bluuebunny/crossref_metadata_2025 huggingface dataset:
The Impact of Different Carrier Salts on the Physical Characteristics of the MSFR
Authors: Cheng Maosong , Cheng Jinhui , Dai Zhimin , Chen Kun
Published: February 2014

Molten Salt Fast Reactors (MSFRs) have been the subject of periodic investigations since the early 1960s. MSFRs can be employed to extend fissile resource availability through uranium-plutonium and thorium-uranium breeding, and are highly flexible to be configured into modified open or full-recycle configuration. A single reactor core design may be capable of performing both fissile resource extension and waste disposition missions. The safety aspects of MSFRs are also innovative. So, in order to use the natural resources more efficiently while offering options for a better management of the nuclear waste, there is currently a renewed interest in MSFRs. In France, a conceptual design of reference MSFR with a fast neutron spectrum and based on the Thorium fuel cycle has been proposed.

For improving and optimizing the design of the MSFR, in the present research, several different fluorides salts are chosen as fuel carrier salts to be evaluated. The evaluation scheme is based on the Scale5.1 code system. And the impact of different carrier salts on neutron spectrum and temperature reactivity coefficient has been evaluated in detail. The results indicate that the LiF-NaF-ThF4-UF4 (47.40-30.10-19.95-2.55) composition may be best in all recommended compositions. A significant amount of design work remains to be performed to develop a workable MSFR system, this work can provide a reference for choosing suitable fuel carrier salts in the design process of the MSFR in the future.
Measurement of Gas Temperature in a Large Enclosure by Tunable Diode Laser Absorption Spectroscopy
Authors: Tian Jun , Hao Zulong , Yu Yu , Li Huagui , He Jianjun , Zhang Yilin , Li Jiayuan
Published: February 2014

In order to study the mechanism of mixed-convection and heat transfer when injecting hot water vapor into the containment, it is necessary to measure temperature distribution inside it. To overcome the shortcomings of traditional thermocouples technique during injection experiments of water vapor, a new temperature measurement method based on tunable diode laser absorption spectroscopy (TDLAS) was proposed in this paper. The TDLAS technique is based on the detection of H2O absorption lines with different energies of low levels, and the spectra can be recorded using a laser detector. The data of H2O absorption lines were selected from HITRAN2004 to obtain a pair of optimal absorption lines. A preliminary design of experimental system was presented as well as parameters selection. According to the problems when applying TDLAS, the corresponding solutions were proposed.
Stability Analysis of the MSBR
Authors: Cheng Maosong , Dai Zhimin
Published: February 2014

The molten salt reactor is one of the six advanced reactor concepts identified by the Generation IV International Forum as a candidate for cooperative development, which is characterized by remarkable advantages in inherent safety, fuel cycle, miniaturization, effective utilization of nuclear resources and proliferation resistance. In 1971, Oak Ridge National Laboratory (ORNL) finished the conceptual design of the single fluid Molten Salt Breeder Reactor (MSBR) based on the design, building and operation of Molten Salt Reactor Experiment (MSRE). According to the conceptual design of the MSBR, a model of stability analysis using point kinetics with the motion of the precursors and coupled with the simplified heat transfer mechanism is presented. The model is validated by MSRE benchmarks and applied to simulate the power and temperature transient phenomena of the MSBR initiated by an abnormal step reactivity insertion, an abnormal ramp reactivity insertion and a period reactivity insertion at full-power equilibrium condition without external control action.

The results of stability analysis for the identified three occurrences in normal operation condition show that the power is self-regulating due to the effect on fuel and graphite temperatures while the temperatures tend to new stable value. So, the negative temperature coefficients of reactivity in the MSBR make the reactor capable of stable self-regulation and the MSBR can self-stabilize to a small reactivity perturbation without any external control action.
A Three-Dimensional Rigid-Body Model for Seismic Analysis of the Pebble-Bed HTR Graphite Core Structure
Authors: Zeng Chuan , Wang Haitao
Published: February 2014

Graphite plays an important role in the pebble-bed high temperature gas-cooled reactors (HTR) as moderator, reflector as well as internal structural material. The HTR core consists of a large number of graphite bricks interconnected with keys. It is required that the structural integrity of the HTR core be maintained when subjected to the seismic load. Hence it is important from the viewpoint of seismic design to investigate the seismic responses of the graphite bricks. Considering the pebble-bed HTR has various graphite shapes, a generalized three-dimensional model with the associated computer code is developed to treat these interconnected graphite bricks with arbitrary shapes. In this model, each brick is treated as a rigid body with six degrees-of-freedom: three translational displacements and three rotations around the brick center of gravity. A nonlinear spring dashpot model is applied to present the collision between adjacent bricks and the interaction forces through the key systems. In the numerical tests, the code is verified by comparing predicted responses with exact solutions for two cases and good agreement is observed. The model is then used for the dynamic analysis of the side reflectors of the pebble-bed HTR core under a given seismic load. The calculated response behaviour of the side reflector column is summarized and discussed.
CFD Analysis of Circumferential Temperature Distribution on Fuel Cladding Surface of China Lead Alloy Cooled Research Reactor (CLEAR-I)
Authors: Chen Zhao , Zhou Guangming , Zhou Tao , Chen Hongli , Bai Yunqing
Published: February 2014

