bluuebunny/crossref_metadata_2025 huggingface dataset:
Development of the Pump-Integrated Intermediate Heat Exchanger in Advanced Loop-Type Sodium-Cooled Fast Reactor for Demonstration
Authors: Amano Katsunori , Enuma Yasuhiro , Futagami Satoshi , Inoue Tomoyuki , Watanabe Sota
Published: October 2016

In the framework of generation IV international forum (GIF), safety design criteria (SDC) and safety design guideline (SDG) for the generation IV sodium-cooled fast reactors (SFRs) have been developed as part of the worldwide deployment of SFRs. Japan Atomic Energy Agency (JAEA) and the Mitsubishi FBR Systems, Inc. (MFBR) have been investigating design study of an advanced loop-type SFR to satisfy SDC in the feasibility study of SDG for SFR. In this study, the ability of the pump-integrated Intermediate Heat Exchanger (IHX) is evaluated as a safety measure for the advanced loop-type SFR. Furthermore, maintainability and reparability of the safety measures are taken into account for the advanced loop-type SFR design study. The pump-integrated IHX has been modified to satisfy these requirements. This paper describes the modifications to withstand severe earthquake, primary coolant leak and sodium-water reaction. Also, this paper includes evaluations of thermal transient, structural vibration with pump rotation and wear-out of IHX tubes for they have been adversely affected by the modifications.
The Effect on the Support Seismic Loads of Primary Equipment With Different Reactor Coolant Pump
Authors: Xianhui Ye , Fengchun Cai , Bin Lan , Yixiong Zhang
Published: October 2016

Seismic analysis of the primary system in nuclear plant design is very important. The change of reactor coolant pump has some impact on the support seismic loads of primary equipment. Using the numerical calculation method, preliminary analysis was completed and some valuable results are obtained. The results show the natural frequency of system should be far away from the spectral peak with optimizing the reactor coolant pump configuration to raise the safety and economy of the plant.
Newspaper Coverage of Nuclear Power in China, From a Public Acceptance Perspective
Authors: Fang Chao , Yang Yanran
Published: October 2016

Environmental problems, especially air pollution and the global warming problems, have promoted the development of nuclear power in China. However, due to the severe consequences and the negative impression of Fukushima accident, the public acceptance remains a big obstacle for the further development of nuclear power. Besides, the protests against nuclear fuel factory and nuclear power plants in China also indicate the low public acceptance towards nuclear power. This issue has attracted a lot of academic attentions and researches to identify different factors that influence nuclear public acceptance. However, few researches have questioned the primary sources that reflect these different factors. This paper will first briefly analyze the progress of public acceptance study and introduces a recent framework of public acceptance in China, which consists of trust, knowledge, benefits (economical and environmental benefits) and risks. Newspaper reports on nuclear power (from 2013 to 20th Nov of 2015) from three Chinese newspaper offices are selected to do content analysis. They are used to analyze the construction of such factors in the newspaper articles. This paper concludes with the finding that the coverage of economical benefits and environmental benefits are rather low and should be attached more importance. Newspaper reports should also include more basic knowledge of nuclear power to let the public better understand nuclear power. In addition, the increasing emphasis on the national benefits makes it a necessary subject that should be studied further in public acceptance in China.
Effects of Cladding Material on Irradiation Performance of Monolithic Mini-Plates
Authors: Ozaltun Hakan
Published: October 2016

Monolithic, plate-type fuels are the proposed fuel form for the conversion of the research and test reactors to achieve higher uranium densities within the reactor core. This fuel type is comprised of a low enrichment, a high density U-10Mo alloy fuel-foil, which is sandwiched between diffusion barriers and encapsulated in a cladding material. To understand the irradiation performance, fuel-plates are being benchmarked for large number of parameters. In this work, effects of the cladding material were studied. In particular, a monolithic fuel-plate with U7Mo foil and Zry-4 cladding was simulated to explore feasibility of using Zircaloy as a surrogate cladding material. For this, a selected mini-plate from RERTR-7 tests was simulated first with as-run irradiation history. By using same irradiation parameters, a second case, a plate with U10Mo fuel and Al6061 cladding was simulated to make a comparative assessment. The results indicated that the plate with Zircaloy cladding would operate roughly 50 °C hotter compared with the plate with Aluminum cladding. Larger displacement profiles along the thickness for the plate with Zircaloy cladding were observed. Higher plastic strains occur for the plate with Aluminum cladding. The results have revealed that any pre-irradiation stresses would be relieved relatively fast in reactor and the fuel-foil would be essentially stress-free during irradiation. The fuel stresses however, develop at reactor shutdown. The plate with Zircaloy cladding would have higher residual stresses due to higher pre-shutdown temperatures. Similarly, the stresses magnitudes are higher in the foil core for the plates with Zircaloy cladding. Finally, pressure on the fuel is significantly higher for the plates with Zircaloy cladding. Overall, employing a Zircaloy as surrogate cladding material did not provide a better thermo-mechanical performance compared with the Aluminum cladding.
Large-Scale Homogeneous Hydrogen-Air-Steam Deflagration Experiment Simulated Using Two Turbulent Flame Speed Closure Models
Authors: Holler Tadej , Jain Varun , Komen Ed M. J. , Kljenak Ivo
Published: October 2016

