bluuebunny/crossref_metadata_2025 huggingface dataset:
Coupling of the 3D Neutron Kinetic Core Model DYN3D With the CFD Software ANSYS CFX
Authors: Grahn Alexander , Kliem Sören , Rohde Ulrich
Published: November 2014

This article presents the implementation of a coupling between the 3D neutron kinetic core model DYN3D and the commercial, general purpose computational fluid dynamics (CFD) software ANSYS-CFX. In the coupling approach, parts of the thermal hydraulic calculation are transferred to CFX for its better ability to simulate the three-dimensional coolant redistribution in the reactor core region. The calculation of the heat transfer from the fuel into the coolant remains with DYN3D, which incorporates well tested and validated heat transfer models for rod-type fuel elements. On the CFX side, the core region is modelled based on the porous body approach. The implementation of the code coupling is verified by comparing test case results with reference solutions of the DYN3D standalone version. Test cases cover mini and full core geometries, control rod movement and partial overcooling transients.
The SCORPIO-VVER Core Monitoring and Surveillance System for VVER Type of Reactors
Authors: Molnar Jozef , Vocka Radim
Published: November 2014

The SCORPIO-VVER core monitoring and surveillance system has proved since the first installation at Dukovany NPP in 1999 to be a valuable tool for the reactor operators and reactor physicists. It is now installed on four units of Dukovany NPP (EDU, Czech Republic), on two units of Bohunice NPP (EBO, Slovak Republic) replacing the original Russian VK3 system and on the full scale plant training simulator at the Centre for training and education of the reactor operators and reactor physicist in Trnava (Slovak Republic). By both Czech and Slovak nuclear regulatory bodies the system was licensed as a Technical Specification Surveillance tool.

Since it’s first installation, the development of SCORPIO-VVER system continues along with the changes in VVER reactors operation. The system is being adapted according the utility needs and several notable improvements in physical modules of the system were introduced. The most significant changes were done in support of the latest optimized Gd bearing fuel assemblies, improvements in the area of core design (neutron physics, core thermal hydraulics and fuel thermal mechanics), adaptation of the system to up-rated unit conditions (uprated power up to 107%), in design and methodology of the limit and technical specifications checking and improvements in the predictive part of the system.

After the currently finished upgrades the SCORPIO-VVER is still in focus of Central European nuclear power plants with the roadmap of upgrades and modifications up to 2016.

This paper shortly describes the system’s main functions, the history of implementation at the VVER-440 type of reactors and deals with the system’s future upgrades and plans to meet the latest requirements of efficient and safety NPP operation.
Turbulent Flow Boundary Layer and Heat Transfer Characteristics for the Intermediate Heat-Exchanger Simulation of a VHTR
Authors: Wu Yannan , Dong Yujie , Yuan Kun
Published: November 2014

In order to improve the thermal efficiency of the intermediate heat exchanger (IHX) in high temperature gas-cooled reactor, the article takes a theoretical calculation and simulation analysis of different turbulence models on the turbulent flow boundary layer of the IHX. And a suitable model for the high temperature and high pressure unit of the IHX is built. According to the boundary layer distribution of different models and the fluid velocity and temperature changes of central region, we find the related characteristics and cut-off point of the boundary layer area and the Poiseuille flow area. At the same time, the paper verifies the three kinds of turbulence model and the formula with solution theory, including the relationship of Nu (Nusselt) number and Re (Reynolds) number and the partial wall friction resistance. These results provides theoretical support for the next step of the heat transfer enhancement research using the artificial roughness elements.
Concept of Pressurized Water Lead-Bismuth-Cooled Fast Reactor (PLFR)
Authors: Takahashi Minoru
Published: November 2014

The innovative concept of pressurized water lead-bismuth-cooled fast reactor (PLFR) has been proposed and studied based on the previous LFR concept: PBWFR. Primary pumps and steam generators that contact lead-bismuth coolant are eliminated. A feedwater is directly injected into the primary coolant of hot lead-bismuth eutectic (LBE) at the outlet of the reactor core under the pressure of 14 MPa as PWR. The specifications of PLFR system are discussed and presented from thermal-hydraulic point of view.
A Double Porosity Model to Describe Both Permeability Change and Dissolution Processes
Authors: Niibori Yuichi , Usui Hideo , Chida Taiji , Mimura Hitoshi
Published: November 2014

Cement is a practical material for constructing the geological disposal system of radioactive wastes. However, such materials alter groundwater up to 13 in pH around the repository, changing the permeability of natural barrier. So far, the authors have examined the relation of permeability change with dissolution process by flowing a high pH solution (NaOH, 0.1 mM) into a bed packed with amorphous silica particles. Here, the particle diameters were adjusted to a size fraction of 74 to 149 μm by sieving. Its specific surface area was estimated as 350 m2/g by the BET method using nitrogen gas. The experimental results showed that the permeability did not immediately change although the soluble silicic acid continuously flowed out of the packed bed.