Because of heterogeneity of sub-channels geometry, a circumferential temperature distribution will exit in the fuel cladding surface during operating condition of a nuclear reactor, which will increase the maximum cladding surface temperature (MCST). However, this phenomenon cannot be obtained in single channel or sub-channel analyses approach. Computational Fluid Dynamics (CFD) is a good method to study this problem. In this paper, the circumferential temperature distribution on the fuel cladding surface of different sub-channels of China Lead-Alloy Cooled Research Reactor (CLEAR-I) were studied using FLUENT 14.0. Obvious circumferential temperature distributions were presented in results and the maximum circumferential temperature difference is 13°C in edge-channels. This study will improve the precision in the MCST calculation of CLEAR-I and be a good guideline to the optimum design of CLEAR-I sub-channels configurations.
Study on the Flow Distribution at the Core Inlet
Authors: Tang Mao , Chen Liang , Liu Yanwu , Zhang Mingqian , Yu Xiaolei
Published: February 2014

Flow distribution at core inlet is an important criteria of reactor hydraulic performance, affected by flow distribution device. Compared with the Generation II PWR, the AP1000 flow distribution device has a great changed, it depends on the flow skirt. In order to strengthen understanding of flow field, obtain the flow distribution at core inlet, and find the key factor which influences the flow distribution, this paper establishes a whole 3D hydraulic model and studies flow field and flow distribution based on CFX.
Assessing the Attractiveness of SMR: An Application of INCAS Model to India
Authors: Locatelli Giorgio , Mancini Mauro , Belloni Pietro
Published: February 2014

Small Modular Reactors (SMRs) have the potential to be an important component of the worldwide nuclear renaissance. Whilst requiring more diluted investment than Large Reactors (LRs), SMRs are simpler build and operate as well as being suitable for deployment in harsh environmental conditions. In addition, useful by-products such as desalinated water and process heat are generated. The economic competitiveness of SMRs with respect to LRs must be carefully evaluated since the economies of scale label these reactors as not economically competitive. As such, a variety of financial and economic models have been developed by the scientific community in order to assess the competitiveness of SMRs. One of these, the INCAS model (Integrated model for the Competitiveness Assessment of SMRs), performs an investment project simulation and assessment of SMR and LR deployment scenarios, providing monetary indicators (e.g. IRR, LUEC, total equity invested) and not-monetary indicators (e.g. design robustness, required spinning reserve). The work in this paper investigates the attractiveness of SMRs for a given scenario, the Indian state, through application of the INCAS model. India is the second most populated country in the world with rapid economic growth and a huge requirement for energy. There is also both good public acceptance and political support for nuclear power in India, important factors favoring the deployment SMRs in particular. India seems particularly suitable for SMR deployment because (i) its energy intensive industrial sites are located far from existing grids, (ii) rapid growth in the region and (iii) the requirement for plants to provide fresh water for the population, as well as for agriculture and industry. The results show that SMRs have roughly the same financial performance of LRs, however they have a competitive advantage as a result of non-financial factors such as co-generation application, higher local content and better management of the spinning reserves in a country with an electricity deficit.
Study on Comparison Between Inconel 617 and GH3128 as Structural Material Candidates for Intermediate Heat Exchanger
Authors: Shi Yunbai , Yuan Kun , Zhao Xi , Wu Yannan
Published: February 2014

Intermediate heat exchangers (IHXs) are applied to transfer high-temperature heat generated in high-temperature gas-cooled reactor (HTGR) core to the secondary circulation loop. The requirements for IHXs are harsh due to its high operation temperature up to 900–1000°C, as well as long lifetime expectation. Inconel 617, an austenitic alloy based on nickel-chromium, is considered the primary candidate as IHX structural material for application in NGNP program in USA. This paper compares the characteristics of two nickel-base super alloys, namely GH3128 and Inconel 617, in terms of chemical composition, tensile strength and creep behavior. The result shows that GH3128, which was originally developed and deployed for aeronautic blast chamber in China, is also promising for IHXs in nuclear heat application, but much tests have to be done in the future before its successful application in nuclear high-temperature IHXs.
Study on the Behaviors of a Conceptual Passive Containment Cooling System
Authors: Wang Jianjun , Luo Beibei , Guo Xueqing , Sun Zhongning
Published: February 2014

In this paper, a closed loop concept, which is composed of two heat exchangers with same scale, pipes, valves and one tank, has been developed as a passive containment cooling system for a large dry concrete containment. The system is designed to maintain the containment integrity by taking the heat drained into the containment following a severe accident, e.g. LOCA or MSLB. Under different conditions in containment, the system may operate in single phase mode or two phase mode. According to the design limitation of containment, the fixed temperature boundary condition is applied to the system analysis. We have developed the codes for the analysis of the system by ourselves. The operating behaviors of the system are studied numerically from startup to long term operation. In the light of the fact that the fraction of steam in the containment may be changing during the accident scenario, it is reasonable that the heat transfer coefficient will be different. Therefore, the sensitivity analysis of the heat transfer coefficient is also performed. Based on the results and corresponding analysis, it can be concluded that the system may be utilized to meet the design purpose for the containment integrity requirement.
Numerical Simulation and Analysis of Temperature Fluctuation in Core Outlet Zone of China Experimental Fast Reactor
Authors: Zhu Huanjun , Xu Yijun
Published: February 2014

