bluuebunny/crossref_metadata_2025 huggingface dataset:
A Three-Dimensional Analysis of the Local Stresses and Strains at the Pellet Ridges in a Horizontal Nuclear Fuel Element
Authors: Gamble Kyle A. L. , Williams Anthony F. , Chan Paul K.
Published: November 2014

A three-dimensional finite element model is being developed for a quarter fuel element, which is equivalent to a full fuel element using symmetry. The model uses the Multiphysics Object-Oriented Simulation Environment (MOOSE) framework developed at Idaho National Laboratory. The model facilitates an in-depth investigation into a variety of deformation phenomena for a horizontal nuclear fuel element including bowing, sagging, and stresses and strains.

This paper presents a preliminary analysis of the local stresses and strains of the sheath (clad) at the pellet-to-pellet interfaces for low, normal and high linear powers. During irradiation the fuel pellets thermally expand and take on an hourglass shape. The hourglassing behaviour leads to higher local stresses and strains in the sheath at the locations of the pellet-to-pellet interfaces. The purpose of this work is to quantify these stresses and strains for varying linear powers, and to illustrate the effect that the material model chosen for the cladding has on the results. Preliminary results are presented for two sheath types: elastic, and elastic including diffusional creep. These models are benchmarked against a validated industry code called ELESTRES. The results indicate that the predicted sheath hoop strain is about half of what is determined by ELESTRES in both the elastic and elastic-creep cases. This highlights the requirement of a pellet cracking model in three-dimensional simulations. The elastic-creep model predicts less stress within the sheath than the elastic model as expected.
Effect of Solidification on Molten Material Jet Behavior in Coolant
Authors: Iwasawa Yuzuru , Abe Yutaka , Kaneko Akiko , Saito Shimpei , Nariai Hideki , Matsuo Eiji , Sakaba Hiroshi , Ebihara Ken-ichi , Koyama Kazuya , Itoh Kazuhiro
Published: November 2014

For the safety design of a Fast Breeder Reactor (FBR), if a Core Disruptive Accident (CDA) occurred hypothetically, it is required to suppress the rapid energy release due to a prompt criticality. Even if the rapid energy release does not occur, there is a possibility that a large amount of fuel melts. Therefore it is important to achieve Post Accident Heat Removal (PAHR). In order to achieve PAHR, it is strongly required that the molten material which is released from a core region gets cool and solidifies in the sodium coolant in a reactor vessel by breaking up. It is considered that the molten fuel is injected into the coolant like a jet. Furthermore, in the actual FBR, the interfacial temperature between the molten fuel jet and the coolant is considered to be lower than the melting point of the molten material. Thus for PAHR in CDA, it is important to understand the interaction between the jet and the coolant in such a condition and to estimate the molten jet behavior quantitatively.

In order to estimate quantitatively the effects of the solidification on the molten jet behavior, we carried out the experiment in which a simulant material was injected into a simulant coolant. In the experiment, we used low melting point alloy (Bi -Sn) and water as the simulant molten material and the simulant coolant respectively. In the experiments, we chose the temperature range including the condition that the interfacial temperature was lower than the melting point of the molten material. The jet breakup and the fragmentation behavior of the molten material jet were observed with a high speed video camera. Then the jet breakup length is estimated form the results. We changed the initial interfacial temperature condition by adjusting temperature of the molten material and the coolant. We also changed the jet velocity by adjusting the height of the nozzle tip from the water surface.

From the experiment, we found that the jet breakup behavior depends greatly on the interfacial temperature and the injection velocity and that the solidification of a molten material jet and the growth of unstable jet surface, which results from the relative velocity of the jet to the coolant, are in a competitive relation for the jet breakup. We also found that when the molten material jet breaks up into fragments, the breakup length is independent of the initial interfacial temperature and the initial injection velocity.
Influence of Two-Phase Flow Structure on Operating Characteristics of the Supersonic Steam Injector
Authors: Sato Kyosuke , Abe Yutaka , Kaneko Akiko , Mori Michitsugu
Published: November 2014

Steam injector (SI) is a passive jet pump with a converging-diverging structure. A SI operates without an electrical power source by direct contact condensation of a supersonic steam flow and a subcooled water jet in a mixing section. Furthermore, there are advantages that a SI has high heat-transfer performance and discharges water at high pressure. Therefore a SI is expected to apply to the safety system that is able to condense steam efficiently and inject water into a core reactor when severe-accident occurs in a nuclear power plant. However it is not cleared about the operating range of a SI which is taken account of the discharged flow structure.

The objective of the present study is to reveal the influence of two-phase flow behavior on operating limits of a SI. The test section of the SI is made by transparent material to observe flow structure in it. The pressure distributions along the flow direction and the discharge pressure were measured by changing the inlet steam pressure and a load on exit of the SI. At the same time, the discharged flow at the diffuser was observed with a high speed camera.

From the observation results, it was confirmed that a boundary which the flow structure changed in the discharged flow. The area from the throat to this boundary is seemed a two-phase region that steam which has not been condensed completely in the mixing nozzle remains. It was found that this boundary moved to upstream as the load on the exit increased and significant pressure rise occurred at the position the boundary reached. Additionally white propagations toward downstream were observed. This propagation is seemed a pressure wave propagation. The velocity of the propagation was estimated by image processing. Assuming a pressure wave propagates at sonic speed, void fraction at the discharged flow was estimated by existing homogenized model.