This study proposes a new mathematical model considering the diffusion and dissolution processes in the inner pore of the particle. This model assumed that each packed particle (74 to 149μm in diameter) consists of the sphere-shaped aggregation of smaller particles (20 nm in diameter). OH− ions diffuse into the pore between such small particles, and simultaneously consumed by the reaction with small particles. The radius of the each packed particle (sphere-shaped aggregation of small particles) was defined by the length from the center of the aggregation to the region where the small particles still remains. Since the outer small particles more easily dissolve than inner small particles because of diffusion process of OH− ions, each packed particle gradually shrinks. The fundamental equations consist of a simple diffusion equation of spherical coordinates of OH− ions considering the reaction term, which is linked by the equation to describe the size change of small particles with time. Here, this model also considered a change (time and space) of the diffusion oefficient caused by the change of the porosity between small particles. Besides, the change of over-all permeability of the packed bed was evaluated by Kozeny-Carman equation and the calculated radii of packed particles. The dissolution rate constant already reported was used.

The calculated result was able to well describe the experimental result, though there was no fitting parameter in the comparison with the experiment results. While the flow paths of underground cannot be simply simulated by a packed bed, this approach suggested that the dynamic behavior of permeability in a natural barrier depends also on non-uniformity of dissolution processes in inner pores (secondary pores) of minerals.
Experimental Study on Restart Performance of Absorber Sphere Pneumatic Conveying
Authors: Zhang He , Li Tianjin , Qi Weiwei , Huang Zhiyong , Bo Hanliang
Published: November 2014

In the high-temperature gas-cooled reactor pebble-bed module (HTR-PM), absorber sphere shutdown system is the second shutdown system. Absorber spheres are transported back to storage vessel by pneumatic conveying when reactor needs to be started. The restart reliability of absorber sphere pneumatic conveying is quite important. Experimental system for absorber sphere pneumatic conveying has been built to investigate the gas-solid flow characteristics of pneumatic conveying. Restart experiments were conducted to study the restart performance of pneumatic conveying. Ambient air was used as source gas and absorber sphere was replaced by glass sphere. The inlet steady gas velocity of the sphere feeder was in the range of 17∼27 m/s. Pneumatic conveying could be restarted in all the experiments, which showed the good reliability of the pneumatic conveying process. The maximum relative deviation of the spheres pile heights in the riser was 4.1%, which showed a good repeatability. Pressure drop of feeder during steady stage was in the range of 4∼8 kPa. Furthermore, the restart experiments were also conducted in the full scale experimental system with helium as the source gas and the pressure of 2 MPa, the results also showed the good reliability of pneumatic conveying process.
Retrieving Fuel Elements by Pneumatic Suction in Pebble-Bed High Temperature Gas-Cooled Reactor
Authors: Wu Bin , Wang Jin-hua , Li Yue , Wang Bing , Liu Ji-guo
Published: November 2014

Spent fuel storage system of pebble-bed high temperature gas-cooled reactor needs to retrieve fuel elements and spent fuel elements from storage tank during fuel reloading condition and some other special status. This function is to be achieved by the negative pressure suction system. Research in depth is needed towards the negative pressure suction system design and experiment in order to reliably supply the system with enough suction capability and meanwhile prevent the fuel element from over-speed impact damage. A comprehensive experimental facility of negative pressure suction was built to investigate and verify the designed system. The facility mainly comprises a fuel canister, a suction tube, a tube feeder, a gas isolator, a Roots blower and pipelines. The negative pressure suction force was provided by the Roots blower and drove the fuel elements out of the canister through the tube. The Roots blower was driven by a frequency converter so that the suction flow rate could be adjusted as wanted. The dynamics model of spherical element in the tube was established. The pressure drop distribution of the negative pressure suction system was also calculated. Then the pressure drops and the sphere velocity were measured at different air flow rates. Based on the experimental results and calculation analysis, parameter requirements for the Roots pump were concluded. Therefore, fuel elements could be successfully retrieved without over-speed impact damage. These results provide useful experience for engineering design of negative pressure suction system in the spent fuel storage system.
Numerical Analysis of Rotordynamic Coefficients of Annular Flow in Canned Motor RCP
Authors: Hu Yaoyu , Wang Dezhong , Yin Junlian , Wang Yujin
Published: November 2014

The operating performance and safety characteristics of canned motor Reactor Coolant Pump (RCP) are vital to the nuclear reactor. The annular flow, which is between the rotor and stator, makes substantial effects on the rotordynamic characteristics of RCP. In this work, the annular flow is simulated by means of 3-dimensional CFD approaches. The annular flow field and the auxiliary impeller are modeled. The results show that the pre-swirl strength at the inlet of the annular flow is evident. Another model without the auxiliary impeller is analyzed under constant pre-swirl ratio with various axial flow rates and whirl speeds. The rotordynamic coefficients of the annular flow are obtained. The results are compared with theoretical predictions. Numerical results show that the fluid in the annular region makes remarkable effects on the rotordynamic characteristics of the canned motor RCP.
The Natural Circulation Characteristic Analysis of AP1000 IRWST
Authors: Xue Ruojun , Xia Bin , Mingyuan Wang
Published: November 2014

