bluuebunny/crossref_metadata_2025 huggingface dataset:
The Study About the Transient Characteristics of Metallic Fuel Sodium Cooled Fast Reactor in the Unprotected Loss of Flow Accident Based on the SAS4A Code
Authors: Qiao Pengrui , Hu Wenjun
Published: October 2017

Unprotected loss of flow accident (ULOF) is the most typical severe accident in sodium cooled fast reactor, which is focused by scholars civil and abroad. Metallic fuel has different safety characteristics with the oxide fuel as the important development direction of future sodium fast reactor, accident analysis of which is also a research focus at home and abroad. This paper bases on one Cooperation Research Project proposed by ANL and organized by IAEA, analyses the Shut-down Removal Test-45R of the metallic fuel sodium cooled fast reactor EBR-II in the US with SAS4A code, to research the transient characteristics of it in ULOF accident. Studies have shown that, metallic fuel sodium cooled fast reactor has very good inherent safety performance, which can reduce the reactor power in ULOF accident through the negative feedback itself.
Set-Point Function Compensation-Based Control for Steam Generator Water Level
Authors: Guo Dan , Xia Hong
Published: October 2017

Steam generator (SG) water level system is a highly complex nonlinear time-varying system. It is complicated at low power levels due to shrink and swell phenomena which must be considered for plant safety and availability. To improve the transient performance of the SG level subject to power adjustments, an innovative set-point function method is put forward in this paper. The set-point functions based on the inverse-control theory and the swell and shrink effect which generate a desirable reference input to the widespread cascade Proportional Integral Derivative (PID) controller of the level control system respectively. The set-point function can apply appropriate control to the feed-water flow rate duly depended on the pivotal time between the power adjustment decision and the real start time of adjustment. Finally, comparative simulation is carried out under the same condition of power adjustment. The simulation results demonstrate that the water level control system added set-point functions can restrain the disturbance and improve the transient performance effectively. The method added the Inverse Control-Based Set-Point (ICSP) function can achieve better control performances than the swell-based set-point (SBSP) function.
A Study on UF6 Transportation Accident Scenarios and Diffusion Model
Authors: Sun Shutang , Li Guoqiang , Sun Hongchao , Zhang Jiangang
Published: October 2017

This paper discussed UF6 leak scenarios in transportation accidents, it’s different from UF6 release accident occur in nuclear fuel cycle facilities. On transport condition, physical changes and direct chemical reaction with vapor will happen after UF6 release, and then diffusion. Combined with physical properties, chemical properties and straight-line Gaussian plume models, this paper established a new diffusion model for UF6. Based on the data of three UF6 release experiments carried by the French Bordeaux National Experimental Base and calculation data of HGSYSTEM/UF6 model, the new model was tested preliminary. The result shows that, this model can describe the diffusion of UF6 in environment accurately, and apply to UF6 transport accident assessment.
Development and Validation of the Parallel All-Speed CFD Code GASFLOW-MPI for Detonation of Premixed H2-Air Mixture in a Hemispherical Balloon
Authors: Xiao Jianjun , Breitung Wolfgang , Kuznetsov Mike , Travis Jack , Redlinger Reinhard
Published: October 2017

The objective was to develop a validated computational fluid dynamics (CFD) based approach for predicting hydrogen detonations and the mechanical loads. Applications of interest were scenarios relevant to hydrogen explosion risk assessment in nuclear power plant under hypothetical severe accident. Model developments were conducted within the framework of the parallel scientific computational tool GASFLOW-MPI thanks to its effectiveness, reliability and robustness in predicting all-speed flows. Validation was completed for hydrogen detonation phenomena in 3-D hemispherical hydrogen cloud. Excellent comparisons between experimental data and model predictions were observed. With the developed detonation modeling capability, the all-speed CFD code GASFLOW-MPI can be applied to model both turbulent dispersion and hydrogen detonation phenomena that occurred in the nuclear reactor containment during severe accident. Further model developments and validations will be performed for flame acceleration (FA) and deflagration to detonation transition (DDT).
Sodium Cooled Fast Bread Reactor HCDA Codes Development
Authors: Tai Shi , Dong-hui Zhang , Wen-jun Hu
Published: October 2017

Liquid metal fast reactor is one of the Gen IV nuclear power, it is necessary to analyze hypothetical core disruptive accident (HCDA) of FBR to ensure that the system can prevent the radioactive material from leaking out. The modified Bethe-Tait model is the primary method to analyze hypothetical core disruptive accident in the world. In order to better analyze the nuclear reactor hypothetical core disruptive accident in China, an improved B-T model is used. At present, on the basis of the improved B-T model, power distribution of the CEFR add to the progress. The results of comparison between the program and SUREX program in France show that the program model can simulate the nuclear reactor hypothetical core disruptive accident in China.
Unprotected Loss of Flow Analysis of a Million Kilowatt Traveling Wave Reactor Core
Authors: Hu Wenjun , Qiao Pengrui
Published: October 2017