From the above, the influence of two-phase flow in discharged flow on the operating limits of the SI is discussed.
The Historical Development of French Nuclear Regulatory Organizations and the Major Factors in Their Changes
Authors: Kimura Kenji , Kimura Hiroshi
Published: November 2014

At the beginning of the French nuclear history, CEA controlled all the nuclear development including its safety. In 1950s, this situation was changed by the participation of many industrial companies, which means industrialization of nuclear technology. This change became clearer when they adopted PWR in 1970. And so they needed and established a system to make safety regulation by putting the multiple actors’ opinions together.

After the accident of Chernobyl, antinuclear public opinion has increased. And in 1997, The Greens obtained the post of the Minister of Ecology. These facts required more independent and transparent regulatory system, so in 1998, a report which proposed to establish a new regulatory organization was published. On the basis of this report, they founded ASN in 2006.

From this French history, we can say that as the relationship between nuclear technology and the society changed, the regulatory body also changed to meet the necessity.
Analysis of Melt Behavior in a Cold Tube by MPS Method
Authors: Chen Ronghua , Oka Yoshiaki , Li Gen , Matsuura Takashi
Published: November 2014

During the severe accident, the core melt could leak out of the reactor pressure vessel through the broken instrument tubes of the RPV lower head. Thus, the melt freezing behavior in an instrument tube is a key factor that concerns the failure of the reactor pressure boundary. In this study, the Moving Particle Semi-implicit (MPS) method was adopted to analyze the melt penetration and solidification behaviors in a tube. The change of melt viscosity with temperature was taken into account in the present MPS method. The influence of surface tension on the melt penetration behavior was also considered. The numerical results had been compared with the upward melt injection experiments. The comparative results showed that the melt penetration lengths were in good agreement with the experiments. The typical melt freezing behaviors were successfully reproduced by MPS method. The crust formed on the surface of the tube increased the melt flowing resistance. The melt velocity also decreased due to the increase of its viscosity. The present results indicate that MPS method has the capacity to analyze the melt freezing behavior in the lower head penetration tubes.
Interim Storage of Spent Nuclear Fuel in the UAE Nuclear Power Plants
Authors: Al Saadi Sara , Yi Yongsun
Published: November 2014

The interim storage options of spent nuclear fuel (SNF) in Barakah nuclear power plants in the UAE were studied in terms of costs and technical issues. Considering the capacity of the spent fuel pools in Barakah nuclear power plants, two scenarios for the interim SNF storage were established. Scenario 1 is ‘minimum use of spent fuel pool’ that SNF will be transferred to dry storage facilities as soon as SNF has been sufficiently cooled down in a pool for the dry storage. Scenario 2 is defined as ‘maximum use of spent fuel pool’ that SNF will be stored in a pool as long as possible till the amount of stored SNF in the pool reaches the capacity of the pools and, then, to be moved for dry storage. For these two scenarios, cost analysis was performed in terms of net present values (NPVs) and levelized unit costs (LUCs). The life cycle of the dry storage was divided into three phases: i) preconstruction phase, ii) construction phase and iii) operation phase. By using data available in literature for the three phases, the total costs were calculated and compared between the two scenarios. For a basic analysis, using the discount rate of 5 % and the required cooling period (Tcool) of 7 years before the SNF transfer to dry storage, LUCs were 184 and 192 $/kg HM for Scenarios 1 and 2, respectively, which were comparable to other analysis results in literature. Then, additional calculations were performed using two different values of the discount rate and the cooling period, respectively. The NPV 1 for Scenario 1 ranges between 175.7 and 413.5 million 2014 $, depending on the discount rate and the cooling period, Tcool. For Scenario 2, NPVs of 85.2 and 237.3 million 2014 $ were obtained for discount rates of 7% and 3%, respectively. The comparisons of the NPVs between the two scenarios showed that Scenario 1 would cost 1.5 to 2.7 times Scenario 2. Technical issues of a dry storage system associated with the site specific conditions in the UAE were also studied. The higher ambient air temperature in the UAE than other countries could affect the cooling capacity of the dry storage by natural convection, which will affect the required cooling period (Tcool) in the spent fuel pool. Also, the harsh environments could have detrimental effects on the integrity of metallic components by degradation phenomena such as pitting, stress corrosion cracking (SCC). This discussion implies that the two aspects related to the harsh environment in the UAE should be studied as early as possible. The environmental and safety impacts associated with the dry storage of SNF were discussed. According to published reports in the USA it seems that there will be no significant environmental impacts of the dry storage for 60 years. However, it is judged that future studies should address the impacts for longer time period than 60 years.
SCDAP Model Improvement With QUENCH-06 Analysis
Authors: Madokoro Hiroshi , Okamoto Koji , Ishiwatari Yuki
Published: November 2014

One of the important severe accident management measures in a Light Water Reactor (LWR) is water injection to the reactor core. Reflooding of the uncovered reactor core is essential to prevent total core degradation. The series of QUENCH tests have been conducted to acquire knowledge of the reflooding. A number of analyses on QUENCH tests have also been done by different computer codes for code validation and improvements.

