bluuebunny/crossref_metadata_2025 huggingface dataset:
The iBR: A Generation III.7 Reactor After the Fukushima Daiichi Accident
Authors: Sato Takashi , Matsumoto Keiji , Kurosaki Toshikazu , Maida Mitsuhiro
Published: November 2014

The paper presents study result on an innovative, intelligent and inexpensive BWR (iBR). It is the first Generation III.7 reactor after the Fukushima Daiichi accident. It has incorporated lessons learned from the Fukushima Daiichi accident and WENRA safety objectives. It has innovative safety to cope with devastating natural disasters including a giant earthquake, a large tsunami and a monster hurricane. The iBR can survive passively such devastation and a very prolonged SBO without any support from the outside of a site up to 7 days even preventing core melt. It, however, is based on the well-established proven ABWR design. NSSS is exactly the same as that of the current ABWR. As for safety design it has a double cylinder RCCV (Mark W containment) and an in-depth hybrid safety system (IDHS). The Mark W containment has double FP confinement barriers and the in-containment filtered venting system (IFVS) that enable passively no emergency evacuation outside the immediate vicinity of the plant for a SA. It has a large volume to hold hydrogen, a core catcher, a passive flooding system and a passive containment cooling system (PCCS) establishing passively practical elimination of containment failure even in a long term. The IDHS consists of 4 division active safety systems for a DBA, 2 division active safety systems for a SA and built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and the PCCS for a SA. IC/PCCS pools have enough capacity for 7 day grace period. The IC/PCCS, core and spent fuel pool are enclosed inside the CV building and protected against a large airplane crash. The iBR can survive a large airplane crash only by the CV building and the built-in passive safety systems therein. The dome of the CV building consists of a single wall made of steel and concrete composite. This single dome structure facilitates a short-term construction period and cost saving. CV diameter is smaller than that of most PWR resulting in the smaller R/B. The iBR is based on the proven ABWR technology and ready for construction. No new technology is incorporated but design concept and philosophy are initiative and innovative.
Development of Deep Hole Drilling Technology for Tube Sheet of Steam Generator
Authors: Ueda Shinichiro , Yagishita Takuya , Maeno Jun , Okuda Masanari
Published: November 2014

Advanced PWR nuclear power plant has very-large steam generators of which tube-sheet is diameter 5000mm / thickness 500 mm / weight 100 ton or larger. Drilling of very thick tube sheet needs deep-hole drilling technique with strict dimensional tolerance control. In addition, the tube sheet has 10,000 or more tube holes, which means that stable drilling operation is mandatory for productivity control. Upon such background, steam generator tube sheet drilling operation needs precise dimensional control technique with highly efficient productivity.

In this study, authors did (1) feasibility-study and development of deep hole drilling machine with the Boring and Trepanning Association (BTA) method, (2) cutting parameter evaluation such as feed rate and cutting velocity, (3) tube hole measurement system development, and (4) drill exchange program development with monitoring drilling machine motor loads and setting alarm level for the loads.

Based on techniques developed by this study, IHI has achieved high-accuracy and stable deep-hole drilling technology for very-thick tube sheets applicable to the very-large steam generators.
3D FEA Elastoplastic Structural Analysis of Hypothetically Cracked Monju FBR Core Support Structures
Authors: Garcia-Rodriguez Daniel , Matsubara Shinichiro
Published: November 2014

In this work the structural reliability of the circumferentially cracked core support mount of Monju Fast Breeder Reactor (FBR) is analyzed using Finite Element Analysis (FEA). The 3D shell model employed was derived after detailed evaluation of the core support mount behavior with a specific 3D solid model. First, elastoplastic static analysis results show that, under nominal operating conditions, the overall structure would be able to survive a total loss of the core support mount. Second, using the double elastic slope method it was inferred that earthquake loading integrity could be warranted up to a crack representing more than 50% of the total circumference. Both results highlight the ample primary loading margins taken in the design of Monju’s reactor core support structures. Furthermore, the developed 3D shell FEA model will be applied to study other extreme cases such as those under severe accident conditions.
Hydrogen Distribution Assessment With CFD Tool in Help of Severe Accident Management
Authors: Philipov Svetlin , Filipov Kalin
Published: November 2014

This paper presents the results of an analysis of the application of CFD tool to help hydrogen management. Some information pointed out the problem of hydrogen generation and distribution. Passive autocatalytic recombiners are the point of interest and mainly PAR units’ location. A severe accident is taken into account regarding the sources of hydrogen generation. The analysis of the severe accident progression is performed with MELCOR code. CFD tool Fluent (ANSYS) is applied to assess hydrogen and steam distribution in the atmosphere of the containment (confinement). The NPP unit of type WWER 440 (V230) is considered but as it is stressed this fact is irrelevant to phenomena and accident management targets.
Analysis of Two-Phase Flow Instability in Multi-Channel System
Authors: Xia Genglei , Peng Minjun , Xue Du
Published: November 2014