The CFD combustion modeling approach based on two combustion models was applied to a hydrogen deflagration experiment conducted in a large-scale confined experimental vessel. The used combustion models were Zimont’s Turbulent Flames Speed Closure (TFC) model and Lipatnikov’s Flame Speed Closure (FSC) model. The conducted simulations are aimed to aid identifying and evaluating the potential hydrogen risks in Nuclear Power Plant (NPP) containment. The simulation results show good agreement with experiment for axial flame propagation using the Lipatnikov combustion model. However substantial overprediction in radial flame propagation is observed using both combustion models, which consequently results also in overprediction of the pressure increase rate and overall combustion energy output. As assumed for a large-scale experiment without any turbulence inducing structures, the combustion took place in low-turbulence regimes, where the Lipatnikov combustion model, due to its inclusion of quasi-laminar source term, has advantage over the Zimont model.
Numerical Simulation of Natural Convection Heat Transfer and Transport in Packed Beds: Mimicking a Proposed New Nuclear Fuel Design
Authors: Noah Olugbenga O. , Slabber Johan F. , Meyer Josua P.
Published: October 2016

The ability of coated particles of enriched uranium dioxide (UO2) fuel to withstand high temperature and contain the fission products in the case of a loss of cooling event is a vital passive safety measures over traditional nuclear fuel active safety system to provide cooling. Hence, it is proposed in this study that Light Water Reactors (LWR) could be made safer by re-designing the fuel in the fuel assembly. A slender geometrical model with tube-to-particle diameter ratio N = 2.503 and porosity ε = 0.546 mimicking the proposed nuclear fuel in the cladding was numerically simulated. This is to investigate the heat transfer characteristics and flow distribution under buoyancy driven force expected in the cladding tube of the proposed nuclear fuel using a commercial code. Random packing of the particles is achieved by Discrete Element Method (DEM) simulation with the aid of Star CCM+. The temperature contour and velocity vector profile obtained can be said to be good illustration of anticipated heat transfer phenomenon to occur in the proposed fuel design. Similarly, heat transfer parameters such as particle-to-fluid heat transfer coefficient, Nusselt number, Grashof number and Rayleigh number were determined from simulated results and are presented. These parameters are of prime importance when analysing the heat transfer performance of a fixed bed reactor.
Effect on Hydrogen Risk of Initial Gas Injection Time During AP1000 Post-Inerting
Authors: Lyu XueFeng , Chen YanLin , Li XiaoBo , Wang ShengFei , Yu Yu , Liu Bin
Published: October 2016

To calculate the hydrogen risk at severe accident of small break of cold leg in 1# steam generation compartment with ADS4 invalid in AP1000 nuclear power plant, and apply the results to level2 probabilistic safety analysis, we study the effect of initial gas injection time on reducing hydrogen risk during AP1000 post-inerting, the initial gas injection times are 300 second, 500second, 700second, respectively. First, analyzing the total elimination of hydrogen by recombiners. Then, analyzing the average hydrogen mole fraction, flame acceleration factor in 1# steam generation compartment and in upper space of containment. Finally, analyzing the pressure and average temperature in containment. The results show that, the premature injection of inert gas can slow down hydrogen diffusion rate from 1# steam generation compartment to the upper space of containment, which causes hydrogen risk rising in 1# steam generation compartment. Post-inerting can ease hydrogen risk in upper space of containment, but can’t ease hydrogen risk in 1# steam generation compartment effectively. The pressure of containment is only relevant to the total mass of inert gas, and the pressure of containment is always less than limit pressure, so the containment breakage is nearly impossible.
The Preliminary Performance Study of ATF With M3 Fuel and Zr-4 Cladding
Authors: Chunyu Yin , Teng Tu , Tang Changbing , Jiao Yongjun
Published: October 2016

Metal matrix microencapsulated (M3) fuel is one of the research directions on Accident Tolerant Fuel (ATF). In this article, it provides one of ATF design which consists of BISO (Bistructural ISOtropic) particles embedded in a zirconium alloy matrix, and the cladding coating with silicon carbon (SiC). The temperature distribution of the ATF element has been built, and then the center temperature has also been calculated based on the operation parameters of the large-scale pressurized-water reactor. Simultaneity, the several factors of fuel failure is preliminary analyzed and calculated, especially the pressure shell failure mechanism.
Proposal for Evaluation With Strain-Based Criterion of Impact Phenomena Between Steel Structure and Tornado Missile
Authors: Sakamoto Yuko , Shirai Koji , Udagawa Toshiko , Kondo Shunsuke
Published: October 2016

