bluuebunny/crossref_metadata_2025 huggingface dataset:
Development of Prediction Technology of Two-Phase Flow Dynamics Under Earthquake Acceleration: 7 — Measurement of Velocity Profile Around Bubble Under Flow Rate Fluctuation
Authors: Takano Jun-ichi , Okachi Satoshi , Monji Hideaki , Kaneko Akiko , Abe Yutaka , Yoshida Hiroyuki , Takase Kazuyuki
Published: February 2014

Earthquake is one of the most serious phenomena for safety of a nuclear reactor, especially in Japan. Therefore, nuclear reactors were contracted considering structural safety for a big earthquake. On the other hand, thermal-fluid behavior of the nuclear reactor is not fully confirmed under the earthquake. The behavior of nuclear reactor may be affected by the fluctuation of gas-liquid two-phase flow caused by the earthquake but not studied enough. For example, the bubble behavior under the fluctuation caused by the earthquake is not clear. In order to clear the two-phase flow behavior under the earthquake, it is studied systematically in the project of “Development of Prediction Technology of Two-Phase Flow Dynamics under Earthquake Acceleration”. In the project, bubbly or plug flow with flow rate fluctuation in a fixed horizontal pipe is studied experimentally. The bubbly or plug flow under the structural vibration but without the flow rate fluctuation is also studied. Here, the structural vibration done by using the pipe on the oscillation table. Moreover, the numerical code which can simulate the flows measured in the experiments is under development. In this paper, the result on the bubbly flow under the flow rate fluctuation is shown. The flow was bubbly flow and/or plug flow in a horizontal circular pipe. The working fluids were water and nitrogen gas. The water was driven by a pump and the flow rate fluctuation was given by a reciprocating piston attached to the main flow loop. In the study, the behavior of horizontal gas-liquid two-phase flow under the flow rate fluctuation is investigated with image processing and particle image velocimetry (PIV). As the result, the bubble’s deformation under the flow rate fluctuation was investigated by the image processing. The velocity profile around the bubble under flow rate fluctuation was measured with PIV. The velocity field around the deformed bubble showed the characteristic profile. Therefore, the velocity field changing with the flow rate fluctuation affected the bubble deformation.
Extension of Load Follow Operation of PWRs Without Boron Adjustment
Authors: Wang Li , Liu Yang , Tai Yun , Zhang Yawei , Tang Zhenpeng
Published: February 2014

Due to development of electric power system, the proportion of nuclear power and its needs for Load follow operation have become large. Then there would be higher requirements to the power control ability of the nuclear power system.

The project with China’s own brands and using the advanced pressurized water reactor (APWR) nuclear power technology, CPR1000, adopts a control method called Mode G, using G banks, N banks, R bank and soluble boron to perform the load follow control. The common Point Model, widely used in conventional core modeling, cannot successfully research the axial power distribution. But in this paper, a two-node dynamic function model was built, taking CPR1000 as an example, considering the coupling coefficient and mutual influence. The transient parameter values were obtained by the steady-state calculation of single channel with its original structure parameter and operation parameter. Then the core system simulation of CPR1000 was modeled. In order to carry out the load follow operation without the boron adjustment, known as BTP mode, the control banks were regrouped to control the average temperature by M banks and axial offset by AO bank. The last, the optimal control implementation was connected to the core simulation platform to operate with the variable daily load. The feasibility and effectiveness of the control policy and the simulation system are validated by the calculation results via one-dimensional in axial computer code APOC.

Results of computations performed for a reference reactor were present, giving the possibility that the optimal control policy could stretches the capability of the reactor to follow an average daily load curve. The paper demonstrated in principle that the BTP mode was feasible for CPR1000.
Simulation on the Decontamination Efficiency and Factor of Radioactive Waste Water Disposed by Supersonic Wave
Authors: Zuo Guangxia , He Bin , Li Rusong , Yu Fengmei , Ma Wenyan
Published: February 2014

We design the structural dimensions of the reactor for the supersonic-based disposing radioactive waste water in this paper. The disposing efficiencies have been simulated under several conditions, and obtained the total α radioactive concentration for the waste water as a function of disposing time, the impact of the supersonic frequency and the time-mean sound energy density on the disposing effectiveness, and the decontamination factor vs. disposing time. The results show that the total α radioactive concentration diminishes as the disposing time increases. The total α radioactive concentration will below the national emission standard, and the disposing time for reaching the emission standard diminishes as the supersonic frequency increases. When the supersonic frequencies are same, the total α radioactive concentration of the disposed radioactive waste water diminishes as the time-mean sound energy density increases under same disposing time, and the decontamination frequency increase as the time-mean sound energy density increases. The total α radioactive concentration of the disposed radioactive waste water diminishes as the time-mean sound energy density increases under other conditions being same. Under the same supersonic conditions, the supersonic decontamination factors for 10Bq/L and 20Bq/L radioactive waste water are same, namely the decontamination factors are independent of the initial radioactive concentration. Moreover, the supersonic decontamination factor for the suspended particles that obey the normal distribution with expectation value of 0.7μm and variance of 0.21 in the waste water is higher than that of the suspended particles that belong to the monodispersity particles. The designed supersonic reactor can satisfy the requirement of the disposing efficiency.
Preliminary Study on the Primary Side Structure Design Improvement Based on CPR1000
Authors: Lin Jianshu , Zhuang Chengjun , Cui Jun , Wang Ting , Zhang Zhenhua
Published: February 2014

