bluuebunny/crossref_metadata_2025 huggingface dataset:
Thermal-Hydraulic Performance of R-134a Boiling at Low Mass Fluxes in a Small Vertical Brazed Plate Heat Exchanger
Authors: Kim Hyun Jin , Liebenberg Leon , Jacobi Anthony M.
Published: October 2017

An experimental investigation was performed to study the heat transfer and pressure drop characteristics of refrigerant R-134a boiling in a chevron-patterned brazed plate heat exchanger (BPHE) at low mass flux. The heat transfer coefficient and pressure drop characteristics are analyzed in relation to varying mass flux (30–50 kgm−2s−1), saturation pressure (675 kPa and 833 kPa), heat flux (0.8 and 2.5 kWm−2), and vapor quality (0.1–0.9). The two-phase pressure drop shows a strong dependence on mass flux and significant saturation temperature drop at high mass flux. The two-phase heat transfer coefficient was both strongly dependent on heat flux (at vapor qualities below 0.4) and on mass flux (at vapor qualities above 0.4). There was also apparent dryout, as depicted by decreased heat transfer at high vapor qualities. These observations suggest that both nucleate and convective boiling mechanisms prevailed. Existing transition correlations however suggest that the experimental data is rather convection-dominant and not a mix of convection and nucleate boiling. The experimental data further strongly suggest the prevalence of both macrochannel and minichannel type flows. Several acknowledged semi-empirical transition criteria were employed to verify our observations. These criteria mostly support our observations that R-134a evaporating at low mass fluxes in a BPHE with a hydraulic diameter of 3.4 mm, has heat transfer and pressure drop characteristics typically indicative of macrochannel as well as minichannel flows. Disagreement however exists with accepted correlations regarding the prevalence of convective or nucleate boiling.
Implementation of Liquid Metal Properties in RELAP5 MOD3.2 for Safety Analysis of Sodium-Cooled Fast Reactors
Authors: Song Jian , Liu Limin , Tang Simiao , Wu Yingwei , Tian Wenxi , Qiu Suizheng , Su Guanghui
Published: October 2017

Due to great deal of operation experience and technology accumulation, sodium cooled fast reactor (SFR) is the most promising among the six Generation IV reactors, which has advantages of breeding nuclear fuel, transmuting long-lived actinides and good safety characteristics. Thermal-hydraulic computer codes will have to be developed, verified, and validated to support the conceptual and final designs of new SFRs. However, work on developing thermal hydraulic analysis code for SFR is very limited in China, while the common software RELAP5 MOD3 is unable to analyze liquid metal systems. So the modified RELAP5 MOD3.2 is being considered as the thermal-hydraulic system code to support the development of the SFRs.

The thermodynamic and transport properties of sodium liquid and vapor have been implemented into the RELAP5 MOD3.2 code, as well as the specific heat transfer correlations for liquid metal. The sodium liquid properties use polynomial equations based on data obtained from Argonne National Laboratory, and the vapor is assumed to be perfect gas. The property equations are acceptably accurate for analysis of SFR, especially for single-phase liquid. New files are added to the fluids directory to generate property tables for new working fluid, which are similar to the table interpolation subroutines for light and heavy water in the original file directory. The method of code modifications are universal for other working fluids and will not affect the code original performance. Some basic verification work for the modified code are carried out. The steam generator of CEFR is analyzed to verify the modified code. The calculated results show that all the water will boil off in the evaporator and the calculated results are in good agreement with the design values. By using modified RELAP5 to model the primary loop of EBR-II fast reactor, the SHRT-17 PLOF test was analyzed. The results show that the natural circulation can be established in the EBR-II primary system after main pumps off to remove the core decay residual heat effectively, and the peak temperature under the safety limits. Moreover, the results computed in this work compared well with the test experimental data for the steady state condition. During the transients, the changing trends of temperature and pressure are similar to experimental data. The discrepancies between calculation and experiment are considered acceptably which need to be improved in the future work. Our work could demonstrate the capability and reliability of the modified RELAP5 for the analysis of SFRs further.
Three Dimensional Modeling of the Thermo-Mechanical Performance of the Fuel Rods of a PWR
Authors: Zhu Wang , Chunyu Zhang , Aolin Li , Cenxi Yuan
Published: October 2017

The fuel rods of pressurized water reactors operate under complex radioactive, thermal and mechanical conditions. Multiphysics has to be taken into account in order to evaluate their performance. Many existing fuel rod codes make great simplifications on analyzing the behavior of fuel rods. The present study develops a three dimensional module within the framework of a general-purpose finite element solver, i.e. ABAQUS, for modeling the thermo-mechanical performance of the fuel rods. A typical fuel rod is modeled and the temperature as well as the stress within the pellets are computed. The results show that the burnup levels have an important influence on the fuel temperature. The swelling of fission products cause dramatically increasing of pellet strain. The change of the cladding stress and radial displacement with the axial length can be reasonably predicted. It is shown that a quick power ramp or a reactivity insertion accident can induce high tensile stress to the outer regime of the pellet and may cause further fragmentation to the pellets.
PSA Research of Transport of New Fuel of HTR-PM
Authors: Renze Wang , Jiangang Zhang , Guoqiang Li , Dajie Zhuang , Dongyuan Meng , Xuexin Wang , Hongchao Sun , Shutang Sun
Published: October 2017

