bluuebunny/crossref_metadata_2025 huggingface dataset:
Enhancement of RPV Critical Heat Flux by Injection Flow
Authors: Chen Wei , Sun Canhui , Geng Jun
Published: October 2017

Under severe accidents, the reactor pressure vessel is flooded with water and the residual heat is removed by two-phase natural circulation through the flow channel between the reactor vessel and thermal insulation. If the heat flux of the outer wall is lower than local critical heat flux, the residual heat can be removed, and if the heat flux of the outer wall is higher than local critical heat flux, the reactor pressure vessel should be molten. For AP-type reactors, like AP1000 and CAP1400, critical heat flux of the reactor pressure vessel is the heat transfer limits of residual heat under severe accidents. Previous studies indicate that after severe accident a two-layer molten pool can be formed, namely metallic layer and oxide layer. Compared with oxide layer, in metallic layer, the heat flux more easily exceeds the heat transfer limits due to its low thermal resistance.

In this study, an approach was proposed to enhance local critical heat flux. This approach is expected to be used in local area around reactor pressure wall, like metallic layer, to increase the reactor pressure vessel intact probability under severe accidents. In this new approach, injection flow channels are added to the present flow channel by adding simple flow pipes from insulation near 60 to 80 degree where exceeding critical heat flux is most likely to happen. The fluid flow under external reactor vessel cooling (ERVC) condition is divided into two parts: one part is from downward (0 degree) to upward (90 degree) along the curved reactor pressure vessel and the other part is from injection pipe (about 70 degree) to upward. The fluid temperature from injection pipe is lower than that from downward due to residual heat from the reactor pressure wall. And hence, the local critical heat flux is likely to increase because of inject turbulence and low fluid temperature.

An experimental facility is conducted to study the mechanism of injection influence on critical heat flux under normal pressure condition. There are two main loops in this facility: one is main loop while the other is injection loop. The test section is an inclined downward heated rectangular channel with its inclined angle varied from 0 degree to 90 degree. Flow and thermal conditions are listed: in main loop, mass flow velocity ranging from 100kg/m2s to 600kg/m2s with fluid temperature from 90 °C to 105 °C; In injection loop, mass flow velocity ranging from 0 to 600 kg/m2s with fluid temperature from 85 °C to 105 °C. Under the above condition, with and without injection flow, critical heat flux experiments were conducted. It indicates that injection velocity has great effect on critical heat flux, while injection subcooled has little effect. The critical heat flux can be increased by 0.07MW/m2 to 0.33MW/m2 depending on various injection velocities and main loop conditions.
The Current Research Status and Technical Development Framework of Multi-Reactor Probabilistic Consequence Assessment
Authors: Ding Hongchun , Tong Jiejuan , Zhang Liguo , Wang Lu , Wang Ruiying
Published: October 2017

The Fukushima Daiichi Accident in March 2011 gave rise to public attention on the possibility and severity of multi-reactor accident, i.e. multiple reactors and spent fuel pools on one site affected by an accident simultaneously or one after the other. Multi-reactor sites are a common phenomenon among nuclear development countries. The current safety assessment and regulatory requirements are established in the light of one reactor in a site. However, the reactors in a site are not independent with each other. Due to these dependencies, there are potential risk comes from multi-reactor accidents. Under this condition, many non-linear components will be involved into the evaluation of radiological consequences. Thus the Site Probabilistic Risk Assessment (SPRA) is not the linear superposition of PRAs in each reactor. From this point of view, SPRA should be developed according to the characteristics of multi-reactor nuclear accidents so as to quantitatively assess the risk of multi-reactor sites. Probabilistic Consequence Assessment (PCA) as part of the Level 3 PRA estimates the offsite consequence combined with the source term and occurrence frequency of accident given by the Level 2 PRA. PCA is an important part and tool to evaluate the overall social risk and verify the safety goals of NPPs. Multi-reactor PCA is an essential technical tool to discuss the multi-reactor issues, such as the verification of safety goal and the development of emergency planning zone (EPZ) in multi-reactor sites. In this paper, the current research status about multi-reactor PCA in main nuclear countries and nuclear organizations are elaborated and a technical development framework is proposed aimed at the multi-reactor site based on MELCOR Accident Consequence Code System (MACCS).
A Modeling Method of the Branch Probability of the Temperature-Induced SGTR Phenomenon
Authors: Liu Zibin , Guo Dingqing , Zhang Bing , Wang Jinkai
Published: October 2017

The phenomenon of temperature-induced steam generator tube rupture (TI-SGTR) is a typical phenomenon in the severe accident process of nuclear power plants. The occurrence of the phenomenon may result in the radioactive material bypass the containment, causing a large radioactive release. This paper investigates modeling methods of the phenomenon of temperature-induced SGTR in level 2 PSA and presents an optimizing modeling method to calculate the probability of branching probability of TI-SGTR, aiming at improving the rationality and veracity of level 2 PSA.
Investigation on Accident Progression and Melt Behavior at the Fukushima Daiichi Units 1 and 2 Using MELCOR Code
Authors: Li Gen , Zheng Shan , Yamaji Akifumi , Chong Daotong , Yan Junjie
Published: October 2017