In order to enhance the efficiency and obtain more safety margin in the new type nuclear power plant design, it is of great necessity and significance to study the potential probability of vessel structure design improvement which is based on CPR1000 NP. In this paper, LB LOCA is chosen as the typical accident with CATHARE GB code and the DRM applied. Sensitivity study of in-vessel structure, length of fuel rods and the water inventory are performed. The result shows that by reducing the length between the vessel bottom and the core inlet, extending the fuel rod length and increasing primary water inventory, the PCT declines respectively. More than 200 K temperature benefit can be obtained when the three modifications mentioned above combined together. Due to the sufficient margin increased, a proper power uprating is feasible.
Development of Numerical Simulation for Jet Breakup Behavior in Complicated Structure of BWR Lower Plenum: 1 — Preliminary Analysis of Jet Breakup Behavior in Complicated Structure by TPFIT
Authors: Suzuki Takayuki , Yoshida Hiroyuki , Nagase Fumihisa , Abe Yutaka , Kaneko Akiko
Published: February 2014

In order to improve the safety of Boiling Water Reactor (BWR), it is required to know the behavior of the plant when an accident occurred as can be seen at Fukushima Daiichi nuclear power plant accident. Especially, it is important to estimate the behavior of molten core jet in the lower part of the containment vessel at severe accident.

In the BWR lower plenum, the flow characteristics of molten core jet are affected by many complicated structures, such as control rod guide tubes, instrument guide tubes and core support plate. However, it is difficult to evaluate these effects on molten core jet experimentally. Therefore, we considered that multi-phase computational fluid dynamics approach is the best way to estimate the effects on molten core jet by complicated structure.

The objective of this study is to develop the evaluation method for the flow characteristic of molten core jet including the effects of the complicated structures in the lower plenum. So we are developing a simulation method to estimate the behavior of molten core jet falling down through the core support plate to the lower plenum of the BWR. The method has been developed based on interface tracking method code TPFIT (Two Phase Flow simulation code with Interface Tracking). To verify and validate the applicability of the developed method in detail, it is necessary to obtain the experimental data that can be compared with detailed numerical results by the TPFIT. Thus, in this study, we are carrying out experimental works by use of multi-phase flow visualization technique. In the experiments, time series of interface shapes are observed by high speed camera and velocity profiles in/out of the jet will be measured by the PIV method.

In this paper, the outline of the developing method based on the TPFIT was explained. And, the developing method was applied to preliminary experiment with/without modeled complicated structures. As the results, predicted interface shapes were almost agreed with measured data. However, predicted falling down velocity of the jet was lower than measured data. We considered causes of this underestimation and improved the method and simulation conditions to resolve this problem.
The Measurement and Analysis of CEFR Sodium Void Reactivity Effect
Authors: Zhou Keyuan , Yu Hong , Hu Yun , Chen Xiaoliang , Gang Zhi , Wang Shixi , Li Zehua
Published: February 2014

Measuring of Sodium Void Reactivity Effect (SVRE), one of the most important tests in China Experimental Fast Reactor (CEFR) physical start-up, is described in the paper, including test method, test results and evaluation of test results. The results met to the test requirement and the sign base of CEFR TIB Accident Special Inspect System. The calculation analysis of CEFR SVRE test has been completed, which provides data support before the test and verifies the reliability of the calculation systems after the test. The technology for analysis and measuring of SVRE in sodium-cooled fast reactor has been accumulated through the research of this test.
Numerical Investigation of Flow and Heat Transfer Characteristics in Wire-Wrap Tight Lattice 19-Rod Bundle
Authors: Li Minggang , Wang Jun , Nie Changhua , Yan Xiao , Huang Yanping , Yang Zumao , Xie Feng
Published: February 2014

Flow and heat transfer characteristics in wire-wrap tight lattice rod bundle have been investigated through CFD code ANSYS CFX 13.0. The bundle consists of 19 fuel rods with triangular tight lattice configuration. The rod ratio of rod pitch to rod diameter is 1.167. Four wires with a diameter of 0.5 mm are helically wrapped on the surface of each fuel rod. The ratio of wire-wrap helical pitch to the rod diameter is varied from 27.5 to 52.5.

Through simulating wire-wrap 3-rod bundle with tetrahedron and hexahedron grid systems, the grid system which applies to simulating the wire-wrap tight lattice rod bundle has been obtained.

The predicted results of eddy viscosity based turbulence models (k–ε, SST) and Reynolds stress turbulence models (BSL, SSG) are compared with each other and several experimental correlations for friction factor and Nusselt number. The predicted results of all the turbulence models are almost the same in some respects, but the friction factor predicted by the eddy viscosity models is higher than that predicted by the RSM.

