bluuebunny/crossref_metadata_2025 huggingface dataset:
Investigation of the Modified RELAP5 MOD 3.2 Capability to Simulate the Transients of the Fluoride-Salt High-Temperature Reactors
Authors: Liu Limin , Yan Qiqi , Zhang Dalin , Qiu Suizheng , Su G. H.
Published: October 2016

Fluoride-salt-cooled high-temperature reactors (FHRs) are a new concept that uses solid fuel and employs high temperature liquid salts as both primary and secondary side working fluids. The thermo-physical properties of the fluoride salts and specific heat transfer correlations have been implemented into the RELAP5-MOD3.2 code to enable the code to simulate the transient thermal-hydraulic response in accidents. The thermo-physical properties calculated by the modified code have been benchmarked versus the experimental data. The results of thermal-hydraulic responses of the PB-AHTR in the steady state and simple loss of forced flow from the publications using the RELAP5-3D are used to verify the ability of the modified RELAP5-MOD3.2 to simulate the transients of the FHRs. The code can predict the variations of the thermal-hydraulic parameters well whereas some results may distinguish from that calculated by the RELAP5-3D. The unprotected loss of flow is analyzed by the modified code, indicating that the passive residual heat removal system can mitigate the consequence of the accidents.
Investigation of Thermal Efficiency of Generic Pressure-Channel Reactors With Steam Reheat
Authors: Zhou Yifeng , Ponomaryov Paul , Mazza Cristina , Pioro Igor
Published: October 2016

Currently, i.e., in 2016, 4361 nuclear-power reactors operate in the world. 96.6% of these reactors are water-cooled (373 reactors (280 PWRs, 78 BWRs and 15 LGRs are cooled with light water and 48 reactors — PHWRs are cooled with heavy water. 15% of all water-cooled reactors are pressure-channel or pressure-tube design, the rest — pressure-vessel design. All current NPPs with water-cooled reactors have relatively low thermal efficiencies within 30–36% compared to that of current NPPs with AGRs (42%) and SFR (40%) and compared to that of modern advanced thermal power plants: combined-cycle plants (up to 62%) and supercritical-pressure coal-fired plants (up to 55%).

Therefore, it is very important to propose ways of improvement of thermal efficiency for this largest group of nuclear-power reactors. It should be noted that among six Generation-IV nuclear-reactor concepts one concept is a SCWR, which might reach thermal efficiencies within the range of 45–50% and even beyond. However, this concept has been never tested, and the most difficult problem on the way of implementation of this type of reactor is the reliability of materials at supercritical pressures and temperatures, very aggressive reactor coolant – supercritical water, and high neutron flux. Up till now, no experiments on behavior of various core materials at these conditions have been reported so far in the open literature.

As an interim way of thermal-efficiency improvement for water-cooled NPPs nuclear steam reheat can be considered. However, this way is more appropriate only for pressure-channel reactors, for example, CANDU-type or PHWRs. Moreover, in the 60’s and 70’s, Russia, the USA and some other countries have developed and implemented the nuclear steam reheat in subcritical-pressure experimental boiling reactors.

Therefore, an objective of the current paper is to summarize this experience and to estimate effect of a number of parameters on thermal efficiencies of a generic pressure-channel reactors with nuclear steam reheat. For this purpose the DE-TOP program has been used.
Development of Numerical Simulation Method for Melt Relocation Behavior in Nuclear Reactors: Validation of Applicability for Actual Core Support Structures
Authors: Yamashita Susumu , Tokushima Kazuyuki , Kurata Masaki , Takase Kazuyuki , Yoshida Hiroyuki
Published: October 2016

In order to precisely investigate molten core relocation behavior in the Fukushima Daiichi nuclear power station, 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. At the moment, multicomponent analysis method which can be treated any number of component as a fluid or solid body, Zr-water reaction model and simple radiation heat transfer model were implemented and showed that multicomponent melt flow and its solidification were confirmed in the simplified core structure system. However, the validation of the JUPITER using high temperature molten material has not been performed yet. In this paper, in order to evaluate the validity of the JUPITER, especially, for high temperature melt relocation experiment, we compared between numerical and experimental results for that system. As a result, qualitatively reasonable result was obtained. And also we performed melt relocation simulation on actual core structures designed by three dimensional CAD (Computer-Aided Design) and then we estimated phenomena which might be actually occurred in SAs.
Implementation of Fault Tolerant Control Scheme for Point Kinetic Reactor Model With Temperature Feedback in Matlab
Authors: He Huan
Published: October 2016