Probabilistic safety assessment (PSA) method was explored to be introduced in radiation risk assessment of radioactive material transport (RMT), and then radiation risk of road transport (RT) of new fuel element (FE) of commercial High Temperature Reactor-Pebblebed Modules (HTR-PM) was analyzed. By accident scenario analysis, mechanical analysis and criticality analysis, both accident conditions of package radiation level hoist and criticality were chosen for accident frequency analysis. Since the results show that accident frequency of package radiation level hoist is low and accident frequency of criticality is extremely low, criticality accident can be neglected in the sequential risk assessment. The consequence of typical accident scenarios was estimated, and the results show that the maximum external exposure dose arising from the accident for emergency workers is 0.55mSv, and 4.55×10−3mSv for public people around, which is acceptable. The global radiation risk is 1.24×10−10person·Sv/(vehicle·each transport), for which impact accident contributes the maximum percent.
CFD Simulation of Passive Containment Cooling System in Hot Leg SB-LOCA for 1000MW PWR
Authors: Li Jingya , Zhang Xiaoying
Published: October 2017

The passive cooling system (PCCS) for reactor containment is a security system that can be used to cool the atmosphere and reduce pressure inside of containment in case of temperature and pressure increase caused by vapor injection, which requires no external power because it works only with natural forces. However, as the driving forces from natural physical phenomena are of low amplitude, uncertainties and instabilities in the physical process can cause failure of the system. This article aims to establish a CFD simulation model for the Passive Containment Cooling System of 1000MW PWR using Code_Saturne and FLUENT software. The comparison of 4 different models based respectively on mixture model, COPAIN test, Uchida correlation and Chilton-Colburn analogy which simulate the condensing effect and the improvement of source code are based on a 3D simulation of PCCS system. To simulate the thermal-hydraulic condition in the containment after LOCA accident caused by a double-ended main pipe rupture, a high temperature vapor with the given mass flow rate are supposed to be the source of energy and mass into containment. Meanwhile the surface of three condensing island applies the wall condensation model.

The simulation results show similar transient process obtained with the 4 models, while the difference between the transient simulation and the steady-state analysis of three models is less than 3%. The large mass flow rate of water loss status inside the containment cause a high flow rate of vapor which could be uniformly mixed with air in a short time. For the self-condensing efficiency of 3 groups of PCCS system, the non-centrosymmetric injection position resulting that the condensing efficiency is slightly higher for the two heat exchanger groups nearby. During the first 2400s of simulation time, more than 75.69% of the vapor is condensed, indicating that for the occurrence of condensation at the wall mainly driven by natural convection, the effect of thermodynamic siphon could improve the flow of gas mixture inside the tubes when the velocity of mixture is not large enough, so that the vapor could smoothly enter the tube and reach the internal cooling surface then to be condensed. Besides, PCCS ensure the containment internal pressure maintained below 2 bar and the temperature maintained below 380K during 3600s.
Study on Periodic Shielding Performance Measurements of Spent Fuel Transportation Package
Authors: Sun Hongchao , Li Guoqiang , Wang Xuexin , Zhuang Dajie , Wang Renze , Sun Shutang , Dongyuan Meng
Published: October 2017

The radioactive activity of spent nuclear fuel is high, and the transportation safety is concerned by public and specialist. The periodic radiation shielding performance measurements of spent fuels package is important content to ensure transportation safety of spent fuels. The radiation shielding performance of package must meet the requirements of “Regulations for the safe transport of radioactive material” (GB11806-2004). However, some of the problems and difficulties reflected in practice need to be solved, such as the measurements results of neutron radiation level of spent fuels package outer are not always reliable. In this paper, the periodic shielding performance measurements of one type of spent fuel transportation package are presented. The monitoring results of using both the neutron multi-sphere spectrometer and portable neutron measurement instrument are compared, and the Monte Carlo simulation is done to verify the measurements results. Some factors are discussed, and an optimized scheme is recommended.
Constrained Least Squares Method Used in Nuclear Accident Source Term Inversion
Authors: Xuan Wang , Fenglei Du , Xiaodong Huang
Published: October 2017

Through experience feedback of the Chernobyl and Fukushima nuclear accidents, source term determination including release rates and release curves, becomes a key issue in nuclear accident impact assessment. When faced with monitoring device failure, use of off-site monitoring data and atmospheric dispersion model simulation results to inverse the source term has been proven an effective means in many case studies.