As for the fast reactor, the temperature fluctuation in core outlet zone in normal operation was an important and typical thermal hydraulic phenomenon. In certain conditions, these temperature fluctuations can lead to thermal-mechanical damage to the Upper Core Structures (UCS). In this paper, the temperature fluctuation in Core Outlet Region of China Experimental Fast Reactor (CEFR) was numerically simulated and researched using the CFD software FLUENT. In the simulation process, LES turbulence method was used for the turbulent model selection. For the mesh generation, 1/4 sector model was built considering the layout of the symmetry and very small-size cell was used for satisfied with the Corant number. At the same time the core outlet temperatures and flows of the coolant in each subassembly under reactor rated power was used as the boundary conditions. The simulation shows that the temperature fluctuation was mainly concentrated in the zone between the fuel subassembly and control rod assembly for the large temperature difference of these subassemblies CEFR was operated in normal conditions. The largest fluctuation amplitude was 19°C and the remarkable frequency was below 5Hz. It belongs to typically low frequency fluctuation. The conclusion was useful for further experimental work and reactor operation.
The Analysis of AP1000 Depressurization During Small Break LOCA
Authors: He Shuangji
Published: February 2014

Based on two different ways-the fluid energy equation and the fluid compressibility, the lumped parameter equations of Reactor Coolant System (RCS) depressurization during small break LOCA (SBLOCA) are derived and the two different derivations could verify each other by giving the identical result. And then, the transient equation is derived from the steady equation and is solved numerically. By comparing the results of steady and transient equations with that of NOTRUMP, it proves the equation’s rationality and validity. These lumped parameter equations could give the quantitative insight into the influence of the venting of water or steam through the Automatic Depressurization System (ADS) valves/break during LOCA depressurization phase.

During the SBLOCA depressurization phase of AP1000, the ADS4 valves are the most important venting paths. There are four ADS4 valves, which are mounted respectively on two sides of hot legs. Westinghouse originally assumed that the most limiting single failure was one of ADS4-B valves failure to open. But in truth, the most limiting failure should be one of ADS4-A valves failure to open, which has been validated by SNERDI with NOTRUMP calculations. However, it’s not easy to analyze the reason until it’s recognized that the key factor is the venting of steam, which is the theoretic conclusion above. This paper will give some insight into the limiting single failure criterion of SBLOCA, based on the theoretic conclusion above and the sensitivity analysis of NOTRUMP calculations.
Validation of a Hybrid Surface Renewal Theory Based HTC Model for the Simulation of Condensation Induced Water Hammer
Authors: Ceuca Sabin Cristian , Macián-Juan Rafael
Published: February 2014

A hybrid Heat Transfer Coefficient (HTC) model has been developed based on the Surface Renewal Theory and implemented into a system code. The model accounts for both large and small eddies for the removal of heat from the two-phase interface towards the subcooled liquid bulk. The hybrid approach for the calculation of the Surface Renewal Period will increase the applicability region of the individual Surface Renewal Theory (SRT) based HTC for various flow regimes. Validation of the new Hybrid HTC model is performed against experimental data.
Interference Fit Design for Rotor Components of the Active Magnetic Bearing
Authors: Liu Xingnan , Yang Guojun , Xu Yang , Li Yue , Shi Zhengang , Yu Suyuan
Published: February 2014

The method to design the interference fit for the rotor components of the active magnetic bearing (AMB) was proposed. The rotor components include the thrust collar, the radial bearing rotor components and the sensor rotor components. Four main influencing factors and the machining accuracy are considered during the designing, such as the centrifugal force, the loadings, the thermal expansion coefficient, the electromagnetic force, and the coupling surface roughness. Some calculation examples were presented. These examples are base on the AMB of a helium blower, which is going to be applied in the High Temperature Reactor-Pebblebed Modules (HTR-PM) in Rongcheng City, Shandong province. Both the analytical calculation and the finite element analysis were carried out, and then were compared. The method of design based on the simple analytic formula was proved to be available.
Modeling of Thermal Fatigue Crack for Enhancement of Electromagnetic Nondestructive Evaluation of Nuclear Power Plant
Authors: Wang Jing , Yusa Noritaka , Pan Hongliang , Kemppainen Mika , Virkkunen Iikka , Hashizume Hidetoshi
Published: February 2014

The present study investigates modeling of thermal fatigue crack by the finite element method from the view point of eddy current testing. Three artificially produced thermal fatigue cracks introduced into Inconel 600 specimens are prepared. Eddy current signals are gathered by a differential type plus point probe with 25 kHz and 100 kHz. Subsequent destructive tests show the profile of these cracks. In the numerical simulation thermal fatigue crack is modeled as a region with constant width, uniform conductivity and real profile. Results of the study combining the previous study on modeling of thermal fatigue crack in type 304 stainless steel specimen reveal that thermal fatigue crack, unlike stress corrosion crack, generally should be modeled as an almost nonconductive region regardless of frequencies. Furthermore the resistance of thermal fatigue crack defined as width divided by conductivity is almost unchanged, even though the appropriate conductivity and width are changing with the frequency.
The Thermal-Mechanical Analysis of the BIT HCCB Blanket for CFETR
Authors: Ma Xuebing , Liu Songlin , Li Jia , Pu Yong , Chen Xiangcun
Published: February 2014