As a highly safety-critical system, it is insufficient for the Nuclear Power Plant (NPP) Instrumentation and Control (I&C) system to simply rely on a conventional control schemes or controllers which only satisfy stability and performance specification to the perturbation of the nominal plant. Since the current operating or newly-built I&C systems are based on transferring or adapting modern high performance electronic devices, it provides the hardware foundation and possibility to incorporate more advanced control systems into nuclear systems to achieve higher safety and stable performance, even in unexpected faulty situations. Active Fault Tolerant Control (FTC) is one of the choices for such advanced control. Active FTC encompasses the following components: 1. nominal controller design, 2. sensors and actuators fault detection and isolation, and 3. fault estimation and fault accommodation. In this research, approaches for each component are integrated into an active FTC scheme. Following this, the active FTC scheme is applied to a point kinetic rector model with fuel temperature and coolant temperature effect to reactivity. Simulation results show that the active FTC scheme designed in this research can effectively track the global power set point, even under situations with single fault from actuator or sensors.
Strategies to Prevent Benign Transients From Becoming Serious Accidents
Authors: Miranda Samuel
Published: October 2016

Nuclear safety criteria are based upon the concept that plant situations that are expected to have a high frequency of occurrence must not pose a danger to the public, and that plant situations that pose the greatest danger to the public must be limited to situations that have a very low expected frequency of occurrence. This concept is implemented by grouping plant situations (or events) into categories that are defined according to their expected frequencies of occurrence. It is important not to allow the boundaries of these categories to be crossed. Plant designs cannot allow low-consequence events of high expected frequency to develop into events of more serious, high-consequence categories.

The development of this system of frequency-based categorization is discussed, followed by an evaluation of various methods that have been proposed and applied to show how minor events could be prevented from becoming major events.
Scaling Analysis and Facility Design for Stability Investigation During Accidents in a PWR-Type SMR
Authors: Yan Yikuan , Shi Shanbin , Ishii Mamoru
Published: October 2016

Small modular reactor (SMR) concept has been developed as one of the key solutions for the growing demand of safe and clean energy. SMR designs can be applied extensively in areas such as sea water desalination and small-scale power generation etc. Unlike conventional light water reactors, most SMRs greatly simplify the structure of reactor pressure vessel, usually eliminate pumps and use natural circulation to cool down the core and transfer energy. However, flow instability may easily occur and affect the entire two phase natural circulation, which is of great importance for the start-up and normal operation process of BWR-type SMRs. For PWR-type SMRs, two-phase natural circulation could exist during accidents such as small break loss of coolant accident (SBLOCA) and loss of heat sink. Current research aims to experimentally investigate potential flow instabilities related to natural circulation for a PWR-type SMR during the accidents. For current research, the NuScale reactor design is selected as the research prototype.

In this paper, the design and scaling analysis of a scaled PWR-type experimental facility are provided. In order to experimentally study the natural circulation behavior of PWR-type SMR during accidental scenarios, detailed scaling analyses are necessary to ensure that the scaled phenomena could be obtained in a laboratory test facility. A three-level scaling method is used to get the scaling ratios derived from various non-dimensional numbers. An ideally scaled facility is first accomplished based on derived scaling ratios. RELAP5 simulations of both steady state and transient cases for the ideally scaled facility are performed and compared to the prototype to ensure the accuracy of the scaling analysis. Then the ideally scaled facility is modified under engineering considerations and an engineering scaled facility is designed. Similar RELAP5 analyses are performed on the engineering scaled facility and the results match well with those in the prototype and ideally scaled facility.
Experimental and Simulative Investigation of ADS-4 Depressurization Process in AP1000
Authors: Xiang Y. , Wu Y. W. , Qiu S. Z. , Tian W. X. , Su G. H.
Published: October 2016

Depressurization and Entrainment TEst Loop (ADETEL) modeled after AP1000 with a scaling ratio of 1:5.6 was constructed to investigate the entrainment and depressurization behavior after the actuation of ADS-4 valves. The entrainment and depressurization features were investigated under different initial pressure, mixture liquid level in the pressure vessel and heating power. The entrainment deposition effect of reactor internals was also investigated. The test data reveals that large amount of water are entrained through the ADS-4 branch line within a short period of time. The liquid entrainment rate and the reduced rate of the mixture liquid level in the pressure vessel increase with the initial system pressure. It is notable that the core uncovery was experienced when the initial pressure was set as 0.5MPa in current experimental conditions. The reactor internals have little effect on the entrained mass and the mixture liquid levels in the pressure vessel. The correlation has been fitted after the Welter.et al’ model for the entrainment onset and entrainment rate, which can be applied in RELAP5 as the simulation investigation.
An Experimental Investigation on Drainage Performance of a Drain Device in a Steam Generator of PWR
Authors: Gong Shengjie , Mei Yong , Gu Hanyang , Ying Binbin , Song Yinxi
Published: October 2016