In Japan, nuclear power plants must be protected from tornado missiles that are prescribed by Nuclear Regular Authority (NRA). When evaluating the structural integrity of steel structures in the plant with impact analysis by numerical code, strain-based criteria are appropriate because the tornado missiles have huge impact energy and may cause large deformation of the structures. As one of the strain-based criteria, the Japan Society of Mechanical Engineers (JSME) prescribes limiting triaxial strain for severe accident of Pressurized Water Reactor (PWR) steel containment. To confirm whether or not this criterion is appropriate to the evaluation of the impact phenomena between the steel structures and the tornado missiles, a free drop impact experiment to steel plates (carbon steel and austenitic stainless steel) was carried out with heavy weights imitated on one of the tornado missiles, followed by an impact analysis of the experiment with AUTODYN code and the JSME strain-based criterion. Consequently, it was confirmed that the strain-based criterion of JSME standard was for evaluating the fracture of steel structures caused by tornado missiles.
Research on Thermal Hydraulic Characteristics of Passive Containment Cooling System for Ship Nuclear Power Platform
Authors: Dai Chunhui , Wu Jun , Tan Sichao , Zhao Zhenxing , Xiao Qi , Liu Zhouyang , Liu Yong
Published: October 2016

Ship nuclear power platform is a small and movable power plant on the sea, aiming at generating electric energy and producing fresh water, it provides support for the national energy strategy. Subsequent to a loss of coolant accident (LOCA), steam is vented in the reactor containment following vaporization of liquid and/or steam expansion. The temperature as well as pressure in the condensation rises synchronously. For removing heat and reducing pressure inside containment subsequent to a LOCA, the Passive containment cooling system of Ship nuclear power platform is designed. In order to establish and maintain the passive heat removing channel, steam condenses on the containment condenser tube surface, coupling natural convection of the seawater inside the tubes. The heat transfer mechanism of Passive containment cooling system is very complex. To solve this problem, a three dimensional heat exchanging/one dimensional natural circulation coupling numerical computing method is proposed to obtained the safety performance of the reactor containment. Models of heat exchanging process between steam which contains non-condensable gas inside the reactor containment and sea water outside are firstly established. Then the thermal-hydraulic characteristics of the steam and sea water beside the heat transfer tubes are obtained by a simulation which is carried out in a LOCA.
Pilot Study of Diagnosis on CPR1000 With Dynamic Uncertain Causality Graph
Authors: Zhao Yue , Dong Chunling , Zhang Qin
Published: October 2016

To develop a computerized intelligent system for CPR1000 with functions of fault diagnosis (FD) for safe operation of nuclear power plant (NPP) counts a great deal. Traditional procedures, earlier systems and classic uncertain artificial intelligence methodologies cannot apply well to some severe conditions like Three Mile Island (TMI) accident. To remedy these defectives, in this paper, the methodology of Dynamic Uncertain Causality Graph (DUCG) is brought in for diagnosis of NPP. Causal graphs of DUCG is used to symbolize logical relationships and inserted weighting factor can balance the representation of multi-valued causal relationships and avoid the cases of exponential scale-up of conditional probability table (CPT). With graph-based inference scheme, precise and compact results of FD can be computed and laid out in graphic forms. A DUCG model centering seven faults of secondary circuit in CPR1000 is constructed based on systematic analysis of Unit-I of Ningde NPP of China General Nuclear Power Group (CGNPG). Verification experiment is carried out for fault steam generator (SG) loss feed-water with sensor measurements from corresponding simulation machine NS-FSS manufactured by China Nuclear Power Simulation Technology Co, Ltd (CNPSC). Superiorities of DUCG in FD emerge by contrast with other techniques, which lay theoretical foundation for the promotion of development of CPR1000 intelligent system.
Performance Analysis of Closed Brayton Cycle System for Small Modular Reactor
Authors: Li Zhi , Yang Xiaoyong , Wang Jie
Published: October 2016

Small modular reactors have potential applications in remote or underdeveloped areas for their compact configurations, low maintenance requirements, flexible deployments, and long lifetimes, etc. A small high temperature gas cooled reactor coupled with closed Brayton cycle is presented in this paper. The modular reactor has inherent safety and closed Brayton cycle has high energy conversion efficiency and compact configuration, they can even be integrated into a mobile power plant. A thermodynamic model is established to analyze the effect of parameters, like pressure ratio and ambient temperature, on optimization of design and the system’s performance. The results show that in design condition, temperature ratio, which is effected by ambient temperature and cooling method, is a key parameter. High temperature ratio leads to high cycle efficiency. Compression ratio and effectiveness of recuperator also influence cycle’s efficiency and reactor inlet temperature. As compression ratio increasing, efficiency would decrease after reaching a maximum. In operating condition, changes of ambient temperature would not only affect the temperature ratio, but also cause a deviation of compressor and turbine from their design points, which would lead to a further effect on compression ratio and efficiency. A comparison is made between the operating condition and the design condition to show the effect of temperature ratio variation on compressor, turbine and cycle’s efficiency. It will be helpful to evaluate the performance of small modular reactor in different areas and seasons. The results will be helpful to understand and design the small modular reactor.
Testing the Impact on Emergency Diesel Performance of Water in Diesel Fuel
Authors: Cryer Robert , O’Connell Mark
Published: October 2016

This paper focuses on testing results of the impact of water contamination in diesel fuel on the ability of an emergency diesel generator (EDG) to successfully start and operate during an emergency. This testing program resulted from the discovery of degraded vent pipes on diesel fuel feed tanks that could have allowed rain water to enter and collect at the bottom of the diesel fuel system and potentially prevent satisfactory engine start-up and operation. The nuclear regulator notified the nuclear plant of a potential yellow finding.