Ledinegg instability is one of the most important static instabilities for two phase flow system, especially in microchannel systems. In this paper, the force circulation two phase flow instability in vertical multi-channel system is performed by the best estimate system computer code RELAP5. The process and inherent reason of flow instability between multichannel system (FIBM) and flow excursion in forced circulation parallel channel system are analyzed. The effects of main operating parameters related to static onset of flow instability are investigated. Inlet subcooling, inlet restrictor, and saturation pressure are sensitive to the stability of parallel channel system.
On the Prospects of Using Deaeratorless Thermal Schemes With Direct Contact Low-Pressure Reheaters at New Nuclear Power Plants
Authors: Gotovsky M. A. , Ermolov V. F. , Mikhailov V. E. , Sukhorukov Yu G. , Trifonov N. N.
Published: November 2014

Now in Russia a new NPP “Brest – OD - 300” is developing which is to become a head in a series of fast reactors cooled by liquid lead or lead-bismuth alloy. Russian abbreviation OD is translated as Pilot-Demonstrational. The ideas are developed to use in this power plant deaeratorless thermal circuit. Such the schemes are widely and successfully used for conventional power plants in Russia.

Deaeratorless thermal schemes are based on the use of direct contact low-pressure reheaters. These reheaters have deaeration ability. Such schemes improve power plants efficiency by 1.5 %. Some other deaerator functions are distributed among other elements of plant. In particular, the presence of an external, independent supply of water, which is available at nuclear power plants and much higher than the supply of water in the deaerator.

Danger of return flow exists for direct-contact low pressure reheaters. But its design allows to eliminate completely the possibility of dangerous return flow. Compliance with safety was verified by calculation and in the operated power plants.
Study on the Dynamic Characteristics of the Water-Lubricated Tilting-Pad Radial Bearing Considering Temperature-Viscosity Effect
Authors: Li Yu , Wang Dezhong , Gu Weiguo
Published: November 2014

One water-lubricated tilting-pad radial bearing is studied in this paper to learn the effect of temperature-viscosity effect caused by the water friction loss on the surface of the pads on the carrying capacity and dynamic characteristics of the bearing. The Reynolds equation considering the turbulent stress term and the energy equation are established based on a simplified model of the tilting-pad radial bearing. The influence of the viscosity change to the dynamic characteristics of the bearing is analyzed while considering the temperature rise in case of water friction loss.

Result shows that the temperature and water friction loss will increase in a large degree when shaft neck eccentricity is large. The carrying capacity of the bearing will decrease when the temperature increases. However, the rise of the temperature will not change the stiffness and damping coefficients. The maximal rise of the temperature reaches 293.15K (20 °C) when shaft neck eccentricity angle is 180°and eccentricity ratio is 0.9 leading to the reduction of bearing capacity by 8%. Thus, to maintain the stability and avoid the obvious temperature rise by water friction, the load force upon the shaft should be less than 280kN when the surrounding temperature is 353.15K (80°C) and shaft neck eccentricity angle is 180° and eccentricity ratio is less than 0.8.
Study on the Rod Drop Performance of High-Temperature Gas-Cooled Reactor
Authors: Yan He , Diao Xingzhong
Published: November 2014

In this paper, the theoretical study and experimental investigation on the rod drop performance of high-temperature gas-cooled reactor (HTGR) pebble-bed module have been presented. The control rod drive mechanisms (CRDMs), serving as the first shutdown system of the reactor, are positioned above the reactor pressure vessel. When the reactor is operated at the power regulation mode, the control rods are pulled up-and-down in their channels around the reactor core. The CRDM provides a fail-safe operational mode for the control rod system. If the reactor emergency shutdown is required the control rods could drop into their channels by gravity. Thus the key factor, emergency insertion time of the whole control rod stroke, which represents the inherent safety of the CRDM, is crucially important and should be measured precisely. In the final objective of ensuring reliability of the CRDM, a full size drive line had been built and tested to obtain the overall performance function of the CRDM. Every component of the CRDM test line was simulated at the scale 1:1, including a 15 meters high test bench that was used as the substitution of the pressure vessel. At current stage, the rod drop performance had been experimental investigated at ambient temperature and pressure. The emergency insertion time of an 8 meters stroke was measured to be less than 50 seconds. A mathematical model of CRDM also had been developed. The rod motion characteristic equations show that the rod dropping speed approaches to a constant during the emergency insertion. The theoretical results are in agreement with the test results.
Investigation on Coupling Behaviors of Thermal-Hydraulics/Neutronics Under Asymmetrical Inlet Conditions
Authors: Zheng Yong , Peng Min-jun , Xia Geng-lei , Li Ren
Published: November 2014

The reactor core is a complex system involving the reactor physics, thermal hydraulics and many other aspects. That means the distribution of the core power largely determines the profile of the thermal parameters, meanwhile the local thermal-hydraulics condition will in turn affect the neutronics calculation by moderator temperature effect and Doppler effects. Issues coupling the thermal-hydraulics with neutronics of nuclear plants still challenge the design, safety and the operation of LWR few years ago. Fortunately, the recent availability of powerful computer and computational techniques has enlarged the capabilities of making more realistic simulations of complex phenomena in NPPs.