Traveling wave reactor (TWR) is an innovation concept nuclear reactor, through the once-through deep burning, the proliferation of fuel can be achieved and the utilization rate of Uranium can be increased. TWR has the characteristics of long lifetime, deep burn up and nuclear nonproliferation, because of its physical character, which makes it to be an attractive innovation concept fast reactor. The China institute of atomic energy (CIAE) has designed a million kilowatt TWR core based on a breeding and burn principle, which has considered the current technological level of sodium cooled fast reactor. In this paper, based on the TWR core design scheme, considered the design of fuel assembly, neutronics and thermal-hydraulic, analyzed the Unprotected loss of flow (ULOF) accident in the TWR core with the SAS4A code, through which research about the transient safety characteristics of a million kilowatt travelling wave reactor core has been done. Analysis shows that the peak temperature of fuel, cladding and coolant in the TWR core have a certain margin from the safety limits through the negative feedback of itself in the ULOF accident, the core of the million kilowatt TWR demonstrates a good safety performance.
Research on Emergency System of Injection Estimation of Reactor Core Recovery and Reactor Core Decay Heat Removal
Authors: Feiye Liao , Pingting Jiang , Wang Liu , Dongyu He
Published: October 2017

One of the lessons learned from Fukushima accident is that the existing procedures used in Nuclear Power Plants (NPPs) are not executed effectively and quickly enough after such an extended accident, for the accident is complex and people are too nervous in such a situation. Thus, emergency system that helps to raise diagnosis efficiency is necessary. In the paper, a quick diagnosis system on injection estimation of reactor core recovery and decay heat removal injection estimation is developed to meet the urgent needs and strengthen requirements for the training and application among utilities and nuclear regulators. The system will assist regulators to quickly know whether the currently flow will probably recover the reactor core, or whether the current injection capacity is sufficient to quench and recover the reactor core, directly after input present parameters into the system. In the system, Matlab method is used, and intuitive insights are considered, which is propitious to give immediate graphical interface and reduce possibility of human error.
Analysis of Fast Reactor Fuel Irradiation Behavior in the MA Recycle System
Authors: Ozawa Takayuki
Published: October 2017

A recycle system for minor actinides (MAs) is currently studied to reduce the degree of hazard and the amount of high-level radioactive wastes. In this system, MAs will be recycled by reprocessing and irradiating as mixed oxide (MOX) with plutonium (Pu) and uranium (U) in a fast reactor. It was reported that MA-containing influences up to MA content less than ∼3 wt.% of the heavy metal amount on thermal fuel properties, i.e. melting temperature and thermal conductivity, would be slight. However, MA content of MOX fuels is expected to be ∼5 wt.% in the future recycle system for MAs, and MAs might affect irradiation behavior of MA-MOX fuels. The main influences of MA-containing would be increase of fuel temperature and cladding stress, and the important irradiation behavior of MA-MOX would be fuel restructuring, redistribution, helium (He) generation and cladding corrosion. Several irradiation experiments were performed in experimental fast reactor Joyo to study MA-containing influence on fuel behavior during irradiation. In addition, CEPTAR.V2 was developed so that fuel properties and analysis models would be included to evaluate the MA-MOX fuel behavior during irradiation. In this study, the MA-containing influences were evaluated with CEPTAR.V2 by using the results of highly americium (Am) containing MOX irradiation experiment, B8-HAM, performed in Joyo, and as a consequence, the irradiation behavior of Am-MOX fuels could be precisely analyzed and revealed as the influences of Am content.
The Effect of IRWST Simulation Method on the Analysis of Small Break LOCA Using RELAP5
Authors: Xu Caihong , Shi Guobao , Cao Kemei , Cai Xiaoyu , Qi Zhanfei
Published: October 2017

The In-containment Refueling Water Storage Tank (IRWST) provides low-pressure safety injection flow for passive CAP1400 Nuclear Power Plant (NPP) during Loss-Of-Coolant-Accident (LOCA) and subsequent Long Term Core Cooling (LTCC). The Passive Residual Heat Removal Heat Exchanger (PRHR HX) and the spargers of Automatic Depressurization System (ADS) stage 1∼3 are submerged in the IRWST. During small break LOCA, heat and mass are delivered through PRHR HX and ADS spargers to IRWST, and IRWST is heated up before its safety injection. However, numerical and experimental investigation has shown that IRWST is not perfect mixing, and thermal stratification exists. During ADS-4/IRWST initiation phase, the temperature of IRWST injection flow is of great importance, and is affected greatly by IRWST simulation method when modeling with system code like RELAP5. In this paper, two different IRWST simulation methods where one use multi channels in horizontal direction while the other use only one, are analyzed for CAP1400 SBLOCA with RE-LAP5, and their effects are compared. Finally, the better method which uses only one channel in horizontal direction is recommended.
Numerical Investigation on the Gas Filling Process of Hydrogen-Air-Steam Mixtures in a Semi-Confined Experimental Pipe
Authors: Li Min , Xu Youyou , Wen Xiaojian , Liu Songlin , Luo Guangnan
Published: October 2017

Hydrogen gathering in the containment may occur followed by a severe accident in a nuclear power plant. A flammable mixture can be formed when hydrogen is mixed with air. The ignition of the gas mixture could threaten the integrity of the containment. In order to provide technology base and experiment data for optimization of hydrogen safety technology of Chinese advanced pressurized water reactor CAP1400, a major project regarding hydrogen safety research of pressurized water reactor containment is underway. As an important part of the project, an experimental facility (A4Q-DH) for the study of hydrogen combustion will be built. Gas displacement method is used to filling the premixed hydrogen-air-steam mixtures into the experimental pipe. Flow behaviors of the gases in the pipe are complicated because fluid flow can be disturbed by the built-in obstacles and gas density can be changed with variation of gas composition concentration. Therefore, it is necessary to evaluate the effectiveness of the gas filling method.