The effect of wire-wrap on pressure drop, friction factor, secondary flow, heat transfer, velocity distribution and temperature distribution in different subchannels (interior, edge and corner) has been analyzed by comparing with those of the bare rod bundle. The effect of wire-wrap pitch on the flow and heat transfer characteristics has also been studied.
Behavior of the Ag-In-Cd Alloy Control Rod Under Irradiation
Authors: Xiao Hongxing , Long Chongsheng , Chen Le , Liang Bo
Published: February 2014

Increase of control rod diameters may occur under Pressurized Water Reactors (PWR) operational conditions. This increase constitutes a limitation on the life-time of control rod assemblies and the safety of reactor. In order to understand the contribution of the irradiation to this damage, SIMROD (simulated the irradiated control rod: Silver-indium-cadmium alloy with the tin additive which aim to simulate the effect of irradiation) and non-irradiated control rod are investigated. The density, microstructure, thermal diffusivity, conductivity and steady compressive creep rate have been measured by Archimedes principle method, optical microscope (OM), scanning electron microscope (SEM), laser bombard method and creep testing machine in this work. The results show that the thermal diffusivity and conductivity of SIMROD alloy increases with the increase of temperature as well as the non-irradiated control rod alloy, fcc alloy of Ag-In-Cd will transform to an hcp phase similar to the zeta phase of the silver alloys while the Sn as the additive. In order to understand the creep mechanism of as-cast alloys, the stress exponent n and activation energy Qc are calculated by the data of stress and temperature dependence of steady creep rate. Grain boundary sliding is the rate controlling mechanism for the alloy at the temperature of 300°C and dislocation creep is the rate controlling mechanism for the alloy at the temperatures of 350–400°C, respectively.
Modelling of QUENCH 03 and QUENCH 06 Experiments Using RELAP/SCDAPSIM Code
Authors: Kaliatka Tadas , Ušpuras Eugenijus , Kaliatka Algirdas
Published: February 2014

An important accident management measure for controlling severe accident transients in Light Water Reactors is the injection of water to cool the degrading core. Flooding of the overheated core, which causes quenching of the fuel rods, is considered a worst-case scenario regarding hydrogen generation rates which should not exceed safety-relevant critical values.

Within the frame of the QUENCH test-program the loss of coolant accidents with the following flooding of overheated core in Light Water Reactors is analysed using an experimental facility. The modelling of QUENCH-03 and QUENCH-06 experiments was performed with RELAP/SCDAPSIM computer code. The observed calculation results showed that thermal properties of shroud materials (heat losses through the shroud) and electrical power of fuel simulators are the main source of uncertainty in the calculations.

The main idea of this article is modification of input parameters to receive the best agreement with the measurements for the selected QUENCH test. Modified input parameters are used in the input deck for another QUENCH test. The good agreement between calculation results and measurements of both QUENCH tests demonstrated the correctness of modified parameters and legitimacy with the real physical processes.
Research on Reactor Coolant Pump Fault Diagnosis Method Based on Multi-Sensor Data Fusion
Authors: He Pan , Liu Caixue , Ai Qiong
Published: February 2014

Usually, more than one sensor is placed to collect vibration signals for reactor coolant pump condition monitoring. The traditional method of reactor coolant pump fault diagnosis does not make full use of the relativity of all vibration signals. In order to make full use of all vibration signals, multi-sensor data fusion is introduced to reactor coolant pump fault diagnosis and a universal reactor coolant pump fault diagnosis model is built up. The reactor coolant pump vibration data fusion diagnosis model is divided into three modules. The three modules are the data level fusion module, the BP (back-propagation) neural networks feature level fusion diagnosis module, the D-S (dempster-shafer) evidence theory decision level fusion module. The data level fusion module is to eliminate the disturbance and extract the feature information about reactor coolant pump faults. The feature information handled by the data level fusion module is used as the inputs of BP neural networks. The neural networks feature level fusion diagnosis module is composed by more than one BP neural networks in condition that the number of input nodes is too large. The feature information is divided into several troops and input into BP neural networks respectively. The outputs of neural networks serve as the basic probability assignment of D-S evidence theory. The D-S evidence theory decision level fusion module fuses the outputs of neural networks and gives the final fusion diagnosis result. The experiment results show that multi-sensor data fusion is successful and promising in reactor coolant pump fault diagnosis.
Failure Analysis and Strategy Investigation of the Rupture on the Gland Seal Steam Drain Pipeline of the Reheat Stop Valve
Authors: Zhao Liang , Li Guangfu , Hu Jianqun , Zhao Weidong , Chen Mingjun
Published: February 2014

Macroscopical inspection and analysis of the pipe component, microstructure, fracture and corrosion product were conducted on the pipe parent metal, heat affected zone and the weld zone of the ruptured reheat stop valve shaft seal steam pipe of Qinshan III. The main failure cause could be concluded that oxidation corrosion on the pipe inner surface near the valve with high temperature steam cause the wall thickness largely reduced, and corrosion fatigue due to pipeline vibration eventually lead to fracture. The thinning rate of the failure pipe is assessed. There are total 32 similar pipelines in the two units of Qinshan III. The preventive repair strategies for this kind of pipelines are investigated. Wall thickness measurement and evaluation should be conducted in the outage sixth years after the pipeline replacement. The inspection location is 30mm∼150mm extent downstream of reheat stop valve outlet nozzle. The pipe section should be replaced in accordingly outage when the minimum measured wall thickness is less than 2mm.
Research Progress on Irradiation Effects of RPV Materials
Authors: Feng Mingquan , Mo Huajun , Li Guoyun , Tong Mingyan , Liu Xiaosong
Published: February 2014

Reactor pressure vessel (RPV) is an important safety component which holds reactor core and keep on high temperature and high pressure coolant. The irradiated region of RPV is subjected to neutron damage so that the fracture toughness decrease and RTNDT of materials with neutron fluence increase. According to fracture toughness requirements based on the rule, RPV beltline materials must have Charpy upper-shelf energy of no less than 102J initially and must maintain upper-shelf energy throughout the life of the vessel of no less than 68J.