The natural convection passive residual heat removal system is an essential system of AP1000 nuclear power plants, which works in the non-LOCA accident. It can discharge the waste heat in the reactor core timely, and prevent the reactor accident. The In-Containment Refueling Water Storage Tank (IRWST) is one of the most important equipments of passive residual heat removal system, which supplies the hot trap to the system. The purpose of the paper is to research the natural circulation characteristic of fluid in IRWST. In this paper, FLUENT hydrodynamics analysis software was used to simulate the flow and heat-transfer characteristics of natural convection. Temperature field and flow field in different time or locations were compared to analyze the course of natural circulation. The error was evaluated by comparing the average temperature on the outlet of heat exchanger and the design temperature. The result showed that FLUENT could simulate the flow and heat-transfer characteristics of natural convection, and the error of the simulation was acceptable. In conclusion, natural convection heat exchanger of the study can discharge the waste heat in the reactor core timely. Due to the complex structure of the heat exchanger and the IRWST, the “heat-transfer dead zones” and “flow dead zones” are generated in local parts. The presence of dead zones affects the formation of natural convection. With the development of natural circulation, the temperature field and flow field in the IRWST keep stable and the temperate rise rate become slowly.
Development of Supercritical Water Heat-Transfer Correlation for Vertically-Upward Internally-Ribbed Tubes
Authors: Zhang Weiqiang , Li Huixiong , Zhang Qing , Zhang Yifang , Wang Tai
Published: November 2014

The investigation on the heat transfer characteristics for supercritical pressure water (SCW) is of value for the development of the supercritical water-cooled nuclear reactor (SCWR). As an important heat transfer enhancement element, heat transfer for SCW in internally-ribbed tubes was still not solved, though lots of experimental studies have been published and a great many heat transfer correlations were proposed.

This paper presented an analysis of heat transfer in the internally-ribbed tubes, through comparing heat transfer correlations for SCW gained from different internally-ribbed tubes under the same operating condition. It was found that all existing heat transfer correlations reported could not been well applied for various internally-ribbed tubes with large deviation between prediction results and experimental values, because rib geometry had a great influence on heat transfer of internally-ribbed tubes. On the basis of experimental data collected from open literature for internally-ribbed tubes, a new general calculation correlation of heat transfer coefficient for SCW was developed for various internally-ribbed tubes by combining an optimized empirical correlation for vertically-upward smooth tubes and four dimensionless numbers of rib geometry.

The results show that the calculated values of the new present correlation is in reasonable agreement with available experimental data collected. Moreover, the new correlation was verified well by experiment data of two new-type internally-ribbed tubes performed beyond the above experimental database.
Elaboration of the System Based Code Concept — Activities in JSME and ASME: (4) Joint Efforts of JSME and ASME
Authors: Asayama Tai , Takaya Shigeru , Morishita Masaki , Schaaf Frank
Published: November 2014

This paper describes the ongoing activities at the Joint Task Group for System Based Code established in 2012 by the Japan Society of Mechanical Engineers (JSME) and the American Society of Mechanical Engineers (ASME) in the ASME Boiler and Pressure Vessel Code Committee. The Joint Task Group aims at developing alternative rules for ASME Boiler and Pressure Vessel Code Section XI Division 3, inservice inspection requirements for liquid metal reactors. The alternative rules will be developed based on the System Based Code concept which was originally proposed in Japan and is being elaborated both in JSME and ASME. The alternative rules are for sodium-cooled fast reactors where some of the components could have difficulties in conforming to the current requirements primarily due to accessibility. The alternative requirements would consist of a set of relieved requirements and a logic flow through which the applicability of them is judged. The logic flow considers both component structural integrity and the plant safety goals. The issuance of a Code Case is envisioned around 2016. Further efforts to integrate the process into a new framework being developed in Section XI would cover various types of reactors.
Development of Instrumentation System for Severe Accidents
Authors: Wada Shohei , Murata Akira , Arita Setsuo , Fushimi Atsushi , Suzuki Hirotsugu , Fujishima Yasutake
Published: November 2014

The accident at Tokyo Electric Power Company’s Fukushima Dai-ichi Nuclear Power Plant (TF-1 accident) caused severe situations and resulted in a difficulty to measure important parameters to monitor plant conditions. Therefore, we have studied to select the important parameters (SA parameters) that should be monitored at the severe accident and are developing the Severe Accident Instrumentations and Monitoring Systems (SA-keisou) that could measure the parameters in severe accident conditions. (SA-keisou in Japanese means “severe accident instrumentation and monitoring systems). In this study, the boiled water reactor (BWR) is mainly explained.

