bluuebunny/crossref_metadata_2025 huggingface dataset:
Study on the Core Cooling Scheme After Accident Shutdown of the Pebble-Bed Modular High Temperature Gas-Cooled Reactor
Authors: Zheng Yanhua , Shi Lei , Chen Fubing
Published: October 2013

One of the most important properties of the modular high temperature gas-cooled reactor is that the decay heat in the core can be carried out solely by means of passive physical mechanism after shutdown due to accidents. The maximum fuel temperature is guaranteed not to exceed the design limitation, so as to the integrity of the fuel particles and the ability of retaining fission product will keep well. Nonetheless, the auxiliary active core cooling should be design to help removing the decay heat and keeping the reactor in an appropriate condition effectively and quickly in case of reactor scram due to any transient and the main helium blower or steam generator unusable.

Based on the preliminary design of the 250 MW pebble-bed modular high temperature gas-cooled reactor, assuming that the core cooling will be started up 1 hour after the scram, different core cooling schemes are studied in this paper. After the reactor shutdown, a certain degree of natural convection will come into being in the core due to the non-uniform temperature distribution, which will accordingly change the core temperature distribution and in turn influence the outlet hot helium temperature. Different cooling flow rates are also analyzed, and the important parameters, such as the fuel temperature, outlet hot helium temperature and the pressure vessel temperature, are studied in detail. A feasible core cooling scheme, as well as the reasonable design parameters could be determined based on the analysis. It is suggested that, considering the temperature limitation of the structure material, the coolant flow direction should be same as that of the normal operation, and the flow rate could not be too large.
Identification of Limiting Case Between DBA and SBDBA (CL Break Area Sensitivity): A New Model for the Boron Injection System
Authors: Gonzalez Raul , Petruzzi Alessandro , D’Auria Francesco , Mazzantini Oscar
Published: October 2013

Atucha-2 is a Siemens-designed PHWR reactor under construction in the Republic of Argentina. Its geometrical complexity and (e.g., oblique Control Rods, Positive Void Coefficient) required a developed and validated complex three dimensional (3D) neutron kinetics (NK) coupled thermal hydraulic (TH) model. Reactor shut-down is obtained by oblique CRs and, during accidental conditions, by an emergency shut-down system (JDJ) injecting a highly concentrated boron solution (boron clouds) in the moderator tank, the boron clouds reconstruction is obtained using a CFD (CFX) code calculation. A complete LBLOCA calculation implies the application of the RELAP5-3D© system code. Within the framework of the third Agreement “NA-SA – University of Pisa” a new RELAP5-3D control system for the boron injection system was developed and implemented in the validated coupled RELAP5-3D/NESTLE model of the Atucha 2 NPP. The aim of this activity is to find out the limiting case (maximum break area size) for the Peak Cladding Temperature for LOCAs under fixed boundary conditions.
DNB Prediction Method Using Local Void Fraction Applied to a PWR Rod Bundle With Mixing Vane Grid
Authors: Sasakawa Tatsuya , Ikeno Tsutomu , Kataoka Isao
Published: October 2013

A new code coupled between a sub-channel analysis code and a computational multi-fluid dynamics (CMFD) code was applied to a PWR rod bundle with mixing vane grid. The code was developed to predict the departure from nucleate boiling (DNB). This is a new technology of CMFD based on abundant experience and models developed for sub-channel analysis: CMFD computed void distribution to fit the average value calculated by two-phase models in sub-channel analysis code. A new source term represented centripetal motion of small bubbles in the wake behind a rising vapor slug. In order to apply the code to a PWR rod bundle, effects of local mass flux, pressure and mixing vane grid was modeled. The present method was applied to the analysis for DNB tests of simulated PWR fuel assembly with mixing vane grids. The result showed that, using only a critical void fraction, reasonable prediction was achieved in a wide range of flow condition and a variety of flow regimes.
Suitability of Eutectic Field’s Metal for Use in an Electromagnetohydrodynamically-Enhanced Experimental Two-Phase Flow Loop
Authors: Lipchitz A. , Laurent Lilian , Harvel G. D.
Published: October 2013

Several Generation IV nuclear reactors, such as sodium fast reactors and lead-bismuth fast reactors, use liquid metal as a coolant. In order to better understand and improve the thermal hydraulics of liquid metal cooled GEN IV nuclear reactors liquid metal flow needs to be studied in experimental circulation loops. Experimental circulation loops are often located in a laboratory setting. However, studying liquid metal two phase flow in laboratory settings can be difficult due to the high temperatures and safety hazards involved with traditional liquid metals such as sodium and lead-bismuth. One solution is to use a low melt metal alloy that is as benign as reasonably achievable. Field’s metal is a eutectic alloy of 51% Indium, 32.5% Bismuth and 16.5% Tin by weight and has a melting point of 335K making it ideal for use in a laboratory setting. A study is undertaken to determine its suitability to use in a two-phase experimental flow loop enhanced by magnetohydrodynamic forces. The study investigated its reactivity with air and water, its ability to be influenced by magnetic fields, its ability to flow, and its ease of manufacture. The experiments melted reference samples of Field’s metal and observed its behaviour in a glass beaker, submerged in water and an inclined stainless steel pipe. Then Field’s metal was manufactured in the laboratory and compared to the sample using the same set of experiments and standards. To determine Field’s metal degree of magnetism permanent neodymium magnets were used. Their strength was determined using a Gaussmeter. All experiments were recorded using a COHU digital camera. Image analysis was then performed on the video to determine any movements initiated by the magnetic field forces. In conclusion, Field’s metal is more than suitable for use in experimental settings as it is non-reactive, non-toxic, simple to manufacture, easy to use, and responds to a magnetic force.
Study on Pressure Relief Transient for High Temperature Gas Cooled Reactor
Authors: Wang Yan
Published: October 2013