The initial analysis effort focused on the use of diesel engine combustion software (Ricardo’s WAVE© 1D engine performance simulation software). Two medium-speed diesel engine models were analyzed with added water content ranging from 10% to 40% water in the diesel fuel. The analyses demonstrated that the engines could start and operate with those percentages of water in the fuel, but that the engine output would experience a power loss or derate proportional to the water content. The regulator was not convinced that the analysis was sufficient.

The validity of the analytical findings above was demonstrated by full-scale tests conducted by MPR Associates on a large diesel engine. To accomplish this, Entergy, the nuclear power plant owner constructed a simulation of the diesel fuel supply system at a facility having the same make and model EDG. Water was introduced into the diesel fuel day tank by two different approaches; slow trickle flow and large slugs of water. These conditions simulated either a steady rainfall while the EDG was operating or a large volume of water collected in the system while the EDG was in standby.

Under both water contamination scenarios, which consisted of more than 50 hours of testing, the diesel engine and generator set responded with negligible loss of frequency or voltage. Further, the engine was confirmed to have undergone no increased wear or degradation as a result of the high levels of water in the combustion chambers. Following a detailed review of the test program and the successful results, the regulator concluded that neither a non-cited violation nor a penalty was warranted.
Moisture Separation for Nuclear Turbine Plant Efficiency and Reliability Improvement
Authors: Takahashi Ryota , Fujita Issaku , Nakamura Taichi , Kasahara Jiro , Yoshii Taichi , Inoue Naoki , Nishida Keigo
Published: October 2016

Feasibility of Rankine cycle improvement for LWR power plant is studied. In LWR, the main steam is saturated steam, thus steam turbine is mostly operated in wet region. Therefore, the moisture separation plays a key role to achieve better turbine plant efficiency. Typical reheating configuration for present LWR plant is single reheating after HPT by MSR.

On the other hand, it is well known that the increasing of numbers of moisture separation or reheat improves the turbine plant efficiency. Double reheat cycle is common for fossil power plant and also has been proposed for LWR. However, double reheat cycle has not been adopted to actual nuclear power plant yet since the double reheat configuration makes heating and drain system complicated, and larger MSR shell and diameter piping are required due to the increase of steam volume flow.

In this paper, improved Rankine cycle with high pressure moisture separator (HP-MS) and improved low pressure loss MSR (LP-MSR) is studied in order to increase the plant efficiency. Using CFD analysis and scale model test, we confirmed pressure loss and moisture separation characteristic of LP-MSR. The performance of improved Rankine cycle with this new MSR was calculated and the results show that the electrical output of typical LWR Plant may increase 1% or more from single reheat cycle.
Qualification of Water-Filled Tank due to Tornado Missile Impact
Authors: Schmitt Timothy A. , Al-Shawaf Taha , Rahman Syed M. , Cristobal Angelo
Published: October 2016

As part of the post-Fukushima accident scenario, the qualification of a water-filled cylindrical tank subjected to tornado missile impact was required to ensure the availability of water inventory in the tank to mitigate the post-accident effects.

Most of the classical tornado missile impact analysis and design involves using empirical formulas that have been developed based on tests. It is recognized that water backed structures provide additional resistance to perforation of the missile due to the mass and properties of the water. Therefore, a finite element analysis was used to qualify the tank for two controlling postulated missiles, namely, 2 ½” diameter schedule 40 pipe and bolted wood decking.

The location on the tank for the missile impact, angle of impact and orientation of the missile were selected to develop the most critical response. The analysis was performed using the LS-DYNA computer program. The true stress-strain material properties were used for both the tank material and the missile types. These material properties were given a bilinear elastic-plastic curve.

It was determined that, even if an impact at the thinnest section at the top of the tank occurs and the missile penetrates, the remaining inventory of water in the tank will be sufficient to mitigate the needs for a post-Fukushima scenario. Impact on the lower elevation of the tank was investigated for any potential failure or tearing of the tank wall. The maximum of equivalent (Von Mises) stress, shear stress, and plastic strain were calculated. The results show that these values are less than the limiting values with additional available margin. Consequently the analysis shows that the tank will survive a hit in the lower portions, and the water inventory of the tank is sufficient to mitigate the effect of a post-Fukushima scenario should a missile penetrate the thinner, upper section of the tank.
AP1000® Submergence Testing of Service Level I Protective Coatings
Authors: Beenken Matthew R. , Rocco Christina L. , Andreychek Timothy S.
Published: October 2016

A portion of the Service Level (SL) I coating systems inside the AP1000®1 containment become submerged during a design basis loss-of-coolant accident (LOCA) as a result of containment flood-up. The Design Basis Accident (DBA) qualification of SL I coatings for the AP1000 did not initially include post-DBA submergence conditions.