In this study, the modeling of the QUENCH-06 experiment was performed with RELAP/SCDAPSIM computer code. The input deck was modified and the SCDAP model was improved in order to represent the experimental facility more precisely. The uncertainties regarding the electrical resistance distribution in the heater rod system and the thermal properties of the shroud insulator were assessed, respectively.

The main purpose of this study is to identify the models to be improved. The rather good agreement between the calculation results and the measurement was acquired than the past studies [1]. In addition, more accurate modeling of the electrical resistance and the thermal properties of shroud insulator was done and their importance was indicated. The temperature profile and oxide thickness still showed similar tendencies with the original case. Further improvements are required mainly in the heat transfer model and the oxidation model in the SCDAP code.
The Influence of Two-Phase Flow on Pore Clogging by Fine Particle Settlement During Ex-Vessel Debris Bed Formation in Severe Accident
Authors: Kim Eunho , Park Jin Ho , Kim Moo Hwan , Park Hyun Sun
Published: November 2014

In severe accident, the assurance of long term coolability of the corium melt is one of the key issues of the severe accident management (SAM) for the mitigation and termination of the accident progression. The Korean NPPs adopts the SAM strategy that water floods into cavity levels up to the upper part of the reactor vessel to provide the external reactor vessel cooling for the in-vessel retention of corium (IVR-ERVC). However, if the IVR-ERVC strategy fails and the corium relocation onto the cavity occurs, the long term coolability of the ex-vessel corium on the cavity floor should be assured. However, this scenario takes a little account of the effect of the two-phase flow from the corium particles and pre-settled debris bed on the cavity floor. Recently there begins research concerning the effect of two-phase flow on the bed formation, like the particle relocation by self-leveling phenomenon reporting that the cavity pool convection by the two phase flow from the debris bed largely distorts the settling trajectories of the debris by their own sizes. In the perspective, the present study aims to investigate the pore clogging by the fine debris particles in two-phase conditions at the DAVINCI facility, POSTECH, Korea. In the experiments, pre-defined debris particles with various diameters and size distributions are plunged into a water pool in where two-phase flow regime is simulated by the gas injected from the bottom of the test section. The configuration of the resultant debris bed is analyzed to identify the role of two-phase flow on the settlement of fine particles and clogging of debris bed pores. In the test, it was observed that the resultant bed shape and particle distribution was significantly affected by the two-phase bubble behavior during the particle settlement. It is also showed that in the two-phase test the fine particles are more settled in the outer region compared to the pre-settled debris bed region, illustrating the potential positive effect against the coolability degradation due to the fine particle pore clogging phenomena.
An Experimental Investigation of Steam Condensation in the Presence of Non-Condensable Gases for Nuclear Plant Containment
Authors: Su Jiqiang , Sun Zhongning , Zhou Yanmin , Yan Chaoxing , Xu Guangzhan , Wang Yang
Published: November 2014

The condensation heat transfer occurring in containment atmospheres during the loss of coolant accident (LOCA), is one of the most important areas in research related to the safety of nuclear reactors. In the advanced Generation III and III+ nuclear reactors, decay heat is removed by passive containment cooling system (PCCS). For the system, the study of condensation of steam in the presence of non-condensable gases is prior to be investigated because when LOCA happens steam flashes into the containment which contains air and other non-condensable gases (helium, argon, etc.). An experimental investigation has been conducted to evaluate the steam heat removal capacity over a vertical tube external surface with air. Condensation heat transfer coefficients have been obtained under the total pressure ranging from 0.4MPa to 0.6MPa, the wall subcooling ranging from 13 to 25°C and air mass fraction ranging from 0.07 to 0.52. The influence of the wall subcooling on the steam condensation heat transfer with the fixed pressure and air mass fraction have been researched. The effect of wall subcooling on condensation heat transfer coefficient with air is negative. The developed empirical correlation for the heat transfer coefficient covered all data points within 15%.
Two-Phase Slug Flow in a Narrow Rectangular Channel Under Inclined Conditions
Authors: Wang Yang , Yan Changqi , Sun Licheng , Yan Chaoxing , Zhou Yanmin , Su Jiqiang
Published: November 2014

The characteristics of two-phase slug flow in a narrow rectangular channel with cross section of 3.25 mm × 43 mm under vertical and inclined conditions are investigated using a high speed video camera system. It is found that the velocity of Taylor bubble in vertical continuous slug flow could be well predicted by the Nicklin et al. (1962) correlation, in which C0 is given by the correlation of Ishii (1977), and the drift velocity given by the correlation of Sadatomi et al. (1982) or Clanet et al. (2004). For low two-phase superficial velocity (FrTP ≤ 3.5), the Taylor bubble velocities gradually increase with the increasing in inclination angles and almost approximate the maximum value for θ = 30°. For high two-phase superficial velocity (FrTP > 3.5), the influence of the inclination angles on the Taylor bubble velocity is insignificant, and the bubble velocity under vertical condition is slight lower than those under inclined conditions. For the inclined cases, the nose of Taylor bubble is deviated from the centerline and its position is the function of the two-phase superficial velocity as well as the inclination angle.
Uses of Supercritical Fluids and Their Characteristics Within Deteriorated Heat Transfer Region
Authors: Gupta Sahil , Saltanov Eugene , Pioro Igor
Published: November 2014

Taking into account the expected increase in global energy demands and increasing climate change issues there is a pressing need to develop new environmentally sustainable energy systems. Nuclear energy will play a big part in being part of the energy mix since it offers a relatively clean, safe and reliable source of energy. However, opportunities for building new generation nuclear systems will depend on their economic and safety attractiveness as well as flexibility in design to adapt to different countries and situational needs. Keeping these objectives in mind, a framework for international cooperation was set forth in a charter of Generation IV International Forum (GIF, 2002). The main design goals for the Generation IV nuclear system concept were to improve economic gains, enhance safety, extend sustainability, and strengthen proliferation resistance.