After a supposed accident lead the reactor to shutdown, the residual heat generated from the decay will be removed out by the cooling system. A cooling water flow will be pumped into the steam generator for cooling the reactor after the pressure of the secondary side of the steam generator is decreased to a lower design value with opening of the pressure relief valve which is installed on the live-steam pipeline. But the temperature of the heat-exchange tubes and the tube plate at the outlet of the steam generator whose temperature in the steady-state operation is 570° high, need to be cooled down to about 200° for the material limitation before the cooling water is injected, to avoid the cold-shock damage on the components. In this paper, exampled as the 200MWe high temperature gas cooled reactor (HTR-PM), a general thermal-hydraulic system analysis code is used for modeling and simulating numerically the pressure relief transient after the accident. The results show the structure components of the steam generator will be cooled down effectively to mitigate the cold-shock damage from the cooling water during the transient with the designed pressure value of 1MPa where the pressure relief valve in 30mm diameter will be closed. It provides supports for the evaluation on the feasibility of cooling water injection and also for the material stress analysis on some relevant components in the steam generator design.
Heat Sink Issue: A Specific Problematic Issue for NPP Implementation in Warm Countries With Extreme Climatic Conditions
Authors: Rodionov Andrei , Mattei Jean-Marie
Published: October 2013

With the nuclear renaissance, many new countries start to consider nuclear energy as a reliable, sustainable energy source and initiated their own programs for a peaceful application.

One of the major issues during the site selection and design of a NPP is the heat sink issue, i.e. availability of reliable and efficient recipient of heat (air or water) in relation with:

• The waste heat removal from the condenser of turbine generator during normal operation of the NPP.

• The assurance of the nuclear fuel residual heat removal and safety important equipment cooling in abnormal and accidental situations.

There are several newcomer countries that have extreme climatic conditions in combination with a shallow sea and absence of an important reliable cold water source inside the country (groundwater, rivers, lakes, etc.). This requires particular consideration of the heat sink issue for the siting survey and design requirements.

The article discusses advantages and disadvantages of possible heat sink solutions, proposes site survey criteria and provides a list of safety and performance issues to be addressed throughout siting and design stages.
Modeling of Dryout in Core Debris Beds of Conical and Cylindrical Geometries
Authors: Takasuo Eveliina , Hovi Ville , Ilvonen Mikko , Holmström Stefan
Published: October 2013

A porous particle bed consisting of core debris may be formed as a result of a core melt accident in a nuclear power plant. The coolability of conical (heap-like) and cylindrical (evenly-distributed) ex-vessel debris beds have been investigated in the COOLOCE experiments at VTT. The experiments have been modeled by using the MEWA severe accident analysis code. The main objectives of the modeling were (1) to validate the simulation results against the experiments by comparing the dryout power density predicted by the code to the experimental results and (2) to evaluate the effect of geometry on the coolability by examining the flow field and the development of dryout in the two geometries. In addition to the MEWA simulations, 3D demonstration calculations of the particle bed dryout process have been performed using the in-house code PORFLO. It was found that the simulation and experimental results are in a relatively good agreement. The results suggest that the coolability of the conical debris bed is poorer than that of the cylindrical bed due to the greater height of the conical configuration.
Dynamic Heat-Exchanger Model for Any Combination of Water and Steam States
Authors: Kim Eunkyeong , Yoshida Takuya , Yashiki Tatsurou
Published: October 2013

The purpose of this study is to propose a dynamic heat transfer model for predicting transient heat recovery steam generator (HRSG) behaviors involving phase changes in heat exchanger tubes. The model deals with any combination of phase states by switching the equations for heat transfer coefficient, specific volume, and friction factor corresponding to their physical characteristics. The model also constrains the change of mass flow calculated by momentum balance to satisfy thermodynamic relationships which are neglected by conventional models. The simulation results show that the proposed model predicts the transient pressure drop, outlet mass flow changes and the reduction in heat transfer coefficient caused by dryout during heating or evaporating processes. In addition, the model improves the accuracy of mass flow transients compared to those obtained by conventional models.
Mechanism of Upward Fuel Discharge During Core Disruptive Accident in Sodium-Cooled Fast Reactors
Authors: Matsuba Ken-ichi , Isozaki Mikio , Kamiyama Kenji , Tobita Yoshiharu
Published: October 2013