CFETR is an “ITER-like” China Fusion Engineering Test Reactor. The BIT (Breeder insider tube) HCCB (Helium cooled ceramic breeder) blanket has been designed as one option for CFETR. Its unique feature is that a BIT structure is adopted for the blanket tritium breeder unit. The breeder unit is an assembly of three coaxial coil-pipes, internal coil-pipe is filled with the tritium-breeding material, and the coaxial coil-pipe is embeded in the beryllium pebble. In order assess the feasibility of the BIT HCCB blanket design, a 3D finite element model slice of the BIT HCCB blanket is developed, and the thermo-hydraulic calculation was performed, and thermo-mechanical computations were carried out. And then the primary stress and secondary stress were evaluated according to Structural Design Criteria for ITER In-vessel Components (SDC-IC). This paper presents relevant analyses and results. The preliminary results shows current design satisfies with allowable limit of material and requirement of SDC-IC code.
Improvement of Temperature Evaluation Model of Biological Shielding Concrete for HTTR Test Simulating LOFC With VCS Inactive
Authors: Takada Shoji , Yanagi Shunki , Iigaki Kazuhiko , Shinohara Masanori , Tochio Daisuke , Shimazaki Yosuke , Ono Masato , Sawa Kazuhiro
Published: February 2014

HTTR is a helium gas cooled graphite-moderated HTGR with the rated power 30 MWt and the maximum reactor outlet coolant temperature 950°C. The vessel cooling system (VCS), which is composed of thermal reflector plates, cooling panel composed of fins connected between adjacent water cooling tubes, removes decay heat from reactor core by heat transfer of thermal radiation, conduction and natural convection in case of loss of forced cooling (LOFC). The metallic supports are embedded in the biological shielding concrete to support the fins of VCS. To verify the inherent safety features of HTGR, the LOFC test is planned by using HTTR with the VCS inactive from an initial reactor power of 9 MWt under the condition of LOFC while the reactor shut-down system disabled. In this test, the temperature distribution in the biological shielding concrete is prospected locally higher around the support because of thermal conduction in the support. A 2-dimensional symmetrical model was improved to simulate the heat transfer to the concrete through the VCS support in addition to the heat transfer thermal radiation and natural convection. The model simulated the water cooling tubes setting horizontally at the same pitch with actual configuration. The numerical results were verified in comparison with the measured data acquired from the test, in which the RPV was heated up to around 110 °C without nuclear heating with the VCS inactive, to show that the temperature is locally high but kept sufficiently low around the support in the concrete due to sufficient thermal conductivity to the cold temperature region.
Optimization of CANDU6 Heat Transport System Inventory Control During On-Power Refueling
Authors: Wu Zhigang , Lin Chuanqing , Liu Yang
Published: February 2014

The Qinshan Phase III (PHWR) Nuclear Power Plant has two 728 MWe CANDU6 reactors and it was found that the pressurizer steam bleed valves operated frequently during on-power refueling. The pressurizer steam bleed valves started to leak internally due to steam cutting, and the internal leakage further lead to frequent operation of degasser condenser spray valves. By investigating the response of CANDU 6 reactors during refueling operations, it was confirmed that the fueling machine flow injection caused the pressurizer level to rise during online refueling, compressing the pressurizer steam space, which is the root cause of PHT system over-pressurization. Based on the root cause analysis, a solution was proposed by adding a feed-forward control term to the PHT system inventory control logic, and the feasibility of this improvement was verified in unit online tests. After implementing the design modification, the system performances also proved the optimized logic change is effective. The research on this subject is not only beneficial to improving the safe and reliable operation of the PHT system in CANDU 6 reactors, but also provides a reference for improving the design of CANDU reactors of the next generation.
Feasibility Study on Application of a Self-Formed Flow Field Downstream of Elbows to Divertor Cooling
Authors: Kodate Shoichi , Kubo Tatsuya , Ebara Shinji , Hashizume Hidetoshi
Published: February 2014

In this study, the characteristic of the swirling flow was analyzed in detail in terms of flow field by means of a visualization experiment using matched refractive index PIV measurement to evaluate the applicability of the swirling flow generated downstream of a three-dimensionally connected dual elbow to the divertor cooling. The dual elbow used in the experiment comprises two 90-degree elbows with the same curvature connected directly in three-dimensional configuration. From the experiment, it was found that strong swirling velocity component appears locally near the pipe wall downstream of the second elbow. Moreover, although the strength of the swirling flow changed gradually as it flowed downstream, it attenuated little even 8D downstream of the dual elbow, where D was the diameter of the piping. Therefore, this swirling flow is expected to survive for a considerable distance downstream of the elbow, and the applicability of this flow field to divertor cooling can be promising. Furthermore turbulence quantities such as Reynolds stress were analyzed in terms of heat transfer performance. Since there were some regions where larger Reynolds stress than a developed turbulent pipe flow was observed near the pipe wall, high heat transfer is expected there.
Proposal of a Novel Blanket System Using LiF-BeF2-LiI for a Fusion Reactor
Authors: Shishido Hiroki , Han Youngmin , Matsui Kentaro , Yusa Noritaka , Hashizume Hidetoshi
Published: February 2014

