bluuebunny/crossref_metadata_2025 huggingface dataset:
Thermal Hydraulic Analysis of Pressurized Thermal Shock for Loviisa NPP Using Apros Simulation Code
Authors: Žemulis Gintaras , Junninen Pasi , Kytömäki Petri
Published: October 2018

Loviisa Nuclear Power Plant consists of two VVER-440 type pressurized water reactor units and it is owned by a Finnish energy company Fortum Power and Heat Oy. Only some years after Loviisa Unit 1 start-up, during early 1980’s, some reactor pressure vessel embrittlement by fast neutron irradiation was recognized. Embrittlement of the reactor pressure vessel materials increases the risk of a fast fracture under emergency core cooling conditions. This led to an extensive study of pressurized thermal shock (PTS) scenarios, which included experimental research, large set of probabilistic and deterministic safety analyses, code development and plant modifications to both lower overcooling transient probability and to mitigate it’s consequences. Current operating licenses for Loviisa 1 and 2 Units reactor pressure vessels were applied by Fortum in 2012 to cover operation up to the years 2027 and 2030, respectively. The operating licenses were granted as applied by the Finnish Radiation and Nuclear Safety Authority (STUK) according to application. However a know-how of PTS scenarios shall be maintained. This becomes even more important in a light of any big power plant automation modernization projects targeting also safety systems. Currently Fortum is involved in an automation renewal project also known under its ELSA-project name. The project is scheduled to be ready in 2018.

The project targets at replacing parts of old analogue automation with new digital systems and also extending their diversity by introducing new safety functions for emergency management scenarios. As a result, changes to some emergency operating procedures are inevitable. Due to the changes to emergency operating procedures PTS-scenarios for Loviisa NPP have been also updated.

In 2016 Fortum launched PTS-project in connection with ELSA-project. Previous PTS-scenario related thermal hydraulic analyses were based on the original 80’s decade analyses, which were subsequently revised many times following other modernization projects that took place at Loviisa NPP. In this new round of PTS-analyses the most important cases were recalculated using Apros (Advanced Process Simulator). ELSA-project introduced changes are not only emergency procedures specific but also involve safety system related process and automation logic modifications. The changes affected therefore boundary conditions needed in the analyses. In total probabilistic safety analyses included over 160 different PTS-sequences. The most limiting sequences were selected for deterministic analyses. PTS-analyses were performed in co-operation with Platom Oy and plant owner Fortum Power and Heat Oy. The main results of PTS-analyses are presented in the paper.
Numerical Study of Effect of Contact Angles on Flow Boiling Employing Thermal LBM Simulation
Authors: Sun Tingzhen , Liu Qian , Gui Nan , Yang Xingtuan , Tu Jiyuan , Jiang Shengyao
Published: October 2018

The influence of contact angle on bubble growth and detachment is investigated in this paper. The phase-change Lattice Boltzmann Method (LBM) which includes a SRT pseudo-potential LB model and a thermal LB model is used to simulate the flow boiling in vertical tube. To verify the correctness of the model, the coexistence curve obtained from the LBM simulations is compared to the analytical one. Then the relation between the diameter of bubble detachment and the wall superheat is compared with the empirical relation. The effect of contact angle on the bubble growth is investigated. The bubble growth with different contact angles is calculated. The bubble growth processes with different contact angles and the detached shape are shown in this paper. The bubble equivalent diameter and the length of contact line is investigated. The bubble equivalent diameter curve shows that bubble growth can be divided into two stages, initially isothermal growth stage, followed by isobaric growth stage. Bubbles show different characteristics at those two stages. The length of contact line curve shows there is a period of stagnation in the process of bubble growing up. This phenomenon can be explained by the theory of dynamic contact angle.
Investigation of Supercritical CO2 Thermal Hydraulic Characteristics in a Printed Circuit Heat Exchanger
Authors: Fu Wen , Ma Xizhen , Li Peiyue , Zhang Minghui , Li Sheng
Published: October 2018