In this paper, gas filling processes for the experimental pipe with different obstacles and gas composition concentrations were simulated using computational fluid dynamics software ANSYS Fluent. The results indicated that hydrogen-air-steam mixtures can be uniformly distributed in the experimental pipe within tens of seconds. The obstacles with modest blockage ratio in the pipe are conducive to shorten the required gas filling time. The hindering effect of annular obstacles is greater than the one of circular and square obstacles. The time required for air to achieve uniformly distribution increases with the increase of the inlet concentration of steam and hydrogen. However, the time required for hydrogen and steam to be evenly distributed in the pipe are relatively close regardless of the shape and blockage ratio of obstacle and the inlet gas concentration.
A New Numerical Method for Fuel Temperature Calculation
Authors: Xin Jin , Liu Xiaohan , Wei Xiaoyan
Published: October 2017

For most fuel rod codes, the time independent heat conduction equation, which is a steady heat conduction equation, is applied in fuel temperature calculation. However, it can affect the fuel temperature prediction in II condition, which the linear power has much change in some seconds. For improving the fuel temperature prediction in II condition, this paper gives a new numerical method, which combines classical thermal conduction integration method and the difference applied in time partial derivative. For guaranteeing the numerical method’s stability and convergence rate, the multi-dimension Newton-Raphson procedure are applied in fuel temperature calculation. This paper describes the theoretical deduction of the numerical method, and Halden fuel thermal conductivity model applied in fuel temperature calculation.

In order to verify new numerical method’s correctness, stability and convergence rate, the comparison between numerical solution and analytic solution is performed in 4 hypothetical conditions that the power transient duration is respectively 3s, 15s, 30s and 120s, the linear power changes from 15kW/m to 45 kW/m, and the fuel pellet surface temperature changes from 400 degree to 750 degree. And fuel density, specific heat and thermal conductivity are assumed as constants so that there exists analytic solution in this condition. The 4 hypothetical conditions have covered the worst II condition. According to the results in 4 hypothetical conditions, the fuel centerline temperature relative difference between numerical solution and theoretical solution is less than 0.6%, and the iterations are less than 5. So the numerical method possesses excellent correctness, stability and convergence, and this method has much potential in application in fuel rod code.
Study of Minor Actinides Transmutation in PWR MOX Fuel
Authors: Chen Shengli , Yuan Cenxi , Wu Jingxia , Zou Yaolei
Published: October 2017

The management of long-lived radionuclides in spent fuel is a key issue to achieve the closed nuclear fuel cycle and the sustainable development of nuclear energy. Partitioning-Transmutation is supposed to treat efficiently the long-lived radionuclides. Accordingly, the study of transmutation for long-lived Minor Actinides (MAs) is a significant work for the post-processing of spent fuel.

In the present work, the transmutations in Pressurized Water Reactor (PWR) Mixed OXide (MOX) fuel are investigated through the Monte Carlo based code RMC. Two kinds of MAs are incorporated homogeneously into two initial concentrations MOX fuel assembly.

The results indicate an overall nice efficiency of transmutation in both initial MOX concentrations, especially for two MAs primarily generated in the UOX fuel, 237Np and 241Am. In addition, the inclusion of 237Np has no large influence on other MAs, while the transmutation efficiency of 237Np is excellent. The transmutation of MAs in MOX fuel depletion is expected to be an efficient nuclear spent fuel management method.
Parametric Study of CSR1000 Thermal Hydraulic Stability Using Response Matrix Method
Authors: Ali Shahzad Muhammad , Tao Zhou , Juan Chen , Liang Liu , Xiao Zejun
Published: October 2017

A parametric study is performed on the thermal hydraulic stability of CSR1000 (a Chinese supercritical water-cooled reactor) using SCTHSAC (Single Channel Thermal Hydraulic Stability Analysis Code). This code uses the response matrix method for analyzing the system stability within time domain, a useful method to reveal detailed information about the operational thermodynamic parameters before and after the onset of instability in a nuclear power reactor. A system response matrix is built on each node of the single coolant channel by linearizing and perturbing conservation differential equations, which is solved to form an array of decay ratios. The maximum decay ratio represents stability of the whole system. A decay ratio map is generated to see the effect of power and inlet orifice diameter ratio on system stability. The parametric study of pressure, inlet temperature and moderator flow fraction gives vital information regarding thermal hydraulic stability of CSR1000 in a sliding pressure startup scheme. Using system response matrix method, a stability surface is plotted against power and flow fraction, providing a useful reference for designing a detailed startup scheme of CSR1000 in the future.
Test Simulation and Numerical Analysis for CAP1400 Reactor CB and CS Bypass Flow
Authors: Yu Fan
Published: October 2017