The elimination of severe power excursion with significant mechanical-energy release during Core Disruptive Accidents (CDAs) is a key issue for the enhanced safety of Sodium-cooled Fast Reactors (SFRs). In order to prevent the formation of a large-scale molten fuel pool within a reactor core, which is one of factors leading to the severe power excursion during CDAs, Japan Atomic Energy Agency (JAEA) is studying the introduction of Fuel Assembly with Inner Duct Structure (FAIDUS). In the current reference design for FAIDUS, the top end of the inner duct is open whereas the bottom end of the inner duct is closed, and therefore it is expected that the molten fuel will be discharged from a reactor core towards an upper sodium plenum through the inner duct. The objective of the present study is to clarify the fundamental mechanism for upward fuel discharge through the inner duct structure in FAIDUS, and thereby to confirm the effectiveness of FAIDUS.

In the previous paper, the possibility of upward discharge of a high-density melt driven by coolant vapor was confirmed by visual observation in the JAEA’s out-of-pile experiment, in which molten Wood’s metal (density at the room temperature: 8700 kg/m3, melting point: 352 K) simulating the molten fuel was injected into a coolant channel (equivalent inner diameter: 30 mm, total height: 2 m, fluid content: water) simulating the inner duct structure. In this paper, the mechanism of upward discharge of a high-density melt driven by coolant vapor pressure and/or flow in this experiment is discussed in terms of the application to reactor conditions. Through this discussion, the following mechanisms were clarified.

1) Coolant vapor pressure is built up within the coolant channel after the melt injection. The magnitude of the pressure buildup becomes larger with increase of melt-enthalpy-injection rate which is defined by the product of melt-mass-injection rate into the coolant channel and melt specific enthalpy.

2) Following the pressure buildup, the melt is discharged upward being driven by the coolant vapor flow directing towards the top opening end of the coolant channel. The upward discharge mass rate becomes higher with the increase of the magnitude of the pressure buildup and therefore melt-enthalpy-injection rate.

From these experimental knowledge, it was suggested that the coolant pressure buildup could act as one of the driving force for the upward discharge of a high-density melt through the inner duct structure in FAIDUS under reactor conditions with higher melt-enthalpy-injection rate than the current experimental condition.
Beyond Design Basis Seismic Analysis for Atucha II Nuclear Plant
Authors: Zhang Tao , Brust Frederick W. , Wilkowski Gery , Xu Heqin , Betervide Alfredo A. , Mazzantini Oscar
Published: October 2013

The Atucha II nuclear power plant is a pressurized heavy water reactor being constructed in Argentina. Nuclear power plants must be designed to maintain their integrity and performance of safety functions for a bounding set of normal operational events as well as abnormal events that might occur during the lifetime of the plant. Seismic fracture mechanics evaluations for the Atucha II plant showed that even with a seismic event with the amplitudes corresponding to an event with a probability of 10−6 per year, that a double-ended guillotine break (DEGB) was pragmatically impossible due to the incredibly high leakage rates and total loss of make-up water inventory. The critical circumferential through-wall flaw size for this case is 94-percent of the circumference. These analyses are performed by placing cracked-pipe-elements into a complete model of the primary cooling system including the reactor pressure vessel, pumps, and steam generators as summarized in the paper.

This paper summarizes these results and further shows how much higher the applied accelerations would have to be to cause a DEGB for an initial circumferential through-wall crack that was 33 percent (about 120°) around the circumference. This flaw length would also be easily detected by leakage and loss of make-up water inventory. These analyses showed that the applied seismic peak-ground accelerations had to exceed 25 g’s for the case of this through-wall-crack to become a DEGB during a single seismic loading event. This is a factor of 80 times higher than the 10−6 seismic event accelerations, or 240 times higher than the SSE accelerations. This suggests there is a huge safety margin for beyond design basis seismic events and Atucha II plant rupture is pragmatically impossible. These surprising results are discussed and could be potentially applicable to other nuclear power plants as well.
Extreme Events: The New Social Design Basis
Authors: Duffey Romney B.
Published: October 2013

Given the consequences of the events at Fukushima [1], it is not correct to adopt “expected deaths”, or probability of release, or severe-damage consequence as a sole basis for design. The paper examines possible approaches for a successful new Social Design Basis that utilizes an integrated risk basis as a relative measure for all technologies and Extreme Event social costs. As an example, a new relative measure for risk based on total consequences is also derived.
Vapor Bubble Condensation Characteristics of Subcooled Flow Boiling in Vertical Rectangular Channel
Authors: Tan Zhi-wei , Pan Liang-ming
Published: October 2013