Chinese manufactures (China First Heavy Industries, China National Erzhong Group Co. and Dongfang Electric Corporation, Shanghai Electric Corporation) have supplied A508-3 steel forging and fabricated RPV for many constructing NPP in China. Nuclear Power Institute of China (NPIC) has being put into practice a program about the irradiation tests and irradiation brettling research of Chinese A508-3 steel from these manufactures. This paper introduces the program and progressing briefly. Specimen, irradiation test parameters, irradiation facility and post irradiation mechanical tests are also described.
Updated Design and Development Route for CH HCCB TBM and its Mockup
Authors: Hu Gang , Feng Kaiming , Zhao Zhou , Zhang Guoshu , Wang Qijie , Cao Qixiang , Chen Yanjing , Ye Xingfu , Wang Pinghuai , Wang Zeming , Yu Dehuai , Tao Shiquan , Wang Shizhong , Tao Haiyan
Published: February 2014

Chinese helium-cooled ceramics breeder test blanket module (CH HCCB TBM) is determined to be tested in ITER machine to get data for fusion reactor design and development in future. Chinese TBM is designed to occupy half of port C with 484mm in torroidal and 1660mm in poloidal. Radial length is 675mm. TBM is composed of box, 12 submodules and independent backplate. Box formed by first wall, grids and caps have 12 caivities to hold submodules. Box and submodules are supported by backplate by welding. Backplate distribute helium with flow rate 1.36kg/s to cool first wall and then part of it go out of TBM by bypass. The rest 0.77kg/s go on to cool caps and girds first and then cool submodules. Submodules with dimensions 250mm×202mm×318mm have independent cooling and purging systems connected to backplate manifold systems. In a submodule, two U-shaped structures hold breeding material Li4SiO4 pebbles. Out of the structure filled beryllium pebbles. Neutronics results show that tritium production is ∼64mg/FPD. Maximum temperature 538°C of structure material occurs in the front of first wall with surface heat flux 0.5MW/m2. Maximum total stress at first wall is 471MPa at 394°C; that in submodules is 426MPa at 400°C; that in backplate is 526MPa at 410°C, In order to explore development technologies for the TBM, a mockup with dimensions 484mm (torroidal)×592mm (poloidal)×675mm (radial) has been designed. The mockup with similar structure ignores bypass and purge gas systems. In the mockup, there’s only one submodule and the other three are replaced by submodule replacements. By discussions and investigations, development route has been decided and the mockup is being fabricated.
Project Development of Experimental Hybrid Reactor on the Basis of a Compact Tokamak for the Disposal of Spent Nuclear Fuel
Authors: Azizov E. A. , Gladush G. G. , Velikhov E. P.
Published: February 2014

The paper describes development of project of hybrid reactors fusion-fission based on compact tokamak. It can have several advantages: such tokamak because of the smallness of aspect ratio A has a smaller volume at the same power as compared with the classical tokamak. At the same time, in their design, one can use database of ITER. The paper presents the results of a long phase of work in this direction, directed to the installation of the industrial scale of interest for use in nuclear power. FNS-2 parameters are R = 2 m, a = 1 m, k = 1.7, B = 3.9 T. Calculations show at neutral injection power ∼ 45 MW, the neutron yield is ∼ 60 MW, it provides a neutron load on the blanket ∼ 0.4 MW/m2. Numerical calculations showed that channel structure blanket cooled by water will transform for the year to 80 kg of minor actinides at subject to the addition of Pu in a 1:1 ratio). When a liquid metal coolant is used, the design of the outer blanket FNS-2 is similar to the block structure of the ITER. During the year this hybrid reactor loading 60 tons of oxides of actinides can recycle 400 kg of minor actinides. This will help dispose of spent nuclear fuel more than 10 LWR-1000 reactors. In this case the thermal power is ∼ 2 GW. The pressure which provides lead pumping through a stainless steel pipe is 1.5 bar at the length of 20 cm.
A NPP KSN DCS Training System With Combined Approach of Simulation and Stimulation
Authors: Zhang Caike , Teng Ye , Ding Shaojie , Jing Yinggang , Cheng Minmin , Hu Baojun
Published: February 2014