This study evaluates the performance of drain device of a steam dryer in steam generator for PWR through experiments with water and air under ambient temperature and pressure. Two different test sections are designed and tested, one of which is equipped with a single drain pipe (test section I) and the other is equipped with double drain pipes (test section II). For the test section I, various slopes of bottom plate (3.5°∼5°) and different numbers of water entrance holes (36 ∼136) are investigated to study their effects on drainage performance of the drain device. For the test section II, only the influence of different entrance holes is studied. In experiments, flow fields in the drain device for different flow rates ranging form 13.8m3/h to 165.7m3/h are recorded using a video camera and the corresponding video images are processed by MATLAB to obtain the free liquid surface configuration. The experimental results show that the free liquid surface height of the test section I with single drain pipe is much bigger than that of the test section II with double drain pipes. In addition, the maximal free liquid surface height in the drain tank increases linearly with the increase of flow rate for both test sections while the number of entrance holes has no obvious effect on the maximal free liquid surface height. For the test section I, the bottom plate inclination has little influence on the maximal free liquid surface height and 4° inclination meets the optimized value.
Study on Thermal Efficiency of SuperCritical Water NPPs
Authors: Mazza Cristina , Ponomaryov Paul , Zhou Yifeng , Pioro Igor
Published: October 2016

As the demand for emission-free energy increases, the continued improvement of Nuclear Power Plants (NPPs) and their thermal efficiencies is crucial to fulfilling that demand. Current NPPs, especially, with water-cooled reactors, have significantly lower thermal efficiencies (32–36%) compared to those of modern advanced thermal power plants (55–62%). Even Generation-III+ water-cooled NPPs will have thermal efficiencies not higher than 37–38%. Therefore, to be competitive on the energy market, new nuclear reactors and NPPs, so-called, Generation-IV concepts, should be designed and commissioned.

The paper discusses the vital role that thermal efficiency plays with respect to how far nuclear reactors can be more cost efficient and competitive. An evaluation of thermal efficiencies has been carried out for SuperCritical Water (SCW) NPPs with Rankine “steam”-turbine power cycle using the IAEA DEsalination Thermodynamic Optimization Program (DE-TOP). Various options for improving thermal efficiencies of SCW NPPs have been studied. This study was performed in support of possible designs of the first experimental SCW reactors.
Development and Technical Supporting System for Smart Grid in China
Authors: Lin Wang , Yujun Zhang
Published: October 2016

The Smart Grid is the further development and improvement of power system, with relating characteristics of informatization, digitization and automation as well as interaction, applications including generating, transmission and dispatching and also transformation and distribution of power. Based on the existing and high speed of communication network, the technical supporting system of Smart Grid in China is built with application of modern equipment technology, control method and decision support.
Stability Validation Based on the OECD/NEA Oskarshamn-2 BWR Benchmark With Toshiba TRAC Version
Authors: Fukunaga Takanori , Yoshida Noriyuki , Mototani Akira
Published: October 2016

We have developed a best-estimate transient code Toshiba TRAC version, a Toshiba advanced version of the TRAC code for BWR. This code is expected to be capable of simulating thermal hydraulic and neutron kinetics phenomena three-dimensionally with high accuracy for BWR transients. A licensing topical report for this code’s validation analyses will be submitted to the Japanese regulatory body for application to future licensing safety analysis. As one of the validation analyses, analysis for BWR stability is also important, and this code’s validation analyses for BWR stability based on the international OECD/NEA Oskarshamn-2 BWR benchmark have been implemented. The Oskarshamn-2 BWR experienced a core instability event, which culminated in diverging power oscillations with a decay ratio greater than 1.3 on February 25, 1999. This paper shows this code’s validation analyses of the core instability event. In conclusion, the initial conditions and the results of the instability event are also in good agreement with the measured data. This code is capable of simulating instability events well.
GRAPE: Graphical RELAP/SCDAPSIM Analysis Platform for Education and Engineering
Authors: Tatsumi Masahiro , Tsujita Kosuke , Sato Kotaro , Tamari Yohei
Published: October 2016

This paper describes development of GRAPE, a plant simulator for introductory education of nuclear engineering and nuclear safety. GRAPE is not a full-featured real-time simulator but a kind of scenario-based analysis and visualization platform using a best-estimate code such as RELAP/SCDAPSIM. The PWR and BWR plant models which can be adopted to the scenarios under severe accident conditions are also developed and maintained.

GRAPE is designed and implemented by taking advantage of recent technologies of the Web such as HTML5, CSS and JavaScript in combination with the modern software architecture. Intuitive graphical user interface was established to meet user needs while maintaining flexibility, extensibility and maintainability. It is also effective for study and research projects as visualization and/or sensitivity platforms since GRAPE can import existing RELAP-based calculation results using a converting utility program to visualize the calculation results. A macro language is available to automate the workflow in sensitivity analysis and make the process more efficient.
Effect of Film Cooling Holes Shape of Helium Turbine Blades in HTGR
Authors: Sun Xiaokai , Peng Wei
Published: October 2016

The effect of holes shape on film cooling was investigated numerically in the present study. Three cases of holes which are sister holes located in the downstream (case 1), upstream (case 2) and side (case 3) were studied when keeping the flow area and flow rate constant. Cylindrical hole was also investigated to make a comparison. The results showed film cooling effectiveness is higher at the blowing ratio M = 0.5 for cylindrical hole; while for the hole with sister holes, film cooling performance is better at corresponding blowing ratio M = 1.0 and it is better than that of the cylindrical hole. The reason is that the interaction between the mainstream and the cooling stream produces the counter-rotating vortex pairs (CVP) at high blowing ratio, which will make the hot mainstream underneath the cooling stream and lifts the cooling stream from the surface for the cylindrical hole. And the three cases of holes with sister holes can weaken the effect of the CVP.
Heat and Mass Transfer Analogy of Steam Condensation With Air Over Vertical External Surface
Authors: Su Jiqiang , Wu Yuxiang , Huang Shuliang , Xu Huiqiang , Zhou Yanmin
Published: October 2016