In this study, the behavior of condensing single vapor bubble in subcooled boiling flow within two different vertical rectangular channels has been numerically investigated by using the VOF (Volume of Fluid) multiphase flow model. The mass and energy transfer model of bubble condensing process induced from the interfacial heat transfer was proposed to describe the interfacial transportation between the two phases. The results of VOF simulations showed good agreements with previous experimental data in the bubble size variation and lifetime. The bubble lifetime was almost proportional to bubble initial size and prolonged by system pressure. With the increase of subcooling, the bubble lifetime reduces significantly, and the effect of mass flux could be negligible. When bubble size increased, the bubble shape tends to be changed in a large-scale channel. The VOF simulation results of deformation have good agreement with those of Kamei’ experiment and the results of MPS (Moving Particle Semi-implicit) simulation in the large-scale channel. Furthermore, the initial bubble size, subcooling of liquid and system pressure play an important role to influence the bubble deformation behaviors significantly. The bubble could deform sharper with the increase of subcooling and initial diameter, or could breakup when the subcooling and the initial diameter reached a certain value at the last bubble stage. Whereas the trend of bubble deformation would be weaken with the increase of system pressure.
Stability Analysis for Reactor Coolant Pump With Vertical Rotor Supported by Fluid Film Bearings
Authors: Hu Yaoyu , Wang Dezhong , Wang Yujin , Zhang Jige , Gu Weiguo
Published: October 2013

A special dynamics model for a vertical rotor-bearing system with two fly wheels is developed, imitating a sort of reactor coolant pump (RCP) under design. The stability and the process of reaching that condition are of special interest since the rotor is supported vertically other than horizontally. And the classical Jeffcott rotor dynamics model is not quite appropriate because the real pump has its two fly wheels outside the region between the two journal bearings and it has a thrust bearing right under the lower fly wheel. A set of dynamics coefficients are introduced to the rotor dynamics model, taking account of the flow confinement in the reduced gap of the canned pump. The stability characteristic is clearly changed due to the additional dynamics coefficients. Numerical simulation shows that the critical speed of the RCP is lowered. The moment of the thrust bearing, which is ignored customarily in rotor dynamics analysis, also influences the dynamic behavior of the rotor system. The stability of the rotor is enhanced while the thrust bearing has its stiffness and damping values in certain ranges.
Development of a New Heat Transfer Model Near the Quench Front in the Reflooding Phase of a Tight Lattice
Authors: Wu Dan , Yu Hongxing , Yu Junchong , Li Jie , Yu Jiyang
Published: October 2013

Heat transfer characteristics near the quench front in a reflooding process are quite complex. Large amount of vapor are generated, and the rod clad temperature drops rapidly to near saturation state. Until now, heat transfer mechanism in this region has not been well understood yet. Best estimate codes like RELAP5, COBRA-TF tend to treat the heat transfer mode near the quench front as transition boiling. However, when calculating the reflooding phase of tight lattice, these codes always under-predict the quench temperature, and also the slop of the temperature drop is predicted to be less steep than the experimental data. In this paper, a new heat transfer model near the quench front in the reflooding phase of a tight lattice is developed. Instead of transition boiling, transient liquid film evaporation is considered to be the main heat transfer mode in this region. It is supposed that heat released near the quench front is through liquid film evaporation. Through comparisons with experimental data, it can be concluded that the new model can better predict the quench temperature and the temperature drop slop.
Finite Element Analysis of Elastomers Using ANSYS
Authors: Kunnil Manoj , Yamarthi David , Kompally Santhosh K.
Published: October 2013

Elastomeric materials have a capability to withstand large deformations and still be able to fully recover their original dimensions. Natural and synthetic elastomers and their derivatives can reach strains as high as 500–1000%. Engineering materials, such as crystalline metals are classified as linear elastic solids, whereas elastomeric materials are considered as nonlinear elastic solids. Elastomers present a very complicated mechanical behavior that exceed the linear elastic theory and contain large deformations, plastic and viscoelastic properties. Finite element (FE) is a powerful tool to analyze such elastomers.

Design of elastomeric systems in an industrial scenario generally requires (i) reliability and (ii) a minimum cycle time. This paper starts with a review of the hyperelastic theory, followed by a detailed discussion on the process involved in the material characterization of hyperelastic material like DuPont™ Viton® fluoroelastomer and polyacrylic elastomer in industrial application point of view. The paper also discusses guidelines to be followed in the various stages of material characterization such as testing, sampling and finite element simulation. Numerical stability issues associated with elastomeric modeling in finite element context and a set of guidelines to be followed in finite element analysis of elastomers are illustrated through a DuPont™ Viton® fluoroelastomer and polyacrylic elastomer pad vibration isolation systems. The above technique has been applied for designing vibration isolation systems for generators.
Recent Advances in Test vs Finite Element Correlations Through Design for Six Sigma Tools
Authors: Kompally Santhosh K. , Ramanath Vinay , Jeevanandan Karthikeyan , Kunnil Manoj
Published: October 2013

In general, thermal generators have a combination of composites and metals in different assemblies. It is important to note that the material properties and interface stiffnesses change during the assembly process. Added to this change, the complex geometry and assembly procedures result in huge variation in material characterizations. These variabilities triggered a requirement of a unique process for material characterization at both component and assembly levels.