In this paper, taking a renovation project as an example, a training system combined of “Simulation” and “Stimulate Simulation” pattern is described. The renovated system is originally composed of a HOLLIAS MACS V6 control cabinet and several HMI workstations. The simple configuration without process equipment signals can not achieve the dynamic operation of the interface, which greatly limits the scope of the training applications. In this renovation project the RINSIM simulation platform is used to develop the process system modeling software and DCS Level 1 modeling software, which is called as the “Simulation” method. Meanwhile, a simulation DCS Level 2 is developed through the DCS HMI simulating configuration software. These two portions form a complete Simulation system. In addition, through the OPC and TCP protocol, the simulation process system modeling software and DCS Level 1 control modeling software can communicate with DCS control cabinet, which forms a Stimulating Simulation system. The two systems which can be switched to one another has a high value in use both for configuration training of I&C and maintenance staff, and also for the system process operation training of operators. This paper gives a basic description of simulation system. It also analyzes the simulation system for multiple operating modes and a variety of DCS simulating solutions, focusing on several key technical difficulties in the simulation system.
Study on Controlling Condensation Heat Transfer by Using Functionalized Heat Transfer Surface
Authors: Yoshizawa Syota , Ikeda Sho , Koizumi Yasuo
Published: February 2014

Condensation heat transfer experiments were performed by using steam at 0.1 MPa. The enhancement of drop-wise condensation heat transfer by functionalizing a heat transfer surface was examined. A gold-plated surface was used to produce the drop wise condensation. Grooved heat transfer surfaces were adopted to functionalize the heat transfer surface. The shape and the size of the groove were rectangular and 2 mm × 2 mm × 2 mm, respectively. When the top parts and the walls of grooves were plated with gold, the heat transfer rate was the highest. The second was the case that all area of the grooves was plated with gold. The third was the case that only the top parts of the grooves were plated with gold. The forth was the plain gold-plated surface. The drop wise condensation was enhanced by 2.6 times at the most when the heat transfer surface was divided into two areas that had different functions of condensing (the gold-plated surface) and draining (the copper-bare surface). Condensate drops on the gold-plated surface were pulled into films on the copper-bare surface. It helped to remove the drops on the gold-plated heat transfer surface and to expose the gold-plated heat transfer surface to steam for much longer time.
Research on Negative Pressure Control Technology in NPP Reactor Building and its Application
Authors: Xie Jianghong
Published: February 2014

The paper mainly elaborates the negative pressure control technology and commissioning approaches for double-wall containment of Russian WWER-1000 nuclear power units. It also carries out an analysis and research on the layered negative pressure technology in the containment. It mainly includes the following three parts:

A Russian WWER-1000 nuclear power unit adopts the structure of double-wall containment for its Reactor building, with independent negative pressure systems for the containment and the annular space between the two walls. The paper mainly elaborates the control methods and limits requirements for the negative pressure in the containment and the annular space under the normal operation condition and in case of design basis accidents, with analysis and argumentation on the design function and operation requirements of the negative pressure system for the containment and the annular space.

In the paper, the design philosophy of layered negative pressure and its feasibility study are analyzed from the aspects of radiated partition, air distribution of the negative pressure system and containment separator for layered negative pressure.

The commissioning methods and technical requirements of negative pressure system in the Reactor building are described in the paper. Problems encountered during commissioning are also addressed and analyzed.

Operations practices prove that the negative pressure control technology for double-wall containment of WWER-1000 nuclear power unit is advanced and reliable, which meets the requirements on nuclear air decontamination emission and radiation protection, and is worthy of study, research and reference.
Analysis of Fluid-Induced Excitation Forces in Triangular Tube Arrays Based on the LES Method
Authors: Jiang Qing-lei , Zhou Ji-yun , Wang Xian-yuan
Published: February 2014

The turbulence-induced excitation and periodic wake shedding are two important flow-induced vibration mechanisms of tube arrays in steam generators and were normally considered to be random excitation and periodic excitation respectively. Recent findings show that the turbulence-induced excitation is actually the quasi-periodic excitation, which is similar to periodic excitation. Since the turbulence-induced excitation generally occurs in the small pitch-diameter ratio (P/D ratio) tube arrays and the periodic excitation occurs only at large pitch-diameter ratios, the turbulence-induced excitation mechanism seems essentially another form of the periodic excitation at small P/D ratios.

To verify this hypothesize and figure out the transition procedure from periodic excitation to quasi-periodic excitation, Numerical simulations were conducted on triangle tube arrays with P/D ratios of 1.47 to 4.0 based on the LES method. Numerical results were compared with experiments results to verify the reliability of numerical models. The lift forces and the force spectra at indifferent P/D ratios were obtained to study the evolution mechanism between turbulence-induced excitation and periodic excitation. The results show that the turbulence-induced excitation and quasi-periodic excitation in tube arrays are essentially other forms of the periodic excitation. The lift forces changed from periodic excitation to quasi-periodic excitation when the P/D ratio was between 2.5 and 3.0, and the lift force nearly complete random excitation when P/D ratios below 1.5. The frequency range of the random excitation is 0 to 50Hz. The peak frequencies of the quasi-periodic excitation were greater than that of periodic excitation and decreased with the increase of P/D ratios.
RCC-M and RSE-M: A Set of Consistent Codes for PWR Design and Operation
Authors: Faidy Claude
Published: February 2014

Inside the French Association for Design, Construction and Surveillance Rules of Nuclear Power Plant Components (AFCEN), RCC-M is dedicated to Design, Material Selection, Fabrication and welding, Examination and RSE-M In-Sercice-Inspection, Surveillance, Monitoring, Maintenance and flaw evaluations.

In a first part, this paper will shortly presents the Code contents and organization of each of them.

In a second part, how they are consistent and flexible to be adapted to international requirements.