In this paper, we have developed a powerful source term inverse system based on constrained least squares algorithm approach and Gaussian atmospheric diffusion model, and we carried out 15 SF6 field trace experiments at a coastal nuclear power plant site in China from August 22, 2007 to August 31,2007, to validate the source term inversion model. From the 15 simulation results, the 5th,8th and 9th inverse results are the best, and the relative error of inverse source term and true source term is between 10.12% and 16.42%, this is mainly because these three experiments are the most successful, and effective sample points of three experiments are above 25, its distribution is also uniform. It is shown that accuracy, number and distribution of the on-site monitoring will directly affect the inversion source term results.

In general, results of source term inversion model are close to the field experiment, within the same magnitude, which shows that constrained least squares algorithm method and Gaussian atmospheric diffusion model can be applied in nuclear accident source term determination. Results of this study show that constrained least squares method is simple and highly efficient and has a high application value.
The Comparison Research About Two Reliability Management Methods in AP1000 Design
Authors: Xu Zhixin , Wang Chengzhang , Liu Jingjing
Published: October 2017

As a kind of Generation-III passive nuclear power plant, AP1000 has applied two kinds of equipment reliability management methods: the equipment Power Production Reliability Classification (R-Classification) method and Design-Reliability Assurance Program (D-RAP). To invest these two methods, the comparison is implemented between the classification principles, judgment basis and implementation process. According to the RCS and CVS systems, the R-Classification and D-RAP results are compared and some suggestions to enhance the NPP reliability managements are proposed.
Supercritical CO2 Power Cycle for Small Modular Reactor
Authors: Guo Zhangpeng , Zhao Yang , Niu Fenglei , Lu Daogang
Published: October 2017

Most Gen IV reactors have high operating temperature to increase power plant efficiency as well as for hydrogen production. The widely used steam Rankine cycle is not suitable for Gen IV Reactors because of low turbine inlet temperature or high turbine inlet temperature (649 °C) with extreme high pressure (34 MPa) by using ultra-supercritical (USC) steam cycle. Supercritical CO2 recompressing cycle can achieve a competitively high thermal efficiency with turbine inlet temperature at 500–600 °C and turbine inlet pressure at 20 MPa. However, this operating pressure is still too high to limit its application in power cycle of nuclear power plant. A combined recompressing cycle and dual expansion turbine technology is employed to reduce the operating pressure for nuclear power plant and increase thermal efficiency. Besides, sensitive analysis is performed for three kinds of S-CO2 power cycle. The results show that the combined recompressing cycle and dual expansion turbine technology is effective and suitable for Gen IV reactors with relatively lower reactor operating pressure (10–15MPa) and high thermal efficiency.
Development of NPP Collaborative Design Management Platform
Authors: Jun Shen , Xiaoyan Li , Yong Wang
Published: October 2017

A collaborative design management platform of Nuclear Power Plant is built by redevelopment on SYSWARE. On the basis of the given overall framework of platform, the realization methods and means of business layers of the platform are discussed in detail. The application example of main parameters optimization of Nuclear Power Plant design is shown, which verifies the basic function of the platform and the feasibility of the knowledge accumulation and reusability. The research result has some reference value for the construction of collaborative design management platform.
Research of Debris Blocking in PWR Vessel After LOCA
Authors: Jiang Yang , Xi Huang
Published: October 2017

One issue about sump strainer downstream Blocking after LOCA has aroused a widespread concern in nuclear safety authorities and nuclear power plants owners. In this paper, the primary loop and the partial second loop of CPR1000 NPP were modeled for calculation based on the RELAP5 code. Direct injection phases of four typical LOCA cases were simulated, the blocking effect of the limiting LOCA in sump recirculation phase was analyzed by sensitivity studies, then the blocking limit for CPR1000 in-vessel was established. Furthermore, the calculation based on the test had been carried out, and the results show that the requirements of the core cooling have been met in actual blocking situation.
Knowledge From Recent Investigation on Flow Regime Characteristics in Debris Bed Formation Behavior Related to SFR Severe Accident Analyses
Authors: Cheng Songbai , Zhang Ting , Wang Shixian , Jiang Guangyu , Lin Shaopeng , Yu Jiahuan , Wang Li
Published: October 2017

