bluuebunny/crossref_metadata_2025_split huggingface dataset:
Thermal Hydraulics Numerical Simulation and Optimized Design of Passive Residual Heat Removal Heat Exchanger in AP1000
Authors: Xia Huining , Lu Daogang , Du Zheng , Zhang Yuhao , Fu Xiaoliang , Ma Zhongying , Yang Yanhua , Liang Min 
Published: February 2014

A model of Passive Residual Heat Removal Heat Exchanger (PRHR HX) and In Containment Refueling Water Storage Tank (IRWST) in AP1000 had been scaled by power-volume theory. FLUENT code analysis for the characteristic of thermal stratification phenomenon and natural convective of the scaled model was performed for 2000s of transients to investigate the temperature field distributions and flow field distributions in IRWST. As for the numerical experiment, the distributions had been achieved at different times.
The centralized layout of PRHR HX, which looks like a circle in a section, makes the temperature of fluid outside tubes higher at the location that closer to the upper of the circle in the horizontal tube section and closer to the center of the circle in the vertical tube section, and there may appear nuclear boiling. Then temperature difference between the tube side and fluid side is smaller, causing the heat exchanger efficiency is worse at these locations.
In order to improve the heat transfer effect of PRHR HX, two schemes had been proposed. First, increase horizontal section length and shorten the vertical section length. Second, Flat layout replaces centralized layout. These models were established and simulated by FLUENT. Heat transfer coefficients of “C-tubes” were comparative analyzed to study which kind of the heat transfer effect was better. It contributes to provide some references for improving the heat transfer effect of PRHR HX.
Comparison of Different Acceptance Standard for Radiographic Rounded Indications in Welds
Authors: Wang Cong , Tang GuoXiang , Zhao JiaGuan , Yang HengLiang 
Published: February 2014

Usually, we do radiographic testing according to different standards. Evaluation and acceptance of rounded indications in different standards are different from each others. This paper will evaluate some radiographs based on ASME BPVC, RCC-M and GB/T3323 respectively, and then a conclusion will be drawn that ASME BPVC in rounded indications’ evaluation and acceptance commonly rely on subjective judgment of radiographs evaluator, lack of quantitative evaluation.
This paper will make a comparison on rounded indication evaluations and acceptances in ASME BPVC, RCC-M and GB/T3323, exploring their differences.
The AP1000® PWR Project Moving Toward Completion in China
Authors: Harrop Grenville 
Published: February 2014

The AP1000® pressurized water reactor (PWR) is the first and only Generation III+ nuclear power plant to be granted design certification by the United States Nuclear Regulatory Commission. The initial deployment of this technology has been the construction of dual AP1000 units in each of two coastal sites in the People’s Republic of China (PRC), at Sanmen (Zhejiang Province) and Haiyang (Shandong Province). The contracts for these units were framed to support the PRC’s intention to achieve self reliance in its nuclear supply infrastructure. Westinghouse is implementing its innovative supply chain strategy, “We Buy Where We Build”™, to promote the technology transfer and increasing levels of localization needed as each unit is constructed.
Since the initial contract award in 2007, the Westinghouse Consortium and the purchasers, State Nuclear Power Technology Corporation of China (SNPTC), the Shandong Nuclear Power Company (SDNPC), and the Sanmen Nuclear Power Company (SMNPC) have worked in harmony to build the units using advanced modular construction techniques that reduce construction timescales and associated risks. First-of-a-kind (FOAK) plant components have been manufactured and delivered, including reactor vessels, steam generators, and other safety equipment. With construction and equipment installation in the final stages, the planning and implementation of the pre-operational testing, system turnover, and commissioning are now underway to prepare for fuel load and future commercial operation.
Magnetic Field Analysis of Active Magnetic Bearings for Helium Blower of HTR-PM
Authors: Guo Leilei , Yang Guojun , Shi Zhengang , Liu Xingnan , Yu Suyuan 
Published: February 2014

High-temperature gas-cooled reactor-pebble bed module (HTR-PM) of China, based on the technology and experience of the 10MW high-temperature gas-cooled reactor (HTR-10), is currently in the design and experiment phase. The helium blower is the key equipment in the first loop of the HTR-PM.
Active magnetic bearings (AMBs) are replacing ordinary bearings as the perfect sustaining assembly for the helium blower because they have several advantages: they are free of contact, do not require lubrication, are not subject to the contamination of wear, have endurance, and control performance very well. So the AMB is the appropriate supporting assembly.
The rotor’s length of the helium blower is about 3.3 m, its weight is about 4000 kg and the rotating speed is 4000 r/min. The maximal radial load of the upper AMB or the lower AMB is about 1950kg, and the maximal axial load is about 4500kg. So the axial and radial AMBs must have enough carrying capacity to support the rotor. It is difficult to design the AMB in limited space. Many factors may restrict the AMB design, for example, clearance between AMB and rotor, electric current, and environment temperature, etc.
In this paper, design of magnetic field will be analyzed. Heat loss of AMB will be discussed. The designing scheme of magnetic field will offer the important theoretical base for AMB design of the helium blower of HTR-PM.
Sobol′ Sensitivity Analysis Using a Neural Network Model of a LB-LOCA in the ZION Nuclear Power Plant With CATHARE-2 V2.5 Code
Authors: Fouet Fabrice , Probst Pierre 
Published: February 2014