During the steam condensation, the presence of non-condensable gases is an important issue affecting the efficiency of the whole thermodynamic process. For this reason, many researchers investigated it by theoretical or experimental methods. A heat and mass transfer analogy model on steam condensation in presence of air over the vertical external surface based on the diffusion layer model is modified in the present paper. Based on previous authors’ experience, the suction effect at the gas-liquid interface and other analogy drawbacks are identified and overcome by supplementing it with more detailed analysis as well as targeted experiments. The experimental data obtained for condensation, outside vertical tube with an external diameter of 38 mm, of air/steam and helium/air/steam mixture, have been used to verify the present heat and mass transfer analogy formulation. By comparing against different available experimental data and previous formulations, the heat and mass transfer analogy formulation is demonstrated to be a accurate enough theoretical approximation. The deviation between predicted values of the new model and experiment results of this paper is less than 15% which has relative higher precision.
Thermal Mixing of Liquid Metal Coolant in an Annular Channel
Authors: Kashinsky O. N. , Lobanov P. D. , Pribaturin N. A. , Kurdyumov A. S. , Vorobyev M. A.
Published: October 2016

Results of experimental studies of mixing of liquids with different temperatures are presented. Experiments were performed in a T-junction connected to an annular channel. Rose’s metal was used as the test liquid. The test section was made of thin wall stainless tubes. The distribution of wall temperatures was measured using a high speed infrared camera. Experiments were performed in transient regimes.

Results obtained provide experimental data necessary for validation of thermal hydraulic codes for heavy liquid metals.
Implementation of Small Punch Testing and Automated Ball Indentation in the Process of Irradiated NPP Materials Degradation Evaluation
Authors: Kopriva Radim , Eliasova Ivana , Kytka Milos
Published: October 2016

In the terms of nuclear power plant operational life management, current trend of components lifetime extension requires precise and credible information of structural material degradation. Present-day standard conventional methods of mechanical testing are usually based on the use of large specimens and higher consumption of testing material, whose availability and volume is often limited.

For determination of material properties, sampling of the necessary volume of material is in most cases connected with affecting the integrity or even destruction of the assessed component. Moreover, several components are not usually covered surveillance programs, e.g. reactor pressure vessel internals.

Innovative testing methods of Small Punch Testing (SPT) and Automated Ball Indentation Test (ABIT) are based on the determination of material properties from miniaturized testing specimens and their semi-destructive approach is very promising for the possibility of present data base of irradiated materials testing results enlargement and enable the option of component in-situ testing (ABI testing).

Paper describes the process of implementation of Small Punch Testing and Automated Ball Indentation Testing techniques in the evaluation of mechanical properties of irradiated NPP structural materials degradation. Presentation depicts all necessary steps carried out at the Hot Cell facility of UJV Rez, a. s. (Integrity and Technical Engineering Division) for the employment of these innovative testing techniques to the portfolio of accredited mechanical tests performed within the frame of surveillance programs of WWER type reactors — e.g. preliminary design, modification of the testing equipment, fractography analyses etc.

Detailed comparison between the results of standard conventional testing techniques and SPT and ABIT for both initial unirradiated and irradiated state for the material A533B (JRQ) is included for determination of suitability of used methods for testing of materials within the surveillance programs of WWER type reactors.
Experiment Research of Vapor Bubble Stochastic Characteristics in a Narrow Rectangular Channel
Authors: Li Shaodan , Li Yong , Lin Yuansheng , Wei Zhiguo , Li Bangming , Wang Junrong
Published: October 2016

In most of the previous studies, the vapor bubble parameters were averaging processed in order to establish the empirical-based correlations or mechanism-based models. While it should be noted that the thermal hydraulic parameters around a bubble in a subcooled flow boiling channel are unsteady, that means the vapor bubbles behaves some stochastic characteristics. In the present research, a high speed visualization experiment was conducted in a subcooled flow boiling narrow rectangular channel to investigate the vapor bubble behavior. The working fluid in the experiment was deionized water. The obtained bubble images were processed by a digital image processing program in order to acquire the bubble parameters. The stochastic features of the bubbles were analyzed based on the experimental results. Two types of bubble behaviors were observed under different working conditions, which results in two types of bubble stochastic features. The results shown that the distribution function of the bubble maximum diameter in a specific nucleation site can be expressed by the normal distribution, while in the whole range of the observation window the distribution of the bubble maximum diameter was expressed by the lognormal distribution. The distribution of the second type bubble diameter depends on the local bubble nucleation site and the upstream bubble behaviors.
Heat Transfer Simulation of the Fuel Transport Cask for Spherical Fuel Elements in Molten Salt Reactor
Authors: Liu Yang , Wang Jun
Published: October 2016

Fuel transport is an indispensable task for nuclear power plants. For pressurized water reactors (PWR) and boiling water reactors (BWR), many research projects have been completed for designing and testing the transport casks for fresh fuel as well as spent fuel [1–3]. To ensure the safety of nuclear fuel during the transportation, many aspects should be analyzed and examined for the casks with fuel inside, such as heat transfer and temperature calculation, radiation protection, nonproliferation issues, etc. The transport cask discussed in this paper is especially for new spherical fuel elements, which should be designed in accordance with the stipulations in the GB11806 Regulations for the Safe Transport of Radioactive Material [4].