This study is a part of the results of the collaborative project by domestic electric power companies and plant manufacturers that is carried out as the Safety Enhancement for LWRs program by Agency for Natural Resources and Energy.
Application of Leakage-Flux Analysis for Three-Phase Induction Motors
Authors: Nakayama Kohichi , Watanabe Yukio , Osaki Kenji , Hegi Kotaro
Published: November 2014

A simple and highly reliable motor diagnostic method was developed. The method monitors low-frequency components of leakage flux from motors. Small wear of bearing housing was detected by the method which could not be detected by the conventional vibration or electric current monitoring methods.
Buckling Analysis of Tube for Steam Generator According to ASME B&PV Code
Authors: Kong Jungpyo , Ha Changhoon , Cho Minki , Yeo Seunghoon , Park Tae-Jung
Published: November 2014

The U-tubes installed inside the steam generator experience high temperature and pressure as a role of heat transfer. Specially, during the secondary side hydrostatic test which ensures the integrity of steam generator, the U-tube is subjected to high external pressure. The purpose of this paper is to investigate the allowable external pressure of the U-tube in steam generator.

In the ASME B&PV Code Section III [1], the allowable external pressure is determined by the rules of NB-3133. Alternatively, NB-3228 analysis may be applied. In order to determining the allowable external pressure of steam generator tube, the buckling analysis is performed. The analysis consists of the collapse pressure and elastic instability pressure analyses. In this study, these pressures are determined by finite element analysis (FEA) using ANSYS computer program.

The non-linear static analysis is performed with ideally elastic-plastic material properties for the collapse analysis. On the other hand, the elastic instability pressure is calculated by eigenvalue analysis in elastic range. These allowable pressures are found to be 24.1 ksi and 10.5 ksi. Therefore the lower pressure of 24.1 ksi is the allowable external pressure of tube. In addition, the results of analysis are compared with other research [6] and Det Norske Veritas (DNV) offshore standards [11].

In conclusion, the results of buckling analysis are well matched with other research [6] and standard [11]. For steam generator tubes, the collapse pressure is dominant factor in failure. Also, the collapse pressure is largely influenced by the ovality of tube.
Analysis of Accident Progression of Fukushima Daiichi NPP by the SAMPSON Code
Authors: Naitoh Masanori , Pellegrini Marco , Suzuki Hiroaki , Mizouchi Hideo , Okada Hidetoshi
Published: November 2014

This paper describes analysis results of the early phase accident progression of the Fukushima Daiichi Nuclear Power Plant (NPP) Unit 1 by the severe accident analysis code SAMPSON.

The isolation condensers were the only devices for decay heat removal at Unit 1, but they stopped after the loss of AC and DC powers. Since there were no decay heat removal for about 14 hours after their termination until the start of alternative water injection into the core by the fire engine, the core melt and the reactor pressure vessel (RPV) bottom failure occurred resulting in large amount of fission products release into the environment. The original SAMPSON was improved by adding new modellings for the phenomena which have been deemed specific to the Fukushima Daiichi NPP: (1) deterioration of SRV gaskets and (2) buckling of in-core-monitor housings which caused the early steam leakage from the core into the drywell, and (3) melt of the in-core-monitor housings in the lower plenum of the RPV.

The analysis results showed that (1) 55.3% of UO2 of the initial loading and 66.1% of the core material including UO2, zircaloy, steel and control materials had melted down into the pedestal of the drywell, (2) the amount of Hydrogen generated by Zr-H2O reaction was 686 kg, (3) amount of Cs element released from fuels was 61 kg which was 72% of the total Cs element which was included in fuels at the initiation of the accident, and (4) 18.3% of the corium which fell into the pedestal was one large lump and the 81.7% was particulate corium.
A Comparison of Creep-Fatigue Assessment and Modelling Methods
Authors: Pohja Rami H. , Holmström Stefan B.
Published: November 2014

Design codes, such as RCC-MRx and ASME III NH, for generation IV nuclear reactors use interaction diagram based method for creep-fatigue assessment. In the interaction diagram the fatigue damage is expressed as the ratio of design cycles over the allowable amount of cycles in service and the creep damage as the ratio of time in service over the design life. With this approach it is assumed that these quantities can be added linearly to represent the combined creep-fatigue damage accumulation. Failure is assumed to occur when the sum of the damage reaches a specified value, usually unity or less. The fatigue damage fraction should naturally be unity when no creep damage is present and creep damage should be unity when no fatigue damage is present. However, strict fatigue limits and safety factors used for creep rupture strengths as well as different approaches to relaxation calculation can cause a situation where creep-fatigue test data plotted according to the design rules are three orders of magnitude away from the interaction diagram unity line. Thus, utilizing the interaction diagram methods for predicting the number of creep-fatigue cycles may be inaccurate and from design point of view these methods may be overly conservative. In this paper the results of creep-fatigue tests carried out for austenitic stainless steel 316 and heat resistant ferritic-martensitic steel P91, which are included in the design codes, such as RCC-MRx, are assessed using the interaction diagram method with different levels of criteria for the creep and fatigue fractions. The test results are also compared against the predictions of a recently developed simplified creep-fatigue model which predicts the creep-fatigue damage as a function of strain range, temperature and hold period duration with little amount of fitting parameters. The Φ-model utilizes the creep rupture strength and ultimate tensile strength (UTS) of the material in question as base for the creep-fatigue prediction. Furthermore, challenge of acquiring representative creep damage fractions from the dynamic material response, i.e. cyclic softening with P91 steel, for the interaction diagram based assessment is discussed.
Two-Phase Flow Studies in Boiling Single Channel Flow Using Wire Mesh Tomography (WMT) and Ultrasound Velocity Profile (UVP)
Authors: Hamdani Ari , Nguyen Thang Tat , Ito Daisuke , Tsuzuki Nobuyoshi , Kikura Hiroshige
Published: November 2014