Core barrel and core shroud are important equipment in reactor. During reactor operation, there should be a portion of fluid flowing between core barrel (CB) and core shroud (CS). And the by-pass flow share should be limited. So there are several by-pass flow holes on the bottom of CS, which should ensure by-pass flow share meet the requirements. Through test and theory analysis for the holes and resistance in this paper, obtain the relationship of holes dimension and resistance, calculate by-pass flow share of actual reactor, to verify the design meet the requirements and provide support for the following reactor design.
Fault Tree Analysis Based on Dynamic Uncertain Causality Graph
Authors: Zhou Zhenxu , Zhang Qin
Published: October 2017

Fault Tree Analysis (FTA) has been widely applied to large, complex industrial systems like nuclear power plants, chemical systems, and weapon systems. Events in classical FTA are assumed binary-state and s-independent but multi-state, dependencies and logic cycles may exist within FTs. Moreover, causalities in FTA are assumed deterministic, while sometimes they may be uncertain. This paper applies Dynamic Uncertain Causality Graph (DUCG) in FTA to overcome aforementioned issues. This paper shows that any FT can be mapped into a DUCG graph. And with DUCG representation model and algorithm, additional modeling and analytical power are obtained. Multi-value, dependencies, logic cycles, and non-deterministic causalities in FTA are solved. This paper also depicts how to calculate the importance measurement, predict failure, and diagnose fault. The results reveal the effectiveness and feasibility of this methodology.
Analysis of the Containment Thermal-Hydraulic Behavior of a Small Reactor During Loss of Coolant Accident
Authors: Lin Qian , Zhang Weizhong
Published: October 2017

The containment thermal hydraulics of a small reactor during loss of coolant accident (LOCA) is studied by a lumped parameter one-dimensional model and a three-dimensional model. The capability of a kind of heat exchanger type passive containment cooling system (PCCS) is analyzed by the one-dimensional model. The calculation results show that, the decay heat can be removed and the containment pressure can be decreased by the proposed PCCS.

The steam and non-condensable gas (the air) distribution in the containment is investigated, the mixing and stratification behaviors are analyzed for several different cases, in which the PCCS and condenser are located at higher, base or lower position. The sensitivity analysis of the PCCS elevation shows that, in despite of the different gas stratification, the containment pressures are nearly the same. Similar conclusions can be obtained by the one-dimensional model and three-dimensional model. The preliminary results may indicate that, the designed PCCS and condenser can be located at a lower part, which will be benefit for the economy of the small reactor or meet other requirements.
Studies on the Mitigation Strategies for Combustible Gases in a NPP
Authors: Chen Meilan , Chen Peng , Liao Yehong
Published: October 2017

This paper studies the mitigation strategies for combustible gases in a NPP. Based on the PSA results and engineering experiences, one typical case and one bounding case are chosen for the study of the control systems for the combustible gases. The combustible gases control strategies for AP1000 and EPR are also discussed. The studies find that, the combustible gases control system should be designed along with containment heat removal systems. The containment heat removal system will cool the containment and condense the vapor in the containment, which will facilitate the effectiveness of the igniters. The advantage of the passive recombiners system is the ability to eliminate hydrogen beyond the flammability limit. But the recombination rate is slow, and at the early stage of the accidents, the hydrogen will still be at a high level. In this condition, the operation of the containment heat removal systems should be designed carefully to eliminate the negative hydrogen risks.
Application of a 3D Discrete Ordinates-Monte Carlo Coupling Method on CAP1400 Cavity Streaming Calculation
Authors: Zheng Zheng , Li Hui , Wang Mengqi
Published: October 2017

Neutrons and photons produced from reactor core during operation pass through the pressure vessel, reach the reactor cavity, and form the reactor cavity streaming. Reactor cavity streaming dose rates calculation during normal operation is important for the evaluation and control of the equipment dose rates in the nuclear power plant. Because reactor is great in dimension and complex in geometry, neutrons and photons fluence rates declined by several orders from reactor core to outside. Cavity streaming calculation is a deep penetration calculation with heavy computation load which is difficult to converge.