The Transport Cask for spherical fuel elements used in molten salt reactor (MSR) should be designed in accordance with the safety standards for transport of radioactive material. It is necessary to evaluate the thermal performance of the transport cask separately in normal transport condition and in accident transient. The MSR fuel sphere elements cask is in a circular cylinder shape and composed of inner container and outer shell cask. The objective of the thermal analysis of the cask under hypothetical accident conditions is to demonstrate that the cask containment boundary structural components are maintained within their safe operating temperature ranges. The heat transfer process (conduction, convection, and radiation) is simulated by ANSYS-CFX in this paper and it is demonstrated that the components of cask are maintained in safe operating temperature ranges. The calculation results are below limit temperatures, indicating that the thermal design of the cask could meet the Standard Regulations. The result is also compared with the fire test, which shows the calculation model is conservative and rational.
Three Dimensional Numerical Simulation of Thermal-Hydraulic Behaviors of CSNS Decoupled Poisoned Hydrogen Moderator With Non-Uniform Heat Source
Authors: Tong Jian-Fei , Wang Jiao-Long , Zhang Hao-Chun , Ji Yu , Tan He-Ping
Published: October 2016

The Chinese Spallation Neutron Source (CSNS) is the national major scientific equipment, which will be completed around 2018. The moderator is the core of the CSNS parts, and CSNS uses three types of moderator to meet the requirements of different neutron wavelength and pulse shapes, and the structure and flow field of Decoupled Poisoned Hydrogen Moderator (DPHM) is the most complicated. Currently, Computational Fluid Dynamics (CFD) is the primary tool for thermal design of DPHM to ensure the reliability of calculation result and precision of the engineering requirements. In the current work, three-dimensional simulation of fluid-solid conjugate heat transfer and flow for container pipe, poisoned plate and moderator material were carried out, by compiling corresponding UDF program for ANSYS FLUENT to describe the non-uniform heat source distribution accurately. Thermal-Hydraulic behaviors such as pressure disturbance, pressure rise and temperature distribution cooling capacity are investigated.
Research on the Occurring Location of ONB in Narrow Rectangular Channel Under Natural Circulation by Factorial Analysis Method
Authors: Qi Shi , Zhou Tao , Chen Baixu , Xiao Zejun
Published: October 2016

This paper describes the research on the location of Onset of Nucleate Boiling (ONB) in Natural circulation. Selecting the narrow rectangular natural circulation loop as the research object, the occurring location of ONB is analyzed by experiment. By the method of factorial analysis, the influences of different factors and the factor interaction to the location of ONB are studied. The results show that, the inlet subcooling has the biggest influence, whose contribution percentage is 49.8%. The second important influence is the gap size, whose contribution percentage is 25.2%. The third important influence is the heat flux, whose contribution percentage is 24.4%. The influence of factor interaction to the location of ONB are negligible.
Assessment on an Intelligent Controller of Marine Condenser
Authors: Xian Gong , Jianhua Sun , Hai Zou
Published: October 2016

At present, simulation research on condenser mostly focuses on the one of power plant. While little attention is paid on the marine one. The marine condenser is different from the one of power plant. One difference between them is that marine condenser limited by the hull space is much smaller than the one of power plant. Another difference is that marine condenser usually runs in condition that steam load changes greatly and frequently. In addition, the temperature of cooling water changes with seasons and maritime space. When the ship makes an ocean-going voyage, the temperature of cooling water varies obviously, which has a negative effect on the control system. Therefore, simulation research to the dynamic characteristic of marine condenser is carried on by this paper.

The vacuum and the condensate supercooling degree are the main operating parameters for a condenser. In this paper, a mathematical model for marine condenser is established. The dynamic variation of each parameter is monitored by this model. Then an expert control system is designed. The condenser vacuum and supercooling degree are controlled by the flow rates of cooling water, air-blend and steam jet for deaerating. The deaerating steam bubbles which are not absorbed completely by condensate water, have a negative impaction on the condenser vacuum. The deaerating steam bubbles heating the condensation water to be exactly saturated is calculated. Referring to it, the fluctuation of condenser vacuum is filtered. So the negative influence on condenser control system can be remitted. The temperature of water replenishing is usually lower than that of condensate water in condenser heat well. And the water is replenished into the heat well part. So water replenishing would increase the supercooling degree of condensate water.

