter increases. This is contradictory with the results obtained from steady state critical flow experiment. Comprehensive analysis shows that the diameter effect observed in transient critical flow experiment is transient effect caused by system pressure change, and the mechanism is the dynamic unbalance in the transient period. It is presumed that for the same test section with a certain diameter, if the pressure change velocity is different, the measured critical flow rate will be different. The conclusion is validated in the pressurization and depressurization blowdown test. It| is arrived that nozzle diameter has no direct effect on critical flow rate, and thus the discrepancy on diameter effect is clarified.
Modeling and Thermal Calculation of Fuel Rod With SiCf/SiC Cladding for LWRs
Authors: Xinyuan Shang , Shaoyang Zhang , Aimin Zhang
Published: October 2017

SiC fiber composite material as fuel cladding is proposed to increase the power density and maximum allowable fuel burnup in light water reactors. Empirical models about thermal properties of the SiC material are developed as a function of operating temperature and neutron fluence. A fuel rod modeling code frapcon2-SiC based on frapcon2 is compiled to predict the performance of SiC cladding when operating. Comparison of the behavior between the SiCf/SiC cladding and Zr-4 cladding in different thickness reveals that higher temperature will get due to the poor thermal conductivity of the SiC. As far as the Conclusion shows that Poor thermal performance of SiC cladding makes the design margin becomes smaller.
Effect of Steam Injection on Last-Stage Blade Temperature of Low Pressure Steam Turbine at Low Volume Flow Condition
Authors: Ma Can , Wu Jun , Lin Yuansheng
Published: October 2017

In the nuclear power plant, the last stage of the low pressure steam turbine is characterized by long blades. These long blades operate under severe working conditions with wet steam flow and strong mechanical stress. At the start up and shut down operating condition where the volume flow is extremely low, the last stage blades operate in ventilation conditions where there is significant reverse flow in the exhaust and the last stage. In such a condition, the reverse flow would cause significant increase in the blade temperature. In addition, the rotating reverse flow would increase the vibration of the rotor blade. Such temperature increase and enhanced vibration can cause blade damage and force the machine to be shut down. In previous work, the steam injection in the last stage has been proposed as a promising method to decrease the reverse flow at low volume flow conditions, which reduces the stall cell size in the last stage blade. This work investigates the effect of the steam injection process on the blade temperature distribution by conducting three-dimensional flow simulations. Various steam injection configurations are compared in this work and the major consideration to be noted in the design process is discussed.
Research on Nickel-Plated Hydrogen-Absorption Device in Fuel Rod and Performance Testing
Authors: Qin Zhou , Jiwei Li , Yu Dang , Yang Ding
Published: October 2017

The hydrogen may be introduced into the fuel rod during the process of production and manufacture. During the operation in reactor, the irradiated fuel pellets also produce radioactive isotopes of hydrogen and tritium. Under the operating condition in pile, the hydrogen in fuel rod will enter the zirconium alloy cladding tube forming hydride, lead to the hydrogen brittleness of cladding tube, and severe cases can lead to the cladding tube broken. The radioactive tritium inside fuel rod has high activity, and it possibly goes through the cladding tube by diffusion penetration into the reactor coolant. With the reactor in waste water or steam waste emissions to the environment, such as lead to tritium radiation safety problems of environmental pollution. Thus, reduce the hydrogen source and tritium pressure in fuel rod, is the way to reduce the hydrogen absorption effect and the release of tritium to coolant. By conducting the Zr-4 alloy nickel-plated hydrogen-absorption device design research, through nickel plating process on the surface of Zr-4 alloy structure parts, eliminating the influence of the oxide film to maintain its excellent absorbing hydrogen isotope activity. During the design operating temperature conditions of fuel rods, the reaction of zirconium hydride has lower hydrogen balance pressure, while the gas cavity kept low pressure hydrogen isotope, can significantly reduce the hydrogen pickup of fuel rod zirconium alloy cladding tube and reduce the tritium permeation emissions by cladding tube. Through nickel-plated hydrogen-absorption device structure design, manufacture, performance testing, analysis and evaluation, demonstrates that the flat plate and cross nickel-plated hydrogen-absorption device can meet the expected effect.
The Application of Heat Pipe Discharge Containment Heat Experimental Study: New Type of Passive Containment Heat Removal System Concept Design
Authors: Qin Haiqi , Lu Daogang , Wang Shengfei
Published: October 2017

Practice has proved that nuclear power technology development and operation of nuclear power is a clean, safe and large-scale provided stable power. AP1000 uses a large number of passive safety technologies. Passive residual heat removal system is an important part, in the long-term cooling stage of nuclear reactor normal operating conditions or accident conditions, to prevent the core meltdown. The research of this paper is to solve the long-term discharge of residual heat of the containment in the accident condition of nuclear power plant. Based on the passive heat removal system of AP1000, combined with the heat transfer characteristics and advantages of heat pipes, the PRHR system is further improved on the basis of the present situation, and a conceptual design of passive containment residual heat removal system is proposed. In order to further verify the feasibility of the conceptual design, we make a simplified simulation of small containment test bench to carry out experimental verification and give a detailed experimental design.
The European Utility Requirements for Advanced Light Water Reactors (EUR): Recent Achievements and New Challenges
Authors: Jacquart Guillaume , Kymäläinen Olli , Vivier Jean-François , Ferraro Giovanni , Vieilletoile Emmanuel
Published: October 2017

Since 1991, the European Utility Requirements (EUR) organisation has been actively developing and promoting harmonised technical specifications for the new mid- and large-size LWR designs to be proposed by the vendors in Europe.