Studies on debris bed formation behavior are of crucial importance for the improved evaluation of core relocation and debris bed coolability that might be encountered in a Core Disruptive Accident (CDA) of Sodium-cooled Fast Reactors (SFR). Motivated to clarify the characteristics of flow regimes underlying this behavior, both experimental investigations and empirical-model development are being performed at the Sun Yat-sen University. As for the experimental study, several series of simulated experiments are being conducted by discharging various solid particles into water pools. To obtain a comprehensive understanding, a variety of experimental parameters, including particle size (0.256∼8mm), particle density (glass, alumina, zirconia, steel and lead), particle shape (spherical and irregularly-shaped), water depth (0∼60cm), particle release pipe diameter (10∼30mm) as well as the particle release height (110∼130cm) were varied. It is found that due to the different interaction mechanisms between solid particles and water pool, four kinds of flow regimes, termed respectively as the particle-suspension regime, the pool-convection dominant regime, the transitional regime and the particle-inertial dominant regime, were identifiable. As for the empirical-model development, by using dimensional analysis technique, a regime map (base map), which is restricted to predictions of spherical particles up until now, was recently suggested. It is shown that a respectable agreement between experiments and the regime-map predictions could be obtained. This work, which gives a large palette of favorable data and insight for a better understanding and an improved estimation of CDAs in SFRs, is expected to benefit future analyses and verifications of SFR severe accident analysis codes in China.
Three Dimensional CFD Simulation of Condensation Inside Inclined Tubes
Authors: Moonesi Shabestary Amirhosein , Krepper Eckhard , Lucas Dirk , Höhne Thomas
Published: October 2017

The current paper comprises CFD-modelling and simulation of condensation and heat transfer inside horizontal pipes. Designs of future nuclear boiling water reactor concepts are equipped with emergency cooling systems which are passive systems for heat removal. The emergency cooling system consists of slightly inclined horizontal pipes which are immersed in a tank of subcooled water. At normal operation conditions, the pipes are filled with water and no heat transfer to the secondary side of the condenser occurs. In the case of an accident the water level in the core is decreasing, steam comes in the emergency pipes and due to the subcooled water around the pipe, this steam will condense. The emergency condenser acts as a strong heat sink which is responsible for a quick depressurization of the reactor core when any accident happens. The actual project is defined in order to model all these processes which happen in the emergency cooling systems. The most focus of the project is on detection of different morphologies such as annular flow, stratified flow, slug flow and plug flow. The first step is the investigation of condensation inside a horizontal tube by considering the direct contact condensation (DCC). Therefore, at the inlet of the pipe an annular flow is assumed. In this step, the Algebraic Interfacial Area Density (AIAD) model is used in order to simulate the interface. The second step is the extension of the model to consider wall condensation effect as well which is closer to the reality. In this step, the inlet is pure steam and due to the wall condensation, a liquid film occurs near the wall which leads to annular flow. The last step will be modelling of different morphologies which are occurring inside the tube during the condensation via using the Generalized Two-Phase Flow (GENTOP) model extended by heat and mass transfer. By using GENTOP the dispersed phase is able to be considered and simulated. Finally, the results of the simulations will be validated by experimental data which will be available in HZDR. In this paper the results of the first part has been presented.
Natural Convection Heat Transfer From Vertical 7x7 Rod Bundles in Liquid Sodium
Authors: Hata Koichi , Fukuda Katsuya , Mizuuchi Tohru
Published: October 2017

Natural convection heat transfer from vertical 7×7 rod bundle in liquid sodium was numerically analyzed to optimize the thermal-hydraulic design for the bundle geometry with equilateral square array, ESA. The unsteady laminar three dimensional basic equations for natural convection heat transfer caused by a step heat flux were numerically solved until the solution reaches a steady-state. The PHOENICS code was used for the calculation considering the temperature dependence of thermo-physical properties concerned. The 7×7 test rods for diameter (D = 7.6 mm), heated length (L = 200 mm) and L/d (= 26.32) were used in this work. The surface heat fluxes for each cylinder were equally given for a modified Rayleigh number, (Rf,L)ij and (Rf,L)Nx×Ny,S/D, ranging from 3.08×104 to 4.28×107 (q = 1×104∼7×106 W/m2) in liquid temperature (TL = 673.15 K). The values of S/D, which are ratios of the diameter of flow channel for bundle geometry to the rod diameter, for vertical 7×7 rod bundle were ranged from 1.8 to 6 on the bundle geometry with equilateral square array. The spatial distribution of average Nusselt numbers for a vertical single cylinder of a rod bundle, (Nuav)ij, and average Nusselt numbers for a vertical rod bundle, (Nuav,B)Nx×Ny,S/D, were clarified. The average value of Nusselt number, (Nuav)ij and (Nuav,B)Nx×Ny,S/D, for the bundle geometry with various values of S/D were calculated to examine the effect of array size, bundle geometry, S/D, (Rf,L)ij and (Rf,L)Nx×Ny,S/D on heat transfer. The bundle geometry for the higher (Nuav,B)Nx×Ny,S/D value under the condition of S/D = constant was examined. The general correlations for natural convection heat transfer from a vertical Nx×Ny rod bundle with the equilateral square and triangle arrays including the effects of array size, (Rf,L)Nx×Ny,S/D and S/D were derived. The correlations for vertical Nx×Ny rod bundles can describe the theoretical values of (Nuav,B)Nx×Ny,S/D for each bundle geometry in the wide analytical range of S/D (= 1.8 to 6) and the modified Rayleigh number ((Rf,L)Nx×Ny,S/D = 3.08×104 to 4.28×107) within −9.49 to 10.6 % differences.
Preliminary Study on Risk-Informed LBLOCA of CPR1000
Authors: Song Jianyang , Liu Jingquan , Wang Ting , Liu Pingping , Lin Zhikang
Published: October 2017