This paper adopts the simulation experiment method to gain the expert knowledge of condenser control. According to power level, inlet temperature of cooling water, the messages of operating modes and sensors, the control system regulates the regulating valves for turbo circulating pump of cooling water, air extractor and steam jet deaerator to change the flow rates of cooling water, air-blend and steam jet for deaerating, so as to achieve automatic control of condenser vacuum and supercooling degree. Finally, two PID controllers are added as well, to counteract the adverse effect, enhance anti-interference ability and optimize control performance of intelligent control system. Simulations of varying load and cooling water temperature are done. Results show that the intelligent control system could effectively control condenser vacuum and supercooling degree under different power level and variable temperature of cooling water condition.
3D Linear Bifurcation Analysis of Steel Containment Vessel
Authors: Schmidt Jay , Laskewitz Bernd
Published: October 2016

The AP1000® Containment Vessel (CV) is a freestanding steel containment designed to protect the public from radiation release. The CV consists of 2 ellipsoidal heads connected by a cylindrical shell and is constructed of carbon steel. The AP1000 plant design has four large penetrations (two airlocks and two equipment hatches) located in approximately the same quadrant of the circumference of the shell which imposes asymmetric effects in the shell structure. The CV is designed and constructed in accordance with ASME Boiler and Pressure Vessel Code, Section III, Subsection NE.

Traditionally, the local and global stability of freestanding steel containments have been designed by use of formulae using conservative assumptions based on an axisymmetric structure. ASME Code Case N-284 “Metal Containment Shell Buckling Design Methods, Class MC Section III, Division 1” outlines methodology for satisfying the stability of the CV using two approaches. Section 1710 provides a stress based buckling approach using detailed formulae that assumes an axisymmetric structure. The second approach provides guidance and acceptability based on a linear bifurcation analysis (2D (1720) or 3D (1730)). Due to the asymmetric structure of the CV, the 3D linear bifurcation method delivers the most accurate results.

The methodology and assumptions implemented by Westinghouse to qualify stability of the CV via Code Case N-284 are outlined. Also, the procedure to properly amplify the stresses as required by N-284 is included as justification of the methods used. This justification was thoroughly investigated by the Nuclear Regulatory Commission (NRC) and deemed acceptable.
Experimental Study on Heat Transfer Characteristics of Circular Jet Impingement Boiling on the Variety of Structured Copper Surfaces in Stagnation Zone
Authors: Modak Mayank , Nirgude Vishal , Sharma Avadhesh K. , Sahu Santosh K.
Published: October 2016

In the present work an attempt has been made to study the heat transfer characteristics of single circular jet on a variety of enhanced surfaces. In the present investigation three different copper target surfaces of various surface modifications: bare copper surface, pin fin enhancement surface and a flat surface coated with alumina porous layer. The heat transfer performance of each surface is studied in two phase boiling operation at different flow rates (3959 < Re < 7900). The comparison indicates that both the surface modification have enhanced the boiling heat transfer rates.
The Comparison of the Transducer Circuits for the Two-Phase Flow Conductivity Probe
Authors: Yao Zhipeng , Ge Liangjun , Abbas Kamil , Shiliang Zhou
Published: October 2016

The void fraction of two-phase flow can be measured by conductivity probe due to the difference of conductivity between air and water. It’s difficult to detect the conductivity of the probe directly so transducer circuits should be used to transfer the conductivity to voltage, and the voltage can be sampled by DAQ device. Two transducer circuits are designed for two-phase flow double sensor conductivity probe. One is a simple DC circuit, it’s made up of a constant DC voltage and a resistance, and the resistance is cascaded with the probe. Voltage drop across the resistance is depend on conductivity of surrounding media of probe tip. The other is AC circuit that utilizes a DDS (Direct Digital Synthesizer) as an excitation power supply which links to a transimpedance amplifier through the probe. The current that flow to the amplifier would change according to the surrounding’s medium of the probe. The transimpedance amplifier transfers the current to voltage. Then we use AD8302, an amplitude detection chip, to detect the amplitude of the s output.

The experiments show that the DC circuit has a higher SNC (signal to noise ratio) and shorter responding time. However, the needle of probe is covered with electrochemical corrosion products and its performance degrades significantly when DC circuit is used and the probe is immersed in 80°C hot water for 8 hours. This kind of phenomenon is not appeared on the probe when DDS-based circuit is used. The result of the experiment verifies the SNC and responding time of DDS-based circuit is approximately equal to that of the DC circuit.
Analysing Laminar Flow Numerical Simulation Error of Fluent in Circular Tube Based on Hagen Poiseuille Formula
Authors: Wang Yaoxin , Zhou Tao , Zhou Lanyu , Fang Xiaolu
Published: October 2016