The objective of this work is to investigate characteristics of co-current boiling flow in a circular pipe with an inner diameter of 52 mm by using wire mesh tomography (WMT) and ultrasonic velocity profile (UVP). The inner wall of pipe is modified by adding fins on the inner pipe’s wall. This modification is intended to change the flow behavior into swirling flow in boiling flow. Firstly, the effect of wall modification on flow behavior is investigated by numerical calculation. Secondly, two-phase flow is investigated experimentally using UVP and WMT. In experiments, local time-average void fraction is measured using WMT and velocity profile is measured using UVP. Furthermore, these measured data, both void fraction and velocity profile, will give information about changing in flow pattern caused by modified inner pipe’s wall.
The Surge Line Stress Analysis Model Setup With the Consideration of Thermal Stratification
Authors: Wang Jianjun , Ge Zengfang , Sun Zhongning , Yan Changqi
Published: November 2014

In this paper, we deal with a typical pressurizer surge line in a conventional pressurized water reactor (PWR). This study is performed to develop an understanding of thermal stratification phenomenon, which may occur in the surge line during either normal condition or transient process, in the pressurizer surge line. The pressurizer surge line model of Daya Bay nuclear power plant is used as base analysis model, in which the hot leg is taken into account. The transient temperature distribution required to assess the phenomenon along the pressurizer surge line is obtained through CFD analysis technology using ANSYS FLUENT. The temperature loads are transferred to ANSYS Mechanical for stress evaluation for the heat up transient process. Subsequently, the usage factor is calculated on the basis of ASME Section-III design curve. The possible mitigation scheme for the thermal stratification phenomenon of changing the layout angles is also simulated and analyzed in detail. The results show that the thermal stratification phenomenon will occur both in normal operating condition and in heat up transient process. The circumfluent effect makes the thermal stratification phenomenon exhibit unique profile due to the introduction of the hot leg. The continuous spray mass flow rate may influence both the temperature difference and the occurrence range for the thermal stratification phenomenon. The stress analysis incorporating both temperature load and pressure load is performed for pressurizer surge line model with hot leg for the conservative and complete heat up case.
Cross-Comparison of One-Dimensional Thermal-Hydraulic Codes on Natural Circulation Analysis of NACIE Loop Test for Lead-Alloy Cooled Advanced Nuclear Energy Systems (LACANES)
Authors: Shin Yong-Hoon , Hwang Il Soon , Polidori Massimiliano , Meloni Paride , Casamassima Vincenzo , Cornet Stephanie , Barucca Luciana , Balestri Davide , Jin Ming , Viellieber Mathias
Published: November 2014

As one of the Generation-IV reactor concepts, lead-alloy-cooled advanced nuclear energy systems (LACANES) have been studied worldwide in order to utilize the advantages of good heat transfer properties, neutron transparency and chemical inertness with air and water. Since the Fukushima accident, the passive safety aspect of the LACANES is increasingly emphasized due to outstanding natural circulation capability. To investigate the thermal-hydraulic capability of LBE, an international cooperation has been performed under OECD/NEA program, under the guidance of the Nuclear Science Committee by a task force named as Lead Alloy Cooled Advanced Nuclear Energy Systems (LACANES) since 2007. This international collaboration had dealt with computational benchmarking of isothermal LBE forced convection tests in the phase I, and the working group published a guideline for using one-dimensional system codes to simulate LBE forced circulation test results from HELIOS loop. The phase II was started after that, to give an additional guideline in the case of natural circulation. NACIE, one of benchmarking targets for the phase II which is a rectangular-shape loop located at ENEA-Brasimone Research Centre, Italy. NACIE test results were benchmarked by each participant using their one-dimensional thermal-hydraulic codes, and they are to follow the guideline from the LACANES phase I for regions where hydraulic loss occurs. Due to the selection of hydraulic loss coefficient relations by users, the cross-comparison results of international participants showed some discrepancies and the estimated mass flow rates had 13% of maximum error. Also, the future R&D areas are identified.
Spent Fuel Pool Under Severe Accident Conditions
Authors: Jaeckel Bernd , Birchley Jonathan , Fernandez-Moguel Leticia
Published: November 2014