And finally in a third part, a short comparison with ASME Corresponding Codes: BPV Section III and XI. The conclusion of the paper will present the major objectives and topics of development of these Codes for new international projects in connection with Code Harmonization by Regulators (MDEP), Standard Development Organization board (SDO) and Industry Association (CORDEL).
Fatigue of Stainless Steel Components: Toward Codified Rules Improvements
Authors: Faidy Claude
Published: February 2014

During the past 30 years many fatigue tests and fatigue analysis improvements have been developed in France in order to improve Codified Fatigue Rules of RCC-M, RSE-M and RCC-MRx.

This paper will present comments and proposals for development of these rules associated to Gaps and Needs in order to finalize and justify the AFCEN Codes new rules.

Recently 3 new international R&D results confirm possible un-conservative fatigue material data:

- High cycle fatigue in air for stainless steel,

- Environmental effects on fatigue S-N curve for all materials, and in particular stainless steels,

- Fatigue Crack Growth law under PWR environment for stainless steel.

In front of these new results, AFCEN is working on a 1st set of rules based on existing knowledge:

- Air fatigue curve

- PWR Environmental effects

A periodic up-dating of AFCEN proposed rules will be done using French and International R&D programs with a particular attention on harmonization with other Code rules developed in USA, Japan and Germany, in particular.
Laser Decontamination of Paint Coated Concrete in Nuclear Plants
Authors: Anthofer Anton , Lippmann Wolfgang , Hurtado Antonio
Published: February 2014

Laser technology offers an efficient decontamination of surfaces contaminated by Polychlorinated Biphenyls (PCB). PCB is a component of protective coatings on concrete walls in NPP’s up to the 1980s-years. State of the art is a manual and mechanic ablation, which then poses danger of a secondary contamination and needs a second treatment in a hazardous waste incinerator. A 10 kW diode laser in continuous wave mode and wavelength of 915–1030 nm rises up the surface temperature for ablation of coatings and thermal decomposition in one process step. Meanwhile, decontamination rates of 6.345 m2/h have been operated using a laser spot of 45 × 10 mm2. An experimental facility with a three-barrier-system to contain toxic material has been designed and constructed. First experimental investigations use epoxy wall coatings for an optimization of the process and to calculate a simulation of the ablation process. Further experimental investigations for ablation of PCB-coatings can be operated safely. A second development is an in-situ measurement system to detect the thermal decomposition of PCB by a laser induced fluorescence (LIF) system by the project partner TU Bergakademie Freiberg (TUBAF), Germany.
Numerical Analysis on Heat Transfer-Characteristics of Supercritical Pressure Water in a Heated Tube Based on Three Dimensional Two-Fluid Model
Authors: Ose Yasuo , Suzuki Takayuki , Misawa Takeharu , Yoshida Hiroyuki , Takase Kazuyuki
Published: February 2014

In Japan Atomic Energy Agency (JAEA), the three-dimensional two-fluid model analysis code considering the supercritical pressure water based on ACE-3D (Advanced Code for Evaluation of 3-Dimensional two-phase flow) has been developed to establish the thermal-hydraulics design by numerical analytical approach for the Super Critical Water Reactor (SCWR). In this paper, in order to evaluate the prediction accuracy of ACE-3D for the heat transfer-characteristics at the pseudo critical point, a numerical analysis of the supercritical water using ACE-3D based on the three dimensional two-fluid model has been conducted for simulating the experiments in a heated tube with both upward and downward flow. For the turbulence model in this analysis, both the standard k-ε model and the low-Reynolds number type k-ε model which uses the Launder-Sharma model were examined to investigate the influence of the turbulence model on the heat transfer-characteristics near the heated wall near the pseudo critical point. As a result, it was found that the numerical results of wall temperature using the low-Reynolds number k-ε model for upward flow in a heated tube were in good agreement with experimental data compared with that of using the standard k-ε model.
On Efficient Passive Cooling of Control Rod Drivers
Authors: Gershuni A. N. , Nishchik A. P. , Razumovskiy V. G. , Pioro I. L.
Published: February 2014

Experimental research of natural convection and the ways of its suppression in an annular vertical channel to simulate the conditions of cooling the control rod drivers of the reactor protection system (RPS) in its so-called wet design, where the drivers are cooled by primary circuit water supplied due to the system that includes branched pipelines, valves, pump, heat exchanger, etc., is reported. Reliability of the drivers depends upon their temperature ensured by operation of an active multi-element cooling system. Its replacement by an available passive cooling system is possible only under significant suppression of natural convection in control rod channel filled with primary coolant. The methods of suppression of natural convection proposed in the work have demonstrated the possibility both of minimization of axial heat transfer and of almost complete elimination of temperature non-uniformity and oscillation inside the channel under the conditions of free travel of moving element (control rod) in it.