This study proposes LiF-BeF2-LiI as a molten salt for a liquid blanket system of a fusion reactor. Firstly, the melting point of LiF-BeF2-LiI is measured as a function of the molar ratio of LiI. The experimental results confirmed that the melting point decreases to 398.2 °C when LiF:BeF2:LiI = 19:9.5:71.5, which is consistent with an earlier report. Experiments to evaluate the density and viscosity of LiF-BeF2-LiI are also carried out. The experiments revealed that the density of LiF-BeF2-LiI increases with the molar ratio of LiI, and the viscosity of LiF-BeF2-LiI (LiF:BeF2:LiI = 2:1:1) is somewhat larger than that of LiF-BeF2 (LiF:BeF2 = 2:1) but much smaller than that of LiF-BeF2 (LiF:BeF2 = 3:2). Monte Carlo numerical simulations using MVP-2.0 with JENDL-4.0 data library are conducted on the basis of the measured density to discuss the feasibility of a blanket system using LiF-BeF2-LiI. The numerical simulations demonstrate that LiF-BeF2-LiI can realize TBR>1.4, which is required for a fusion blanket system, with Be pebble layer and enriching 6Li.
A Survey on the Quantitative Evaluation Methods of Software Reliability of Digital I&C Systems at NPPs
Authors: Liu Ying , Yang Ming , Li Fengjun , Zhang Chao , Hashim Muhammad
Published: February 2014

The instrumentation and control (I&C) systems are important parts of the Nuclear Power Plants (NPPs). They are the key to safety, reliability and economic operation of the nuclear power generating units. With the development of systems at NPPs. Software reliability is important because software failure is usually considered as one form of common cause failures in digital I&C systems. The reliability analysis of the digital I&C systems, particularly quantitative evaluation on the software reliability analysis of the digital I&C systems belongs to a great challenge. This paper provides an overview of the ongoing research activities on a reliability analysis of digital I&C systems of NPPs, quantitative evaluation methods of software reliability, the advantages and disadvantages of the approaches, and research status at home and abroad. China is using the untraditional reliability analysis method of Multilevel Flow Model (MFM) to evaluate software reliability in digital I&C systems.
A Study on the Influences of Reactivity Feedback Mechanisms in China Experimental Fast Reactor Unprotected Transients
Authors: Wu Yuanyu , Yu Hong , Ren Lixia , Hu Wenjun , Qian Hongtao
Published: February 2014

Inherent safety properties of reactor have always played an important role in severe accidents preventing and consequences mitigation. With proper design, reactivity feedback mechanisms can bring benign reactivity feedbacks to the reactor core during unprotected transients, thus contributing to the severe accidents mitigation. In overpower transients, the increasing power causes the fuel temperature to increase, which directly brings fuel Doppler feedback and core axial expansion feedback. In unprotected loss-of-flow accidents, as the flow rate decreases, the mismatch of power and flow causes the increase of coolant temperature, thus directly resulting in the coolant reactivity, core radial expansion as well as the control rod driveline expansion feedbacks. Through the simulation of China Experimental Fast Reactor (CEFR) unprotected transients, the influences of different reactivity feedback mechanisms have been investigated and analyzed. The coolant reactivity exhibits significant negative feedback and makes the dominant contribution in controlling the reactivity in both UTOP and ULOF transients.
Effect of FCI Electrical Conductivity on MHD Flow in Channels With Multi-Layer Inserts for Fusion Blanket
Authors: Zhang Heng , Zhou Tao , Meng Zi , Chen Hongli
Published: February 2014

Fusion blanket is a key component for energy transformation and extraction in fusion reactor, in many kinds of blanket designs, the flow channel insert (FCI) made of a silicon carbide composite (SiCf/SiC) is a key element, which serves as electric and thermal insulator. In the high temperature liquid metal blanket (HTL) of fusion-based hydrogen production reactor (named FDS-III),. The multilayer flow channel inserts (MFCI) are put into the breeding zone to achieve the LiPb outlet temperature around 1000°C for high efficient production of hydrogen. However, the flow of liquid metal meandering through MFCI will cause strong magnetohydrodynamic (MHD) effect under the influence of the intense fusion magnetic field and the liquid metal in the neighboring channels will effect each other. When the FCI conductivity varies, the influence varies coresspondingly. So MHD effect of different FCI electrical conductivity is a key issue which should be concerned in HTL blanket.