The accident progression and melt behavior at the Fukushima Daiichi Units 1 and 2 were investigated using MELCOR 2.1. In the modeling the lower head failure mechanism by penetration tube rupture and ejection was modeled. In the modeling of Unit 2, according to the latest findings by TEPCO investigation, the possibilities of torus room flooding, RCIC piping leakage and thermal stratification in suppression pool were taken into account. The analysis results indicate that for Unit 1 when considering penetration tube failure, a part of debris still remained in the lower head after debris discharge; otherwise all the debris discharged out. The present MELCOR modeling of Unit 2 well reproduced the RPV and PCV pressure. A part of the core was damaged and the debris that slumped into the lower head was sufficiently cooled down. The pressure vessel kept intact.
Application of the Burnable Poison in the Design of Space Nuclear Reactor
Authors: SanBing Wang , Qilin Xie , ChaoHui He
Published: October 2017

The previous research showed that the application of burnable poison was helpful to improve the criticality safety of space nuclear reactor (SNR). In order to analyze the worth of burnable poison in the SNR’s design, a model of homogeneous reactor had firstly been built based on the design of SAFE400. Comparing its difference with the real design of SAFE400 through criticality calculation, the precise of our model had been verified. Then the influence of the criticality parameters and immersion accident character parameters for this model had been analyzed for the application of the different burnable poisons (such as samarium, europium or gadolinium). The calculation results had shown that the application of the most of the burnable poisons would soften the neutron spectrum and induced a decrement of reactor’s keff in the beginning of life. However, the immersion accident analysis gave out another result that only the reactor using gadolinium could ensure the criticality safety of reactor after it made its initial keff equal with the design value. Meanwhile, compared with the initial design of SAFE400, in one hand, the burn-up results had shown that the decrement of homogeneous reactor’s reactivity using gadolinium as burnable poison was deceased after the 10 years full power operation; in other hand, its neutron spectrum became more softer with the operation time; and what’s more important, the amount of the burnable poison was not decreased with burn-up during its service life-time. These results implied that the application of the burnable poison (especially for gadolinium) could highly ensure the criticality safety and stable operation of SNR.
Study on Water Hammer in the Overpressure Discharge Pipe of Nuclear Power Plant
Authors: Xu Qingmu , Cai Kun , Qin Jie , Yuan Junkai , Li Juan
Published: October 2017

Water hammer phenomenon is a significant pressure wave in pipe system caused by momentum change when the moving fluid is forced to stop or change direction instantaneously. Common causes of water hammer are sudden valve closing at the end of a pipeline system, pump failure, check valve slam etc. The steam transportation pipeline system may also be vulnerable to water hammer when it confronts with the situation where liquid and steam co-exist. Water hammer often occurs when steam condenses into water in a horizontal section of steam piping. Then steam “picks up” water to form a high-velocity “slug” and create extra stress to pipe. When steam is trapped into sub-cooled water, the collapse of vapor cavity can lead to collision of two columns of liquid, resulting in a large rise in pressure which will damage pipes, supporting structures and hydraulic machinery. Nuclear power plant is composed of complex equipments and piping systems, lots of which contain both liquid and steam. Hence, there is a potential threat of occurrence of water hammer to the normal operation of systems. Thus, this phenomenon needs to be well investigated and prevented with some effective methods.

For the purpose of overpressure relief under severe accidents, the spent fuel pool cooling system of CAP1000 series nuclear power plant provides a discharge passage from containment to spent fuel pool. When the containment pressure exceeds the control value, valve is opened to discharge high-temperature and high-pressure steam until the pressure drops to a safety value. During this process, serious water hammer happens, causing pressure rise beyond the design pressure and further leading to damages to pipes and structures. Therefore, water hammer of overpressure discharge pipeline in CAP1000 plant is studied in this work. On the basis of verification of the capabilities of computational code RELAP5/MOD3.3, hydraulic transient of water hammer is simulated under different conditions. It is indicated that after steam discharge stops, residual steam in pipe condenses because of contact with sub-cooled water in spent fuel pool. Subsequently, the rapid backflow and vapor cavity lead to a severe water hammer. The detailed analysis has shown that water temperature of spent fuel pool has a decisive influence on the mechanism of water hammer phenomenon, including collision of liquid column to valve disc and cavity collapse in the horizontal pipe. The collision and separation of liquid column result in relatively lower pressure amplitude.
Forced Convection Heat Transfer for Helium Gas Under Exponentially Decreasing Flow Condition
Authors: Liu Qiusheng , Wang Li , Shibahara Makoto , Fukuda Katsuya
Published: October 2017