The obtained results widen the possibilities of substitution of the active systems of cooling the RPS drivers by reliable passive systems, such as high-performance heat-transfer systems of evaporation-condensation type with heat pipes or two-phase thermosyphons as heat-transferring elements.
Nonlinear Crest-Cut-Off Method for Reducing the Stress Concentration of Bolt With Many Threads: Application in the Main Bolt of HTGR’s RPV
Authors: Zhao Jiaqing , Zhang Zhengming
Published: February 2014

For the standard thread, severe stress concentration appears in the root of first several threads which share the major part of bolt load, and this is also an important issue in the design of the main bolt thread in the High Temperature Gas-cooled Reactor’s (HTGR’s) Reactor Pressure Vessel (RPV). The linear crest-cutoff method (LCCOM) linearly cuts down the height of the engaged threads near the bolt head, and it could reduce the stress concentration at the roots of first several threads. However, as revealed in finite element simulation, when there are as many as forty threads in the bolt, even though the axial force shared by the first thread could be decreased by LCCOM, the axial force shared by subsequent threads are still very high, which are often larger than that by the first thread. To settle this problem, a nonlinear crest-cut-off method (NCCOM) is proposed, which employs the curved thread profile of quadratic polynomial function, instead of tapered profile of linear function in LCCOM. The proposed curved thread profile has one additional degree of freedom, and it could also be degenerated to tapered profile. As for the main bolt of forty threads in the HTGR’s RPV, the suitable parameters of curved profile are determined by the intensive numerical simulations. The results show that the proposed NCCOM yields lower axial force of the first several threads, and produces lower stress at the roots of threads in the bolt compared with LCCOM.
The Testing and Application of Continuous-Energy Neutron Cross Section Library HENDL-ADS/MC
Authors: Chen Chong , Zou Jun , Xu Dezheng , Zeng Qin , Wang Minghuang
Published: February 2014

A point-wise cross-section data library HENDL-ADS/MC (Hybrid Evaluated Nuclear Data Library) has been produced by FDS team to do the nuclear analysis for the ADS system. The HENDL-ADS/MC library contained 408 nuclide cross-section files including actinides, fission products and structural materials for neutron energy up to 150 MeV. The nuclear library also contained several sub-libraries with different temperatures. A series of neutron integral experiments and critical safety benchmarks have been performed to test the availability and reliability of the HENDL-ADS/MC data library. To validate and qualify the reliability of the high neutron energy cross section for HENDL-ADS/MC library further, a series of high neutron shielding experiments have been performed using MCNP. The testing results indicated the accuracy and reliability of HENDL-ADS/MC library.
Countermeasures to Improve LWR Safety Level in Design and Operation by Accident Analyses at Fukushima Daiichi Nuclear Power Plants
Authors: Shimizu Kazumasa , Hamada Yuhei , Sakashita Hiroto , Mori Michitsugu
Published: February 2014

The 2011 off the Pacific coast of Tohoku Earthquake occurred on March 11. The earthquake attacked the Fukushima Daiichi nuclear power station with six boiling water reactors (BWRs), three out of which, units 1 through 3 in rated operation except for three reactors of units 4 through 6 in scheduled periodic inspection, automatically shut down in response to the intense seismic motion. Emergency diesel generators started to pump water to cool reactors, and an hour later, the back-up generators lost their all functions by the station blackout resulting from tsunami flooding. In this situation at the unit 1, the isolation condenser system (IC) should have made a critical role to keep the reactor pressure and water level to be safety by removing the decay heat by natural circulation. In fact at the unit 1 during the accident, IC valves were closed by fail-safe and could not have shown the ability of the designed function.

An accident report gave general descriptions of the causes and results of accidents, but not the quantitative data indicative of details; therefore, it seems difficult to identify the specific problems in plant operations. Even in this case, if an appropriate analysis code is available for reproducing events based on the reports, it will be possible to determine individual data quantitatively and identify problems in plant operations.

In our work, we used the nuclear reactor thermal-hydraulic code RETRAN-3D/MOD4, which has been approved and licensed by U.S. Nuclear Regulatory Commission, to model light water reactors (LWRs) and reproduce the circumstances of the 2011 Fukushima Daiichi nuclear accident as the simulation code. Here, we subjected transition analyses of the process on the core-meltdown accident, and put forward the system to prevent the accident, where the accident analysis report was employed to simulate conditions of the accident. It could enable us to suggest adequate operation procedures suitable for LWR to avoid the severe accident, and to propose countermeasures to improve LWR safety level in design and operation.
Development of Co-Reduced Valve Seat for Nuclear Power Plants
Authors: Tazawa Toshiyuki , Saito Masahiro , Kojima Nobuo
Published: February 2014

The overlaying of Co-based alloy (e.g. “Stellite”) has been applied to valve seats in power plants to improve wear and abrasion resistance. However, using of the Co-based alloy should be avoided to valves which involve with reactor water in nuclear power plant because of activating to Co-60, and increment of radiation dose. Therefore, the alternative material to Co-based alloy has been required for these valves.

Hardfacing is usually applied to improve abrasion resistance. Several high hardness materials without Co element were investigated, and it was found a Fe-based alloy as an alternative to Co-based alloy for the overlaying of valve seats.

Mechanical tests of this Fe-based alloy formed by Plasma Transfer Arc (PTA) welding were carried out. Obtained values of tensile strength, Vickers hardness and Charpy impact value of the Fe-based alloy were almost equivalent to Co-based alloy. In addition, mock-up valve specimen was prepared, and sliding test was performed in simulated BWR plants environment. As a result, the amount of leakage was little enough when the Fe-based alloy is overlaid on one side seat part, and the radiation exposure from the valve seat can be reduced by 78 percent compared with the case of using the Co-based alloy for both side of overlaying.