Submergence testing was performed using a standard test (Reference 2). Test articles were carbon steel coupons coated with one of two coating SL I coating systems; an untopcoated inorganic zinc system (Carboline Carbozinc® 11 HSN2) and a system consisting of inorganic zinc and an epoxy (Carboline Carboguard® 890N3) topcoat. Half of the inorganic zinc samples tested were irradiated to a total accumulated dose of approximately 1×109 rads prior to submergence testing.

Autoclaves were used to simulate the post-LOCA environment inside of the AP1000 containment. The working fluid was a boric acid solution buffered with trisodium phosphate (TSP) to a pH of approximately 7.8. The pressure and fluid temperature inside of the autoclaves was regulated to begin at ambient conditions followed by a pressurization and heat up following pressure and temperature conditions calculated for a cold leg break LOCA for the AP1000 plant, plus an added 10% margin, and the slowly decrease over the 30 day test period to saturation temperature at about 20 psia, simulating a cool-down of the AP1000 plant. Coolant samples of the autoclave inventory were taken and evaluated for dissolved chemical species at specific time intervals during the test.

Following exposure to the submerged test environment, each coupon was visually inspected immediately following removal from the autoclaves and was inspected again several days after completion of the test cycle for signs of blistering, rusting, lifting, peeling, discoloration, softening, etching, wrinkling, cracking, swelling, dissolving, delamination, and changes in gloss relative to their observed pre-test condition.
Numerical Investigation on Heat Transfer to Supercritical CO2 in Vertical Annular Channel
Authors: Zhao Zhenxing , Wu Jun , Bai Fan , Xiao Qi , Dai Chunhui , Liu Zhouyang , Liu Yong , Tao Mo
Published: October 2016

The special fluid flow and heat transfer characteristics of supercritical CO2 in a vertical annular channel have been numerically investigated. The AKN k-ε model was selected to model the turbulent flow and heat transfer of supercritical fluid. The three heating types were individual outer-wall heating, simultaneous outer/inner walls heating and outer-wall heating (inner-wall cooling) separately. The local heat transfer coefficients were obtained to investigate the influence of inner-wall thermal boundary conditions, supercritical fluid mass flux, fluid temperature and flow direction on outer-wall heat transfer phenomenon. The mechanisms of abnormal heat transfer and primary influence factors were analyzed by the detailed information on the flow, turbulence and thermal fields. When the supercritical fluid is in the large-property-variation (LPV) region and flows upward, the inner-wall thermal boundary condition obviously affects the heat transfer characteristics of outer wall. When supercritical fluid flows downward, the inner-wall boundary condition hardly affects the heat transfer phenomena of outer wall. The increase of inner-wall heating heat flux will result in the larger deterioration region and heat transfer decline on outer wall when the other conditions remain unchanged. When the heat transfer deterioration also appears on the inner wall with the increase in the inner-wall heat flux, the outer-wall heat transfer no longer decreases, but the deterioration region abruptly increases. However, as inner-wall cooling heat flux increases, the heat transfer deterioration phenomenon on outer wall will weaken gradually.
Modelling for Resuspension of Multilayer Particles
Authors: Zhang Tianqi , Peng Wei , Yu Suyuan
Published: October 2016

The resuspension of graphite dust in High Temperature Gas-cooled Reactors is not only an essential way to estimate the concentration of particles in the normal operating core, but also a key phenomenon related to the release of radioactive products in depressurization accidents and the safety evaluation of environment. Precise resuspension modelling is an important work in safety analysis of HTRs. Current models were based on a single particle sitting on the substrate, however, multilayers of graphite dust deposit in the core during decades of operation. This paper uses recursion to deduce Rock’n’Roll model on each layer, and finally get the resuspension fraction changing with time and flow velocity. The results showed a “lag effect” of resuspension compared with those of Biasi’s modified model. Compared with the results of multilayer resuspension, the original monolayer model overestimated the resuspension rate and fraction.
Investigations About Released Debris and Corrosion Products and Their Impact to Head Loss Build Up and Decay Heat Removal Under Water-Chemical PWR-LOCA Conditions
Authors: Renger Stefan , Alt Sören , Kästner Wolfgang , Seeliger André
Published: October 2016

Investigations about the release, transportation and deposit of fibrous insulation material (FIM), corrosion products as well as resulting compounds and debris mixtures become more important to reactor safety research, when considering long-term behavior of emergency core cooling systems (ECCS) during loss of coolant accidents (LOCA). Debris released by a leakage jet leads to head loss buildups at the sump strainers, the debris filters and the spacers of fuel assemblies. However, these complex processes may influence the decay heat transfer out of the reactor core.