This paper covers the details of a 6-stage DFSS methodology, which involves filling the above-stated gaps by performing mechanical tests at component and sub-assembly levels, followed by series of finite element correlations at various stages of design cycle. This paper emphasizes a DFSS-based probabilistic approach, developed with a built-in validation for evaluating finite element variables to match with assembly tests. This paper also discusses the success of this DFSS-based process in bench marking with two test cases.
A Comparative Study of Direct Soil Model and Soil Spring Model in Common Basemat Analysis of APR1400 Nuclear Power Plant
Authors: Ahn Seong Moon , Lee Dong Seung , Park Jae Wan , Basu Bhaskar
Published: October 2013

It is common practice to analyze the basemat using soil springs to represent the supporting soil media. A uniformly loaded mat foundation analyzed for a uniform spring constant will indicate uniform deflection. However, in reality, the behavior of the basemat is influenced by the flexibility of the supporting soil media. The settlement profiles of the basemat demonstrate smaller displacements at the edge and larger displacements toward the center. The spring constants used in the analysis should also consider stiffness effect of superstructure and the effect due to loading pattern.

For APR1400 Nuclear Power Plant, the mat foundation of the Reactor Containment Building and the Auxiliary Building is a common reinforced concrete mat covering an area approximately 12,210 square meter (110 m × 111 m) with a thickness of 3.05 m in the Auxiliary Building area and various thicknesses from 3.35 m to 10.06 m in the Reactor Containment Building area.

In this study, the soil layers are directly modeled into the structural analysis model using finite/infinite solid elements to represent the soil properties and then the analysis results are compared with the analysis using spring constants by iterative method, and uniform spring constant by conventional method. The results of analysis using the Direct Soil Model are similar to the results calculated by the Iterative Soil Spring Model, but have wide variance with the results calculated by the Uniform Soil Spring Model.
Theoretical Study on the Flow Instability of Supercritical Water in the Parallel Channels
Authors: Su Yali , Feng Jian , Tian Wenxi , Qiu Suizheng , Su Guanghui
Published: October 2013

For the flow of the supercritical water (SCW), the fierce variation of density and specific volume possibly cause flow instability. Based on the structure of parallel channels, mathematical and physical models were established to simulate the flow and heat transfer characteristics of the supercritical water in the parallel channels with semi-implicit scheme and staggered mesh scheme. Flow instability of super-critical water was obtained by using the little perturbation method. Pseudo-subcooling number (NSUB) and pseudo-phase change number (NPCH) are defined based on the property of SCW. The marginal stability boundary (MSB) is obtained with using the NSUB and NPCH. The effects of mass flow rate, inlet temperature and system pressure on the flow instability boundary were also investigated. When increasing the mass flows and system pressure, decreasing the heat flux, the stability in the parallel channels increases. The effect of inlet temperature in the low pseudo-subcooling number region is different from that in high pseudo-subcooling number region.
Improved Regulatory Oversight and Immediate Retrofits for Operating Pressurized Heavy Water Reactors for Beyond Design Basis Accidents
Authors: Nijhawan Sunil
Published: October 2013

The most painful lessons that engineers will learn from reviews of the Fukushima disaster relates to the unacknowledged failure of regulators, designers and utilities in better retrofitting existing reactors in a timely manner to better withstand and mitigate known severe accident related challenges to reactor and containment integrity. Commercial nuclear power reactors have operated for over 50 years, and the first severe accident progression studies and understanding of related phenomena began to mature over 30 years ago. However, some very basic accident prevention, mitigation and management measures have not been required by the regulators and hence not initiated by the utilities.

In public reviews that inevitably followed, severe accidents in power reactors at Three Mile Island, Chernobyl and Fukushima have been often presented as site specific aberrations in design, operations, safety culture and acts of God beyond mortal imagination. If a sincere soul searching, regulatory overhaul and actual, effective, timely and far reaching measures are not taken for operating reactors, the world is bound to witness recurring severe core damage scenarios followed by series of studies that will predictably conclude that such severe accidents can happen only in other jurisdictions and in other designs and that ‘our’ reactors are ‘safe’ and only minor enhancements are required to placate the anxious public or inquiring regulators.

The paper examines the role played by regulators and other Canadian stakeholders in failure to ensure that the so-called residual risk from operating nuclear reactors is minimized in a timely manner. It presents specific examples of known deficiencies in PHWR designs that may degrade a sustained loss of power from a recoverable outcome into significantly more severe consequences. It discusses ideas for potential design retrofits and reiterates the need for more open, concerted and cooperative efforts internationally in accident progression and consequence analyses and supporting experiments.