To achieve high thermal efficiencies of up to 45–50%, the use of SuperCritical Fluids (SCFs) as working fluids in heat transfer cycles is proposed. An important step towards development of SCF applications in novel Generation IV Nuclear Power Plant (NPP) designs and in other industries; is to understand the thermal hydraulic behavior and prediction of Heat Transfer Coefficients (HTCs) at supercritical (SC) conditions.

Heat transfer under SC turbulent conditions is generally very complex and is extremely sensitive to the test geometry and operational flow parameters. Detailed sets of experiments have been conducted around the world in tubes and other geometries with SCFs to study the basics of heat transfer. By variation of operational parameters and test geometries, fundamental heat transfer data sets are collected that will help in our understanding of SC heat transfer phenomena.

Applications for SCFs are not limited to power industry; recent advancements have indicated the use of SCFs in a much wider range of applications due to its unique and attractive heat transfer characteristics. In this paper, proposed uses of SCFs are presented and basic concept of DHT within SCFs is analyzed.
Flow Instability Inside Heat Transfer Tubes of Moisture Separator Reheater
Authors: Manabe Jun
Published: November 2014

This paper presents some consideration and analysis on the flow instability inside the tubes of the moisture separator reheater (MSR), which is a combined equipment of the mist separator and the in-tube side condensation multi-tube type heat exchanger for superheating the exhaust from the high pressure turbine to the low pressure turbine. The condensed water in the tubes will be exhausted by the friction of excess steam vented through the tubes, and with such design, the flow in the tubes will be in an annular gas-liquid two-phase regime, ensuring the steady drainage of condensed water without subcool and the associated temperature oscillation at the tube end. The mechanism of the unstable two-phase flow dynamics in the tubes is discussed. The analysis and the results presented in this paper will be of interests to the related researchers and engineering applications.
Safety Analyses for Pb-Bi-Cooled Direct Contact Boiling Water Fast Reactor (PBWFR)
Authors: Tian Yonghong , Tian Wenxi , Zhang Dalin , Su Guanghui , Qiu Suizheng
Published: November 2014

Pb-Bi-cooled direct contact boiling water fast reactor (PBWFR) can produce steam from the direct contact of feedwater and lead bismuth eutectic (LBE) in the chimney of 3 m high, which eliminates the bulky and flimsy steam generators. Moreover, as the coolant LBE is driven by the buoyancy of steam bubbles, the primary pump is not necessary in the reactor. The conceptual design makes the reactor simple, compact and economical. Owing to the large thermal expansion coefficient of LBE and good performance of steam lift pump, the reactor is expected to have excellent passive safety. A computer code was developed to investigate the thermal-hydraulic behaviors and safety performance of PBWFR. In order to test and verify its safety performance, the following accidents without scram were simulated with the code: Unprotected rod run-out transient over power (UTOP), unprotected loss of heat sink (ULOHS), simultaneous UTOP and ULOHS accident. The results show that PBWFR has very good inherent safety due to the satisfactory neutron and thermal-physical properties of LBE. Cladding materials turn to be the key factor to restrict its safety performance and UTOP is more risky for PBWFR than the others.
Character of Liquid Metal Embrittlement of the Steel T91 in Contact With Lead-Bismuth Eutectic
Authors: Hojna Anna , Di Gabriele Fosca , Klecka Jakub
Published: November 2014

This work summarizes results of the work carried out in CVR (Centrum výzkumu Rez) on the evaluation of the susceptibility to LME (Liquid Metal Ebrittlement) of the ferritic/martensitic steel T91 in contact with LBE (Lead-Bismuth Eutectic) at 300 and 350°C. The work, carried out in the past few years, through tensile and fracture toughness tests, revealed a marked sensitivity to embrittlement for the T91 in contact with LBE. Tests were carried out in well defined conditions and according to ASTM standards. It was observed that the LBE decreased the fracture toughness, JIC, by more than 30%, compared to the value in air. The results are discussed based on examinations of the fracture surface evidencing LME occurrence. The fracture toughness measured in LBE is proposed to be the appropriate value for fracture resistance evaluation, applied to extension of a pre-existing LME crack.
Safety Evaluation of KAERI Pyroprocess Facilities
Authors: You G. S. , Choung W. M. , Ku J. H. , Moon S. I. , Kim H. D.
Published: November 2014

KAERI (Korea Atomic Energy Research Institute) has been studing the pyroprocess since 1997. For a demonstration of the pyroprocess, KAERI developed two facilities, the ACPF (Advanced spent fuel Conditioning Process Facility) and PRIDE (PyRoprocess-Integrated inactive DEmonstration facility), by 2012. From 2013, KAERI has been studying a pre-conceptual design of the ESPF (Engineering-Scale Pyroprocess Facility).