In this work, the numerical study aims at clarifying the MHD effect for liquid metal LiPb in the multi-layer flow channel inserts of different conductivity by using a code named MTC-H 2.0. The MHD flows with Special geometry of the HTL flow channels with MFCI were considered, velocity and FCI electrical conductivity’s effect on electromagnetic coupling are discussed.
Experimental Study of a Gas Separator for MSR Gas Removal System
Authors: Zhang Nana , Yan Changqi , Sun Licheng
Published: February 2014

An independent gas removal system was firstly put up by Oak Ridge National Laboratory (ORNL) for Molten Salt Breeder Reactor (MSBR) to remove fission gas such as Xenon etc. from molten salt (LiF–BeF–UF), where the gas separator is a key equipment for which high separation efficiency and low pressure drop are required. Swirl vanes and recovery vanes are installed around the inlet and outlet of the gas separator. With reference to the design of ORNL, a new gas separator was designed for a conceptual design of a small molten salt research reactor.

An experimental loop was also fabricated to test the local pressure drop and separation efficiency of the new designed gas separator using of water and air as working fluid. The separation efficiency is measured by comparing the gas flow rate from the bubble generator and that at the outlet of the separator using of high speed video camera. The tests were conducted with the water flow rates covering 10–30 m3/h and an average void fraction of 0.3%. Besides, the shape of gas cores is found necessary to estimate the working performance of gas separators. High separation efficiency comes with a stable and straight gas core, and the gas will be taken off from two ends of the gas separator. What’s more, the result shows that low concentration of gas does not increase the pressure drop compared with the condition of liquid only, except when it exceeds a certain level. Finally, local flow resistance of the gas separator was also obtained under low gas concentration condition, showing that the resistance factor is nearly unvaried with a value of 5.
RELAP5 Core Modeling Study for Level 1 PRA Thermal-Hydraulic Analyses
Authors: Yang Changjiang
Published: February 2014

Thermal-hydraulic (T/H) analyses are used to support the level 1 Probabilistic Risk Assessment (PRA) success criteria and the manual operation time. To address the multiple-failure accident scenarios that are considered in the PRA, usually numerous T/H analyses were performed. So it is meaningful to develop a relative simple T/H model with acceptable accuracy for level 1 PRA T/H analyses. To achieve this object, the core modeling effects on the core damage progression were studied according to ASME/ANS RA-Sa-2009. Two types of core modeling methods were studied, including single channel core modeling and multi-channel core modeling. For the single channel core modeling, the study was focused on the axial nodes number effect. For the multi-channel core modeling, the cross-flow effects were studied. Several cases were calculated on a 3-loop PWR medium size break LOCA core damage scenario with Relap5/MOD3.2. Some key parameters related to the core state, such as peak cladding temperature (PCT), core water level and coolant inventory, were compared and analyzed. A kind of core modeling for level 1 PRA T/H analyses was suggested at the end.
Instrument Calibration in the Nuclear Power Plant Refueling Cycle Extension Project
Authors: Wang Yugui
Published: February 2014

Nuclear power plant refueling cycle extension is an effective way of nuclear power plant to improve economic benefit. This paper, about the Qinshan Nuclear Power Plant phase II 1&2 unit, combination the practical refueling period experience of 12 months refueling period was extended to 18 months, focus on the instrument calibration extension methods of the refueling cycle extension project. Instrument in nuclear power plant running plays very important role, various instruments in the extension period after extended whether their properties still remain available and reliable, require detailed and reliable demonstration. Considering the Qinshan Phase II instrument equipments are purchased from multinational nations and different manufacturers, in this paper, the various demonstration methods are analyzed and compared, from all these methods, we chose the best one as the primary method in the actual demonstration, and use the others as supplemented methods by the demonstration strategies, and the main methods are detailed in the description and examples, so these can ensure that the instrument refueling cycle extension strategy is reliability.
Investigation on Iodine Release Behavior During the Operation of High Temperature Engineering Test Reactor (HTTR)
Authors: Ueta Shohei , Inoi Hiroyuki , Mizutani Yoshitaka , Ohashi Hirofumi , Iwatsuki Jin , Sakaba Nariaki , Sawa Kazuhiro
Published: February 2014

Japan Atomic Energy Agency (JAEA) has planned to investigate on iodine release behavior from fuel through the testing operation of High Temperature Engineering Test Reactor (HTTR) in order to contribute to the reasonable estimation of the radiation exposure necessary for the realization of HTGR in the future. In this test, the fractional release of iodine will be measured and evaluated by measuring xenon isotopes, the daughter nuclides of iodine isotopes, in the primary coolant sampling under the loss-of-forced cooling (LOFC) test by which the primary coolant circulator is shut down and/or the manual scram test of HTTR. In parallel, the local area of primary coolant circuit where iodine is plated-out will be evaluated. This paper describes the testing plan and the preliminary analytical study on the release behavior of iodine and xenon isotopes through the operation of HTTR.
Transient Analysis of Passive Decay Heat Removal System for CEFR
Authors: Qian H. , Li Z. , Ren L.
Published: February 2014

A passive decay heat removal system (DHRS) has been installed on Chinese Experimental Fast Reactor (CEFR). To well predict the thermal-hydraulics behavior of DHRS in transient, an integrated model has been developed for CEFR by using OASIS code. The model included the main thermal transfer system and DHRS circuit. The transient analysis of loss of off-site power (LOSP) accident with various initial steady states has been performed. The calculation results show that the initial steady state does not essentially effect on the peak cladding temperature in the core.
Safety Analysis on Mitigation Overpressure ATWS in VVER-1000 NPP
Authors: Qi Ting , Yang Changjiang
Published: February 2014