Printed circuit heat exchangers are considered for use as the intermediate heat exchangers (IHXs) in high temperature gas-cooled reactors (HTGRs), molten salts reactors (MSRs) and other advanced reactors. A printed circuit heat exchanger (PCHE) is a highly integrated plate-type compact heat exchanger with high-temperature, high-pressure applications and high compactness. A PCHE is built based on the technology of chemical etching and diffusion bonding. A PCHE with supercritical carbon dioxide (CO2) as the working fluid was designed in this study based on the theory correlations. Three-dimensional numerical analysis was then conducted to investigate the heat transfer and pressure drop characteristics of supercritical CO2 in the designed printed circuit heat exchanger using commercial CFD code, FLUENT. The distributions of temperature and velocity through the channel were modeled. The influences of Reynolds number on heat transfer and pressure drop were analyzed. The numerical results agree well with the theory calculations.
An Experimental Study on a Straight-Channel Printed Circuit Heat Exchanger for Supercritical CO2 Power Cycle Applications
Authors: Xu Aiwei , Huang Yanping , Wang Junfeng
Published: October 2018

A kind of compact plate-type heat exchanger, namely, Printed Circuit Heat Exchanger (PCHE) is one of the attractive options for S-CO2 Brayton Cycles, Because it can withstand higher temperature and pressure and has high surface-area-to-volume. The experiments were conducted for a NPIC straight-channel PCHE. In current study, we chose water as cold side fluid and S-CO2 as hot side fluid. Firstly, we fixed the cold side operation parameters to study heat transfer and pressure drop characteristics of the S-CO2 fluid side. Then we fixed the hot side operation parameters to study the pressure drop characteristics of the water side. Finally, existing heat transfer and friction correlations were used to compared with NPIC straight-channel PCHE experimental data and new correlations were developed.
Analysis on Pellet-Cladding Interaction of Fuel Rod During Power Ramp of NHR200-II
Authors: Kang Tianshan , Li Songyang , Wang Dingqu , Jiang Yueyuan , Li Weihua
Published: October 2018

In order to ensure the safety of fuel rods in nuclear reactor, it is necessary to consider the condition of the power ramp during reactor operation, which may cause breakage risk of fuel rod at the time of pellet-cladding interaction (PCI) appearing. To analyze this phenomenon and reduce the risk, a performance analysis model for fuel rod is developed to carry out the steady state and transient simulation using commercial software COSMOL. The full model has three main computational models, which are heat transfer model, mechanical model and internal pressure model. The calculation results show that the stress and strain of NHR200-II fuel rods has enough margins during the power ramp process, which indicates that the risks of fuel rod damage are very low.
Development of the LMFBR Subchannel Analysis Code ATHAS-LMR and Analysis of Blockage Accident
Authors: Du Peng , Shan Jianqiang , Zhang Bo
Published: October 2018

ATHAS is the subchannel analysis code, independently developed by Nuclear Safety and Operation Laboratory (NUSOL) of Xi'an Jiao Tong University (XJTU), the original version of the ATHAS is only suitable for pressurized water reactor. This paper is aimed at the characteristics of sodium cooled fast reactor core mentioned above, developing the analysis program AYHAS-LMR, which is applicable for LMFBR.

In the design of sodium cold fast reactor, flow blockage must be considered to meet the design criteria. In view of this, a new model is proposed in this paper and added in ATHAS-LMR, considering the flow blockage which changes the flow area and the wetted perimeter. This model is based on the original grid model, we use the grid to simulate the presence of block, increase the resistance of the blockage area artificially. Then, ATHAS-LMR is able to simulate the flow blockage in LMFBR.

In this paper, simulations of FFM5B blockage test were done for the high flow case with ATHAS-LMR, and we compare the simulation results with COMMIX-1. The result indicates that ATHAS-LMR is suitable for the analysis of the flow blockage.
We Never Built Small Modular Reactors (SMRs), but What Do We Know About Modularization in Construction?
Authors: Mignacca Benito , Locatelli Giorgio , Alaassar Mahmoud , Invernizzi Diletta Colette
Published: October 2018