A similar but newer scenario implies that the boric acid coolant in pressurized water reactors (PWR) can support corrosion processes at hot-dip galvanized installations, leading to a significant release of ionic zinc into the coolant during the sump recirculation phase. A long-term change of chemical properties of the coolant (e.g. pH, zinc ion concentration) has to be considered in safety analyses. Chemical analyses showed that the solubility of zinc in boric acid coolant is inversely proportional to the coolant temperature. Consequently, zinc ions can be dissolved at lower temperatures in the containment sump. Precipitations of zinc borate (ZBP) are possible at hot spots in the reactor core. The ZBP can be formed as a flocculent disperse phase in the coolant or as solidified layers at hot fuel rod surfaces. Layer spalling could lead to the release of further solid particles into coolant flow inside the reactor core.

In several joint research projects between the Zittau/Görlitz University of Applied Sciences (HSZG) and the Helmholtz-Zentrum Dresden-Rossendorf (HZDR) investigations on the impact of these processes to the head loss buildup and the heat transfer from core were done at laboratory and semi-technical scaled test facilities. Generic experiments showed the formation of ZBP in heating rod configurations. The ZBP may remain in the core structures or can be transported on debris filter cakes in upstream and downstream components of ECCS and influence the head loss. After this, research addressed the systematical clarification of physico-chemical mechanisms and their influence on thermal-hydraulic-dynamic processes occurring as a consequence of flow induced corrosion effects during LOCA.

This paper includes a description of the most important involved test facilities, applied measuring techniques, an overview of boundary conditions considered experimentally and selected results.
An Experimental Study of Small Sized Reactors Containment Heat Transfer Phenomena Under MSLB Accident
Authors: Wang YeYun , Wang ShengFei
Published: October 2016

Main Steam Line Break (MSLB) accident will lead to the release of high-energy steam inside the containment, then cause the pressure and temperature of containment rise, which threatening containment integrity and the normal operation of the containment equipment. Containment equipment layout will affect the flow of steam, while the break position, steam temperature and other factors also affect the temperature distribution inside the containment. Test bench was reconstructed on the original foundation, but analogs of Pressure Vessel and CMT have not been arranged in the simulated reactor containment at this stage. Experiments with steam as the working fluid were carried out at ambient pressure environment. Steam injected into the containment and after reaching steady state data acquisition card got temperature values of each measuring point. The main purpose of this work is to study the temperature distribution inside containment and heat transfer phenomenon of steel shell in MSLB accident. Experimental results show that the temperature stratification appears in the containment, although the temperature distribution in the shell is different under different operating conditions.
Utilization of Thorium in LWR Fuels Aiming at Thermal Conductivity Improvements
Authors: Kubáň Jan , Škoda Radek
Published: October 2016

One of the main drawbacks of uranium dioxide, which is used in almost all nuclear power reactors, is its low thermal conductivity. As a consequence, temperature at the center of fuel pellet is relatively high, because heat is poorly conducted away. To reach a higher level of safety, maximal temperature in any fuel pellet is one of the main limiting parameters, which restrict the fuel thermal output. This paper deals with the use of thorium in LWR fuels with the objective of fuel pellet maximal temperature reduction.

Research investigating homogenous distribution of thorium dioxide (thoria) in uranium dioxide fuel has already been done and did not lead to considerable thermal conductivity improvements. The aim of this study is to investigate heterogeneous distribution of thorium in commonly used uranium dioxide fuel in the form of uranium and thorium pellets placed together.
Hafnium in Nuclear Fuel Cladding Used Both as Oxidation Protection and Burnable Absorber
Authors: Závorka Jiři , Škoda Radek
Published: October 2016

The problem with higher nuclear fuel enrichment is its high initial reactivity. It has a negative effect on the peaking factor, which is one of the license conditions. The second major problem is the ability to control the reactivity of the reactor, and thereby maintaining the multiplication factor in the core equal to 1. Long-term control of the reactivity in PWR reactors is typically conducted by the concentrated boric acid (H3BO3) in the coolant; its highest possible concentration is determined by the requirement to maintain a negative reactivity coefficient. Another option are burnable absorbers.

This work deals with usage of hafnium as an advanced type of burnable absorber. Based on the model of computing code UWB1 for the study of burnable absorbers, a new cladding of nuclear fuel is designed with a thin protective layer made of hafnium. This cladding is used as a burnable absorber that helps reducing excess of fuel reactivity and prolongs the life of the fuel assemblies, which increases economic coefficient of the use of nuclear power plants. This cladding would also work as a protective layer increasing endurance and safety of nuclear power plants. Today zirconium alloys are exclusively used for this purpose. The main disadvantages of zirconium alloys include rapid high temperature oxidation of zirconium — a highly exothermic reaction between zirconium and water steam at temperatures above 800 °C. During this reaction hydrogen and inconsiderable amount of heat are released. Hydrogen excess, released heat, and damaged cover of fuel may deepen the severity and consequences of possible accidents. Another disadvantage of zirconium alloys is their gradual oxidation under standard operating conditions and ZrH formation, which leads to cladding embrittlement.
Evaluation of Jet Impact Region and Fluid Force Generated From Ruptured Pipes: 2 — Evaluation of Fluid Force Using Computational Fluid Dynamics Analysis
Authors: Takahashi Shiro , Xu Qiang , Takamura Noriyuki , Morita Ryo , Uchiyama Yuta , Watanabe Shun
Published: October 2016