Three dimensional Discrete Ordinates and Monte Carlo (SN-MC) coupling method combines the advantage of the SN method with high efficiency and the MC method with fine geometrical modeling. The SN-MC coupling method decreases the tally errors and increases the efficiency of the MC method effectively by using MC surface source generated by the SN fluence rates. In this paper, the theoretical model of the 3D SN-MC coupling method is presented. In order to fulfill the coupling calculation, a 3D Discrete Ordinates code is modified to output angular fluence rates, a link code DO2MC is developed to calculate cummulative distribution functions of source particle variables on surface source, and a source subroutine is written for a 3D Monte Carlo code. The 3D SN-MC coupling method is applied on the calculation of the CAP1400 cavity streaming neutron and photon dose rates. Numerical results show that the 3D SN-MC coupling codes are correct, the relative errors of the results are less than 20% compared with those of the MC bootstrapping method, and the efficiency is greatly enhanced.
The Preliminary Shielding Analysis of CAP1400 Based on JMCT-S Code
Authors: Zhou Yan , Zheng Zheng , Mei Qiliang
Published: October 2017

JMCT-S[1] 3-D high resolution Monte Carlo neutron and photon transport code is proposed for shielding design due to its significant advantages such as complex geometry modeling, accurate calculation method, high efficient parallel computing and visible mass data analysis. To demonstrate its feasibility and benefits in engineering, this paper presents the verification of JMCT-S code based on primary shielding analysis for CAP1400 nuclear power station.

Firstly, by employing JMCT-S code this paper completes the modeling and simulations of CAP1400 nuclear power station with full-core pin-by-pin and reactor pressure vessel internals inside of primary shield wall[2]. Secondly, neutron fluence rates are calculated based on the equilibrium cycle information. Particularly, the full-core neutron source is calculated from 3-D power distribution, fissionable nuclide fractions, fission spectrum, particle numbers and energy released per fission. Finally, JMCT-S code is verified by comparing results with those of the MCNP[3] code.

According to the calculation results, JMCT-S code has similar accuracy to the MCNP code. The relative error of neutron fluence rate in active core is within ±1.5%, and deviation of neutron fluence rate for fast neutron fluence rate on inner surface of reactor pressure vessel is less than ±10%. Consequently, JMCT-S code is reliable in shielding design of reactor plant which lays the foundation for usage of JMCT-S code.
Introduction of Three Methods Used for the Nuclear Accident Diagnosis
Authors: Chen Peng , Xu Weifeng , Yang Fangqing , Liao Yehong
Published: October 2017

In order to deal with the nuclear severe accidents, the severe accident management systems are popularly considered and developed at home and abroad recently. A severe accident management system usually includes these functional parts: accident monitor, accident diagnosis, accident simulation, accident prognosis and SAMG support. Here, the accident diagnosis part is mainly concerned, and three nuclear accident diagnosis methods are introduced here, including BP neural network method, SDG expert diagnosis technique and artificial diagnosis method, which are also applied to a severe accident management system developed by us. In this paper, firstly, the severe accident management system developed by us will be introduced briefly. Then, three accident diagnosis methods for nuclear power plant (NPP) are showed and described in detail. At last, two cases including LOCA and SGTR accidents are used for the verification of these accident diagnosis methods and some analyzing results and conclusions are given. The results show that the three diagnosis methods are very useful for the accident diagnosis of NPP, which can diagnose the accident type accurately and offer much information or support to the severe accident management system and operators. The paper offers some reference significance for the research of accident diagnosis methods and the development of severe accident management system.
Application of Monte-Carlo Simulation to Design of Sampler and Detector in Radiation Monitoring System
Authors: Sugihara Kei , Sakai Hirotaka , Hattori Kanako , Tanaka Genki , Hayashi Mitsunobu , Ito Toshiaki , Oda Naotaka
Published: October 2017

In this study, the applicability of Monte-Carlo code PHITS(1) to the equipment design of sampler and detector in the radiation monitoring system was evaluated by comparing calculation results with experimental results obtained by actual measurements of radioactive materials.

In modeling a simulation configuration, reproducing the energy distribution of beta-ray emitted from specific nuclide by means of Fermi Function was performed as well as geometric arrangement of the detector in the sampler volume. The reproducing and geometric arrangement proved that the calculation results are in excellent matching with actual experimental results. Moreover, reproducing the Gaussian energy distribution to the radiation energy deposition was performed according to experimental results obtained by the multi-channel analyzer.

Through the modeling and the Monte-Carlo simulation, key parameters for equipment design were identified and evaluated.

Based on the results, it was confirmed that the Monte-Carlo simulation is capable of supporting the evaluation of the equipment design.
The Application of Fault Tree Method in Systematic Multiple Spurious Operation Analysis in Fire PSA
Authors: Yuehua Hu , Lin Li , Zhen Yan , Lei Song
Published: October 2017