The possibility of a spent fuel severe accident has received increasing attention in the last decade, and in particular following the Fukushima accident. Several large scale experiments and also separate effect tests have been conducted to obtain a data base for model development and code validation. The outcome of the Sandia BWR Fuel Project was used to define the flow parameters adjusted for the low pressure and the increased flow resistance due to the presence of the spent fuel racks which resulted in reduced buoyancy driven natural circulation flow compared with reactor geometry. The possibility of a zirconium fire, using the flow parameters obtained from the spent fuel experiments, is investigated in the present work. The important outcome of the study is that spent fuel will degrade if temperatures above 800 K are reached. In partial loss of coolant accidents the flow through the lower bottom nozzle is blocked and because there is no cross flow possible due to the spent fuel racks the coolant flow in the upper dry part of the spent fuel is limited by the steam production in the lower still wetted part of the fuel. This accident scenario leads to the fastest heat up in a postulated spent fuel accident. The influence of different kind of spent fuel storage (hot neighbour and cold neighbour) is investigated. An important factor in these calculations is the radial heat transfer to the neighbouring fuel assemblies. In the present work limits of the spent fuel storage under accident conditions (minimum allowed water levelin the pool) and total loss of coolant (maximum coolable decay heat per fuel assembly) are shown and explained.
Design and Development of Virtual Assembly Prototype System for CLEAR-I
Authors: Shang Leiming , Yang Zihui , Zhao Jinbo , He Tao , Hu Liqin
Published: November 2014

China LEAd-based research Reactor (CLEAR-I), which is one part of Accelerator Driven subcritical System (ADS) proposed by Chinese Academy of Sciences. Assembly scenario simulation is part of the CLEAR-I R&D project, enabling their suitability for the environment or scenario key points to be verified, taking into account unexpected situations and providing technical answers. Based on virtual reality technology, a virtual assembly prototype system for CLEAR-I has been developed for training, ergonomics evaluation, safety and security purposes. Main capabilities of system include: 1) CAD-based virtual scene modeling; 2) immersive experience in virtual environment; 3) interactive virtual assembly simulation of complex components. The paper describes the system architecture, main methods and its preliminary application to CLEAR-I.
Modelling of the Impact of Local Effects on Fuel-Cladding Interaction During Power Ramp
Authors: Klouzal Jan , Dostál Martin
Published: November 2014

The power increase rate of the reactor is often derived using the fuel performance code. Too restrictive rates are not desirable since they lead to the loss of production. On the other hand fast increase or not well controlled axial power oscillation may result in the rod failure due to pellet-cladding interaction.

Most of the currently used fuel performance codes treat the stack of the fuel pellets using a simplified “1.5D” approach where individual pellets are not distinguished and the fuel stack is taken to be symmetrical. In reality, several effects must be taken into account when more accurate description is required. These local effects include contact at pellet-pellet interface, fuel pellet cracking under thermal stress, fabrication defects pellets or azimuthal asymmetry in the heat generation or heat transfer conditions due to rod bowing or presence of control rods.

Detailed models of the local phenomena are therefore being developed at ÚJV Řež using the ABAQUS 6.12 code and used to improve the predictions of the codes routinely used for the core design assessment. For example the impact of the use of the advanced pellet materials on the peak loading that the cladding will experience during the power ramp has been quantified.
Experimental Investigation Into Buckling Failure of Slender Metal Plates in Severe Accident Conditions
Authors: Jo Byeongnam , Sagawa Wataru , Okamoto Koji
Published: November 2014

Buckling failure load of stainless steel columns under compressive stress was experimentally measured in severe accident conditions, which addresses the accidents in Fukushima Daiichi nuclear power plants. Firstly, buckling failure load defined as load which causes failure of the column (plastic collapse) was measured in a wide range of temperatures from 25 °C up to 1200 °C. The load values measured in this study were compared to numerical estimations by eigenvalue simulations (for an ideal column) and by nonlinear simulations (for a column with initial bending). Two different methods for measurement of the buckling failure load were employed to examine the effect of thermal history on buckling failure. Different load values were obtained from two methods in high temperature conditions over 800 °C. The difference in the buckling failure load between two methods increased with temperature, which was explained by the effect of creep at high temperatures. Moreover, the influence of asymmetric temperature profiles along a plate column was also explored with regard to the failure mode and the buckling failure load. In present study, all of the buckling processes were visualized by a high speed camera.
Cold Water Injection and Rod Ejection Analysis of Annular Fueled PWRs by a Hybrid Lumped Parameter Model
Authors: Duarte Juliana P. , Rivero José de Jesus , Alvim Antonio Carlos M. , Piqueira José Roberto C. , Frutuoso e Melo Paulo F. F.
Published: November 2014

Annular fuels are being studied to increase the power of advanced third-generation reactors by 50%. This paper aimed to analyze transient scenarios through a hybrid lumped parameter-finite difference model in a pressurized water reactor with annular fuel. The model used in this work is more detailed than the double lumped parameter one, but still simple enough to model some transients in PWR fuels, as rod ejection accident and cold water insertion accident. The heat transfer equations are solved by the numerical semi-implicit Crank-Nicolson method together with point kinetics equations with six groups of delayed neutrons and a lumped parameter model for the reactor coolant.