The safety assessment approach of nuclear power plants (NPPs) has been evolved with the technological progress and the lessons learned from the major events. Recently, the risk-informed analysis methodology combined probabilistic safety assessment (PSA) and traditional deterministic methodology has been a hot topic. Following the Risk-Informed Analysis Methodology, the PCT margin of CPR1000 Nuclear Power Plant was re-evaluated in this paper. In the PSA analysis, 162 probabilistic sequences had been identified after LBLOCA occurs. Then 18 probabilistically significant sequences were selected for the deterministic methodology analysis with deterministic realistic method (DRM) for CPR1000 NPP. With calculated PCT of each dominant sequence, a load spectrum of PCT for LBLOCA was generated. Then the risk-informed PCT margin can be evaluated by two different methods, namely the expecting value estimation method and the sequence probability coverage method. In conclusion, it was found that the PCT margin evaluated by the Risk-Informed Analysis Methodology can be greater than that of the deterministic DRM methodology by 16∼34°C.
Performances Analyses of Crud of EPR Fuel Assembly
Authors: Wang Xuming , Yuan Cenxi , Ye Chen
Published: October 2017

Taishan European Pressurized Water Reactor (EPR) is a third generation advanced pressurized water reactor (PWR), which adopts the third generation advanced fuel assembly (AFA-3G-LE) from AREVA for the first time. As suggested by American Electric Power Research Institute (EPRI), an EPRI level III crud risk assessment is necessary for new type of plants. Because crud induced power offset (CIPS) and crud induced local corrosion (CILC) can lead to axial offset anomaly (AOA) and fuel cladding failure, respectively.

A EPRI level III CIPS/CILC risk assessment for Taishan EPR is performed with a new framework of simulation by using sub-channel code FLICA, crud code BOA, and Monte Carlo transport code Tripoli-4. Such framework enables a self-consistent calculation, including a detailed description on neutronics contributed by boron. The validation of present work is confirmed because of the good agreement with the experienced data of EPRI. The results show that AFA-3G-LE has a good performance on crud risk assessment. Even in the worst case, the boron-10 deposition (2.6 g) and the maximum thickness of crud (59 μm) are lower than the low risk threshold, 31.33 g and 75 μm, respectively. Hence, It is expected that Taishan EPR has a very low risk on CIPS and CILC.
CFD Application to Subcooled Boiling Analyses on the First Wall of WCCB Blanket for CFETR
Authors: Jiang Kecheng , Ma Xuebin , Liu Songlin
Published: October 2017

In this paper, by using the CFD method, the efforts have been tried to conduct the investigation on the subcooled flow boiling in the first wall channel of Water cooled ceramic breeder (WCCB) blanket for CFETR. The detailed 3D distribution of temperature and vapor in the flowing passage have been presented under the heat flux of 0.5MW/m2 and 1MW/m2, respectively. Due to the high heat flux from plasma, the vapor distribution in the channel decreases from the plasma side to the breeder side along the radial direction. Undoubtedly, the volume fraction of vapor increases along the flowing direction because of the heating. Besides, the distribution of the channel wall along the toroidal direction presents the U-shaped tendency. As demonstrated by the results, the vapor is more likely to be generated at the corner of the square channel, and this can easily cause the Critical Heat Flux (CHF), which will destroy the structural integrity and materials melting. To avoid the enrichment of vapor in the corner, the optimization on the flowing channel has been performed by smoothing the corner. The results show that the volume fraction of vapor can be effectively decreased compared with the original square channel. Moreover, from the perspective of thermal hydraulics, the circular tube is the optimized channel which can not only avoid the concentration of vapor, but also can decrease the peak volume fraction.
Visualization Study on Onset of Nucleate Boiling for Natural Circulation Flow in a Narrow Rectangular Channel
Authors: Tian Chunping , Cao Xiaxin , Wang Jianjun , Yan Changqi , Yang Kuan , Zhu Zhiqiang , Xu Jianjun
Published: October 2017

Experiments on onset of nucleate boiling (ONB) have been performed for the natural circulation flow in a one-sided heated narrow rectangular channel. With fast camera putting into use, the visualization study was conducted to capture the location of ONB. The experiments were performed at pressure of 0.2 MPa and 0.3 MPa, with subcooling temperature ranging from 40 to 60 K, and heating flux ranging from 70 to 220 kW/m2. Differing from conventional channel, the superheats of wall surface in the narrow rectangular channel needed for ONB is larger, and can be predicted with Thom’s correlation in developed flow region. Moreover, the superheat needed for ONB will decrease for the existence of entrance effect. The visualization study indicated that the traditional determination of ONB will bring in more error. In this study, valuable data and image have been provided for the error analysis of ONB.
Control Scheme Research of 10MW Fluoride Salt Cooled High Temperature Experiment Reactor
Authors: Ruan Jian , Zou Yang , Li Minghai , Xu Hongjie
Published: October 2017

Fluoride salt cooled High temperature Reactor (FHR) is a kind of Gen-IV reactor which possesses many attractive features, such as high temperature, low pressure etc. Thermal-hydraulic features of molten salt are different from coolants of traditional reactors, which dominate operation transient behavior of FHR. However, as a new type reactor with sphere fuel element and fluoride salt coolant, FHR has inadequate operating experience and data used for reactor control and the design of power regulating system. Therefore, research of power regulation strategy is very important for FHR in automatic control operation and commercial application.