In this paper, the safety evaluation methods and results for these three facilities are described.
Formulation and Reliability Feature Analysis of Analog, Digital and Hybrid I&C Architectures for Research Reactors
Authors: Ur Rahman Khalil , Heo Gyunyoung , Kim Man Cheol , Zubair Muhammad
Published: November 2014

Optimized design of Instrumentation and Control (I&C), having rationally good reliability features like high availability, low spurious trip rate and low cost, is very important for safety and economical use of nuclear reactors. The demand for optimization becomes more significant in case of small size educational or research reactor due to less margin of cost. The aim of study, in this article, is to develop analog, digital and hybrid architectures for reactor protection I&C system and identification of suitable one for research reactor. The reliability feature analysis of proposed architectures has been performed in three parts (i) analytical analysis for redundancy (ii) Bayesian network model for estimation channel failure λdc and (iii) unavailability or Probability of Failure on Demand (PFD) for each I&C system architecture. It is found here that 2oo3 (2 out of 3) architecture is highly suitable for research reactor as it has low nuisance process trip rate, rational availability (reasonable high) and low probable to CCF failure, if compared with other ones. Results have shown that hybrid configuration can perform better than analog, digital architectures. This research would be helpful for the designers to consider appropriate model with defined cost and availability for their design.
Generic Experimental Investigations of Thermohydraulic Instabilities With Void Fraction Measurement at Natural Circulation Test Facility Geneva
Authors: Cloppenborg Tim , Schuster Christoph , Hurtado Antonio
Published: November 2014

Passive systems like natural circulation (NC) loops can offer reliable and cost efficient alternatives to common active systems for decay heat removal in nuclear power plants. During the transition between stable single and stable two phase flows, instabilities e. g. flashing and geysering may occur in the riser due to low system pressure and saturation temperature conditions. These instabilities may cause severe stress to the system components. This paper presented some results of the study on the decay heat removal system based on natural circulation, performed on the open loop NC test facility GENEVA, built at TU Dresden in 2013. 16 probes were used to determine void fraction along the riser on nine different levels in high time and spatial resolution, and stability maps was created for riser with inner diameters of 20 mm and 38 mm and up to 85 kW evaporator power.
Rapid Fabrication of Li2TiO3 Pebbles via Microwave-Induced Combustion Process and 3D Printing
Authors: Zhou Qilai , Gao Yue , Mou Yang , Xue Lihong , Li Heping , Yan Youwei
Published: November 2014

Li2TiO3 has been recognized as one of the most promising tritium breeding materials for D-T fusion reactor blanket. In this study, ultra-fine Li2TiO3 powder was prepared rapidly by microwave-induced solution combustion synthesis (MSCS) using nitrates of lithium and titanate as raw materials, citric acid as fuel. The as-synthesized Li2TiO3 powder exhibits an average crystalline size as small as 20 nm with uniform distribution. Computer-assisted 3D printing technology was employed to fabricate Li2TiO3 ceramic breeding pebbles. This computer-assisted process is precise, efficient and controllable, which offers an alternative for the mass production of Li2TiO3 pebbles. The pebbles exhibit good sphericity, relatively small grain size, preferable sinterability and crushing load strength.
FloEFD Calculation Assessment to Experimental Data in Nuclear Safety Case
Authors: Philipov Svetlin , Popov Dimitar , Batakliev Krasimir
Published: November 2014

This paper presents the results of comparison between CFD code FloEFD and an experiment in Kozloduy NPP concerning coolant mixing. This comparison is done with purpose to find out behavior of FloEFD to manage complex tasks. Real experimental data is used from a performed experiment in the frame of an international project on nuclear safety. The main assessment point is the temperature distribution of the coolant at core inlet. A representative parameter is considered axis of minimal mixing. Also, temperature distribution in the down chamber at level 2.5m from the bottom is assessed.
Development of a PWR-W and an AP1000® Containment Building 3D Model With a CFD Code for Best-Estimate Thermal-Hydraulic Analysis
Authors: Jiménez Gonzalo , Bocanegra Rafael , Fernández Kevin , Queral César , Montero-Mayorga Javier
Published: November 2014

The simulation of design basis accidents in a containment building is usually conducted with a lumped parameter model. The codes normally used by Westinghouse Electric Company (WEC) for that license analysis are WGOTHIC or COCO, which are suitable to provide an adequate estimation of the overall peak temperature and pressure of the containment. However, for the detailed study of the thermal-hydraulic behavior in every room and compartment of the containment building, it could be more convenient to model the containment with a more detailed 3D representation of the geometry of the whole building.