The model takes into account in an approximate way the hot spot by using a composed peaking factor equal to 2.5. The reactivity feedback is taken into account by considering the Doppler effect of fuel temperature, and also moderator temperature variation. The results were compared with solid fuel performance and showed that the annular fuel reached considerable lower fuel temperature profiles even for 150% power, as compared to 100% power for solid fuel, thus showing that this kind of fuel has a better safety performance for the transients analyzed. The rod ejection accident showed that feedback effects can lead the reactor to a new safe steady state condition.
A Lower Bound for the Dryout Quality in Annular Flow
Authors: Manning J. P. , Walker S. P. , Hewitt G. F.
Published: November 2014

The mechanism responsible for Critical Heat Flux (CHF) depends on the flow regime. In the annular flow regime it is normally assumed that CHF occurs when the liquid film dries out. The quality at the onset of annular flow varies, but is generally a few percent, and phenomenological models to predict CHF are routinely applied at qualities above this value. In this paper we will demonstrate that annular flow film dryout cannot occur until a quality significantly greater than this. This finding means that for a large fraction of the annular flow regime the film dryout mechanism cannot be responsible for CHF. This finding provides guidance as to under what circumstances such phenomenological models may properly be used.
Investigation on the Backflow Induced Pre-Rotation of RCP
Authors: Yin Junlian , Li Jingjing , Chen Hui , Hu Yaoyu , Wang Dezhong
Published: November 2014

The pre-rotation phenomenon found at the inlet pipe of a pump under small flow rate was observed but the mechanism was not discovered. To do this, an unsteady CFD simulation with the flow rate decreasing from the best design point to a small flow rate was carried out. The numerical results show that a critical point when pre-rotation occurs exists. The flow pattern evolution on the S1 and S2 stream surface, as well as the pressure distribution at in the inlet, was analyzed. It is shown that it is the reverse flow appeared near the shroud at the leading edge which leads to the occurrence of pre-rotation. The pre-rotation only exist in the periphery of the inlet pipe and the propagation length is limited.
Thermal Hydraulic Facility for Full Length Bundle CHF Experiment
Authors: Qin Shengjie , Lang Xuemei , Zhuo Wenbin , Li Pengzhou , Liu Ye
Published: November 2014

One of the important limitations in nuclear reactor design is to avoid departure of nucleate boiling (DNB) and dry-out (DO), which may cause fuel assembly cladding failure. Critical heat flux (CHF) under different condition is a complicated physical phenomenon which could not be predicted by analytical method. Experiment is the common way to acquire the CHF correlations for complicated geometry like bundle channels. To simulate the thermal hydraulic condition in an actual reactor, different scaling and geometry rod bundles are applied. The facilities of NPIC (Nuclear Power Institute of China) which carry out the CHF experiments of full length rod bundle and primary experimental results are introduced in the article.
AGN-201M: Using Low Power Training Reactors in Education
Authors: Škoda Radek
Published: November 2014

The AGN reactor type was supposed to be a workhorse of nuclear education in the sixties, with dozens of units being produced and sold around the USA and overseas. However, nowadays only 4 units are fully operational, the rest has been mothballed or decommissioned. In 2010 Texas A&M University made the AGN-201M again part of its education program, after years of reactor inactivity. This very recent practical experience from Texas A&M shows that AGN-201M can be successfully used for undergraduate and graduate nuclear education and new educational methods can be implemented on the reactor. The paper demonstrates new experimental classes developed for the reactor and may serve as an idea for other mothballed AGNs or training reactors worldwide.
Small Modular Reactor and Large Nuclear Reactor Fuel Cost Comparison
Authors: Pannier Christopher P. , Škoda Radek
Published: November 2014

Small modular reactors (SMRs) offer simple, standardized, and safe modular designs for new nuclear reactor construction. Factory built SMRs promise competitive economy when compared with the current reactor fleet. Construction cost of a majority of the projects, which are mostly in their design stages, is not publicly available, but variable costs can be determined from fuel enrichment, average burn-up, and plant thermal efficiency, which are published design parameters for many near-term SMR projects. This paper gives a simulation of the fuel cost of electricity generation for selected SMRs and large reactors, including calculation of optimal tails assay in the uranium enrichment process. The fuel costs of several SMR designs are compared between one another and with current generation large reactor designs providing a rough comparison of the long-term economics of a new nuclear reactor project. SMRs are predicted to have higher fuel costs than large reactors. Particularly, integral pressurized water reactors (iPWRs) are shown to have from 15% to 60% higher fuel costs than large reactors. Fuel cost sensitivities to reactor design parameters are presented.
Effect of Support Friction in Thermal Stress of VVER-1000 RCS for NOP Condition
Authors: Nguyen Truong Quang , Namgung Ihn
Published: November 2014