Firstly, using Fluent laminar model simulate volumetric flow rate and pressure differential. second, using Hagen Poiseuille formula calculate coefficient of dynamic viscosity. Then we can calculate laminar flow numerical simulation error of fluent by contrasting the difference of coefficient of dynamic viscosity between laminar simulative results of Fluent and calculated results of Hagen Poiseuille formula. Geometric model is Circular pipe. Fluid medium is water. The study found three conclusions. Fluent model relies on simplified mathematical formula, which can not accurately describe real flow situation, so Fluent simulation results is not accurate; When inlet velocity is at the range of 0.06m/s to 0.15m/s, Given coefficient of dynamic viscosity by Fluent model is always smaller than calculated coefficient of dynamic viscosity by Hagen Poiseuille formula; Laminar flow numerical simulation error of Fluent in circular tube gradually increase along with the increase of inlet velocity at the range of 0.06m/s to 0.15m/s.
Experimental Study on Heat Transfer Enhancement Method Using Porous Material With High Porosity
Authors: Asari Yuta , Takeda Tetsuaki , Funatani Shumpei
Published: October 2016

Heat transfer enhancement method using fins is well known and widely used. However, it needs to consider the loss of structural strength on the heat transfer surface by attaching fins. Therefore, heat transfer enhancement method using porous material had been proposed as the alternative. This paper’s objective is to study the pressure drop and heat transfer enhancement in a horizontal circular pipe after insertion of high porosity material. This material was cut off in three blocks and inserted with interval in a horizontal pipe to suppress the pressure drop. We had conducted to study the influences of change with porosity and wire diameter on the heat removal and the pressure drop. Experimental results show that inserting the porous material with porosity of 0.997 into the horizontal smooth pipe with a constant heat flux can enhance the heat transfer by about 30%.
Investigation on Thermal Treatments of High-Resistance Austenitic Stainless Steels Employed as Structural Materials in Generation IV Reactors
Authors: Mazzei Giulia B. , Lo Frano Rosa , Stubbins James F.
Published: October 2016

Many R&D programs are being conducted worldwide to study and, in some cases, develop materials with the properties required by Gen IV reactor systems. These advanced reactor systems require materials having sufficient dimensional stability and mechanical properties to meet the anticipated near- and long-term service demands under extreme conditions. These conditions are most challenging in environments with high levels of irradiation damage coupled with requirements for corrosion resistance and severe mechanical loading conditions. Based on past experience, austenitic alloys based around 316SS compositions, 15% Cr-15% Ni-Ti, have not been able to meet these demanding conditions at elevated temperatures. Ferritic-martensitic steels and ferritic-martensitic ODS steels have been proposed for these conditions. However, long-term fast neutron irradiation results show that austenitic alloys of the 15%Cr–25%Ni-Ti-Nb class have excellent swelling resistance at temperatures up to 600°C and at doses larger than 150 dpa. A major reason for this excellent behaviour is the thermal-mechanical treatment of the alloy and the presence of both Ti and Nb to doubly stabilize the precipitate structure.

This material could represent an effective option for claddings in SFR and other advanced reactor applications. For this reason, in consideration of 60-y-of-service requirement, predictions of long-term materials behaviour will be carried out taking into account ageing effects (from 500° to 700°C). A complete analysis aiming to characterize the microstructure of the as-received material, cold-worked at 20%, and also after thermal treatment is under way. Ageing tests as well as TEM and SEM analysis of the microstructural evolution as a function of aging condition will be performed. The evolution of some phases in the common nucleation and growth sites will be presented.
Cryogenic Hydrogen Cooling of Heated Moderator Vessel
Authors: Nilsson Per , Pucilowski Mateusz , Bessler Yannick
Published: October 2016

In ESS, a pulsed proton beam of 5 MW mean power will hit a tungsten target to generate neutrons by spallation. The pulses are 2.86 ms long and occur with 14 Hz; the power within a pulse is 125 MW. Only centimeters from the target, the neutrons are moderated by liquid hydrogen in aluminium vessels.

The deposited power heats the surrounding structures and fluids which are circulated and cooled. The hydrogen is operating at 15 bar and average temperature between 17 and 21 K, i.e. above the critical pressure 12.8 bar, but below the critical temperature 32.9 K. During the pulses, the peak heat deposition in the aluminium is 15 W/cm3 and in the hydrogen 4 W/cm3.

If the cooling of the aluminium is neglected during one pulse, the temperature increases to 34 K. That is above the critical temperature, where physical properties change strongly with temperature. Therefore the conjugate heat transfer has to be investigated in detail.