Nuclear power plants are designed to avoid damage to their safety installations because of jet impingement when a pipe is ruptured. We have investigated evaluation methods for the design basis of protection of plants against effects of postulated pipe rupture using computational fluid dynamics (CFD) analysis. The steam jet tests obtained using particle image velocimetry (PIV) were conducted in order to verify the CFD analysis. Spread of steam jets could be visualized and the shapes of the steam jets obtained by analysis were almost the same as those by tests. The spread angle of free jet was investigated using CFD analysis. We also measured jet fluid force when a cylindrical structure was installed downstream from the jet nozzle. Steam jet fluid force obtained by analysis was almost the same as that by tests. We judged the CFD analysis to be applicable to evaluation of jet fluid force generated from ruptured pipes.
Fault Diagnosis of Feedwater Pump in Nuclear Power Plants Using Parameter-Optimized Support Vector Machine
Authors: Yu Rui , Li Xianling , Tao Mo , Ke Zhiwu
Published: October 2016

The condition monitoring of the feedwater pump in secondary circuit is critical to the safe operation of the nuclear power plant. This article presents a fault diagnosis method of feedwater pump by using parameter-optimized support vector machine (SVM). While the fault features of feedwater pump are reflected from the power spectrum of the vibration signals, we trained and diagnosed the fault feature table with support vector machine. The optimal penalty factor C and kernel parameter γ of support vector machine are selected by grid search and k-fold cross validation. Then the faults are diagnosed by the SVM model under the optimal parameters. Diagnostic results show that the parameter-optimized SVM method achieves higher diagnostic accuracy than the PNN method, exhibiting superior performance to effectively diagnose the faults of feedwater pump.
Analysis and Research on Natural Circulation Capacity of HFETR
Authors: Xu Taozhong , Deng Caiyu , Xiang Yuxin
Published: October 2016

Natural circulation is being used as an important circulation to remove reactor residual heat. In the core of High Flux Engineering Trial Reactor of China (HFETR), the coolant is driven by pumps normally and flows from upside to downside in the core. When HFETR is shut down or runs in low power, the natural circulation between the hot water in the core and the cold water in the reflector inside the pressure vessel is established to cool down the core. Since the natural circulation processed only in the pressure vessel, the accident pumps need to be turned on periodically to remove reactor residual heat. The inversion of flow direction in HFETR and internal natural circulation lead to a different natural circulation establishment process from traditional reactor in which coolant flows form down to top normally. In this paper the transition between the natural circulation and forced circulation is analyzed by RELAP5/MOD3 code. The results showed that the accident pump could be turned off in the power of 850kW; The time, at which the accident pump needs to be turned on to transit the natural circulation to forced circulation, is decided by the temperature of the water in top of pressure vessel, and a formula between temperature of the water in the top of pressure vessel and the reactor power was obtained. The research results have theoretical and practical value to the full use of the natural circulation ability, as well as the safety of the engineering reactors or similar test facilities.
Characteristic Confirmation Test by Using HTTR and Investigation of Absorbing Thermal Load Fluctuation
Authors: Honda Yuki , Tochio Daisuke , Sato Hiroyuki , Nakagawa Shigeaki , Ono Masato , Fujiwara Yusuke , Hamamoto Shimpei , Iigaki Kazuhiko , Takada Shoji
Published: October 2016

The characteristic confirmation test has been demonstrating by using High Temperature engineering Test Reactor (HTTR). The nuclear heat supply performance test, which is one of the characteristic confirmation test is planned to be carried out after restarting of HTTR. Towards the realization of the industrial utilization of a High-Temperature Gas-cooled Reactor (HTGR) cogeneration system as an extension of a nuclear plant, it is important to ensure the reactor safety in the case that thermal-load of the heat application system is fluctuated or lost. The preliminary analysis for the thermal load fluctuation test, which is one of the nuclear heat supply performance test has been investigated. In the analysis, the reactor outlet temperature can continue to be stable against the reactor inlet temperature changing by the thermal fluctuation. It means that HTGR have the capability of absorbing the thermal fluctuation. This paper focuses on the investigation of the mechanism of absorbing the thermal fluctuation. With the reactor inlet temperature increasing, the graphite moderator reactivity keeps negative though the fuel reactivity becomes active. The large negative graphite moderator reactivity enhances the capability of the absorbing thermal fluctuation. In addition, in the middle of the core, the graphite moderator reactivity insertion trend is inverted. This trend is unique to HTGR because of the large temperature difference between top and bottom of HTGR core.
Transient Physics Procedure and Preliminary Characteristics Analysis of Aqueous Homogeneous Reactor
Authors: Wang Liangzi , Wei Yanqin , Xiang Hongzhi , Li Xiangyang
Published: October 2016