The current study deals with the development of an integrated thermal-hydraulics/neutronics model for Qinshan phase II NPP project reactor for the analysis of specific plant transients in which the neutronic response of the core is important, application of RELAP5-HD making use of the Helios code to derive the macroscopic cross-sections. Based on the coupled model, the steady state calculation and the transient simulation, involving the abnormal operation mode with asymmetrical coolant flux and temperature on the inlet of reactor, have been performed. The results show that the values obtained from coupled code RELAP5-HD calculation are in good agreement with the available experimental data, and the calculated accident parameters curves can predict all major trends of the transient. Steady state and transient condition calculation results are in accordance with the theoretical analysis from the aspect of coupled thermal-hydraulics/neutronics, this demonstrated a successful best estimate coupled RELAP5-HD model of Qinshan phase II NPP reactor has been developed, and the established model will provide a good foundation for the further analysis of the primary loop. It also can be concluded that the more accurate CFD method coupling three dimensional neutron kinetics code based on neutron diffusion method are necessary for steady-state calculation and analysis of transient/accident conditions when asymmetrical processes take place in the core. It is worth mentioning that RELAP5-HD code has already programmed the human-machine interface and the interface for coupling with other code, hence RELAP5-HD code has a broad application prospect in PWRs safety analysis.
Elaboration of the System Based Code Concept — Activities in JSME and ASME: (2) Development of Evaluation Tools Based on LRFD
Authors: Machida Hideo , Asayama Tai , Watanabe Daigo , Hojo Kiminobu , Hayashi Masaaki
Published: November 2014

In the System Based Code (SBC), a reliability target is defined according to the importance of components (risk and/or failure probability), and the grade of material, design, manufacture and maintenance are chosen to satisfy the reliability target. Therefore, a reliability assessment of the components plays an important role in the concept of SBC. In the reliability assessment guidelines of SBC under development in JSME, the Load and Resistance Factor Design (LRFD) method and the Monte Carlo method will be applied to reliability assessment. Along with a development plan of the SBC, examination of the target reliability according to the importance of components and preparation of the LRFD methods and assessment of partial safety factor for some failure modes are underway. Until now, the LRFD methods have been developed for burst due to internal pressure, plastic collapse due to membrane and bending stress, fatigue, plastic collapse of a pipe having a circumferential crack, and buckling of a thin wall cylinder. This paper describes the action plans for the development of the reliability assessment methods and examination results till date.
Coupled Simulations of Natural and Forced Circulation Tests in NACIE Facility Using RELAP5 and ANSYS Fluent Codes
Authors: Martelli D. , Forgione N. , Barone G. , Del Nevo A. , Di Piazza I. , Tarantino M.
Published: November 2014

In this work the activity performed at the DICI (Dipartimento di Ingegneria Civile e Industriale) of the Pisa University in collaboration with the ENEA Brasimone Research Centre is presented. In particular the document deals with the application of an in-house developed coupling methodology between a modified version of RELAP5/Mod3.3 and Fluent commercial CFD code, to the NACIE (Natural Circulation Experiment) LBE experimental loop (built and located at the ENEA Brasimone research centre).

The first part of the document treats the description of the NACIE loop type facility, while in the second part, the developed coupling tool is presented and the obtained numerical results are compared to stand alone RELAP5 results and to data obtained from the NACIE experimental campaign. The experimental tests are performed varying the argon flow rate and the electric power supplied to the heater and both natural and assisted circulation tests are investigated. The numerical model set-up is based on a two-way explicit coupling scheme and 2D and 3D geometrical domain were investigated.

Comparative analyses among numerical and experimental results showed good agreement, giving positive feedback on the feasibility and capability of the developed coupling methodology.
Fuel Dispersion and Flow Blockage Analyses for MYRRHA-FASTEF Reactor by SIMMER Code
Authors: Bandini Giacomino , Eboli Marica , Forgione Nicola
Published: November 2014

This work illustrates the 3D set up model and the results concerning the recent analysis of fuel dispersion in the MYRRHA-FASTEF reactor performed with SIMMER code within the EU-FP7 SEARCH Project. Under severe accidental conditions, the release of fuel in the primary system can occur in case of fuel rod clad failure and degradation. Two cases were therefore taken into account, an imposed fuel release to study key parameters which influence the dispersion phenomenon and a coolant flow blockage in a fuel assembly.

The reactor was simulated by a 3D Cartesian geometry with 65×63×42 cell mesh. Steady-state and transient analyses were performed by SIMMER-IV. Steady-state analysis was performed in order to assess the correct operability of the code and of the model. The results were compared with the design values. The most significant results obtained for temperature trends and profiles, velocity and mass flow rate trends are reported. Transient results were also analysed, i.e. fuel dispersion transients were simulated, comparing the effect of fuel porosity on the fuel dispersion inside the pool. In addition, the effects of the release position and the fuel particle dimension on the dispersion phenomenon were also investigated.

The final section of the paper describes the effects of a flow blockage on the core degradation and dispersion of fuel particles in the primary circuit of the MYRRHA reactor. This simulation, with fuel porosity equal to 5%, started after a preliminary steady state condition. The mass flow rate in one of the inner fuel assemblies was then reduced to about 10% of the initial value.