Note: Product names mentioned herein may be trademarks of their respective companies.
Antibacterial properties and cytocompatibility of tantalum oxide coatings with different silver content
Authors: Huang Heng-Li , Chang Yin-Yu , Chen Hung-Jui , Chou Yu-Kai , Lai Chih-Ho , Chen Michael Y. C.
Published: February 2014

Tantalum (Ta) oxides and their coatings have been proved to increase their applications in the biomedical fields by improving osseointegration and wear resistance. In this study, Ta oxide coatings containing different proportions of Ag are deposited on SS304 materials. A twin-gun magnetron sputtering system is used to deposit the tantalum oxide-Ag coating. In this study, Staphylococcus aureus, which exhibits physiological commensalism on the human skin, nares, and mucosal and oral areas, is chosen as the model for in vitro antibacterial analyses via a fluorescence staining method using Syto9. The cytocompatibility and adhesive morphology of human skin fibroblast cells (CCD-966SK) on the coatings are also determined by using the microculture tetrazolium assay. This study shows that Ta2O5 and Ta2O5-Ag coatings with 12.5 at. % of Ag exhibit improved antibacterial effects against S. aureus and have good skin fibroblast cell cellular biocompatibility.
Surface Expression of ω-Transaminase in Escherichia coli
Authors: Gustavsson Martin , Muraleedharan Madhu Nair , Larsson Gen
Published: February 2014

Chiral amines are important for the chemical and pharmaceutical industries, and there is rapidly growing interest to use transaminases for their synthesis. Since the cost of the enzyme is an important factor for process economy, the use of whole-cell biocatalysts is attractive, since expensive purification and immobilization steps can be avoided. Display of the protein on the cell surface provides a possible way to reduce the mass transfer limitations of such biocatalysts. However, transaminases need to dimerize in order to become active, and furthermore, they require the cofactor pyridoxal phosphate; consequently, successful transaminase surface expression has not been reported thus far. In this work, we produced an
Arthrobacter citreus
ω-transaminase in
Escherichia coli
using a surface display vector based on the autotransporter
a
dhesin
i
nvolved in
d
iffuse
a
dherence (AIDA-I), which has previously been used for display of dimeric proteins. The correct localization of the transaminase in the
E. coli
outer membrane and its orientation toward the cell exterior were verified. Furthermore, transaminase activity was detected exclusively in the outer membrane protein fraction, showing that successful dimerization had occurred. The transaminase was found to be present in both full-length and proteolytically degraded forms. The removal of this proteolysis is considered to be the main obstacle to achieving sufficient whole-cell transaminase activity.
Ste18p Is a Positive Control Element in the Mating Process of Candida albicans
Authors: Lu Hui , Sun Yuan , Jiang Yuan-Ying , Whiteway Malcolm
Published: February 2014

Heterotrimeric G proteins are an important class of eukaryotic signaling molecules that have been identified as central elements in the pheromone response pathways of many fungi. In the fungal pathogen
Candida albicans
, the
STE18
gene (
ORF19.6551.1
) encodes a potential γ subunit of a heterotrimeric G protein; this protein contains the C-terminal CAAX box characteristic of γ subunits and has sequence similarity to γ subunits implicated in the mating pathways of a variety of fungi. Disruption of this gene was shown to cause sterility of
MTL
a
mating cells and to block pheromone-induced gene expression and shmoo formation; deletion of just the CAAX box residues is sufficient to inactivate Ste18 function in the mating process. Intriguingly, ectopic expression behind the strong
ACT1
promoter of either the Gα or the Gβ subunit of the heterotrimeric G protein is able to suppress the mating defect caused by deletion of the Gγ subunit and restore both pheromone-induced gene expression and morphology changes.
Draft Genome Sequence of Petroleum Hydrocarbon-Degrading Pseudomonas aeruginosa Strain PK6, Isolated from the Saurashtra Region of Gujarat, India
Authors: Patel P. A. , Kothari V. V. , Kothari C. R. , Faldu P. R. , Domadia K. K. , Rawal C. M. , Bhimani H. D. , Parmar N. R. , Nathani N. M. , Koringa P. G. , Joshi C. G. , Kothari R. K.
Published: February 2014

Pseudomonas aeruginosa
strain PK6, a potential petroleum hydrocarbon-degrading soil bacterium, was isolated from a site contaminated by a petroleum hydrocarbon spill from an automobile service station in Junagadh, Gujarat, India. Here, we provide the 6.04-Mb draft genome sequence of strain PK6, which has genes encoding enzymes for potential and related metabolic pathways of the strain.
Twin athlete brothers with open physes operated for ACL reconstruction on the same day, but with different elapsed times after injury: a 5-year follow-up
Authors: Astur Diego Costa , Lauxen Daniel , Ejnisman Benno , Cohen Moises
Published: February 2014

We present the case of twin brothers with open physes who practiced judo to a high level and were operated on the same day for anterior cruciate ligament reconstruction. One of them was injured a year before surgery, and the other was injured a month before the procedure. The brother who chose to undergo a conservative treatment sustained meniscus injury afterwards and showed lower objective results when evaluated 5 years after surgery.