Systematic multiple spurious operation review is the important basement of Fire PSA Component Selection and Fire Induced Risk Modeling. It is a task of systematically evaluation of all plant systems and identification of equipments susceptible to potential spurious actuation that may challenge safe shutdown capability via spurious operations, which can ensure that the fire PSA model addresses the impacts of single spurious operations (SSOs) and multiple spurious operations (MSOs). In this paper, different methodologies of systematic multiple spurious operation review are studied and it is found that two methodologies are given in literature published at home and abroad: one methodology is development of comprehensive plant specific MSO list by panel review. This methodology is based on generic list of MSOs combined with plant specific information review. This methodology proposes high requirements for the expert panel and hard to avoid missing of some scenarios. Another methodology is review of all plant systems and components to receive SSOs, and then using MSO matrix to evaluate and encode each unique pair to receive MSOs that disable the system function. This methodology can ensure the integrity of MSO list, but the matrixes are too complicated. In this paper, the fault tree method is employed to improve the matrix method in the MSO evaluation process of the latter methodology. Reactor Coolant System (RCS) is taken as an example to describe the application of this method to CAP1000. Engineering practice shows that the fault tree method can effectively avoid omissions of MSO scenarios, and it is more useful and simple for tertiary or quaternary MSO evaluation, as well as more convenient for further MSO evaluation due to cross system issues. Furthermore, MSO fault trees can be directly called by fire PSA model for evaluation, which can lay a foundation for the follow-up work of fire PSA.
Radiation Shielding Analysis of a Spent Fuel Dry Storage Cask for FA300 Spent Fuel Assemblies
Authors: Wang Mengqi , Pan Nan , Li Hui , Jia Baojun
Published: October 2017

Spent fuel dry storage technology is one of the most important intermediate storage technologies for spent fuel, because of its high security, good economic and easy to expand the scale. This article aims at designing a spent fuel dry storage cask which can contain 21 FA300 spent fuel assemblies.

The spent fuel dry storage cask is designed as concrete cask structure, which has the advantages of low manufacturing cost and simple manufacturing technology. Ventilation channels are designed for heating transfer, because the concrete is not a good thermal conductivity material. And labyrinth structure is designed for the ventilation channel to reduce the cavity streaming.

Radiation sources in spent fuel assemblies are mainly produced from fission products, actinides and their daughters located inside the effective fuel region, and other activation products in structure materials, which are calculated by ORIGEN.

The source and geometry of this problem are complex, and this is a real world deep penetration and streaming problem. Discrete ordinate method has great advantage in solving the deep penetration problem. Based on three-dimensional discrete ordinate code TORT, radiation shielding design method for spent fuel dry storage cask is studied, including main shield cask, cover lid, and ventilation channel. The results show that this spent fuel dry storage cask containing 21 FA300 spent fuel (cooling time: 10 years) assemblies can satisfy the requirement of dose rate limits in GB18871.
Analysis of the Onset Power of Heat Transfer Deterioration in Supercritical Water Based on Grey Correlation
Authors: Dongliang Ma , Tao Zhou , Shi Qi , Zejun Xiao
Published: October 2017

On the basis of the experimental data about the onset power of heat transfer deterioration (HTD) in supercritical water, using the analysis method of grey correlation model, select the mass flow rate, pipe diameter and the system pressure, analysis the connection degree of grey correlation between the onset power of HTD and these factors in supercritical water. Through the analysis we can conclude that: Firstly, mass flow rate has the largest correlation degree with the onset power of HTD, so they have the closest relationship; and secondly, the pipe diameter has the smaller correlation degree with the onset power of HTD; on the finally, the system pressure has the minimal correlation degree with the onset power of HTD. When selecting different resolution, to have different influence on the size of the calculated correlation degree. When the selection of resolution is smaller, the difference between the correlation degrees of each factor is greater. When the selection of resolution is 0.01, it can effectively distinguish the grey relational degree of each factor. Through the analysis of the grey correlation degree about the onset power of HTD, it can provide theoretical basis for the prediction of the onset power of HTD in supercritical water.
CFD Study on Specifics of Flow and Heat Transfer in Vertical Bare Tubes Cooled With Water at Supercritical Pressures
Authors: Zvorykin Alexander , Fialko Nataliia , Julii Sherenkovskyi , Aleshko Sergey , Meranova Natalia , Hanzha Mark , Bashkir Ivanna , Stryzheus Svitlana , Voitenko Artem , Pioro Igor
Published: October 2017

The paper presents results of a study on flow and temperature fields in bare tubes cooled with SuperCritical Water (SCW). This study is based on a Computational Fluid Dynamics (CFD) simulation with the FLUENT code for upward flows in vertical tubes with heated length of 4 m and an inner diameter of 10 mm. Operating conditions were: Mass flux – G ≈ 500 and 1000 kg/m2s; heat flux – q = 189 – 826 kW/m2; and inlet coolant temperature – Tin = 320–360°C.

CFD predictions were compared with experimental data in this study. All three heat-transfer regimes: 1) normal heat transfer; 2) improved heat transfer; and 3) deteriorated heat transfer; were considered. The obtained results show that within normal and improved heat transfer CFD predicts experimental values reasonably well. However, within conditions of deteriorated heat transfer CFD predictions are less satisfactory.