Knowledge of the heat transfer phenomenon during flow decay transient condition is important for the safety assessment of very high temperature reactor (VHTR) during the loss of coolant accident. In this study, transient heat transfer from a horizontal cylinder to helium gas under exponentially decreasing flow rate condition was experimentally studied. The experiment was performed by using a forced convection heat transfer test loop. A flow control value with its control system was used to realize the flow decay condition. Helium gas was used as coolant and platinum cylinder with 1 mm in diameter was used as the test heater. A uniform heat generation rate was added to the cylinder by a power source. The cylinder temperature was maintained at an initial value under a definite initial flow rate of the helium gas. Then, the mass flow rate of the helium gas starts to decrease exponentially with different time constants ranged from 4.3 s to 15.4 s. The initial flow velocity ranged from 10 m/s to 4 m/s. The surface temperature, heat flux, and heat transfer coefficient were measured during the flow decay transient process under wide experimental conditions such as initial flow rate, flow decay time constant. It was found that the temperature of the test heater shows rapid increase during this process, the increasing rate of the temperature is higher for a shorter time constant. The heat transfer coefficient versus time during the flow rate decreasing process was also obtained. The transient heat transfer process during exponentially decreasing flow rate condition was clarified based on the experimental data.
Research on Impact of Pressurizer Heater Block on SGTR Overfill Analysis
Authors: Lixin Liu
Published: October 2017

Steam generator overfill due to steam generator tube rupture have been analyzed for CAP1400 plant, in which two cases were simulated, including pressurizer heater switching off available and unavailable under low pressurizer water level. The results indicate that it has a certain margin to steam generator (SG) overfill for the ruptured steam generator under steam generator tube rupture (SGTR) accident. Analysis results have also shown that the SG overfill margin will be slightly decreased in case of pressurizer heater swiching off unavailable, however, the impact is not significant.
Modeling the Natural Circulation Influenced by Earthquake
Authors: Zhang Yinxing , Gao Puzhen , Zhou Jian , Chen Hanying , Ji Peng
Published: October 2017

More and more attentions have been paid to the impact of the earthquake on the reactor thermal hydraulics after the Fukushima accident. In order to figure out the seismic effect on the flow in primary loop, the natural circulation in primary loop of CPR1000 influenced by earthquake is modeled by learning from the study on natural circulation under ocean conditions. Natural circulation ability at earthquake, which includes longitudinal earthquake and transversal earthquake, is dynamically simulated by a program written with MATLAB, respectively. Besides, the safety of the system is discussed based on the calculated results.

The research results show that, once the longitudinal earthquake occurs, in the earlier stage of earthquake, the peak and valley values of the coolant mass flow rate decrease gradually and the amplitude will be a constant as time goes on because of flow fluctuation in seismic condition. On the contrary, as for transversal earthquake, the peak and valley values of the coolant mass flow rate increase gradually in the earlier stage of earthquake. However, the amplitude will be a constant as time goes on, just like that in the longitudinal earthquake.

For both of the two conditions, they have the same characteristics. The constant mentioned above grows up with the increase of the earthquake acceleration peak value and the period; the fluctuation period of mass flow rate is generally the same as the period of earthquake acceleration and the mass flow rate changes steadily; the average mass flow rate decreases slightly with the increase of the earthquake acceleration peak value and the period.

Both of the vibration period and the peak of seismic wave have disturbing effect on reactor coolant system natural circulation. And the increase of the peak of the earthquake acceleration as well as the period of the earthquake acceleration will reduce the natural circulation ability. In the range of research parameters, the system is considered to be safe according to the calculation results of core coolant temperature.
Accident Safety Evaluation Method for Spent Fuel Dry Storage Facilities
Authors: Yubin Zhang , Yong Ouyang , Yuwei Zhou , Jinlin Liu
Published: October 2017

Dry storage is one of the ways to store spent fuel in the middle of the reactor, which can effectively alleviate the pressure of the storage on the spent fuel pool of nuclear power plant. This paper tries to combine the site of dry storage facilities and the design characteristics to explain and discuss the safety evaluation method under the accident condition, from the mechanical analysis, critical safety, the decay heat removal, the shielding design and so on. Then according to the operating procedures and the accident condition that may be occurred, put forward some possible ways of monitoring and measures of safety protection should be added.
A New Low Enrichment Uranium Core Design of MNSR
Authors: Mengjiao Wang , Yiguo Li , Xiaobo Wu , Dan Peng , Jingyan Hong , Qian Hao , Jinhua Zhang , Jin Lu
Published: October 2017

The Miniature Neutron Source Reactor (MNSR) is a low-power research reactor, which uses 90% high enriched uranium (HEU) fuel. However, due to the nuclear safety risk, and according to the principle of nuclear non-proliferation, MNSR must be gradually converted from HEU to low enriched uranium (LEU), which means the LEU fuel with U-235 enrichment less than 20% should be used.