Safety analysis on operating nuclear power plant (NPP) plays an important role on nuclear energy application, especially after the severe accident of Fukushima plants. This paper focuses on one of the operating NPPs in China, the TianWan NPP Unit1&2. TianWan Unit1&2 belong to VVER-1000 NPP type, which have special design characteristics and different safety migrating methods comparing with other domestic PWR NPPs in China. Calculations and analyses were made to give thermal hydraulics support to Level-1 PSA of this VVER type PWR on the anticipated transient without scram (ATWS) accidents. The calculation of loss of main feed water ATWS was carried out using the RELAP5 code to evaluate the intrinsic safety mainly impacted by fuel and moderator. The model also considers reactivity introduction caused by the change of boron concentration to find out the influence of the emergency boron injection system (JDH) on mitigating the coolant system over pressure. The paper gives out the success criteria of the safety valves of pressurizer (PRZ) with the critical moderator temperature coefficient (MTC) value according to the condition of coolant system pressure being under the pressure limit. It is indicated that the JDH system can play an important role on mitigating the over pressure of coolant system in the late phase of the transient.
On the Influence of Controller Parameters on the Performance of Rotating System With Magnetic Bearings
Authors: He Ying , Sun Zhe , Zhao Jingjing , Shi Zhengang , Shi Lei , Yu Suyuan
Published: February 2014

Rotating machineries, in High Temperature Gas-cooled Reactor (HTGR), are usually supported by magnetic bearings. In the paper, we provide a research on the relationship between the performance of machinery and the controller parameters. More specifically, this paper concentrates on the modified Proportion-Integral-Differential (PID) controller, which is characterized by five parameters: two poles, two zeros and the gain. The system performance is evaluated by the closed-loop poles. To perform quantitative investigation, a flexible-rotor-magnetic-bearing system model is established. Based on this model, the Bode plot and root locus method are applied to analyze the effect of controller’s parameters (locations of zeros and poles along with gain) on the closed-loop performance (stability, damping ratios of closed-loop poles). The results of this paper can offer some design criterion of such controllers and help the users of rotating machinery with magnetic bearings to understand the relationship between the performance of machinery and the controller parameters.
The First Criticality Experiment of CEFR
Authors: Yang Xiaoyan , Yu Hong , Gang Zhi , Zhou Keyuan , Zhao Jinkun , Chen Yiyu , Hu Yun , Tian Hechun , Li Zehua
Published: February 2014

As the first fast reactor in China, the first criticality of CEFR (China Experimental Fast Reactor) was successfully achieved on 21st July 2010. The first criticality test of CEFR consists of two processes: the fuel loading for net core and the criticality process. In this paper, some detailed information about this test is introduced, including methods, procedures, results, etc. The test results are compared with the theoretical analysis. Comparison shows that the test results match the theoretical analysis very well, which demonstrates that the codes utilized for theoretical analysis are capable of the basic calculation of CEFR.
Calculation of Bowing Reactivity of Core Assemblies in Liquid Metal Fast Reactors
Authors: Huo Xingkai , Yu Hong , Cao Pan , Xu Li , Hu Yun , Yang Xiaoyan , Zhang Jian
Published: February 2014

In liquid-metal-cooled fast reactors, the temperature rise and its gradients over the core region, sometimes with an addition of the accumulated irradiation creep and swelling, result in a radial core expansion and the bowing of subassemblies, both of which lead to the radial displacements of fuel and reflector materials.

In spite of the small magnitude of such displacements, the reactivity change due to radial core expansion and bowing of subassemblies during a transient, such as unprotected loss-of-flow (ULOF), is significant in magnitude and plays one of the most important roles among all inherent safety features; the focus is on the change of the power-to-flow ratio (P/F) from 1.0 to 2.0, during which a large temperature gradient is induced.

A 3-D code, named PERMOV, is developed for the calculation of such reactivity feedback, at present as a part of the Neutronics Analysis System (NAS); NAS is a 3-D nodal code, independently developed by researchers in China Experimental Fast Reactor; and at the same time, the code PERMOV has a high portability, making it available to run with other mature core physics codes.
Neutronics Design Study on Ignition Zone of Travelling Wave Reactor
Authors: Zhang Jian , Yu Hong , Gang Zhi , Zhou Keyuan , Hu Yun , Duan Tianying
Published: February 2014

Traveling wave reactor is a kind of nuclear reactor that can convert fertile material into fissile fuel as it runs using the process of nuclear transmutation. In the ignition stage of traveling wave reactor, the core performance is especially complex, since the fissile fuel and fertile material is put in different regions at the beginning. And the distribution of power density will change severely with burn-up during the reactor operation. It is an important part of the traveling wave reactor study to optimize the design of the ignition stage.