Taking station blackout as example of an instigator of a severe accident, the paper discusses certain elements of the current PHWR designs that accelerate the onset and progression of core damage and present substantially degraded opportunities for mitigation and control. It demonstrates how in many cases, a sustained loss of power in a PHWR may cause a containment bypass with early and unacceptable off-site consequences. Many issues discussed in this paper are common to other reactor designs and regulatory jurisdictions.
Thermal Hydraulics Behavior Around a Single Heated Rod With Adhesion Products of Sodium-Water Reaction
Authors: Ohta Junki , Kudoh Hideyuki , Sugiyama Ken-ichiro
Published: October 2013

A safety assessment on Sodium-Water reaction in a steam generator (SG) of the SFBR is also one of significant issue. If a heat transfer tube is damaged by some reasons, the high pressure water or steam would leak into the sodium side as a high-temperature and high-corrosive sound-speed jet, and it might give rise to the ruptures of surrounding heat transfer tubes caused by overheating. On the suspicion of secondary failure of tubes, we need to understand the detail of heat transfer characterics on its surface. To grasp these phenomena mechanically and systematically, we studied in the case of low-speed jet. In our previous study, an experiment on the heat transfer between the tube and the jet was carried out by using a horizontal heated-rod and injecting argon gas-steam mixture (reaction gas) under the condition of low sodium temperature and low-speed gas jet [ICONE18-29388]. The result showed that the heat transfer between the rod and the reaction jet was decreased by the reaction products, adhered on the rod surface. But from the reasons of the sodium purity and the rod design, the correct adhesion condition was not declared. In the present study, we got over these problems. A simple method for purification of the used sodium which includes the reaction products was first established. Then the local heat transfer characteristics of the reaction jet were investigated with the more points of measuring temperature on the rod surface than the previous study by using the purified sodium. As soon as injecting the gas mixture into the sodium pool, the heat transfer to the rod surface was reproducibly decreased by the adhesion of the reaction products on it. In the first 10 second from the injection of the gas mixture, the local heat transfer at the lateral side of the rod was decreased obviously. About 10s later from the injection of the gas mixture, the heat transfer at the backward side also began to decrease. At the front side of the rod, the heat transfer was less affected by the chemical reaction. Thus, it was confirmed that the local heat transfer is closely influenced by the position of the adhesion products, and its products tend to adhere on the lateral and backward sides.
Validation Results of TRACG05 Against BWR and ABWR Start-Up Test Data
Authors: Ochi Hitoshi , Takii Taichi , Kudo Yoshiro , Kagiyama Rui , Yamada Takeshi , Isono Kimiko , Nishijima Mitsuko , Kaneko Junichi
Published: October 2013

Studies have been made on application of TRACG05, which is the latest version of TRACG code, to Best Estimate Plus Uncertainty (BEPU) approach for Anticipated Operational Occurrences (AOO) in BWR and ABWR. This paper presents validation results of TRACG05 against BWR and ABWR start-up test data.

TRACG05 analyses of BWR and ABWR start-up tests, such as generator load rejection (pressurization event) and recirculation pump trip (core flow decrease event), were performed using standard plant nodalization and nominal inputs.

In addition to the nominal cases, sensitivity studies and statistical analyses were conducted. The sensitivity studies include the sensitivities to TRACG05 model uncertainties, which have relatively high importance being tabulated in a Phenomena Identification and Ranking Table (PIRT), and nodalizaton. The sensitivity analyses include the random variation of TRACG05 model uncertainties, which have High ranking in PIRT, and initial core power. These uncertainties were preliminarily evaluated from comparisons of calculated results to separate effects test data, integral test data and component test data.

TRACG05 calculated results of the nominal cases agree to the start-up test data with reasonable accuracy. For statistical analyses, the standard deviations of the key parameters related to the safety parameters (critical power ratio, etc) for AOO were calculated at each time step. The test data basically fall inside the 2 sigma bands. It indicates that the differences between calculated results and test data are within the uncertainties of TRACG05 model and initial core power that are taken into account for safety analyses in BEPU process.

It is concluded that the TRACG05 code using the standard nodalization, inputs and model uncertainties is validated against the full-scale plant data and is capable of predicting the AOO transients.
OECD/NEA ROSA Project Experiment on Steam Condensation in PWR Horizontal Legs During Large-Break LOCA
Authors: Takeda Takeshi , Ohtsu Iwao , Nakamura Hideo
Published: October 2013

Separate-effect experiment simulating steam condensation on emergency core cooling system (ECCS) water in PWR cold legs during reflood phase of large-break loss-of-coolant accident (LBLOCA) was conducted in OECD/NEA ROSA Project using the Large Scale Test Facility (LSTF). A test section was furnished in the downstream of the LSTF break unit horizontally attached to the cold leg. The boundary test conditions were defined based on PWR LBLOCA analysis by RELAP5/MOD3.2.1.2 code considering typical Japanese safety analysis conditions. Significant condensation of steam appeared in a short distance from the simulated ECCS injection point, and the steam temperature in the test section decreased immediately after the initiation of the ECCS water injection. Fluid temperature distribution at 50 mm downstream from the ECCS injection point was significantly non-uniform, but became almost uniform in less than 350 mm. Total steam condensation rate estimated from the difference between steam flow rates at the test section inlet and outlet was in proportion to the simulated ECCS water mass flux until the complete condensation of steam. Inlet steam was completely condensed if inlet steam mass flux is less than 195 kg/(m2s) when the simulated ECCS water mass flux is 148 kg/(m2s); equivalent to full high-pressure injection with single-failure low-pressure injection conditions. Clear images of high-speed video camera were obtained on droplet behaviors through the viewer at 200 mm downstream from the ECCS injection point, especially for annular mist flow. The number of flowing droplets decreased with increasing distance from the ECCS injection point.
Validation of TRACG05 for Application to BWR AOO
Authors: Matsui Asuka , Tamitani Masashi , Kudo Yoshiro , Takano Sho , Iwamoto Tatsuya , Nishijima Mitsuko , Kaneko Junichi , Ochi Hitoshi , Takii Taichi , Soneda Hideo
Published: October 2013