The prototype MNSR of China Institute of Atomic Energy has completed the transformation, but other commercial MNSRs have not finished, which is different with the prototype in the application and structure. Therefore, using MCNP code to simulate, calculate and optimization design LEU core has been done in this issue. Firstly, UO2 with U-235 enrichment of 12.5% was selected as the fuel pellet of LEU core, keeping the rest of the core unchanged. The Φ, excess reactivity and the worth of the central control rod are calculated and analyzed. The results show that the commercial MNSR of LEU conversion is feasible. Secondly, in this paper, through changing the fuel elements and the arrangement method, the new low enriched uranium (NLEU) core was designed to improve Φ/P ratio of the core, the proportion of thermal neutrons and the worth of the control rod. UO2 with U-235 enrichment of 19.75% was selected as the fuel pellet of the NLEU, NLEU not only meets the design parameters, but in many parameters, NLEU is better than LEU. The fuel element quantity is reduced by 43%, from original 344 to 196; reducing the amount of U-235 loading; improving the Φ/P ratio and the thermal neutron fraction is increased. The results show that the NLEU optimizes some parameters, simplifies the core structure, saves the construction cost, improves the nuclear safety and is more suitable for the application of MNSR.
Newly Developed Coupling Scheme in RMC With Internal Thermal Feedback and OTF Doppler-Broadening Method
Authors: Yexin Ouwen , Huang Shanfang , Wang Kan
Published: October 2017

RMC (Reactor Monte Carlo)[1] is a self-developed Monte Carlo code for nuclear reactor analysis by Reactor Engineering Analysis Lab (REAL), Tsinghua University. On the basis of the self-developed subchannel module (RMC-TH) and Monte Carlo Cell Tally, the internal coupling interface is developed, which combines both input files to one and realizes the fast mesh correspondence process using the cell expansion technology for repeated structure with thermal-hydraulics feedback. It breaks through the bottleneck of geometrical extensibility for coupled code. On-the-fly Doppler broadening method is adopted as the way to consider the temperature effect on microscopic cross section, which only needs the 0 K cross section library so that the memory cost can be apparently reduced. Steady state simulation analysis are performed on PWR fuel pin and 17×17 assembly model, and the results show the feasibility, accuracy and efficiency of the coupling methodology. Therefore, a promising technology roadmap for the large-scale and geometrically universal nuclear reactor in both steady-state and transient conditions with thermal-hydraulic feedback are established. The roadmap can be further applied to neutronics-thermal-hydraulics-depletion coupling in multi-physics simulation process.
Local Single-Phase Heat Transfer Research of Natural Circulation in Narrow Rectangular Channel
Authors: Zhu Zhiqiang , Cao Xiaxin , Yan Changqi , Tian Chunping
Published: October 2017

In order to explore and analyze the heat transfer characteristics in narrow rectangular channel, experiments on local single-phase heat transfer of natural circulation in a one-side heating narrow rectangular channel have been conducted under vertical and inclined condition. The thermotechnical parameters such as inlet temperature, heat flux and inclination angle varies during the experiments. The width of the flow channel is 40 mm and the narrow gap is 2 mm. It is heated from one side with a homogeneous and constant heat flux and the working medium is deionized water.

Based on the experimental results, under vertical condition, the driving force in the loop goes up and the Reynolds number also increases when the inlet temperature is elevated, which causes an increase in local Nusselt number. When the heat flux rises, the local Nusselt number increases and the heat transfer temperature difference increases. The local Nusselts number is influenced by entrance effect and the entrance region length is computed for laminar and turbulent flow. Under inclined condition, with the inclination angle from −30° to 30°, it is found that when the inclination angle is positive, the local Nusselt number in fully developed region is larger than that under vertical condition and increases with the angle value, even though the Reynolds number decreases by the effect of incline. This phenomenon is explained by giving an analysis of the natural convection, which is characterized by the normal Grashof number, in the direction perpendicular to the heating plat. Moreover, the variation of heat transfer is also interpreted on the basis of field coordination principle. However, when the inclination angle is negative, the heat transfer shows no obvious difference between vertical condition and inclined condition.
Numerical Simulation of Premixed Hydrogen-Air Combustion Behavior in a Closed Vessel
Authors: Xu Youyou , Li Min , Wen Xiaojian , Liu Songlin
Published: October 2017

A closed vessel of 1.9 m length and 0.9 m in internal diameter is being designed and will be constructed to experimentally investigate combustion and propagation behavior of the hydrogen-air mixture. Before the experiment is performed, the benchmark and pre-analyses on the hydrogen combustion and propagation behavior in this vessel are carried out by using FLUENT CFD code. The benchmark results indicated that the peak overpressure verse time agrees well with the experimental results of published data. Furthermore, the effect factors such as the hydrogen concentrations, the ignition position, the initial temperature and the pressure were simulated. The results show that the ignition position, initial temperature and pressure have significantly influence on hydrogen behavior. In addition, the overpressure increases with the equivalent ratio and reaches the peak in the vicinity of stoichiometric ratio. However, the overpressure peak is considerably lower than the design pressure of the vessel.
Seismic Risk Analysis of the 380VAC Emergency Electrical Power Distribution Cabinets of a Nuclear Power Plant in China
Authors: Yang Jianfeng , Wang Handing , Zhang Xiaoming , Feng Bingchen , Wang Weijin , Wang Zhao
Published: October 2017