The main objective of this project is to obtain a standard PWR Westinghouse as well as an AP1000® containment model for a CFD code to analyze the thermal-hydraulic detailed behavior during a design basis accident. In this paper the development and testing of both containment models is presented.
Research on Modular Modeling Method of Reactor Coolant System Based on THEATRe
Authors: Shi Bo , Tian Zhao-Fei
Published: November 2014

At present, research on the reactor coolant system is less yet, though modular modeling method has been widely used in the second-loop system of reactor. This paper takes the reactor coolant system of Qinshan-1 nuclear power plant as the object of study, analyses and researches on modular modeling method of reactor coolant system based on THEATRe, which is a large Thermal-Hydraulic real time simulation software developed by GSE Company and adopts NMNP (Nodal Momentum Nodal Pressure) solving method. This research establishes the modular model of the reactor coolant system equipments (including reactor core, main coolant pump, pressurizer, steam generator) using the THEATRe code. Due to each module is wrote into through different input cards, they can be solved by using their own matrix of velocity-pressure to guarantee the independence of the numerical calculation for different modular modules. THEATRe code does not have its own TDV like relap-5, meanwhile it also needs to ensure the pressurizer module can play a role in the multi-pressure node system. So this paper modifies solving method of the THEATRe source code to get suitable pressure boundary and flux boundary for RCS equipment modular module, and selects reasonable time step and data exchange frequency to achieve the data exchange of boundary pressure, flux and enthalpy among the equipment modules, which lays the foundation of establishing the real-time modular simulation model of the reactor coolant system in the future.
Helical Pitch Optimization of Double-Tube Once-Through Steam Generator
Authors: Fang Huawei , Wei Xinyu , Yan Shoujun , Wan Jiashuang , Zhao Fuyu
Published: November 2014

Once-Through Steam Generator (OTSG) is widely used in nuclear reactor system due to its advantages of compactness. The heat transfer performance of DOTSG is studied in this paper. In order to minimize the DOTSG volume and reduce the pressure drop of coolant, the pitch of inner helical tube is optimized with Pontryagin Maximum Principle (PMP). The double-tube is divided to three regions according to the coolant phase in secondary side. With given heat transfer load, choosing a combination function of minimum tube length and minimum pressure drop constructed with linear weighted method as objective function, the pitch optimization proceeds from superheated region to boiling region, and then to sub-cooled region in sequence, using Maximum Principle and gradient method. Then the pitch and temperature distribution along the axis is obtained respectively. The results show that the optimal pitch keeps constant along the axial direction in sub-cooled region and superheated region, but varies in boiling region. In boiling region, compared with minimum tube length optimization, the optimal tube length is 6.4% longer while the pressure drop is 36.3% smaller; and compared with minimum pressure drop optimization, the optimal pressure drop is 29.1% larger while the optimal tube length is 4.6% smaller. With the optimal pitch, the temperature distribution is in agreement with the general physic rules, which proves the correctness and the feasibility of the Maximum Principle method used for the structural optimization of DOTSG in this paper.
Validation of THACS for Sodium Cooled Fast Reactor Based on Benchmark Analysis of EBR-II
Authors: Nina Yue , Zaiyong Ma , Benxue Hu , Qiu Suizheng , Su Guanghui
Published: November 2014

In this paper the thermal-hydraulic characteristics of the primary loop of the Experimental Breeder Reactor (EBR-II), including the temperature and the flow characteristics of the core, the intermediate heat exchanger (IHX) and the experiment subassembly XX09 and XX10, were analyzed with the transient thermal-hydraulic code THACS. The THACS code contains the core, the pumps, IHX, the sodium pool and some other modules, and each module could operate separately. All of the primary–loop components are simulated one-dimensional, and in the core calculation the incompressible model for the single phase. The multiple-channel model is applied to simulate the core subassemblies, including the average, hot, XX09, XX10, the reflector and the blanket channels. The neutron physics is calculated with the point reactor kinetics, and the reactivity feedbacks caused by the Doppler effect, coolant density, axial expansion of fuel rods and radial expansion of core are considered. Two tests, namely the SHRT-17 and SHRT-45R tests, are simulated to validate our tools and models. The THACS simulation results show that the EBR-II type sodium cooled fast reactor could shut down automatically relying on inherent negative feedbacks in the two tests.
The CFD Analysis of Hydrogen Behavior in HTR-PM During the Water-Ingress Accident Caused by Double-Ended Guillotine Break of Two Steam Generator Heating Tubes
Authors: Zhang Ruidong , Sun Ximing , Dong Yujie
Published: November 2014

Water ingress into the primary circuit of HTR-PM (High Temperature Gas-cooled Reactor Pebble-bed Modular) is one of the beyond-design-basis accidents, which will lead to chemical corrosion of graphite and metal material in the reactor, resulting in the generation of combustible gases such as hydrogen and carbonic oxide. Besides, the rise of pressure in the reactor containment caused by the incoming of abundant of gas will threaten to the safety of radioactive material releasing. Therefore, it is essential to analyze the gas behavior in the containment of HTR-PM under the condition of water ingress.