A code programmed in Fortran platform is used for investigating the system transient behavior, control logic and strategy. Based on the transient analysis code OCFHR for FHR, power control logic strategy is studied on a model of 10 MW Fluoride salt cooled High Temperature Experiment Reactor. OCFHR is a specialized code in FHR transient analysis, which contains point reactor model, simplified core thermal-hydraulic model, molten salt-salt exchanger and molten salt-air exchanger with a tube-shell type, control rod system and power regulation and control model. The control module of OCFHR uses the incremental PID controller to regulate control parameters and adopts the compound mode of control rod adjustment, load adjustment and molten salt flow adjustment, so that it can adjust the control rod position, primary and secondary molten salt flow rate and air flow rate of load at different operating power levels.

Two kinds of steady operation strategies are studied in this paper, which are a) constant outlet coolant temperature and b) constant average coolant temperature. The power level is regulated by control rod while the working temperatures are adjusted by shifting the load with weight coefficients of power and temperature deviations.

The results show that the incremental PID controller with optimized parameters can achieve the control requirement. Both of temperature control strategies gain great performances under 10%FPand 50%FP power regulation. The target power is reached quickly and accurately by using the incremental PID controller while the temperature control is very time-consuming. Compared with b), strategy a) has less temperature overshoot but larger power overshoot and longer adjusting time. The step wise power regulation for FHER is doable when a wide power adjustment range is needed and the simulation 10%FP treated as a step works well. Besides, the preliminary study of varying secondary coolant flow rate also indicates that the secondary loop plays an important part in restraining the deviation of secondary coolant temperatures during the process of balancing the power and load, so it is better to adjust the secondary coolant flow in terms of the power regulation range.
Investigation and Modeling of the In-Tube Heat Transfer of Fluids at Supercritical Pressure During Heating
Authors: Zhao Chen-Ru , Zhang Zhen , Bo Han-Liang , Jiang Pei-Xue
Published: October 2017

Investigations and numerical modelling are performed on the heat transfer to CO2 at supercritical pressure under buoyancy affected conditions during heating in a vertical tube with inner diameter of 2 mm. Numerical modelling are carried out using several low Reynolds number (LRN) k-ε models, including the model due to Launder and Sharma (LS), Abe, Kondoh and Nagano (AKN), Myong and Kasagi (MK) models. The numerical results are compared with the corresponding experimental data and the predicted values using the semi-empirical correlation for convection heat transfer of supercritical fluids without deterioration. The abilities of various LRN models to predict the heat transfer to fluids at supercritical pressures under normal and buoyancy affected heat transfer conditions are evaluated. Detailed information related to the flow and turbulence is presented to get better understanding of the mechanism of the heat transfer deterioration due to buoyancy, as well as the different behavior of various LRN turbulence models in responding to the buoyancy effect, which gives clues in future model improvement and development to predict the buoyancy affected heat transfer more precisely and in a broader range of conditions as they come to be used to simulate the flow and heat transfer in various applications such as in the supercritical pressure water-cooled reactor (SCWR) and the supercritical pressure steam generator in the high temperature gas cooled reactor (HTR).
Experimental Study on Natural Circulation in a Lead Bismuth Eutectic Loop
Authors: Shi Leitai , Su Guanghui , Bing Tan , Chen Ronghua , Tian Wenxi , Qiu Suizheng , Wu Junmei
Published: October 2017

In order to investigate the thermal-hydraulic characteristics of LBE in Accelerator Driven System (ADS), the Natural Circulation Capability Loop (NCCL) test facility was designed and constructed at Xi’An Jiaotong university in 2015. NCCL is a middle-scale experimental loop designed for investigating the natural circulation capacity, gas-lift pump enhancing circulation capacity and heat-transfer characteristic of LBE. For the natural circulation capability experiment, the loop is filled with argon gas at 0.2 MPa before filling LBE from store tank. The maximum temperature of LBE is 500 C°, while the normal operation temperature was maintained at 400 C°. In this paper, the LBE natural circulation characteristics were investigated with experiments in NCCL test facility. The study includes measurements on start-up of natural circulation and capability of natural circulation. Significant natural convection flow was observed in the experiments. It was found that the natural circulation was quickly established and stabilized due to LBE high thermal expansion property. It took only a few minutes to have a stable natural circulation prevailing from cold conditions. At the same time, the temperature difference between heating section and cooling section increase quickly and reach to the maximum value. And in the range of 10 minutes, a steady circulation can be performed. The natural circulation flowrate depends on the loop resistance, and the temperature difference between the hot leg and the cold leg, as determined by the power level and the heat sink capacity. The experiments show that the maximum flowrate for the natural circulation is 0∼0.81 kg/s.
Nuclear Safety Information and Education Strategy for the Public Developed by IRSN: How to Share Technical and Scientific Knowledge With the Public
Authors: Matahri Naoëlle , Choffel de Witte Ilma , Bigot Marie-Pierre
Published: October 2017