TRACG code, coupling a three-dimensional neutron kinetics model for the reactor core with thermal-hydraulics based on two-fluid conservation equations, is a best-estimate (BE) code for BWRs to realistically simulate their transient and accidental behaviors. TRACG05 is the latest version and was originally developed to analyze Reactivity Initiated Accident (RIA). TRACG05 incorporates the same neutronics model of the latest core simulator with a three-group analytic-polynomial nodal expansion method.

In addition to application to RIA safety analyses, TRACG05 has been planned to apply to safety analyses for Anticipated Operational Occurrences (AOOs) in BWRs by using a Best Estimate Plus Uncertainty (BEPU) methodology. To apply BEPU with TRACG05 to BWR AOOs, validations must be performed to evaluate the uncertainties of models relevant to important phenomena by comparing with appropriate test results for BWR AOOs. At first, a PIRT (Phenomena Identification and Ranking Table) was developed for each event scenario in AOOs to identify relevant physical processes and to determine their relative importance. According to the PIRT, an assessment matrix was established for separate effects tests (SETs), component effects tests (CETs), integral effects tests (IETs), and integral BWR plant start-up tests. The assessment matrix related the important phenomena to the test database, which was confirmed that all the important phenomena were covered by all tests specified in the matrix. According to the assessment matrix, comparison analyses have been specified to perform systematic and comprehensive validations of TRACG05 applicability to AOOs. The comparison analyses were done as the integrated code system with the up-stream reactor core design codes, therefore higher quality was enabled to evaluate the safety parameters. As the result, the uncertainties of important models in TRACG05 were determined so as to enable BEPU approaches for AOO safety issues.

Here, as a SET, comparisons between TRACG05 and experimental data of void fraction in a bundle simulating an actual fuel bundle, which is one of the most important models in the application of TRACG05 to AOO analyses are shown. In addition, as pressurization event in AOOs, comparisons between TRACG05 and experimental data of Peach Bottom 2 Turbine Trip Test, which is one of integral tests for a BWR plant, are shown. This is the only test showing large neutron flux increase and strong coupling of neutron kinetics and thermal-hydraulics in the core due to void and Doppler feedbacks. Furthermore, a sensitivity analysis regarding a delay time of control rod (CR) insertion initiation which was the most sensitive uncertainty to the results is also shown.
Semi-Analytical Study of Aerosol Washout by Rain
Authors: Lemaitre Pascal , Nuboer Amandine , Querel Arnaud , Depuydt Guillaume , Porcheron Emmanuel
Published: October 2013

The accidents of Chernobyl and Fukushima have shown the necessity to better understand all the mechanisms implied in the scavenging of aerosol particles released to the atmosphere during a nuclear accident.

Among all the phenomena involved in the deposition of aerosol particles, we focus here on the aerosol particles scavenging by the raindrops below the clouds, also called washout (as opposed to the rainout, which concerns scavenging inside the clouds). The strategy of IRSN to enhance the knowledge and the modelling of any mechanism involved in the washout of aerosol particles by rain spans from environmental studies, to analytical ones.

The semi-analytical approach chosen here is halfway between these two modes of reasoning. A companion paper is also submitted to the conference to present the microphysical approach chosen at IRSN.

In order to perform this study, aerosol particles were dispersed in the TOSQAN chamber, which is a large cylindrical enclosure (4.8 m height with 1.5 m internal diameter). The aerosol particles once dispersed, synthetic rains of different kinds (from stratiform to convective rains) can be activated. Finally, the instantaneous spectral scavenging coefficients are determined from the spectral decrease of aerosol particles concentration in the chamber as a function of time.

In order to be able to produce synthetic rains representative of any tropospheric events, a special generator has been designed; it is based on a vibro-rotative disk. This generator is able to produce monodispersed rains at the top of the TOSQAN chamber with rainfall rates from 7 to 15 mm/h and drops diameters from 0.5 to 2.5 mm injected at velocities close to their terminal one.

During these tests, the spectral aerosol concentration is measured in line with the help of a Welas granulometer. This instrument is based on white light scattering.