In nuclear safety, the Best-Estimate (BE) codes may be used in safety demonstration and licensing, provided that uncertainties are added to the relevant output parameters before comparing them with the acceptance criteria. The uncertainty of output parameters, which comes mainly from the lack of knowledge of the input parameters, is evaluated by estimating the 95% percentile with a high degree of confidence. IRSN, technical support of the French Safety Authority, developed a method of uncertainty propagation. This method has been tested with the BE code used is CATHARE-2 V2.5 in order to evaluate the Peak Cladding Temperature (PCT) of the fuel during a Large Break Loss Of Coolant Accident (LB-LOCA) event, starting from a large number of input parameters.
A sensitivity analysis is needed in order to limit the number of input parameters and to quantify the influence of each one on the response variability of the numerical model. Generally, the Global Sensitivity Analysis (GSA) is done with linear correlation coefficients. This paper presents a new approach to perform a more accurate GSA to determine and to classify the main uncertain parameters: the Sobol′ methodology.
The GSA requires simulating many sets of parameters to propagate uncertainties correctly, which makes of it a time-consuming approach. Therefore, it is natural to replace the complex computer code by an approximate mathematical model, called response surface or surrogate model. We have tested Artificial Neural Network (ANN) methodology for its construction and the Sobol′ methodology for the GSA.
The paper presents a numerical application of the previously described methodology on the ZION reactor, a Westinghouse 4-loop PWR, which has been retained for the BEMUSE international problem [8]. The output is the first maximum PCT of the fuel which depends on 54 input parameters. This application outlined that the methodology could be applied to high-dimensional complex problems.
Preliminary Neutronics Design and Analysis of the Bit Helium Cooling Ceramics Blanket for CFETR
Authors: Li Jia , Liu Songlin , Ma Xuebing , Pu Yong , Chen Xiangcun 
Published: February 2014

CFETR is a Tokamak fusion engineering test reactor whose concept design is being developed in China. It is a key issue for breeding blanket design to attain tritium self-sufficiency as one of important missions of CFETR. This paper presents a preliminary neutronics design and analysis employing a BIT (breeder inside tube) type helium cooling ceramics blanket (HCCB) design concept as one of CFETR blanket design candidates. Firstly, 1D reactor model was designed using ceramic breeder Li4SiO4 and beryllium in pebble for multiplier. The primary blanket parameters were optimized to yield the higher tritium breeder ratio (TBR), including the thickness of outboard breeder blanket, enrichment of Li-6 and ratio of Li4SiO4 to Be. Secondly, based on the optimized blanket parameters and plasma parameters, a detailed 3D neutronics calculation model of 22.5° reactor sector was developed, including blanket modules, shield, divertor, vacuum vessel and TF coil. The gap between blanket modules had been taken into account. Finally, a set of nuclear analyses were carried out addressing the key neutronics issues by Monte Carlo neutron-photon transport code MCNP version 5 and the FENDL-2.1 data library.
The preliminary analysis results showed that the global TBR could achieve 1.21 which satisfied the tritium self-sufficiency demand. Nuclear heat, neutronic flux, and distribution of neutron wall loading (NWL) were also analyzed as source terms of the blanket thermal-hydraulics design and reactor nuclear response.
High Breeding Core of a Supercritical-Pressure Light Water Cooled Fast Reactor
Authors: Yoshida Taishi , Oka Yoshiaki 
Published: February 2014

Breeding of plutonium with light water cooling has been studied for many years, but high breeding to meet growing demand for electricity in a developed country has not been accomplished. The purpose of this study is to investigate a high breeding core of Super FBR (supercritical pressure light water cooled fast breeder reactor) with new fuel assemblies consisting of tightly packed fuel rods without gaps, which leads to low coolant to fuel volume fraction. The plant system of a Super FBR is once-through coolant cycle with high head pumps. The coolant flow rate is low due to the high enthalpy rise in the core. It is compatible with the high pressure drop of the new fuel assemblies. Both neutronic and thermal hydraulic design of the core is considered. The challenge of high breeding with light water cooling is to satisfy negative coolant void reactivity, high breeding and low enrichment simultaneously. The core with new assemblies has been designed with the average coolant density of 248 kg/m3. It is achieved by setting 380C inlet and 500C outlet temperature. For satisfying negative void reactivity, a solid moderator layer composed of zirconium hydride (ZrH) rods are adopted in some blanket assemblies. Cross sections of the blanket fuel assemblies with ZrH rods are prepared with assembly-wise calculation, because the pin-wise collision probability calculation overestimates the breeding. MOX fuel is used for seed fuel assemblies.
Three types of core layouts with “radially heterogeneous”, “radiating” and “scattered” seed assemblies have been considered, and “radiating” layout shows best breeding characteristics among them. The seed assemblies in a “radiating” layout are not radially separated so that more numbers of blanket assemblies can be placed in high neutron flux region of a core. Fraction of blanket fuel assemblies with ZrH rods is selected for high breeding.
Super FBR using the new fuel assemblies achieved both negative void and high plutonium breeding.
Numerical Simulation Analysis on the Flow Mechanism of Passive Containment Cooling System
Authors: Song Yang , Li Wei-Hua , Li Sheng-Qiang 
Published: February 2014

In the third generation PWR, passive containment cooling system was one of the key security parts. If accident happened, the blasting valves in the containment would automatically open and the steam would be injected into the containment. When the pressure reached a certain value, the tank on the top of containment would automatically spray. Driven by gravity, the cooling water would flow along outer surface of containment. Steam inside the containment would condense on inner wall and the pressure in the containment would decrease.
A numerical simulation of steam jet in the passive containment cooling system was carried out using computational fluid dynamics methods. Flow mechanism in the process was studied. A simplified three-dimensional model of the passive containment cooling system was built. The hexahedral structured mesh was made. The RNG k-epsilon turbulence model was used in solving the transient Reynolds Navier-Stokes equations by the Fluent. The distributions of velocity in containment at different time were acquired. Reference would be provided for understanding of flow mechanism of gas jet-flow in containment during the early stage of accident.
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.
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.
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.
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.
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.