According to the research of the operating principle, installation position and running environment of the 380VAC emergency electrical power distribution cabinets (Hereinafter referred to as electrical cabinets) of a nuclear power plant in China, there are three aspects caused by earthquake that seriously affect the safety of the electrical cabinets, including relay chatter, failure of electrical cabinet structure and spatial interactions. Relay chatter refers to contacts of the relay being changed during the period of strong shaking. It may lead to associated circuits malfunction and the equipment failure of the relay control unless it can be effectively reset. The purpose of relay chatter is to find out these relays whose consequences are unacceptable after earthquake and calculate failure probability. Failure of electrical cabinet structure in the earthquake is to carry out seismic fragility evaluation. The goal of seismic fragility evaluation is to assess a given value which describes the ground acceleration capacity and the corresponding uncertainties, and then, the conditional probability of failure as a function of peak ground acceleration [PGA] and a family of fragility curves can be obtained. In this paper, finite element model of the electrical cabinet is established using ANSYS Workbench software. According to the electric cabinets seismic failure mode, we take some of the parameters including the parameters of the floor response spectrum, material strength parameters and so on as the input to calculate the median ground acceleration capacity and the corresponding uncertain parameters. The seismic spatial interactions are defined as the electrical cabinet destroyed due to the surrounding objects failure by falling, collapse, etc. Therefore, if necessary, it is needed to evaluate the seismic fragility of the surrounding objects. Usually through walking down, checking the design drawings or the combination of the above methods, we can find out the surrounding objects for an electric cabinet. So we analyze the seismic risk of the electrical cabinet from the above three aspects. When the results of the above three aspects obtained, we convolute of the electrical cabinet fragility with the seismic hazard curve which represents the frequency of occurrence of earthquake motions at various levels of intensity at the site. Then Monte Carlo sampling is adopted to analyze the uncertainty distribution. In this article, Risk Spectrum Professional software (reference 8) and Risk Spectrum Hazard lite software (reference 9) are used to complete the calculation and get some quantitative seismic risk insights. The above seismic risk insights can support the establishment of seismic probabilistic safety analysis model (Hereinafter referred to as SPSA) for a nuclear power plant, which helps to formulate seismic improvement strategies.
Research on Method and Characteristics of Aerosol Natural Removal in Containment After Accidents
Authors: Fu Yaru , Geng Jun , Sun Dawei , Mei Qiliang , Huang Gaofeng , Pan Nan
Published: October 2017

For passive nuclear power plants, the radioactive aerosols in containment are removed by natural processes after accidents. These processes have high reliability. However, it is very complicated to analyze the behaviors of aerosols in natural processes. The dominant processes include coagulation, sedimentation, diffusionphoresis, and themophoresis. The main work of this paper is summarized as follows: (1) To determine methods of analysis radioactive aerosol natural removal coefficient in containment. (2) To complete comparative analysis natural removal processes to AP1000 and CAP1400 after LOCA. (3) A comprehensive sensitivity analysis is carried out for a number of parameters, including containment free volume, sedimentation area, aerosol size, packing fraction, mass ratio of radioactive and nonradioactive aerosol, and steam condensation rate etc, (4) To complete comparative analysis of natural removal processes between core meltdown and non-meltdown accident sequence after LOCA. The results show that, (1) Removal coefficient by sedimentation of CAP1400 is larger than AP1000, removal coefficient by diffusionphoresis and themophoresis of CAP1400 also smaller than AP1000. (2) In general, the greater the containment free containment, the smaller the aerosol natural removal coefficient, and the greater aerosol size, the more the amount of aerosol removed by sedimentation, in the case of other parameters unchanged. (3) The removal coefficients have few differences after core meltdown and non-meltdown accidents.
Prediction of Subcooled Water Flow Boiling Critical Heat Flux (CHF) at Low Pressure
Authors: Hou Fangxin , Chang Huajian , Zhao Yufeng , Zhang Ming , Chen Peipei , Gao Tianfang
Published: October 2017

In vessel retention (IVR) is one of the key severe accident mitigation strategies to maintain reactor pressure vessel (RPV) integrity. IVR designs utilize the reactor pressure vessel lower head to contain molten fuel and rely on external reactor vessel cooling (ERVC) to remove decay heat. The capacity of ERVC is limited by the critical heat flux (CHF) of flow boiling on the outside of the reactor vessel surface. Therefore, the determination of critical heat flux (CHF) is crucial to predict whether the adoption of IVR would be successful in mitigating severe accidents. In 1999, Celeta et.al proposed a superheated layer vapor replenishment model. In this model they postulated that CHF would occur when the superheated layer was occupied by the vapor blanket coming into contact with the heated wall and they successfully predicted the critical heat flux in subcooled water flow boiling under high mass flux, high liquid subcooling and low/medium pressure conditions. To evaluate the practicability of this model in predicting CHF under IVR conditions, CHF experiments were performed under natural circulation conditions on the experiment facility “Test of External Vessel Surface with Enhanced Cooling” (TESEC). Experiments are conducted in a 30 mm wide, 61mm high rectangular flow channel with a 200 mm long heated surface along the flow direction. Two quartz windows are installed at the sidewalls of the flow channel for visualization. In order to simulate various positions of the reactor lower head, experiments at different inclination angles of the test section were conducted. The high speed visualization data at CHF point at various inclination angles were processed and analyzed by a MATLAB code developed by the author. The vapor blanket thickness at various inclination angles was measured from the visualization data and was also predicted by the Celeta model. By using geometry data from high speed images, CHF values were calculated by Celeta model and compared with the experimental results at various inclination angles. Limitations of the Celeta model in adaptation of predicting CHF under IVR conditions were further discussed.
Numerical Comparison and Analysis of Nuclear Core Design Code System SCAP and MPACT
Authors: Yang Weiyan , Tang Chuntao , Zhang Hongbo , Yang Bo
Published: October 2017