The results show that the SIMMER-IV code is capable of predicting steady-state results in good agreement with the nominal values, also confirming the correctness of the set up model.
Integral Effect Test for Performance Evaluation of the PAFS (Passive Auxiliary Feedwater System) During a SBO (Station Blackout) Transient
Authors: Bae Byoung-Uhn , Kim Seok , Park Yu-Sun , Cho Yun-Je , Kang Kyoung-Ho
Published: November 2014

Station blackout (SBO) accident is considered as one of the most significant design extension conditions (DECs), which has been extensively focused after the Fukushima Dai-chi accident. When the SBO accident occurs in the APR+ (Advance Power Reactor Plus), the PAFS (Passive Auxiliary Feedwater System), which is an advanced safety feature adopted in the APR+, should play a significant role to cool down the core decay heat without any operation of active safety systems. This study focuses on validation of the cooling and operational performance for the PAFS during the SBO transient with utilizing an integral effect test facility, ATLAS-PAFS. In order to simulate the SBO transient of the APR+ as realistically as possible, a pertinent scaling approach was taken into account. The initial steady-state conditions and the sequence of event in the SBO scenario for the APR+ were successfully simulated with the ATLAS-PAFS facility. In the transient simulation, major thermal-hydraulic parameters such as the system pressures, the collapsed water levels, the break flow rate, and the condensate flow rate at the return-water line were measured and investigated. Following the reactor trip at the initiation of the transient, the coolant inventory of the secondary system of the steam generator was reduced by the repeated opening and closing of the MSSV. When the collapsed water level reached 25% of wide range, the PAFS was actuated to cool down the primary system by the condensation heat transfer at the PCHX (Passive Condensation Heat Exchanger). The pressure and the temperature of the reactor coolant system continuously decreased during the heat removal by the PAFS operation. It points out that the PAFS can supply auxiliary feedwater to the steam generator and remove the core decay heat without any active system. From the present experimental result, it could be concluded that the APR+ has the capability of coping with the hypothetical SBO scenario with adopting the PAFS and proper set-points of its operation. This integral effect test data will be used to evaluate the prediction capability of existing safety analysis codes of the MARS, RELAP5 as well as the SPACE code and to identify any code deficiency for a SBO simulation with an operation of the PAFS.
To Analyze the Core Damage Scenario via the RISMC-Based Metatheory and Reliability Theory
Authors: Zheng Gangyang , Gong Yu , Zhang Zhijian , Liu Zibin
Published: November 2014

With “theory of nuclear safety (TONS)”, this paper intends to explain the Core Damage (CD) scenario of a Nuclear Power Plant (NPP) with the systematic methodology, many notions introduced here can be extended to other types of nuclear installations, as well. This systematic methodology combines the Risk-Informed Safety Margin Characterization (RISMC) Metatheory of TONS, and the basic reliability theory. A “metatheory” of such theories, here, is a theory to analyze the Theory of Nuclear Safety (TONS); in its own theory system, it is designed to summarize the safety of a NPP. Meanwhile, the basic reliability theory, which is decided by the authors, is focus on the mission reliability model (a model can be established by Reliability Block Diagram (RBD)); then the related basic concepts, is simple and clear, and quite mature in NPP field.

The present work outlines the traditional reliability theory and the RISMC-based Metatheory, and these two concepts here are taken as the appropriate TONS to analyze the CD Scenario, after that, a renovate or renew TONS, from these two sides, can be introduced to analyze the fundamental safety of NPP.
Stress Analysis of the Canned Nuclear Coolant Pump Based on FSI
Authors: Ji-Ge Zhang , Yue Zhu
Published: November 2014

The reactor coolant pump (RCP) is one of the most important components in nuclear power plants. It operated in high temperature, high-pressure, high speed and radiative environment, so a long-term security and reliable operations is required. Many internal flow analysis of RCPs was carried out, mainly foucs on steady and unsteady flow field at different operating points in RCP. The research about flow passage components, such as the blade of the RCP, is relatively few. When the RCPs operates in the nuclear power plant, the flow field lashed against the impeller of the RCP, results in a network of small cracks is found on the surface of impeller. For example, periodic vibration caused by a break in a guide vane leaded to cracking of two pieces blades of impeller in a large power plant in southwest of China, and this accident caused much economic loss. The computational method of stress due to the hydraulic reason is an important problem of the RCP. In this work, at first the CFD simulation including the case, guide vane, impeller, inlet and outlet at different operation points is studied, and the result of the pressure distribution on impeller blade is loaded on the impeller using fluid-structure interation (FSI) method. The result showed that the maximum von Mises stress appears on the trailing edge close to the impeller hub which has a large change in gradient of stress and which is prone to fatigue failure. The maximum stress on the impeller is mainly in proportion to the operating power. The maximum stress on the impeller have periodical characteristic, which is due to the number of blade of diffuser. All of these equip us with better understand of the fatigue and fracture of RCP, and it make sense to protect the fatigue damage and promote the stability of RCP.
Pressure Drop Analysis of a Re-Entrant Fuel Channel in a Pressure-Channel Type SuperCritical Water-Cooled Reactor
Authors: Maghsoudi S. , Peiman W. , Pioro I. , Gabriel K.
Published: November 2014