The EUR document consists of a comprehensive set of requirements covering the whole Nuclear Power Plant (NPP). It encompasses all aspects (safety, performance, competitiveness) and all parts of a NPP (nuclear island and conventional island). The document can be used by the utilities (guide for design assessment, technical reference for call for bids) and by the vendors, as a technical guide. The harmonisation which is sought after by the EUR aims at delivering the safest and most competitive designs based on common rules shared all over Europe. Fourteen nuclear operators across Europe are members of the organisation.

After the publication of the Revision D of the EUR document (October 2012), the EUR organisation has been extremely active. This presentation describes the main results obtained during the last four years and the new challenges for the coming years (roadmap 2016–2018) in the three following fields.

First, the revision of the EUR document, in order to maintain it at a state-of-the-art level, remains the highest priority for the organisation. Regarding the new revision E of the EUR document [1], expected to be issued in the 2nd quarter of 2017, the presentation describes the most significant updates implemented in many fields among which: revised safety requirements taking into account the most recent European and international safety standards issued by WENRA and IAEA, the lessons learnt from the Fukushima accident, including re-evaluated seismic and external natural hazards approach and the most recent international standards, for example for Instrumentation and Control (I&C).

The assessment of new designs is the second main technical activity of the EUR organisation. The MHI EU-APWR design has been assessed against the revision D between 2012 and 2014. New design assessments are in progress (namely Korean KHNP’s EU-APR and Russian AEP’s VVER TOI) and are planned to be ended in 2017–2018. A new applicant is in the file. The presentation briefly recalls the EUR design assessment objectives and process and the progress of the different assessment projects.

The third topic to be covered by the presentation is the interaction between the EUR and the other stakeholders, in particular the other international organisations (ENISS, WNA/CORDEL, WENRA, IAEA, EPRI/URD). The presentation describes how the EUR organisation is connected to these stakeholders and the corresponding cooperation results and future projects.
Enhanced Thermal Conductivity UO2-GaN Fuel, Proposed Fabrication Methods and Multiphysics Fuel Behavior Analysis
Authors: Zhou Wei , Liu Rong , Zhou Wenzhong
Published: October 2017

In this paper, we first propose a novel composite nuclear fuel of UO2-GaN, which has never been reported before, and then its fully coupled multiphysics fuel performance is investigated using the CAMPUS code developed by ourselves. We propose two different fabrication methods to obtain the UO2-GaN fuel, which are Green Granule/Slug Bisque and Spark Plasma Sintering, respectively, resulting in different fuel thermal conductivities. By comparing two kinds of UO2-GaN fuel which are fabricated by two methods, we found that fuel fabricated by Green Granule/Slug Bisque possesses high thermal conductivity and performs well during the reactor operation. The gap width, gap conductance, fission gas release, plenum pressure, deviation of oxygen to metal ratio and displacement are all studied in this work. The performance of this novel fuel is also compared with the traditional UO2 fuel. The UO2-GaN enhanced thermal conductivity composite fuel shows the potential of decreasing the fuel temperature, and improving fuel performance and reactor safety. This makes GaN a good candidate to fabricate composite fuel with UO2 from the thermal standpoint. However, this work is to conduct an exploratory approach to the effect for the GaN addition to UO2 fuel with very limited data. So, further studies are still needed on GaN’s compatibility with UO2, neutronic behavior, fission product retention capabilities and irradiation performance, both on experimental measurements and numerical simulations.
Research on the Inspection Strategy Change Based on Risk-Informed Method in Nuclear Power Plant
Authors: Shuixiang Ye , Qiongzhe Li , Haiying Xi
Published: October 2017

In order to gain more operational benefits, many nuclear power plants will intend to optimize the generation activity to increase generating capability factor, such as, shortening the refueling schedule by optimize the inspection strategy of relevant to SSCs (structure, system, component) during reactor shutdown. Therefore, this paper will introduce the research on the inspection strategy change based on risk-informed method, and provide detailed discussion on change of the RHR heat exchanger inspection strategy by employed the above mentioned method in one nuclear power plant. The result shows that the optimization of RHR heat exchanger inspection strategy meets all the consideration of the principles in risk-informed decision-making required by RG 1.174 and the optimization proposal is acceptable.