SCAP (SNERDI Core Analysis Package) is a nuclear core design code system newly developed by Shanghai nuclear engineering research and design institute (SNERDI), which includes three main codes: core calculation code CYCAS, automatic link code AutoLink and lattice code PANDA. By numerical comparison and analysis between SCAP system and the next generation code MPACT developed by university of Michigan, on the one hand to validate the accuracy and applicability of SCAP system, on the other hand to analyze the performance in actual nuclear power plant simulation between the second generation and the next generation codes. The benchmark applied to compare is the first cycle of the U.S. in service nuclear power plant Watts Bar # 1. The reference values include the measured data and results calculated by Monte Carlo code KENO. The numerical results indicate that, for the general core parameters, such as critical boron concentration and control rod worth, the accuracy of SCAP and MPACT is similar; for the local parameters, such as peak power factors FΔH and FQ, some biases are observed.
Out-of-Pile Properties Investigation of UO2-BeO Fuel Pellet
Authors: Li Jiwei , Ding Yang , Liu Wentao , Bi Guangwen , Zhao Ruirui , Zhou Qin
Published: October 2017

In order to increasing the thermal conductivity of fuel pellet used in nuclear power plants, a UO2-BeO composite fuel was developed. The fuel pellets with different beryllia addition were manufactured by sintering and the physical properties were tested and compared with uranium dioxide fuel. The metallograph show that the continuous phase of beryllia was formed. The results show that the thermal conductivity was obviously larger and the elasticity modulus and the coefficient of thermal expansion were a little smaller than uranium dioxide. The thermal conductivity and the elasticity modulus were increased and the coefficient of thermal expansion was decreased by increasing the ratio of beryllia contents.
Uncertainty Analysis of SB-LOCA in AP1000 Nuclear Power Plant
Authors: Shi Hao , Cai Qi , Chen Yuqing , Jiang Lizhi
Published: October 2017

Best estimate plus uncertainty methods (BEPUs) could be used to eliminate the over-conservatism and gain more safety margin in the analysis of thermal-hydraulic transient process at nuclear power plant. Based on the Best estimate thermal-hydraulic system code RELAP5/MOD3.2 platform, the best estimate plus uncertainty methods (BEPUs) proposed by GRS (Gesellschaft fur Anlagen- and Reaktorsicherheit) are presented together with applications to a small break loss of coolant accident (SB-LOCA) on the AP1000 Nuclear Power Plant best estimate analysis model. According to the results of uncertainty calculations, the dispersion bands of maximum cladding temperature and the core outlet void fraction are displayed and assessed.
Thermal Hydraulics Analysis of the Fluoride-Salt-Cooled, High-Temperature Reactor
Authors: Fu Yao , Sun Qiang , Zhou Chong , Zou Yang
Published: October 2017

Fluoride-salt-cooled, high-temperature reactor (FHR) technology combines the robust coated particle fuel of high-temperature, gas-cooled reactors with the single phase, high volumetric heat capacity coolant of molten salt reactors and the low-pressure pool-type reactor configuration of sodium fast reactors. This paper discusses one key technology area required to further define and develop the FHR: the thermal hydraulic performance of the core, primary system and second loop. Shanghai Institute of Applied Physics (SINAP) is leading the China Academy of Science (CAS) FHR program. A TMSR-SF1 reactor with a fluoride cooled pebble bed design has been suggested by SINAP, and the design is currently in progress. For this preliminary thermal hydraulic assessment, a TMSR-SF1 system model was developed using RELAP5. The RELAP5 model was used to help define and size systems such as the intermediate coolant salt selecting. A loss of flow transient was also simulated to evaluate the performance of the reactor during an anticipated transient event. A steady-state calculation was carried out and the calculated initial conditions show the influence of different salt. The loss of forced flow (LOFF) transient simulation results show that the passive residual heat removal system can effectively remove all decay heat from the primary loop under this extreme accident scenario. Some initial recommendations for modifying system component designs, such as heat exchanger with different salt and install place of pump, are also discussed.
An Innovative Spherical Fuel Element to Inhibit the Infiltration of Liquid Fluoride Salt in Molten Salt Reactor
Authors: Zhong Yajuan , Lin Jun , Xu Liujun , Jiang Haitao , Zhu Zhiyong
Published: October 2017