Pressure drop calculation and temperature profiles associated with fuel and sheath are important aspects of a nuclear reactor design. The main objective of this paper is to determine the pressure drop in a fuel channel of a SuperCritical Water-cooled Reactor (SCWR). One-dimensional steady-state thermal-hydraulic analysis was conducted. In this study, the pressure drops due to friction, acceleration, local losses, and gravity were calculated at supercritical conditions. The total pressure drop due to all these parameters was between 108 and 121 kPa.
Validation of a 3D Hybrid CFD-DEM Method Based on a Self-Leveling Experiment
Authors: Guo Liancheng , Morita Koji , Tagami Hirotaka , Tobita Yoshiharu
Published: November 2014

The postulated core disruptive accidents (CDAs) are regarded as particular difficulties in the safety analysis of liquid-metal fast reactors (LMFRs). In the CDAs, core debris may settle on the core-support structure and form conic bed mounds. Heat convection and vaporization of coolant sodium will level the debris bed, which is named “self-leveling behavior” of debris bed. To reasonably simulate such transient behavior, as well as thermal-hydraulic phenomena occurring during a CDA, a comprehensive computational tool is needed. The SIMMER code is a successful computer code developed as an advanced tool for CDA analysis of LMFRs. It is a multi-velocity-field, multiphase, multicomponent, Eulerian, fluid dynamics code coupled with a fuel-pin model and a space- and energy-dependent neutron kinetics model. Until now, the code has been successfully applied to simulations of key thermal-hydraulic phenomena involved in CDAs as well as reactor safety assessment. However, strong interactions among rich solid particles as well as particle characteristics in multiphase flows were not taken into consideration for its fluid-dynamics models. Therefore, a hybrid computational method was developed by combining the discrete element method (DEM) with the multi-fluid models to reasonably simulate the particle behaviors, as well as the thermal-hydraulic phenomena of multiphase fluid flows. In this study, 3D numerical simulation of a simplified self-leveling experiment is performed using the hybrid method. Reasonable agreement between simulation results and corresponding experimental data demonstrated the validity of the present method in simulating the self-leveling behavior of debris bed.
Preliminary Severe Accident Analysis of Fusion-Fission Hybrid Reactor Using the MELCOR Code
Authors: Dabin Zhang , Zhou Zhiwei , Xie Heng , Yang Tang
Published: November 2014

The fusion-fission hybrid conceptual reactor is a proposed means of generating power, which adopts a water cooled fission blanket based on ITER. Due to the water cooled fission blanket, safety performance of the hybrid reactor should be considered, including decay heat remove, core uncovered, core meltdown, core degradation, radioactivity releases, etc., similar with the PWRs. The main goal of this work is to develop the fission blanket model by using MELCOR code, and to evaluate the safety performance under severe accidents preliminarily.

Based on MELCOR 1.8.5, some modification is made for the severe accident analysis of fission blanket. Using modified MELCOR code, an analysis model is developed for the fission blanket and the cooling loop. The strategy of the In-Vessel Retention (IVR) using the ex-vessel cooling method is evaluated during a large break LOCA. The calculation results describes the main phenomena during the severe accident progression, including core dry out, meltdown, relocation, etc.. Simulation result is shown that the decay heat in the fission zone can be removed out by the ex-vessel cooling system merely, and the fuel max temperature will not reach the melting point.
Computation of Thermal Striping in the Upper Plenum of PGSFR
Authors: Kim Dong-Eun , Ko Sung-Ho , Choi Seok-Ki , Lee Tae-Ho
Published: November 2014

A computational study of a thermal striping in the upper plenum of the PGSFR (Prototype Generation-IV Sodium-cooled Fast Reactor) being developed at KAERI (Korea Atomic Energy Research Institute) is presented. First, previous experimental and numerical studies on the thermal striping are briefly discussed. Both RANS (Reynolds-Averaged Navier-Stokes) and LES (Large Eddy Simulation) approaches are employed for the simulation of thermal striping in the upper plenum of the PGSFR. For the RANS approach, the conventional k–ε turbulence model is employed and the LES is performed using the WALE (Wall-Adapting Local Eddy-viscosity) model. More than 11.8 million unstructured elements are generated in the upper plenum region of the PGSFR using the ICEM commercial code. From the RANS results, the time-averaged velocity components and temperature field in the complicated upper plenum of PGSFR are calculated. In the LES results, the time history of temperature fluctuation at the several locations of solid walls of UIS (Upper Internal Structure) and IHX (Intermediate Heat Exchanger) are additionally stored. Comparisons of the predicted time-averaged velocity components and temperature between the two methods are also presented. From the temporal variation of temperature at the solid walls, one can find the locations where the thermal stress is large and assess whether the solid structures can endure the thermal stress during the reactor life time.
Chinese Innovative Patent Technologies for the Reactor Pressure Vessel
Authors: Shi Qian , Zhang Tianqi , Fang Xiang , Peng Wei
Published: November 2014