Based on the preliminary design of 250MW HTR-PM, the paper analyzes the transport and distribution of hydrogen, helium and other kinds of gas in the containment under the condition of water ingress accident resulted from a double-ended guillotine break of two steam generator heating tubes which is a typical beyond design basis accident of HTR-PM with the multi-dimensional computational fluid dynamics CFD code GASFLOW. In addition, the density and distribution of hydrogen will be investigated in relation to possible hydrogen explosion.
End-to-End Technology of Modeling a Melt-Structure Interaction During IVMR in VVER With HEFEST-URAN Toolkit
Authors: Filippov A. S. , Drobyshevsky N. I. , Kamenskay D. D. , Kisselev A. E. , Moiseenko E. V.
Published: November 2014

The problem of reactor pressure vessel (RPV) integrity during in-vessel melt retention for VVER type reactor is considered. Comprehensive numerical analysis contains the modeling of full severe accident scenario including thermal and mechanical loads on the RPV. The RPV stress calculation takes into consideration vessel thermal erosion and creep with a long-term strength, and error estimations: uncertainty analysis, sensitivity study. These problems are solved with the presented HEFEST-URAN package including: HEFEST code (2D FEM analysis of melt-structure interaction), HEFEST-M code (2D FEM stress analysis) and Varia code (multivariate calculations with uncertainty/sensitivity statistical assessment). HEFEST-URAN is a subset of the SOCRAT code developed in the Russian Federation. In the paper the HEFEST-URAN methodology and validation are described with some examples. The results of multivariate calculations of IVMR for VVER-440 reactor are presented with sensitivity study of the effects of essential factors defining the RPV thermal loading and long-term strength.
Flow Visualization Experiment of a Swirling Flow Formed Downstream of a Piping With Successive Three Elbow to be Applied to Divertor Cooling
Authors: Kodate Shoichi , Ebara Shinji , Hashizume Hidetoshi
Published: November 2014

As one of potential divertor cooling systems, three-dimensionally connected elbow piping has been proposed. In this study, a visualization experiment was conducted for swirling flows generated downstream of five kinds of piping with successive three three-dimensionally connected elbows in order to evaluate the applicability of these systems to divertor cooling by comparing with the dual elbow case previously obtained in terms of strength of the swirling flow and turbulent statistics. From the experimental results, it was found that the triple elbow piping in which all elbows were connected three–dimensionally, referred to as 3D+3D layout, generated strong swirling velocity components than those of the dual elbow case and it became up to 70 % of the mean flow velocity. Moreover, it did not attenuate even 5D downstream of the triple elbow, where D was the diameter of the piping and the applicability of this flow field to divertor cooling can be promising. In addition, when heat transfer was evaluated in terms of turbulent statistics in the 3D+3D layout, heat transfer enhancement is expected because larger turbulent kinetic energy was observed near the pipe wall than a straight piping case.
Seismic Test on Double-Layer Model of HTR-PM Graphite Structure
Authors: Lai Shigang , Sun Libin , Shi Li , Zhang Zhengming
Published: November 2014

The HTR-PM (High Temperature Reactor Pebble bed Module) reactor core consists of assemblies that include graphite bricks, carbon bricks and keys/dowels. The double-layer structure, which is composed of two layers of bricks connected by keys and dowels, is the basic load-bearing unit of the graphite internals. A series of seismic experiments have been carried out to study the dynamic characteristic of this multi-body structure. As a part of them, the present study aimed to evaluate the nonlinear dynamic response of a full-size graphite/carbon unit and investigate the integrity of this assembly.

Dynamic excitations were applied at the bottom of the double-layer structure. The experimental random-motion excitation ranged from 0.2g to 0.75g. The sine-sweep excitation frequency band ranged from 5–100Hz, sweep rate was 1oct/min, with acceleration magnitudes of 0.3g and 0.5g. The sine beat wave excitation frequencies were 5Hz, 10Hz, 20Hz and 40Hz with acceleration of 0.5g. Response data of acceleration and displacement at certain points on bricks were measured. PSD curves and displacement time-histories were acquired. Results indicated that the fundamental frequency of the assembly is 42∼49Hz and the damping rate is 4.9%. The dynamic response of the assembly exhibited as an integrated component only when the excitation is lower than 0.35g.
Accident Analysis of Fukushima Daiichi Nuclear Power Plant Unit 2 by the SAMPSON Severe Accident Code
Authors: Takahashi Atsuo , Pellegrini Marco , Mizouchi Hideo , Suzuki Hiroaki , Naitoh Masanori
Published: November 2014

The transient process of the accident at the Fukushima Daiichi Nuclear Power Plant Unit 2 was analyzed by the severe accident analysis code, SAMPSON. One of the characteristic phenomena in Unit 2 is that the reactor core isolation cooling system (RCIC) worked for an unexpectedly long time (about 70 h) without batteries and consequently core damage was delayed when compared to Units 1 and 3. The mechanism of how the RCIC worked such a long time is thought to be due to balance between injected water from the RCIC pump and the supplied mixture of steam and water sent to the RCIC turbine. To confirm the RCIC working conditions and reproduce the measured plant properties, such as pressure and water level in the pressure vessel, we introduced a two-phase turbine driven pump model into SAMPSON. In the model, mass flow rate of water injected by the RCIC was calculated through turbine efficiency degradation the originated from the mixture of steam and water flowing to the RCIC turbine. To reproduce the drywell pressure, we assumed that the torus room was flooded by the tsunami and heat was removed from the suppression chamber to the sea water. Although uncertainties, mainly regarding behavior of debris, still remain because of unknown boundary conditions, such as alternative water injection by fire trucks, simulation results by SAMPSON agreed well with the measured values for several days after the scram.
Validation of ATHLET Code by LOCA-Induced Pressure Wave Propagation Tests
Authors: Trosztel István , Tóth Iván , Ézsöl György
Published: November 2014

Propagation of pressure waves inside the reactor vessel after a large break LOCA is an issue since it affects pressure drop across core internals and, as a result, induces stresses in different components, like core barrel, core structures and even fuel. For reactor safety analysis pressure wave propagation is traditionally performed by systems codes. However, strong dispersion among the calculated results calls for test results to validate the calculations.