To inhibit the infiltration of liquid fluoride salt and easy to load and unload, fuel element in molten salt reactor (MSR) was isostatically pressed with an innovative design: A fuel-free low density graphite core of ≤ 30 mm diameter embedded in fuel-zone shell of ≥ 2.5 mm thickness, and then enveloped in a high density graphite shell of ≥ 5 mm thickness. Bulk density of the spherical fuel element can be designed from the range of 1.65–1.80 g/cm3, which is lower than the density of the liquid fluoride salt to make sure the fuel element can float in the MSR to load and unload. Characteristics of mercury infiltration and molten salt infiltration in graphite shell were investigated and compared with A3-3 graphite to identify the infiltration behaviors. The results indicated that the graphite shell has a low porosity about 9%, and an average pore diameter of 100 nm. The fluoride salt occupation of A3-3 was 10 wt% under 6.5 atm, whereas the salt gain did not infiltrate in graphite shell even up to 6.5 atm. It demonstrated that the outside graphite shell could inhibit the infiltration of liquid fluoride salt effectively. At the operating temperature of MSR (700 °C), thermal conductivity of graphite shell was 13.61 W/m K. The coefficient of thermal expansion (CTE) of outside graphite shell lied in 6.01×10−6 K−1 (α⫽) and 6.15×10−6 K−1 (α⊥) at the temperature range of 25–700 °C. The anisotropies factor of graphite shell calculated by CTE maintained below 1.12, which could meet the requirement of the spherical fuel element (below 1.30). The constant isotropic properties of graphite shell are beneficial for the integrity and safety of the spherical fuel element for a MSR.
Small Modular Pressurized Water Reactors Combined With Conventional Thermal Power Plant
Authors: Jiang Huijing , Bai Ning , Zhan Huanfen , Shen Feng , Gao Bin , Wang Xuee
Published: October 2017

Small modular reactor is investigated worldwide with the advantages of lower initial investment and short construction period. Generally, the economy of small modular pressurized water reactor (PWR) is not as good as large PWR, so various applications of small PWR are investigated, such as marine reactor, heat supply and sea water desalination. Limited to the parameters of steam generator, the generating efficiency for the pressurized water reactor nuclear power plant is about 33%, while the steam temperature of supercritical fossil power plant can exceed 600°C and generating efficiency is more than 45%. The essence of a hybrid power plant is to use a fossil fuel to superheat wet steam in an outer steam superheating device, after the steam generator to improve the parameters of working fluid. On one hand, the innovative hybrid nuclear power plant which combines nuclear reactor with conventional thermal energy can improve the efficiency of small PWR. On the other hand, this hybrid power plant has lower carbon emission compared with traditional thermal power plant. This paper describes two different coupling schemes of small pressurized water reactor combined with supercritical thermal power plant using steam turbine.

Efficiency of hybrid power plant is influenced by the coupling scheme, steam parameter of the superheating device outlet, the proportion of nuclear energy, efficiency of assemblies and so on. The plant efficiency becomes higher with the improvement of parameter of the superheating device’s outlet steam, and it is higher when the proportion of nuclear energy becomes lower. Take the 660MWt integrated small pressurized water reactor as an example, when the proportion of nuclear energy accounts for 48%, the thermal efficiency of this innovative hybrid power plant is about 43%,while the net efficiency is 41%, that is much higher than the efficiency of traditional pressurized water reactor, improving about 24 percent. As to the carbon emission, it depends on the coal consumption rate of power supply. The coal consumption rate of this hybrid power plant is 158g /kWh, while the consumption rate of thermal hybrid power plant is 280g/kWh, reducing about 44 percent. Also, the fundamental solutions of technical problems for this innovative hybrid power plant are discussed in the paper. Furthermore, several useful outcomes and suggestions for key equipments are put forward, such as the scheme of a superheating device and high-temperature steam turbine, and the possibility of using a lava boiler as the superheating device to improve the steam parameters after the steam generator is analyzed.
Research on the Effect of a New Flexible Shielding Material on the Seismic Performance of Nuclear Piping
Authors: Liu Feng , Li Yulong , Liu XiaJie , Bai Bing , Lv Yong-hong , Zhan Jie
Published: October 2017

The present paper provide an available FEA method for the structural analysis and seismic design of pipelines coated with flexible shielding materials, in nuclear power plant, subjected to earthquake action. The modal and spectral analysis are carried out by using FEM software ANSYS. By comparing the numerical simulation results of dynamical behavior of the pipelines with or without flexible shielding materials, the effects of the flexible shielding material on the seismic performance of the pipeline system in nuclear power plant are discussed. Finally, a local supporting constraint for reducing seismic effects is put forward, which can significantly decrease the maximum stresses in the pipelines coated with flexible shielding materials.
Simulated Temperature Profiles for Corrosion Products Deposited in ITER Experimental Loop
Authors: Arslan Muhammad , Yan Ma , Ali Shahzad Muhammad , Jinyun Lei , Ahmed Shahroze
Published: October 2017