There is a huge and rapid growing demand for energy in China. The developing of new energy, including nuclear energy, is an important way to solve China’s energy and environment problems. Currently, Nuclear power is playing a more and more important role in China, especially in the coastal areas where the economy is developing rapidly. The main nuclear reactor equipment is the key guarantee to nuclear safety, such as the reactor pressure vessel. With the continuous development of science and technology, the reactor pressure vessel uses more and more of advanced design methods, high performance materials and industrial-proven technologies. As an airtight container who is directly exposed to the great operating pressure of reactor, the reactor pressure vessel plays a crucial role to the safety of reactor. In the paper, by retrieving Chinese Intellectual Property Office patent libraries, 12 Chinese innovative patents for the reactor pressure vessel are introduced, containing four kinds of main technologies: closure, body, nozzle safe end and other technologies. The patent results could lead the corresponding advice to the design of the equipment effectively, to the system arrangement reasonably, and further to the reference for designing a safe nuclear power plants. The paper is to exchange the latest technological progress of reactor pressure vessel, in order to promote technological innovation in the nuclear power field.
3D Core Thermal Hydraulics With the PORFLO Code: Turbulence Modeling and Porous Medium With Porosity Steps
Authors: Ilvonen Mikko , Hovi Ville , Taivassalo Veikko
Published: November 2014

VTT is developing a 3D two-phase flow solver code called PORFLO with the objective of enabling coupled reactor dynamics simulations through a parallel computation and porous medium description of complex geometries. The current features include unstructured mesh usage, 6-equation flow solution, k-ε turbulence modeling and MPI parallelization. The present applications are fuel bundles and cores of VVER and EPR, with a steam generator and core debris bed having been simulated previously. In this work, the main new developments are improved turbulence modeling, improved interpolation of face velocities in porous medium, and new improved modeling of EPR plant. The recent work examines porous medium based macroscopic models for turbulence quantities in the fuel assembly sub-channels. A few modifications in the standard k-ε turbulence model were tested against experimental data from the AGATE, MATiS-H and IVO VVER LDA measurements. In the coupled solution of pressure and velocities, problems were encountered at locations with a large change in the porosity. In the latest development, the pressure weighted momentum interpolation of face velocities was modified to explicitly account for the interphase drag term, in a manner similar to the Coupled Phasic Exchange Algorithm. In this work, the improved two-phase solver was tested in single-phase cases.
HLM Fuel Pin Bundle Characterization in CIRCE Pool Facility
Authors: Tarantino M. , Martelli D. , Di Piazza I. , Forgione N. , Agostini P. , Coccoluto G.
Published: November 2014

Since the Lead-cooled Fast Reactor (LFR) has been conceptualized in the frame of GEN IV International Forum (GIF), ENEA is strongly involved on the HLM technology development.

Currently ENEA has implemented large competencies and capabilities in the field of HLM thermal-hydraulic, coolant technology, material for high temperature applications, corrosion and material protection, heat transfer and removal, component development and testing, remote maintenance, procedure definition and coolant handling.

In this frame the Integral Circulation Experiment (ICE) test section has been installed into the CIRCE pool facility, and suitable experiments have been carried out aiming to deeply investigate the pool thermal-hydraulic behavior of a HLM cooled pool reactor.

In particular a fuel pin bundle simulator (FPS) has been installed in the CIRCE pool. It has been conceived with a thermal power of about 1 MW and a linear power up to 25kW/m, relevant values for a LMFR. It consist of 37 fuel pins (electrically simulated) placed on a hexagonal lattice with a pitch to diameter ratio of 1.8. The pins have a diameter of 8.2mm and active lengths of 1 m. Along the FPS, three spacer grid properly designed by ENEA have been installed.

The FPS has been deeply instrumented by several thermocouples. In particular three sections of the FPS have been instrumented to monitor the heat transfer coefficient along the bundle as well as the cladding temperature in different rank of sub-channels.

A full characterization of the FPS has been experimentally achieved both under forced and natural circulation, and the main results gained during the run are reported into the paper.

Moreover the paper reports a preliminary analysis and discussion of such results, also in comparison with CFD calculations performed by CFX code.
Computational Fluid Dynamics Modeling of Two-Phase Boiling Flow and Critical Heat Flux
Authors: Tentner Adrian , Merzari Elia , Vegendla Prasad
Published: November 2014

This paper presents recent advances in the modeling of two-phase boiling flow and critical heat flux that have been implemented in the Extended Boiling Framework (EBF) [1, 2, 3]. The EBF code was developed as a customized module built on the foundation of the commercial Computational Fluid Dynamics (CFD) code STAR-CD, which provides general two-phase flow modeling capabilities, for the detailed analysis of the two-phase flow and heat transfer phenomena that occur in Boiling Water Reactor (BWR) fuel assemblies. These phenomena include coolant phase changes and multiple flow regimes that directly influence the coolant interaction with the fuel pins and, ultimately, the reactor performance. An effort to expand the EBF two-phase models and to explore their applicability to other CFD codes is currently underway.