The pressure wave propagation following a larger LOCA is being systematically addressed by experiments in the PMK-2 integral-type test facility. In order to capture the high speed propagation of pressure waves special pressure transducers (capable to resolve the pressure variation with a frequency of 4 kHz) have been installed. The first four tests were conducted with rupture disks for opening the break, but a special quick opening valve will be installed for future tests, allowing the adjustment of the opening time between 12 and 50 ms.

The paper presents results of validation of the ATHLET code by the test results. The low-frequency oscillation of the measured system pressure was shown to be caused by flow rate coming from the pressuriser that compensates mass lost via the break: the frequency of the oscillation was slightly under-predicted. The propagation of the first rarefaction wave from the top of the downcomer to the upper plenum is very well calculated by ATHLET: in spite of the first order discretisation no numerical diffusion can be observed. The calculated pressure differences between two different locations in the system are of primary interest, since they define the loads on primary system internals. ATHLET somewhat overestimates the amplitude of the pressure difference pulses, while it fairly well describes the frequency of oscillations. First analyses indicate an effect of the calculated break flow rate. ATHLET calculates a slower attenuation of the pressure oscillations as compared to test results. This can be the consequence of rigid walls assumed in the analysis. The tendency of increasing first pressure peak with increasing system pressure is well predicted by the code.

In summary, it can be stated that ATHLET calculations produce slightly conservative results based on comparison with measured data.
CFD Validation of Void Distribution in a Rod Bundle With Spacer
Authors: Goodheart Kevin , Ylönen Arto , De Cacqueray Victor , Prasser Horst-Michael
Published: November 2014

As CFD (Computational Fluid Dynamics) continues to grow in the nuclear industry the need for validation is essential to this growth. The focus of this paper is to highlight the validation of AREVA’s 2-Phase CFD models in a rod bundle with spacer. The experimental work is based on void distribution measurements using a 64×64 wire mesh sensor at the Paul Scherrer Institute adiabatic test loop fuel rod bundle (SUBFLOW). The 2-Phase models are based on the Eulerian multiphase framework where the interaction models are further developed using user routines in a commercial CFD tool to achieve stability and accuracy in a rod bundle configuration with a spacer. The 2-Phase CFD models coincide with the SUBFLOW experiments in showing the effect of void collecting in the center of the sub-channels slightly downstream of the spacer and further downstream the smaller bubbles migrate toward the rod surfaces whereas the larger bubbles stay in the sub-channels. Good agreement is achieved between CFD and void distribution experiments. The multi-phase CFD method was used by AREVA to improve the performance of the new products GAIA and ATRIUM™ 11.
A CFD Analysis of Flow Blockage Phenomena in ALFRED LFR DEMO Fuel Assembly
Authors: Di Piazza I. , Tarantino M. , Magugliani F. , Alemberti A.
Published: November 2014

A CFD study has been carried out on fluid flow and heat transfer in the HLM-cooled Fuel Pin Bundle of the ALFRED LFR DEMO.

In the context of GEN-IV Heavy Liquid Metal-cooled reactors safety studies, the flow blockage in a Fuel sub-assembly is considered one of the main issues to be addressed and the most important and realistic accident for LFR Fuel Assembly. The present paper is a first step towards a detailed analysis of such phenomena, and a CFD model and approach is presented to have a detailed thermo-fluid dynamic picture in the case of blockage. The closed hexagonal, grid-spaced fuel assembly of the LFR ALFRED has been modeled and computed. At this stage, the details of the spacer grids have not been included, but a conservative analysis has been carried out based on the current main geometrical and physical features. Reactivity feedback, as well as axial power profile, have not been included in this analysis. Results indicate that critical conditions, with clad temperatures around ∼900°C, are reached with blockage larger than 30% in terms of area fraction.

Two main effects can be distinguished: a local effect in the wake/recirculation region downstream the blockage and a global effect due to the lower mass flow rate in the blocked subchannels; the former effect gives rise to a temperature peak behind the blockage and it is dominant for large blockages (>20%), while the latter effect determines a temperature peak at the end of the active region and it is dominant for small blockages (<10%). The blockage area has been placed at the beginning of the active region, so that both over-mentioned phenomena can fully take place. The mass flow rate at the different degree of blockage has been imposed from preliminary system code simulations.

Transient analyses with fully resolved SST-ω turbulence model have been carried out and results indicate that a blockage of ∼15% (in terms of blocked area) leads to a maximum clad temperature around 800 °C, and this condition is reached in a characteristic time of 3–4 s without overshoot. Local clad temperatures around 1000 °C can be reached for blockages of 30% or more.

CFD simulations indicate that Blockages >15% could be detected by putting some thermocouples in the plenum region of the FA.