The results of these experiments highlight the influence of “meteorological” conditions inside the chamber on the washout of the chamber atmosphere, especially when the relative humidity is reaching saturation.
Local Void Fraction in Triangular Arranged Rod Bundle
Authors: Kashinsky O. N. , Lobanov P. D. , Kurdyumov A. S. , Pribaturin N. A. , Volkov S. E.
Published: October 2013

An experimental study of gas-liquid flow in a vertical bundle of 37 rods in a triangular arrangement was performed. An interaction of gas phase with a spacing grid was considered. Gas was injected through a 1.2 mm capillary tube inserted into one of the sub channels of the central rod. Void fraction measurements were performed by a conductivity probe mounted on the central rod Measurements were performed in the flow region just downstream the spacing grid. The processing of the conductivity probe records allowed to obtain three-dimensional distribution of the gas phase around the central rod of the bundle. Measurements performed demonstrated a non-uniform distribution of the gas phase over the circumference of the central rod in the region close to the grid. The data obtained allow to estimate the rate of gas transfer between the sub channels.
Study on Control Method of Natural Circulation by Injection of Helium Gas
Authors: Nomura Masashi , Takeda Tetsuaki , Funatani Shumpei , Shimura Takuya
Published: October 2013

This study is to investigate a control method of natural circulation of air by injection of helium gas. A depressurization accident is one of the design-basis accidents of a Very High Temperature Reactor (VHTR). When the primary pipe rupture accident occurs in the VHTR, air is expected to enter into the reactor pressure vessel from the breach and oxidize in-core graphite structures. Finally, it seems to be probable that the natural circulation flow of air in the reactor pressure vessel produce continuously. In order to predict or analyze the air ingress phenomena during the depressurization accident of the VHTR, therefore, it is important to develop the method for prevention of air ingress during the accident.

The experiment has been carried out regarding natural circulation using a circular tube consisting of the loop type or the reverse U-shaped type. The vertical channel consists of the one side heated tube and the other side cooled tube. The experimental results were obtained as follows. When the temperature difference between the vertical tubes was kept at 52K, the velocity of natural circulation flow became about 12cm/s. During this steady state, a small amount of helium injected to the channel. Then, the flow velocity of natural circulation suddenly decreased. The volume of injected helium is about 3% of the total volume of the channel. The velocity became around zero. After 1500 seconds elapsed, the natural circulation suddenly produced again. The experimental results show that the natural circulation flow of air can be controlled by the method of helium gas injection. This paper also discusses an overview of the method for the prevention of air ingress during the primary pipes rupture accident.
Numerical Flow Analysis of Electromagnetic Fluid Using GSMAC Finite-Element Method
Authors: Yang Kun , Saitoh Yoshitaka , Kawai Hideki , Tsuzuki Nobuyoshi , Kikura Hiroshige , Hashimoto Gaku
Published: October 2013

Numerical calculations for the cavity flow which simulates Joule-heated glass furnace were executed using GSMAC finite element method (GSMAC-FEM). The cavity consists of two main electrode plates on facing sides, two auxiliary electrode at the bottom and constant temperature wall on the top working as a heat sink. Magnetic field effect in Joule-heated glass furnace was verified by flow-electric-magnetic coupling analysis. When the calculation coupled with the flow and electric field was executed, unsteady down flows occurred from the top surface of the cavity. However, from the result of calculation coupled with flow, electric field and magnetic field, a large up flow appeared in the center of the cavity and permanently existed. Consequently, flow behavior including magnetic field was completely different from that excluding magnetic field effect. The results suggest that numerical analysis including magnetic field effect is recommended for accurate understanding of flow behavior in the Joule-heated glass furnaces. On the other hand, the data of velocity field, temperature field and magnetic field are also got by numerical simulation and compared with the result of Japan Atomic Energy Agency.
Analysis of the MSFR Core Neutronics Adopting Different Neutron Transport Models
Authors: Fiorina Carlo , Aufiero Manuele , Cammi Antonio , Guerrieri Claudia , Krepel Jiri , Luzzi Lelio , Mikityuk Konstantin , Ricotti Marco Enrico
Published: October 2013

The Molten Salt Fast Reactor (MSFR) has recently been chosen as the reference circulating-fuel Molten Salt Reactor design in the framework of the Generation IV International Forum. Different from most of the Molten Salt Reactor designs developed and proposed in the past, the MSFR is featured by a fast neutron spectrum. On one hand, this choice leads to a simplified core (no graphite is present) and to improved characteristics in terms of breeding and/or transuranic burning. On the other hand, the removal of graphite leads to a significantly different behavior in terms of both core neutronics and dynamics. In view of the ongoing development of dedicated tools for the MSFR transient simulation, it is useful: 1) to accurately determine the core feedback coefficients, for lumped models and to get an insight into the reactor safety features; and 2) to analyze the degree of approximation related to the use of deterministic multi-group transport and diffusion approaches, in case of multi-dimensional models. The present paper compares the results obtained by means of the deterministic code ERANOS 2.2N with those obtained through the Monte Carlo code PSG2/SERPENT, for different fuel cycle strategies. In particular, the comparison is based on the capability to reproduce nominal reactivity, feedback coefficients, spectra and flux profiles. In the light of these results, the capability of a simple few-group diffusion model to deal with the MSFR neutronics is preliminarily assessed. Such model has been set up by means of the general-purpose finite-element COMSOL Multiphysics software. It is of interest for a subsequent development of multi-physics models able to reproduce the peculiar MSFR transient behavior.