The paper presents results of recent CFD analyses of Critical Heat Flux (CHF) experiments that have measured the axial distribution of wall temperature in two-phase upward flow in a vertical channel with a heated wall. The experiments were designed to produce the onset of CHF in the upper half of the heated channel. The simulated axial distribution of wall temperature is compared with experimental data, illustrating the ability of the extended EBF model to capture the onset of CHF for a wide range of thermal-hydraulic conditions relevant for BWRs. The paper concludes with a discussion of results and plans for future work.
Experimental Investigation of Dynamic Characteristics of Swing and Tilted Disc Check Valves
Authors: Marcinkiewicz Jerzy , Adamkowski Adam , Lewandowski Mariusz , Janicki Waldemar
Published: November 2014

The present work describes an experimental investigation of the dynamic characteristics of check valves, which means experimental examination of closing function and ability to generate pressure transients under different flow decelerations in the pipeline. Two designs of check valves are tested: a swing disc and a tilted disc check valve.

The valves are mounted on discharge pipe of a centrifugal pump. The fluid transient is generated by stopping the pump motor from actual velocity to completely stop in a prescribed time.

Each of the check valves is subjected to tests covering different pressure levels in the upper reservoir, initial flow rates in the pipeline, several decelerations of pump rotation, three settings of torque acting on the valve disc, three values of the moment of friction forces acting on the valve axis and finally free fall of the disc in stagnant water and air.

The test stand, the instrumentation and chosen valves as well as scope and conditions for performing the experiments are described. Selected measured results like angular velocities of the discs, pressures in the pipe at different conditions, and volumetric flow rates are presented and discussed.

The dynamic behaviors of the tested valves were compared with each other.
Monte Carlo Burn-Up Code System MCADS and its Application
Authors: Yan Xuesong , Qi Ji , Yang Lei , Zhang Xunchao , Zhang Yaling
Published: November 2014

A Monte Carlo burn-up code system, MCADS, which combines the Monte Carlo code MCNP and the depletion code LITAC, was developed for evaluation the capabilities of fuel burn-up in reactors. The code can be extremely useful for analysts who perform isotope production, material transformation, and depletion and isotope analyses on complex, non-lattice geometries, and uniform and non-uniform lattices. Three well-documented benchmarking problems including IAEA-ADS, OECD/NEA-PWR and fast reactor were used to validate MCADS. The discrepancy between the MCADS calculation and other codes fell into a reasonable range.
Simulation of Gas Stratification Build-Up in the Containment Under Severe Accident Conditions
Authors: Andreani Michele
Published: November 2014

The presence of hydrogen stratification in a NPP containment in the case of a severe accident is a source of concern, as pockets of the gas in high concentration could lead to a deflagration or detonation risk, which might challenge the containment structural integrity. These issues, as well as the capability of various computer codes to predict the evolution of a representative accident, are addressed in the coordinated projects ERCOSAM of the 7th EURATOM FWP and the project SAMARA sponsored by ROSATOM. The projects aim to establish whether in a test sequence representative of a severe accident in a LWR hydrogen stratification can be established during the initial transient following a loss of coolant accident (LOCA) and whether and how this stratification can be broken down by the operation of Severe Accident Management systems (SAMs): sprays, coolers and Passive Auto-catalytic Recombiners (PARs). Experiments with helium (as simulant of hydrogen) have been performed at “small scale” in TOSQAN (IRSN, Saclay), and “medium scale” in the MISTRA (CEA, Saclay), PANDA (PSI, Villigen) and SPOT ((JSC “Afrikantov OKBM”, Nizhny Novgorod) facilities. The present paper presents the analysis of the initial transient of some tests in the PANDA, TOSQAN and SPOT facilities using the GOTHIC computer code. The work therefore addresses the capability of the code and a relatively coarse mesh to simulate the pressurisation and build-up of steam and helium stratification for conditions representative of a postulated severe accident scenario, properly scaled to the various facilities. The prediction of the pressurisation is excellent, and the position of the gas concentration stratification front at the end of the steam and helium releases is generally well captured.
Role of Zirconium Nitride Formation Under Zirconium-Based Claddings Oxidation in Air During NPP Beyond-Design-Basis Accidents
Authors: Vasiliev Alexander
Published: November 2014

The high-temperature (≈800–2100°K) oxidation of nuclear reactor fuel assembly zirconium-based claddings in the air behaves in different way compared to oxidation in the steam.

First, the heat effect of the chemical reaction of oxidation in the air is approximately two times larger than in the steam.

Second, the kinetics of oxidation in the air is non-parabolic (approximately linear, that is more strong) in contrast to parabolic kinetics of oxidation in the steam.

Third, important factor is the diffusion of oxygen and nitrogen atoms through very porous (transparent) nitride layers formed in α-Zr(O) layer. It resulted in linear kinetics in small-scale tests and strong hydrogen release during reflood phase in QUENCH-16 and PARAMETER-SF4 tests. The role of nitride formation during air ingress phase seems to be extremely important especially if oxygen starvation phase is present.

Fourth, it is necessary to treat adequately the interdiffusion of oxygen and nitrogen in gas mixture containing, oxygen, nitrogen and possibly steam, hydrogen, other non-condensables. This effect influences the time moment of oxygen starvation.