The key characteristics of small modular reactors (SMRs), as their name emphasized, are their size and modularity. Since SMRs are a family of novel reactor designs, there is a gap of empirical knowledge about the cost/benefit analysis of modularization. Conversely, in other sectors (e.g. Oil & Gas) the empirical experience on modularization is much greater. This paper provides a structured knowledge transfer from the general literature (i.e. other major infrastructure) and the Oil & Gas sector to the nuclear power plant construction world. Indeed, in the project management literature, a number of references discuss the costs and benefits determined by the transition from the stick-built construction to modularization, and the main benefits presented in the literature are the reduction of the construction cost and the schedule compression. Additional costs might arise from an increased management hurdle and higher transportation expenses. The paper firstly provides a structured literature review of the benefits and costs of modularization divided into qualitative and quantitative references. In the second part, the paper presents the results of series of interviews with Oil & Gas project managers about the value of modularization in this sector.
An Inverse Method to Estimate Emission Rates of Multi-Radionuclides Based on an Ensemble 4DVar Method With Local Gamma Dose Rate Measurements
Authors: Geng Xiaobing , Xu Mei , Zhang Lijun , Yuan Biao
Published: October 2018

An inverse source estimation method is proposed to reconstruct emission rates of multi-radionuclides using local gamma dose rate measurements under the data assimilation framework. It involves the Proper Orthogonal Decomposition (POD)-based ensemble four-dimensional variational data assimilation (PODEn4DVar) algorithm and a transfer coefficient matrix (TCM) created using FLEXPART, a Lagrangian atmospheric dispersion model. PODEn4DVar is a hybrid data assimilation method that exploits the strengths of both the ensemble Kalman filter (EnKF) and the 4DVar assimilation method. With an explicit expression of control (state) variables in the cost functional, the data assimilation process is substantially simplified than traditional 4D variational method. By setting a unit emission rate and running the ATDM model (FLEXPART in this article) driven by meteorological fields forecasted with WRF, we get the transfer coefficient matrix with the progression of nuclear accident. TCM not only acts as observation operator in PODEn4DVar, but also eliminates the control run in traditional data assimilation framework. The method is tested by twin experiments with ratios of nuclides assumed to be known. With pseudo observations based on Fukushima Daiichi nuclear power plant (FDNPP) accident, most of the emission rates were estimated accurately, except under conditions when wind blew off land toward the sea and at extremely slow wind speeds near the FDNPP. Because of the long duration of accident and variability of meteorological fields, measurements from land only in local area is unable to offer enough information to support emergency response. With abundant measurements of gamma dose rate, emission rates can be reconstructed sequentially with the progression of nuclear accident. Therefore, the proposed method has the potential to be applied to nuclear emergency response after improvement.
Prediction and Sensibility Analysis for Nuclear Safety-Critical Software Reliability of DCS
Authors: Ying Liu , Ya-feng Wang , Bo Pang , Lei Tang , Bo Feng , Guo-hai Cao
Published: October 2018

With the development of control and information technology at NPPs, software reliability is important because software failure is usually considered as one form of common cause failures in Digital I&C Systems (DCS). The reliability analysis of DCS, particularly qualitative and quantitative evaluation on the nuclear safety-critical software reliability belongs to a great challenge. To solve this problem, not only comprehensive evaluation model and stage evaluation models are built in this paper, but also prediction and sensibility analysis are given to the models. It can make basement for evaluating the reliability and safety of DCS.
Design and Non-Proliferation Viability of Small Modular Reactors
Authors: Koreshi Zafar ullah
Published: October 2018

Small Modular Reactors (SMRs) are economically competitive nuclear power systems aimed to provide sustainable clean safe and reliable nuclear energy free from the risk of fissile material proliferation. They are smaller versions of the present-day large nuclear power reactors with additional design simplifications, improved and reliable passive safety systems incorporating innovative concepts. With the intrinsic advantage of high power density and carbon-free emissions, SMRs and especially their innovative features are the signals for a nuclear comeback, or in Dr Alvin Weinberg’s words “the second nuclear era” in many ways. According to some estimates, there could be up to 96 SMRs by 2030. This paper addresses three vital areas to the understanding of the SMR’s in emerging global environments: (i) design, (ii) production of plutonium during operation, and (iii) their scope of applications. A representative, though very small SMR, Toshiba’s innovative 4S design is used for presenting estimates of plutonium production which are applicable to other SMRs as well. To better understand the viability of SMRs, this work considers the emerging developers, exporters and markets where SMRs can make significant improvements to the overall socio-economic development of societies challenged with formidable barriers.
Visualization Study on Droplet-Entrainment in a High-Speed Gas Jet Into a Liquid Pool
Authors: Sugimoto Taro , Saito Shimpei , Kaneko Akiko , Abe Yutaka , Uchibori Akihiro , Ohshima Hiroyuki
Published: October 2018