The main purpose of this research is to investigate the effect of friction in the thermal stress of Reactor Coolant System (RCS) of VVER-1000. RCS is a large system connecting reactor vessel, steam generators and RC Pumps. During the heat-up of reactor, the RCS expand and during cool-down of reactor, it contracts. Because of the heavy weight of reactor and steam generator, the friction at the support of RCS affects the thermal stress of RCS. In this paper how much support friction contributes to the development of thermal stress is assessed in order to investigate the thermal stress and effect of support friction. A quarter-symmetry model of VVER-1000 RCS is developed in ANSYS and meshed with hexahedral elements to ensure better solution accuracies. The model includes reactor vessel, steam generator and reactor coolant pump. Internals of reactor vessel, steam generators and RCPs are represented by point mass to simplify the model. Temperature of inside surface of hot-leg side of reactor vessel to inlet side of steam generator is assumed same uniform hot-leg temperature, and the temperature of inside surface of outlet side of steam generator to reactor vessel is uniform cold-leg temperature. All outside surface are assumed insulated. The analysis includes neither transient thermal loading nor dynamic loadings. The analysis results show that friction at support brings little effect on the peak thermal stress. The peak thermal stress occurs at hot-leg nozzle of reactor pressure vessel and it approached near yield stress. If load combination is included the localized total stress at hot-leg nozzle could go over the yield stress. This peak stress could affect fatigue life in a long run. A recommendation is made that a detailed fatigue analysis of VVER-1000 RCS is necessary.
Study on DNB-Type Critical Heat Flux With Chopper-Cosine Axial Heat Flux Distributions
Authors: Zhao Dawei , Xiong Wanyu , Liu Wenxing , Xu Jianjun
Published: November 2014

Departure from nucleate boiling (DNB) type critical heat flux (CHF) is one of most important thermal criteria for nuclear reactor design. Concerning on the typical chopper-cosine heat flux profile at reactor core, it is of great significance to predict the CHF under non-uniform heating conditions for reactor design and the performance promotion of reactor system. Some correction factors are proposed for the prediction of CHF with non-uniform axial power shapes. In this study, a mechanistic DNB-type CHF model has been developed on the basis of liquid sublayer dryout mechanism. The non-uniform axial heat flux is taken into account of upstream memory effect on boiling crisis in this model. The predictions of present model and Tong’s non-uniform heat flux shape factor method are compared with the experimental results in the vertical tube with chopper-cosine axial heat flux distributions. The comparison results show the present model has fairly good prediction capability for DNB-type CHF under non-uniform heating condition.
The Implementing Geological Disposal Technology Platform: Key Challenges in Research and Development in Radioactive Waste Management
Authors: Delay Jacques , Slovak Jiri , Kowe Raymond
Published: November 2014

The Implementing Geological Disposal of Radioactive Waste Technology Platform (IGD-TP) was launched in November 2009 to tackle the remaining research, development and demonstration (RD&D) challenges with a view to fostering the implementation of geological disposal programmes for high-level and long-lived waste in Europe.

The IGD-TP’s Vision is that “by 2025, the first geological disposal facilities for spent fuel, high-level waste and other long-lived radioactive waste will be operating safely in Europe”. Aside from most of European waste management organisations, the IGD-TP now has 110 members covering most of the RD&D actors in the field of implementing geological disposal of radioactive waste in Europe.

The IGD-TP Strategic Research Agenda (SRA), that defines shared RD&D priorities with an important cooperative added value, is used as a basis for the Euratom programme. It provides a vehicle to emphasise RD&D and networking activities that are important for establishing safety cases and fostering disposal implementation. As the IGD-TP brings together the national organisations which have a mandate to implement geological disposal and act as science providers, its SRA also ensures a balance between fundamental science, implementation-driven RD&D and technological demonstration.

The SRA is in turn supported by a Deployment Plan (DP) for the Joint Activities to be carried out by the Technology Platform with its members and participants. The Joint Activities were derived from the individual SRA Topics and prioritized and assigned a timeline for their implementation. The deployment scheme of the activities is updated on a yearly basis.
Development of Numerical Simulation Method for Relocation Behavior of Molten Materials in Nuclear Reactors: Analysis of Relocation Behavior for Molten Materials With a Simulated Decay Heat Model
Authors: Yamashita Susumu , Takase Kazuyuki , Yoshida Hiroyuki
Published: November 2014

In order to precisely investigate molten core relocation behavior in the Fukushima Daiichi nuclear power plants, we have developed the detailed and phenomenological numerical simulation code named JUPITER for predicting the molten core behavior including solidification and relocation based on the three-dimensional multiphase thermal-hydraulic simulation models. In this paper, in order to distinguish a fuel component from core internals component in the JUPITER code, we added the multicomponent analysis method to the code and carried out to check effectiveness of the multicomponent analysis model based on the numerical simulation of melting behavior of the simulated fuel assemblies and core internals. From the present numerical results, it was confirmed that a newly developed multicomponent analysis method appropriately can predict the relocation behavior of molten materials in complicated structures, that is relocation and solidification behavior for fuel components including the heat source, i.e., simulated decay heat, and the core internals without heat sources from simplified fuel assemblies through the core support plate and simplified control rod guide tubes to the lower head in the reactor pressure vessel (RPV).