Fifth, it is important to take into account the heat effect of nitridation (formation of ZrN) which is not small.

Sixth, at final phase (quenching at water reflood or slow cooling) the reoxidation of nitrides can take place. During this process. nitrogen and hydrogen (in steam atmosphere) is released and additional chemical heat generation is produced.

The development of a special model of Zr-based alloys oxidation in air considering all above-mentioned processes is currently underway, some results are presented below. A numerical scheme is realized to determine layer boundaries relocation and layer transformations in claddings.

The model is implemented to SOCRAT computer modeling code. This model is being verified on separate-effect tests as well as on integral air ingress experiments QUENCH-10, QUENCH-16 and PARAMETER-SF4.
Head Loss Through Fibrous Beds Generated on Different Types of Containment Sump Strainers
Authors: Lee Saya , Abdulsattar Suhaeb , Hassan Yassin A.
Published: November 2014

During a Loss of Coolant Accident (LOCA), the high energy jet from the break may impinge on surrounding surfaces and materials, producing a relatively large amount of fibrous debris (mostly insulation materials). The debris may be transported through the reactor containment and reach the sump strainers. Accumulation of such debris on the strainers’ surface can cause a loss of Net Positive Suction Head (NPSH) and negatively affect the Emergency Core Cooling System (ECCS) capabilities. The U.S. Nuclear Regulatory Commission (U.S.NRC) initiated the Generic Safety Issue (GSI) 191 to understand the physical phenomena involved in this type of event, and help develop the tools to prove the safety and reliability of the existing Light Water Reactors (LWR) under these conditions. Some nuclear power plants have already adopted countermeasures in an attempt to limit the effect of the debris accumulation on the ECCS performance, by replacing or modifying the existing strainer configurations. In this paper, two different strainer designs have been considered and sensitivity analysis was conducted to study the effect of the approach velocity on the pressure drop at the strainer caused by the debris accumulation. The development of the fibrous beds was visually recorded in order to correlate the head loss, the approach velocity, and the thickness of the fibrous bed. The experimental results were compared to semi-empirical models and theoretical models proposed by previous researchers.
CFD Analysis of Single Phase Counter-Current Buoyancy Driven Flows and its Applications to Passive Reactor Design
Authors: Sebilleau Frederic , Kansal Anuj K. , Issa Raad I. , Walker Simon P.
Published: November 2014

Increasingly, nuclear plants rely on natural circulation, for both fault conditions and / or normal power removal. Prediction of such buoyancy-driven flows is needed. However, their complex nature leads to 3D effects in ‘wide’ geometries, making prediction impossible with system codes. Even in slender “pipe-like” geometries countercurrent flow of hot and cold fluid makes a one-dimensional simulation totally misleading. However, simply moving to a three-dimensional CFD treatment is not sufficient. The strong anisotropy of the turbulence and the coexistence of various flow regimes make the choice of an appropriate turbulence model difficult. Countercurrent flow in a pipe might occur when the “natural” buoyant flow was of hot fluid up the pipe, but a feature such as a local heat-sink (an un-insulated valve in the pipe, perhaps) acts as a source of cold fluid, which attempts to flow down the pipe as a counter-current flow. On a different scale, counter current flow such as this would occur for example inside the secondary containment. This countercurrent flow problem captures the complexities of most buoyant flows, and this provides a challenging model problem. In this paper, we describe the design and preliminary analysis of an experimental rig being built to study this. Initial CFD and experimental results are presented.
Use of Level 1 PSA in the Frame of EPR Licensing
Authors: Georgescu Gabriel , Dupuy Patricia , Corenwinder Francois
Published: November 2014

The French “Technical Guidelines for the design and construction of the next generation of NPPs with Pressurized Water Reactors” specify that the safety demonstration has to be achieved in a deterministic way, supplemented by probabilistic methods. In this context, for the EPR reactor of Flamanville (EPR-FA3), the PSA has been used from the beginning of the design. In the frame of the application for commissioning of EPR-FA3, EdF has to provide an “as-build” full scope PSA for the reactor and for the spent fuel pool, covering the internal events, as well as the internal and external hazards of significant impact. Some of these probabilistic studies were developed and evolved during the EPR design (PSA for “internal events”, specific studies for practically eliminated sequences, long term accident sequences…) and were analysed by IRSN, as French Safety Authority (ASN) technical support, at different EPR project stages (initial design, detailed design, construction application….) leading to many design and studies improvements. Today, in order to make the analysis of the application for the commissioning of the EPR-FA3 reactor more effective, the “anticipated” analysis of this application is in progress in France. In this context, the updated versions of the level 1 probabilistic studies for internal events and hazards were analysed by IRSN in 2013. The results and conclusions of this analysis were presented by IRSN early 2014 during a dedicated meeting of French Standing Group of experts for Reactors safety (SGR). The paper presents the analysis performed by IRSN of EdF EPR-FA3 level 1 probabilistic studies, highlighting the role of PSA to the achievement of high level of safety of EPR reactor.