The CFD outcomes of the heat flux effect on the flow and heat transfer of SCW are presented. Specifics of flow within the pseudocritical region (i.e., approximately ±25°C around a pseudocritical point) are discussed. The buoyancy effect is investigated by axial velocity profiles at the medium mass flux of 500 kg/m2s and heat flux of 287 kW/m2.
New Design of Thorium Molten Salt Reactor Core and Preliminary Safety Analysis
Authors: Qi-chang Chen , Sheng-yi Si , Jin-kun Zhao , Hua Bei
Published: October 2017

In order to improve the breeding ratio and core safety, new thorium molten salt reactor (TMSR) core is designed. The new designed TMSR core is composed of hexagon moderator elements, which contain SiC tube to form a central fuel channel and employs BeO as moderator. The composition of the fuel salt, adopted in this core, is also optimized. Based on this core design, steady state and transient safety characteristic of TMSR are preliminarily analyzed using coupled multi-physics code. Power/temperature distribution and reactivity coefficients are analyzed for the steady state core, which demonstrated that the core has flat temperature distribution and large negative power coefficients at all power level. Transient simulations are carried out for power start-up, pump speed variation, loss of heat sink and so forth, the temperature and power response are also analyzed. The results indicate that the TMSR core power and temperature are closely related to the control rod position, velocity of flow and composition of fuel salt, and the new designed TMSR has excellent performance of safety under various operating conditions.
Two Phase Boiling Heat Transfer of Water in Minichannel
Authors: Tian Ye , Huang Wei , Li Pengfei , Cao Simin , Sun Yan
Published: October 2017

Printed Circuit Heat Exchangers (PCHE) is a new type of compact heat exchangers, it will be widely used for nuclear industry due to its higher heat transfer area density, compact size, and design flexibility. The hydraulic diameter of PCHE tubes ranges from 1mm to 2mm which belongs to mini-channel according to Kandlikar and Grande (2003)’s study.[1] In this paper, two-phase flow boiling heat transfer of water in mini-channel is discussed. The most of previous literatures in this field mainly focused on flow boiling of refrigerants, but the main working fluid in PCHE tubes is water. A composite correlation of flow boiling of water through mini-channel has been developed on basis of a database of water in this paper. Mean absolute error (MAE) method is used to evaluate relative error. Comparing with the experimental data, the MAE of the new correlation is 23.4%.
Measurement of the Vapor Blanket Thickness by Image Processing Method
Authors: Hou Fangxin , Chang Huajian , Zhao Yufeng , Zhang Ming , Chen Peipei , Gao Tianfang
Published: October 2017

It is widely believed that the behavior of vapor bubble/blanket over heating surface plays a critical role in determining the critical heat flux (CHF) in the subcooled flow boiling. Various CHF models are based on phenomenon observations of vapor bubble/blanket in the flow channel and use vapor bubble/blanket physical parameters to determine CHF values. In this study, subcooled flow boiling tests were conducted on the experiment facility “Test of External Vessel Surface with Enhanced Cooling” (TESEC). A series of natural circulation subcooled flow boiling CHF experiments is performed in a 30 mm by 61 mm rectangular flow channel with a 200 mm long heated surface along the flow direction at various inclination angles of the test section. With the assistance of high speed video technology, the process of flow boiling in the experiments was recorded and analyzed. A novel image processing method based on a MATLAB code is used to analyze high speed images at 999 frames/second and is able to provide detailed statistical information of vapor behavior on the heating surface. By this process, the static and dynamic information of vapor blanket is obtained at the pre-CHF conditions at various inclination conditions of flow channels (30 to 90 degrees). In addition, the Fast Fourier Transform (FFT) algorithm is used to further analyze the dynamic behavior of the vapor blanket.
Toward an Open-Source Neutronics Code for Circulating-Fuel Reactors
Authors: de Troullioud de Lanversin Julien , Glaser Alexander , Göttsche Malte
Published: October 2017

In circulating fuel reactors, such as the Molten Salt Reactor, the fuel circulates throughout the reactor instead of being immobile as in solid fuel reactors. The vast majority of nuclear simulation codes are primarily designed to simulate solid fuel reactors. Hence, many features unique to circulating fuel reactors, such as fuel injection and removal, cannot be properly modeled with these codes. The work presented here focuses on developing a numerical simulation package that can effectively and accurately model these reactors. This package consists of the coupling of the Monte Carlo particle transport code OpenMC with a modified version of ORIGEN-S, and uses a novel algorithm that calculates the optimal fuel injection and removal schemes for such reactors to achieve certain conditions such as a stable reactivity. We demonstrate our code’s accuracy by benchmarking the coupling module with the MCODE coupling code, and by simulating the operation of the ORNL Denatured Molten Salt Reactor using the coupling and fuel injection modules. The resulting fuel injection scheme is in agreement with the original study of that reactor while offering a much finer resolution for the injection scheme over time. This work is part of a broader project to develop an open-source neutronics code for circulating fuel reactors that will couple OpenMC with an in-house open-source depletion module.
Numerical Simulation of Radiation Aerosol Collection in Mesoscopic Impactor Filters
Authors: Zhang Wei , Niu Fenglei , Qi Houbo , Guo Zhangpeng
Published: October 2017