In France, since 2006, the legal framework has been reinforced in order to have a better compliance with the safety features of nuclear installations but also a relevant communication with the public about the nuclear risks and the nuclear decisions. Starting with the Nuclear Safety Transparency law of 2007 defining transparency in the nuclear field as “the set of provisions adopted to ensure the public’s right to reliable and accessible information on nuclear safety”, afterwards reinforced by the Law on Energy Transition and Renewable Energy in 2015. This law reinforced the transparency provisions, requiring not only transparent one way information but also public participation.

Even before the first transparency law, in France, communication consisted in a classic one way information process based on reporting incidents or events which occurred on Nuclear Power Plants or installations. Next it was decided to publish all technical inspection notifications on nuclear installations on the regulatory website. The French Technical Support Organization IRSN, and the Nuclear safety Authority ASN, promote these reports and publications through press conferences and nowadays also through twitter.

IRSN, way before the transparency requirements, made it one of their priorities to develop different methods and tools for the improvement of communication between Experts and Public promoting visibility and trustworthiness. Thanks to the new legal framework, the development of new tools to inform and engage citizens is accelerated. The traditional tools available are annual reports, newsletters, websites, magazines, Press data center, press conferences, etc...today completed with new tools such as YouTube, twitter, Facebook.

In addition, IRSN developed ways and tools promoting direct contact with the public, such as “Open House Days” allowing the public to discover work on site and to dialogue with Experts. In line with the “face to face” formula, IRSN implemented an Information and Education Strategy for the Public to enhance their Radiation Protection and Nuclear Safety Culture.

The objective of this article is to explain further each method developed and the support used to enhance the Public’s Radiation Protection and Nuclear Safety Culture.
Issues on Radiation Safety Supervision of Radioactive Material Transportation in China
Authors: Xuexin Wang , Weiqiang Wei , Rongyao Tang , Hongchao Sun , Dajie Zhuang , Renze Wang , Shutang Sun , Dongyuan Meng , Feng Yan , Guoqiang Li
Published: October 2017

China has been attaching greater attention to the safety supervision on the transport of radioactive material packages, one key part of which is the radiation monitoring to the radioactive material packages and vehicles. Issues on the radiation monitoring and safety supervision of radioactive material packages transportation discussed include: (1) Determination of proper monitoring location close to package surface; (2) Neutron radiation monitoring; (3) Choosing proper detectors for complicate radiation field monitoring; (4) Optimization of surface contamination monitoring. The requirement in Chinese standard and solution on above issues during practice are also given.
Simulation on Neutron Space-Time Kinetics for Accelerator Driven Sub-Critical System
Authors: Song Yingming , Gao Qingyu , Wang Ke , Guo Yaping , Zhang Lu , Yang Yongwei
Published: October 2017

Monte Carlo transport theory was applied to the variables space and time separated framework of neutron space-time kinetics calculation for Accelerator driven sub-critical reactor. The improved quasi-static approximation was combined with Monte Carlo neutron transport code (IQS/MC) for neutron space-time kinetic process of ADS sub-critical system. Besides, the IQS/MC simulation calculation program with visualization operation platform for ADS sub-critical system was developed simultaneously. The beam transient was analysed simulatedly based on the physical model of CIADS. Three-dimensional grid distributions of relative neutron flux of energy group were separated along time can be obtained by computing energy group separated of neutron flux, meanwhile the totally relative power, fuel temperature and outlet temperature of coolant at the core varied as the time were also obtained. The calculated results of IQS/MC method and point reactor method were compared, which agreed well with the relevant physics laws and verify that the IQS/MC method is applicable to the simulation of ADS neutron space-time kinetics and ADS neutronics transient security analysis.
Research on Flow Characteristics of Top and Vertical Containment Surface With Different Structures
Authors: Xinxin Pan , Jie Liu
Published: October 2017

Following an accident, the passive nuclear power plant CAP1400 mainly uses cooling water evaporation on the outer containment surface to remove the decay heat inside containment to outside environment to ensure the plant to a stable status. But the outside surface of the containment vessel is attached by a large number of different forms of the disorder structure including circumferential welds and longitudinal welds on the containment surface, and small local bulge, which may affect water film flow and coverage outside containment surface. Once a large amount of water splashes and can’t return to the containment surface, the heat removal capacity of the system during and after an accident will be significantly reduced, and the safety of power plants will be threatened.