An experimental setup was designed to find out behavior of deposition thickness of corrosion products. This experimental setup is designed on the bases of ITER (International Thermonuclear Experimental Reactor) loop. The experiment was done for different values of pH to study its effect on the deposition thickness. System pressure was kept at 0.5 MPa with maximum limit of temperature at 150°C. The pH value was kept in the range of 7.00–10.50. Artificial corrosion particles (Fe3O4) were introduced manually in the loop, which was allowed to stabilize. Four test tubes were used to determine the thickness of deposition, and the temperature at these points were measured by thermocouples. It is also very important to determine the temperature distribution around such points, in order to establish its relationship with pH and deposition thickness. Results show that the change of temperature and pH (of coolant) have a strong effect on deposition thickness. The temperature distribution, found in the various components of the experimental loop, will be used to correlate its relation with the deposition thickness and the coolant pH in future studies.
Numerical Simulations of Two-Phase Flow in a Fuel Assembly of Blockage due to LOCA
Authors: Ren Jingwen , Niu Fenglei , Zhuo Weiqian , Wang Da
Published: October 2017

During a LOCA accident, the debris caused by the action of high energy fluid discharged from the break may transport to the containment sump, then may be entrained into the core by the ECCS water. The debris may cause the blockage of fuel assembly. The air may also enter the reactor with water. Numerical simulations are performed to analyze the air-water flow and particle-water flow through the blocked fuel assembly. The pressure drop in fuel assembly will impact the long-term core cooling capability. The effects of different parameters on the pressure drop over the fuel assembly are analyzed. Pressure drop increases as blockage percentage, mass flow rate or inlet velocity increases for both two-phase flow. The decrease of air volume fraction and the increase of particle volume fraction all cause the increase of pressure drop. Pressure drop increases slightly as bubble diameter increases, and the tiny effect of particle diameter on pressure drop was found as particle diameter varying at an increment of 10um.
Simulation on Pellet-Cladding Mechanical Interaction (PCMI) of Accident Tolerant Fuel (ATF) With Coated Cladding
Authors: Deng Yangbin , Qiu Bowen , Wu Yingwei , Zhang Dalin , Tian Wenxi , Qiu Suizheng , Su G. H.
Published: October 2017

In this study, based on the code FROBA (Fuel ROd Behavior Analysis), a thermal-mechanical analysis code initially developed for traditional UO2-Zr fuel elements by our research group, a modified version named FROBA-ATF was developed to perform the fuel performance simulation of ATFs with different claddings, including Zr-4, SiC and Zr-4 coated with SiC. Compared with initial version, the cladding could be divided into arbitrary number control volumes with different materials in the new code, so it can be used to perform the calculation for multilayer coatings. In addition, a new non-rigid PCMI calculation model was established in the new code. Neither of the cladding and the pellet was regarded as the rigid body in this study, which means it can provide more accurate prediction compared with the rigid-fuel model in the initial code when Pellet-cladding Mechanical Interaction (PCMI) happened. The FROBA-ATF code was used to predict PCMI performance of two kind fuels with coated claddings, including the internal-surface coating and external-surface coating. The calculation result indicates that because the coating surface was close to the inner surface of the clad where also was the PCMI surface, the absolute value of the combine pressure of internal-surface coated cladding was substantial larger than that of the external-surface coated cladding, which might be harmful the coating behavior. However, the internal-surface coated mode can provide a protection for alloy due to the isolation from direct contact with fuel pellets, which can result in a much lower equivalent stress of zirconium body during the PCMI.
A CFD-Based Optimization Design of Flow Distribution Device in the Lower Plenum of Intermediate Heat Exchanger of SFR
Authors: Zhang Xiaolong , Tseng Peichi , Yu Jiyang , Saeed Muhammad
Published: October 2017

A computational fluid dynamics based simulation is performed to optimize the design of the flow distribution device in the lower plenum of the intermediate heat exchanger (IHX) of a pool-type sodium-cooled fast reactor (SFR) in this work. As a typical shell and tube heat exchanger, hot primary sodium flows in the IHX from the top and flows over the tube bundles, called shell-side. The secondary sodium (tube-side) runs through heat transfer tubes and its inlet plenum is specified at the bottom. The flow distribution device is arranged in the lower plenum of IHX, to change the flow distribution of the secondary sodium before into the heat transfer tubes. The CFD tool used in the work is ANSYS Fluent code. Two separated flow distribution devices have been simulated and compared. First, the orifice plates, three flow distribution orifice plates with different positions in the cylinder of lower plenum are respectively set as the model 1, 2 and 3. Secondly, the conical disk model, which is arranged at the bottom of the lower plenum, is established as model 4. And changing the size of the conical disk, the model 5 is established to predict the influence of the size of the conical disk on flow distribution. The results show that all of these models have similar velocity distributions at the outlet of lower plenum, which can be divided into three separate regions, where the flow velocity is higher at the inner and outer, and the velocity in the middle is lowest. When the orifice plate is set at the higher position, the overall velocity distribution is more uniform at the outlet. And the larger conical disk could make a more uniform velocity distribution as well.