Typically, the concentration of natural radioactive aerosols in the containment vessel of nuclear power plant is low and in equilibrium. But when a serious nuclear accident occurred, the massive radioactive aerosols would be rapidly released. In order to ensure the integrality of the containment, the pressure inside the containment must be reduced by reducing the concentration of the aerosol. It can cause a serious damage to the atmospheric environment if such radioactive aerosol directly release. In this paper a new mesoscopic impactor filter has been developed which not only can filtrate and collect the aerosol particles but also can decrease the flowing resistance of gas. This paper intends to make numerical simulation to study the regularity of the deposition of aerosols under the laminar condition at different working condition in mesoscopic impactor filters. The 3D model of the filter was established with the commercial software of ICEM CFD and the meshes were divided accurately. The gas phase uses the laminar model and the particles use DPM (Discrete Phase Model). The detailed modeling method is given and the simulation results are analyzed.
Study of the Numerical Solving Methods of the Motion Model of Polydispersed Droplets
Authors: He Xiaoqiang , Liu Qianfeng , Zhao Chenru , Bo Hanliang
Published: October 2017

Moisture separation is one of crucial devices in PWR power plant for it plays an irreplaceable role in eliminating droplets from steam and supplying dry-saturated for turbines. It would be helpful to design and optimize the structure of moisture separator through analyzing the behaviors of droplets. This paper studies the numerical solving methods of the motion model of polydispersed droplets. The equations of model belong to the field of nonlinear stiff ordinary differential equations, thus backward differentiation formula, a kind of multi-steps methods are used which advance in solving stiff different equations. The coefficients of equations involve the velocity and rotation of flow field in separator, and these parameters are given by the output of Fluent. After solving the differential equations we can get the velocity and angle velocity of droplets in different locations and the movement track of droplets in separator, as well as the separation efficiency of moisture separator for polydispeased droplets with special distribution. Finally through the numerical solving of the motion model of polydispersed droplets in chevron-type separator, we find that multi-steps method improves the numerical stability and reduces the steps of iteration under the same calculation precision compared with classical methods such as fourth-order Runge-Kutta method. So it lays the foundation for development of moisture separator program suit for engineering computing.
A Fast Particle Swarm Optimization (PSO) Method for Detecting Control Valve Stiction
Authors: Zhang Dongming , Zhang Guangxin , Wang Xu
Published: October 2017

The control valve is one of the key fluid control components in a nuclear power plant, such as a Main Feedwater Control Valve, Pressurizer Spray Valve, etc. Stiction is one of the most common problems for control valves due to breakout static friction. Control valve stiction may lead to a poor performance, or even oscillations in the control loop. This paper presents a method based on parameters identification of detecting control valve stiction. It can detect the control valve stiction parameters quickly and quantitatively so that provides support for online detection or prospective maintenance of control valves.
Reliability Assessment of Reactor AP1000 Passive ERVC System
Authors: Chengxin Cui , Ting Huang , Lian Chen , Ying Chen , Lei Zhang
Published: October 2017

As the safety requirements level of nuclear power plant is gradually improving, more and more passive systems are used in the advanced type reactors. However, the reliability evaluation work for these passive systems is still in the primary stage. Therefore, in this paper the work is started based on the reliability evaluation process of passive system, simulates the physical process of passive systems by RELAP thermal hydraulics program, and evaluates the reliability of AP1000 External Reactor Vessel Cooling (ERVC) system. After the large number of calculations the cumulative density distribution curve of the temperature in lower head of the reactor pressure vessel is obtained, so by combining with specific success criteria the reliability of AP1000 ERVC system can be obtained, and this reliability evaluation result of ERVC system can be used in the PSA model to better guide the design of nuclear power plants and improve the safety of nuclear power plants.
Experimental Research on the Influence of Inlet Conditions of Corrugated Plate Dryer
Authors: Feng Mao , RuiFeng Tian , BoWen Chen , Wei He
Published: October 2017

Corrugated plate dryer is one of the key components of natural circulation steam generator in the pressurized water reactor nuclear power plant. It play a decisive role in the water separation system to making steam become a good quality after steam flow out of a vertical steam generator. With the increase of the reactor power in the future design, it requires that the performance of corrugated plate dryer must be increased in the characteristics of steam load, separating efficiency, volume and resistance to make sure that steam generators still produce qualified steam at higher steam loads and higher cycle magnitudes.

In this paper, we have done some experimental studies on the corrugated plate dryer. And we concentrated on airflow parameters on the separation performance and resistance performance of dryer.

(1) First of all, we have determined the experimental research program about this study, after this, we finished the design works of the test bench, and then we have completed the construction work based on the design works.

(2) Afterwards, the flow and separation mechanism of the wave plate dryer was experimentally studied by means of parameter measurement and analysis combined with observation and recording. The effects of different inlet humidities and inlet velocity on separation efficiency, pressure drop and re-entrainment were studied.

The above study shows that:

(1) The separation efficiency of corrugated plate dryer is related to inlet humidity and inlet airflow velocity. The separation efficiency of corrugated plate dryer is increase with inlet humidity goes higher, and increased to reach a highest point then goes down when inlet airflow velocity increased.

(2) The re-entrainment is related to inlet humidity. When inlet humidity increased, the inlet airflow critical velocity which lead to re-entrain would decrease.

(3) The pressure drop of dryer is related to the inlet airflow velocity. When the inlet wind velocity increases, the pressure drop increases.