This paper simulates the water flow characteristics on the containment surface. Firstly, a simplified analysis model of water film flow structure is established considering different forms of structures including the circumferential weld, small local bulge and etc. The circumferential weld is simplified with the horizontal protrusion. Local small uplift is simulated with point-like protrusion. Computational fluid dynamics (CFD) software STAR CCM+ is applied in this simulation with the volume of fluid method and surface tension model, and sets contact angle and other parameters according to the characteristics of the containment surface coating, so as to establish the CFD analysis method. The analysis results and water flow test results agree well. And then, the CFD analysis method is used to analyze the flow patterns of water film on the surface of the wall under different conditions.

The results show that the circumferential weld has a certain influence on the initial water film covering only. But the film coverage rate of the water film is stable after the water film is stable, and there is no loss of the water film. For the circumferential weld on the vertical wall, the effect on the water film coverage is little. Water film loss of low-flow is negligible, and water film loss of high-flow is no more than 1%. For the form of local small protrusions, little change of the flow characteristics of water film can be observed.

The structure form of the containment outside wall is various. According to the rules of the water film, including the water film covering width and the water film loss, it can be used as the design input of the accident analysis program WGOTHIC to guide the system design.
Boring Analysis in Solving Out-of-Tolerance of Wall Thickness of Surge Line in Nuclear Power Project
Authors: Zhu Huadong
Published: October 2017

The nuclear reactor coolant surge lines are installed in the reactor building to connect the pressurizer and hot section of RCL (Reactor Coolant Loop). After being adjusted through heating or spraying, the pressure in the pressurizer is transmitted to the main loop through the surge lines so as to maintain stable working pressure inside the main loop. The weld of surge line is required to have in-service inspection. According to the requirements for the inner diameter counter bore indicated in surge line groove machining drawings, the design standard states that “if the weld is subject to service inspection, the length of the counter bore shall be 2Tmin for pipe and Tmin for components and fittings. After the surge lines of XXX project arrived on site, it was found that the back wall thickness of internal boring exceeded tolerance by analyzing the minimum wall thickness in the boring processing area of the surge lines. In order to solve the problem of back wall thickness of these borings less than the designed wall thickness, this paper carries out specific analysis and gives solutions. As for the problem of out-of-tolerance of minimum wall thickness in the boring processing area inside the grooves in the middle section of surge lines; it is recommended to modify the diameter of supposed internal boring of the groove to 367.0.00mm to satisfy the requirement of minimum wall thickness. As for the out-of-tolerance of minimum wall thickness in the boring processing area on the inner wall of groove of the connection-line nozzle in both hot section of main pipe of surge lines and in the pressurizer, it is recommended to modify the angle of inner boring of groove of surge lines. Specific solutions include:

Solution 1:

Modify the internal groove form from being perpendicular to the end face as shown in the design drawing to being 0°–4° included angle between the internal boring face and the end face.

Solution 2:

Correct the inner diameter of pipe section S001F to the designed inner diameter, in order to satisfy the requirement of processing inner diameter and minimum wall thickness. Analyze and calculate the wall thickness based on the corrected inner diameter to obtain the data of wall thickness on the equipment interface end of the surge lines after groove preparation.

Solution 1 and Solution 2 were used simultaneously, which is to modify the groove design on the one hand and ask the manufacturer to correct the inner diameter on the other hand so as to meet the design scope. By applying the above-mentioned combined solutions, we have solved the out-of-tolerance of wall thickness of surge lines. At present, the groove processing technology of surge lines has been approved by Administration and the installation has already been completed on the whole, which testifies to been a successful practice.
Fusion: Go Small to Go Fast
Authors: Locatelli Giorgio
Published: October 2017

This paper presents the business case for small fusion reactors. The conventional view in fusion research is that “economy of scale” means that small reactors are not economically viable compared to their larger counterparts. However, empirical evidence shows how the importance of the “economy of scale” is overrated. Firstly several studies show that large investment projects are usually delivered over budget and late. Large projects (or megaprojects) are more likely to go over budget and larger the project greater the overall risk. On the contrary, small plants are more manageable investments. Firstly, for the same power installed, there is more chance to exploit the advantages from learning and co-siting economies. Since the overall investment is a fraction of a large plant, the overall “bankability” is better and the financing easier. Secondly, small plants are more easily usable for cogeneration and load following. This is becoming a fundamental design criteria for power plants to be delivered after 2030. Lastly, the division of a large investment into smaller investments provides investors with “degrees of freedom” to hedge some of the risks and exploit valuable opportunities. The “Real Options approach” is a mathematical framework able to price these options. In summary, small fusion reactors can represent a more credible and faster route to deployment than large fusion reactors.