In this paper, based on a two-dimensional RZ geometry model, some schemes with different sizes and compositions of the ignition zone, middle ignition zone position design and burnable neutron poison addition are simulated and analyzed. Finally, an optimized core design with multi-zone configuration and burnable neutron poison addition is shown. Some design outlines are introduced for further study.
Research of the Effects on a Nuclear Cable Material Irradiated by Beta/Gamma Irradiation
Authors: Zhong Weixia , Sun Jiansheng , Liu Jinping , Sun Ping
Published: February 2014

In this paper, irradiated samples by different irradiation doses of the beta/gamma radiation from a selected nuclear cable material, which were studied by tensile machine, DSC and DMTA. The research results show that the elongation at break (EAB) drops with absorbed dose of beta/gamma radiation and at the same dose the EAB drop for beta-irradiated samples is higher than that for gamma-irradiated samples, which means that this material becomes more brittler with more irradiation. The OITP value for the irradiated samples exhibits pronounced drop with absorbed beta/gamma irradiation dose. And moreover, the OITP value of these gamma irradiated samples drops faster than that of those beta irradiated samples. DMTA researches show that the storage modulus (E′) and the loss factor (tanδ = E″/E′, E″ represents the loss modulus) of the irradiated samples present higher values with the increase dose of beta /gamma radiation. It is interest that the E′ behaves firstly an increase, then decreases by rising the temperature within the temperature range of −30–20 °C, and this trend becomes pronounced with increase of irradiation dose, combined with the activation energy, correspond to the glass transition process, for all the irradiated samples which hints that the glass transition process may be hindered in terms of a higher activation energy, but the movement unit could be smaller with a lower Tg, as a result of the beta/gamma irradiation.
Bubble Growth During Subcooled Forced Convective Flow Boiling
Authors: Zheng Qiang , Gao Puzhen , Hu Jian
Published: February 2014

The inception, growth and collapse of vapor bubbles were observed and measured by using visual method under subcooled flow nucleation. The test section was a single-side heated rectangular channel by the scale of 2×40×700mm and the working fluid was clean water. The working condition was set as: the inlet subcooling Δ Tin = 330 °C, the mass flux m = 694kg/(m2s), the heat flux q = 210kW/m2 and the absolute pressure p = 0.22MPa. A high speed camera was used to record the bubble behaviors at the speed of 5000fps (frame per second). The results showed that the bubble lifetime was from 0.4ms to 2.2ms and the fraction of bubble with short lifetime was bigger than that of long lifetime. The bubble’s average diameter showed a linear relationship with the lifetime and it was also found that the sliding bubble would enhance heat transfer.
Failure Analysis of the Branch Connection Leakage on Primary Pipeline of the Primary Heat Transport System and the Maintenance Strategy Investigation for the Similar Connections
Authors: Zhao Liang , Zhao Ting , Chu Fengmin , Hu Jianqun , Wang Xin
Published: February 2014

An primary pipeline branch connection of the primary heat transport system leaked in Dec, 2007. Physical and chemical testing are conducted on the leakage location of the branch connection. The primary failure cause was concluded that transgranular stress corrosion cracking (SCC) on stainless steel side of the branch connection dissimilar metal weld lead to the leakage. There are total 8 similar pipelines in each unit of Qinshan III. The preventive repair strategies for this kind of pipelines are investigated. The similar defects could be detected with ultrasonic testing in the planned shutdown outage. The corresponding precaution condition establishing methods for the similar branch connections with 3 typical positions are investigated and defined. A maintenance strategy to repair the dissimilar weld on the branch connection was developed.
Experimental Study on the Fluctuation Characteristics of Pressure Drop and Mass Flux of the Two-Phase Flow in Narrow Channel
Authors: Chen Deqi , Wang Qinghua , Duan Zhengang , Pan Liang-ming
Published: February 2014

In this paper the study focuses on a visual investigation on the gas-water two-phase flow in a vertical circular narrow channel with 2 mm inner diameter under atmospheric pressure. Experiments were carried out with different working conditions, including different gases as gas-phase working fluids such as nitrogen, air, carbon dioxide and argon, and the gas flow rate, Q, varied between 0 ml/s (single liquid phase flow) to 9.0 ml/s, and the liquid mass flux, G, varied between 581.3 kg/m2s to 3201.8 kg/m2s. The influence of liquid mass flux, gas flow rate as well as Eo number and Mo number (using these two non-dimensional parameters to specify the effect of gas-phase properties) on the fluctuation of pressure drop and mass flux were investigated in this study. It is found that the pressure drop increases along with increasing liquid-phase flow rate with identical other working conditions, and the corresponding flow patterns are slug flow even though the liquid-phase flow rates are different. However, the pressure drop decreases at first and then increases along with gas-phase flow rate, with constant liquid flow rate (liquid mass flux), and the corresponding flow patterns include slug flow, slug-annular flow and annular flow. Based on the experimental result, it is also found that the smaller Eo number and Mo number of the gas-phase working fluid, the smaller the fluctuations of the pressure drop and mass flux would be due to the gas-phase working fluid is different.
Bubble Generation From an Orifice in Upward and Downward Liquid Flow
Authors: Kashinsky O. N. , Lobanov P. D. , Chinak A. V. , Vorobyev M. A.
Published: February 2014

The dependencies of detachment bubble diameter on liquid velocity and gas flow rate through a capillary placed in the liquid flow in a channel with cross section of 10×10 mm were obtained. An estimate of bubble detachment frequency was made. The effect of bubble detachment regime on the distribution of bubble size in the flow is discussed.