This work includes 1D principal transient calculations of a general configuration as well as CFD simulations of the heating and cooling of a specific design. The 1D calculations are performed using GNU/Octave and the CFD using ANSYS/CFX. It is concluded that with an inlet temperature of 17 K, the wall temperature can be kept below the critical temperature in the general configuration and sufficient cooling can be ensured in the investigated specific design.
Heat Transfer Coefficient Measurement for Downward-Facing Heat Transfer on Curved Rectangular Water Channel
Authors: Jung Jun Yeong , Jeong Yong Hoon
Published: October 2016

In-Vessel Retention by External Reactor Vessel Cooling (IVR-ERVC) is method of removing the decay heat by cooling reactor vessel after corium relocation, and is also one of severe accident management strategies. Estimating heat transfer coefficients (HTCs) is important to evaluate heat transfer capability of the ERVC. In this study, the HTCs of outer wall of reactor vessel lower head were experimentally measured under the IVR-ERVC situation of Large Loss of Coolant Accident (LLOCA) condition. Experimental equipment was designed to simulate flow boiling condition of ERVC natural circulation, and based on APR+ design. This study focused on effects of real reactor vessel geometry (2.5 m of radius curvature) and material (SA508) for the HTCs. Curved rectangular water channel (test section) was design to simulate water channel which is between the reactor vessel lower head outer wall and thermal insulator. Radius curvature, length, width and gap size of the test section were respectively 2.5 m, 1 m, 0.07 m and 0.15 m. Two connection parts were connected at inlet and outlet of the test section to maintain fluid flow condition, and its cross section geometry was same with one of test section. To simulate vessel lower head outer wall, thin SA508 plate was used as main heater, and test section supported the main heater. Thickness, width, length and radius curvature of the main heater were 1.2 mm, 0.07 m, 1 m and 2.5 m respectively. The main heater was heated by DC rectifier, and applied heat flux was under CHF value. The test section was changed for each experiment. The HTCs of whole reactor vessel lower head (bottom: 0 ° and top: 90 °) were measured by inclining the test section, and experiments were conducted at four angular ranges; 0–22.5, 22.5–45, 45–67.5 and 67.5–90 °. DI water was used as working fluid in this experiment, and all experiments were conducted at 400 kg/m2s of constant mass flux with atmospheric pressure. The working fluid temperatures were measured at two point of water loop by K-type thermocouple. The main heater surface temperatures were measured by IR camera. The main heater was coated by carbon spray to make uniform surface emissivity, and the IR camera emissivity calibration was also conducted with the coated main heater. The HTCs were calculated by measured main heater surface temperature. In this research, the HTC results of 10, 30, 60 and 90 ° inclination angle were presented, and were plotted with wall super heat.
Impact Resistant Design for FLEX Building Equipment Door
Authors: Haider Shakir Y. , Calhoun David J.
Published: October 2016

After the 2011 Japan earthquake and tsunami caused significant damage at Fukushima Daiichi, the Nuclear Regulatory Commission required all US nuclear power plants to have a mitigation strategy for beyond design basis events. Industry-developed response plans, called “FLEX” strategies, deployed new, portable equipment such as diesel generators and cooling pumps. As this new equipment needs to be available after a natural disaster, storage in protected locations is required. Many nuclear plants have recently constructed new storage buildings, or FLEX buildings, as part of their post-Fukushima strategy.

The equipment door is a critical component of a FLEX building. Large enough to drive a semi-trailer truck through, it is required to protect the equipment in case of an earthquake, flood, tornado and also may need to be capable of opening within a few minutes in order to respond during an emergency. The equipment door presented in this paper serves these purposes very effectively. The composite section of the door is capable of shielding the structure from penetration as well as overall dynamic response from tornado missile impact. The door travels on an overhead rail which, being indoors and above the opening, provides reliable door movement in case of snow or ice during winters or in case of debris from wind or tornado. Latches capable of withstanding tornado missile impact forces also restrain the door in case of wind or seismic forces. The door is opened by means of motorized trollies and is also equipped with a backup opening device by means of an air winch in case of a power loss.

The door and the latches that restrain the door from impact are analyzed using ANSYS finite element software. A limit state analysis is performed that identifies the sequence of yield limit states for the components of the door and the door latch as the loading progresses. The analysis continues until the ductility limit state for the latch is reached. Redistribution of stresses within the components of the latch is observed during the analysis. A modal analysis and a direct integration time history analysis is also performed to capture the dynamic response due to impact loading.

Overall, this paper presents a highly robust and reliable design for a FLEX building equipment door that is capable of protecting the contents of the building during a natural disaster and remaining operational during the response after an emergency.
Coupled Reactor Kinetics and Heat Transfer Model for Fluoride Salt-Cooled High-Temperature Reactor Transient Analysis
Authors: Wang Xin , Huff Kathryn D. , Aufiero Manuele , Peterson Per F. , Fratoni Massimiliano
Published: October 2016

Coupled reactor kinetics and heat transfer models have been developed at the University of California, Berkeley (UCB) to study Pebble-Bed, Fluoride-salt-cooled, High-temperature Reactors (PB-FHRs) transient behaviors. This paper discusses a coupled point kinetics model and a two-dimensional diffusion model. The former is based on the point kinetics equations with six groups of delayed neutrons and the lumped capacitance heat transfer equations. To account for the reflector effect on neutron lifetime, additional (fictional) groups of delayed neutrons are added in the point kinetics equations to represent the thermalized neutrons coming back from the reflectors. The latter is based on coupled multi-group neutron diffusion and finite element heat transfer model. Multi-group cross sections and diffusion coefficients are generated using the Monte Carlo code Serpent and defined as input in COMSOL 5.0.