Aqueous Homogeneous Reactor demonstrated in this analysis uses Uranyl Nitrate water solution as fuel; for the use of medical isotope producing, AHR has many advantages like small size, low power, passive safety, low output of nuclear waste, etc. Liquid fuel and radiolytic gas bubbles lead to big negative feedback features and fuel fluctuation; these may result in transient excursion, which is extremely important and could affect the operation stability and design of AHRs. For the analysis of transient characteristic of AHR, we developed this transient physics procedure of AHR, it uses improved quasi-static approximation and is based on Multi-group Monte Carlo method, which could consider spatial effects onto transient behavior and has powerful geometry processing ability; both physical model and preliminary validation results will be presented in this article. This procedure applies as an analyzing tool for AHR’s transient behavior; characteristics of AHR in typical transient process are also presented in the article.
Study on Thermal Characteristics of the Regenerative Heat Exchanger
Authors: Si Junping , Tong Mingyan , Yang Wenhua , Huang Gang
Published: October 2016

The regenerative heat exchanger is widely used in nuclear power plants and research reactors. It is composed of the regeneration section and the cooling section. The heat transfer mainly occurs at the cooling section, while the regeneration section is designed to reduce the temperature difference between the hot and cold fluids and weaken the damage to the heat exchanger due to the existence of thermal stress. Meanwhile, some heat is also can be recovered through the regeneration section. This paper mainly aims to analyze the thermal characteristics of the regenerative heat exchanger according to its structure properties, and provides some suggestions for regenerative heat exchanger design based on the influence of some key factors on thermal characteristics. The results show that improving the outlet temperature in the regeneration section primary side can both reduce the heat exchange areas of the regeneration section and the cooling section, but this will rise thermal shock and increase the operation safety risk. The baffles arrangement will enhance heat exchange capacity, and the heat exchange area decreases with the baffle gap height increasing. With the heat exchange area margin of the regeneration section improvement, the actual power will gradually reduce. The measures, including increasing secondary water flow or taking a corresponding margin about 52.8%∼59.2% that of the regeneration section for the cooling section heat exchange area, can be taken to overcome the adverse effects of the margin on the regenerative heat exchanger. More heat exchange areas of the regeneration section and the cooling section are required to satisfy the rated power with the fouling thermal resistance of the primary water increasing. Moreover, adopting a lower fouling coefficient favors the generative heat exchanger running under the design power.
Development and Verification of Thermal Hydraulic Subchannel Analysis Code for Motion Conditions
Authors: Cai Rong , Huang Siyang , Wang Kai , Tian Wenxi , Qiu Suizheng , Su Guanghui
Published: October 2016

As the conventional core analysis codes are designed for the land-based reactor core, a thermal-hydraulic subchannel analysis code for motion conditions (SACROM) is developed. To evaluate the effect of different motion conditions on coolant flow, the model of additional forces is established. To check the accuracy of the models, the code has been verified by test data, a commercial subchannel code and a CFD code. In the steady-state verification, the ISPRA data were used and the predicted results agree well with the test data. For the transient simulations without motion conditions, the code COBRA-EN was chosen and the results from SACROM fit the results from COBRA-EN well. And CFX code was used to verify the accuracy of the model of additional forces for motion conditions. The results show that the code can be used in the thermal hydraulic characteristics of the reactor core under motion conditions.
Eigenvalue Implicit Sensitivity Calculation and Analysis for Lattice-Physics Calculation
Authors: Liu Yong , Cao Liangzhi , Wu Hongchun , Zu Tiejun , He Qingming
Published: October 2016

Accurate nuclear cross-section sensitivity-coefficient evaluation is important for sensitivity and uncertainty analysis, similarity analysis, cross-section adjustment et al. A cross-section perturbation will affect the lattice-physics calculation results through the transport calculation directly and through the resonance calculation indirectly. The indirect effect was found to be important in some cases in the previous studies. To quantify the indirect effect on the lattice-physics calculation results for subgroup resonance calculation method, a sensitivity and uncertainty analysis code COLEUS was developed based the GPT-based method. The eigenvalue sensitivity to non-resonance nuclide cross sections was investigated. Numerical results show that in the traditional LWR, the sensitivity coefficients will be overestimated if implicit sensitivity is neglected. And in the BWR, the implicit sensitivity will become more important along with the temperature rise. But if resonance fission and resonance capture play a coequal role or the background cross section is big, the implicit sensitivity can be small.
A Fuzzy Control Approach for the Coordinated-Level Control of the Modular-HTGR-Based Nuclear Steam Supplying System
Authors: Yuan Yue , Huang Xiaojin
Published: October 2016

As one of the most popular Generation IV nuclear energy system, High Temperature Gas Cooled Reactor (HTGR) has outstanding inherent safety features. However, the nonlinearity and complexity of the modular-HTGR-based nuclear steam supplying system (NSSS) has put higher requirements to its control system. Based on the basic ideas and theories of fuzzy system, fuzzy control method provides a powerful tool to the control of nonlinear systems. This paper built the T-S fuzzy model of the modular-HTGR-based nuclear steam supplying system (NSSS), and designed the steam temperature T-S fuzzy controller using the parallel distributed compensation (PDC) method. Simulation shows that the designed T-S fuzzy controller has a better effect than the traditional PID control method.