A sodium-cooled fast reactor (SFR) is now under development in Japan. A shell-and-tube type once-through heat exchanger is to be installed to generate steam in the design. Low-pressure hot sodium flows in the shell side and high-pressure water, which heated to become steam, flows in the tube side. It has been anticipated that a pin hole is formed on the tube wall and high-pressure steam blows out from the hole. When a high-pressure steam flows out from the tube hole, a high-speed steam jet is formed in the sodium coolant. Fine sodium droplets are torn off from the sodium surface and entrained into the steam jet. Sodium-water chemical reaction causes an increase of entrained droplet temperature. The hot and high-speed sodium entrained droplets attack the wall of a neighboring tube and cause a wastage on the tube wall, which may lead to a failure propagation.

In Japan Atomic Energy Agency (JAEA), an analysis code for the sodium-water reaction phenomenon, called SERAPHIM, has already been developed. Visualization data is required to validate the liquid entrainment model in this code. Since the flow velocity at the gas leakage is a sonic speed, it is extremely difficult to visualize the inside of the gas jet. Experiments have been carried out to visualize this phenomenon in the past; however, experimental data for model validation has not been entirely obtained due to the above-mentioned difficulty. Thus, the motivation of this study is to examine the possibility of visualization method and to clarify flow structure.

To this end, we first performed the preliminary experiments using simple test facilities. Two types of test sections were used in the experiments: three-dimensional one and two-dimensional one. In the experiment using the three-dimensional one, we tried to visualize a more realistic phenomenon. Through this experiment, the whole gas-jet behavior was clearly captured. However, we found that the detailed droplet-entrainment behavior in a gas jet could not be obtained in this setup, especially at high-velocity conditions. Then, we carried out the experiments using the two-dimensional one. In these experiments, the flow structure of a gas jet was simplified. However, it was difficult to distinguish the liquid film formed on the wall surface of the test section from the entrained droplets. We considered that the liquid film is formed due to the nozzle outlet shape and improved the test section. By experiments with new test section, we succeeded in visualizing entrained droplets of relatively large diameter and calculated droplet diameter distribution. Then, we discussed the mechanism of entrained droplet behavior.
Mechanism of Flashing Phenomena Induced by Microwave Heating
Authors: Fujita Shunya , Abe Yutaka , Kaneko Akiko , Yuasa Tomohisa , Segawa Tomoomi , Yamada Yoshikazu , Kato Yoshiyuki , Ishii Katsunori
Published: October 2018

In a de-nitration conversion processes of the nuclear fuel cycle, mixed oxides (MOX) are produced from the reprocessing solution (plutonium and uranium mixed nitrate solution) of used nuclear fuel. The microwave heating (MH) method has various advantages as one of de-nitration conversion techniques, i.e., this can complete rapidly processes, reduce the waste liquid, and operated remotely and easily by the waveguide. Fine crystal powders are thereby generated, and the manufacture of high-density and high-quality pellets is significant advantage.

The MH method is accompanied with transient boiling phenomena such as overflow and flashing. From the viewpoint of enhancing mass productivity and cost efficiency, in the future, scaling up for the size of de-nitration vessel and shortening processing time are desired. In addition, the safe design of device and the appropriate conditions in microwave irradiation process are required. Hence the extensive understanding of transient boiling phenomena induced by microwave heating is important, and the detailed mechanism of flushing and overflow should be clarified. The aim of this study is to clarify the transient boiling phenomena and detailed mechanism of flashing and overflow. Flashing are affected by physical factors such as solution properties, vessel characteristics, and input energy. Distilled water and different dielectric constant solution as working fluid are used. Dielectric constant was adjusted by the concentration of potassium chloride aqueous solution. A cylindrical vessel kept enough clean by ultrasonic washing machine before experiments was used. The influence of vessel diameter and initial water depth on flashing is examined. As input energy, amount of power supplied and position of the object to be heated in an oven are varied. A high-speed video camera was installed to observe boiling phenomena, and thermography and fiber optic thermometer were to measure the temperature on vessel walls and in a solution, respectively.

As a result, the higher the dielectric constant solution, the less risk of flashing phenomena. In no flashing case, the surface temperature became over 100 °C and the precipitation of potassium chloride was existed. When the high dielectric constant solution was heated by microwave irradiation, the potassium chloride which was precipitated on the inner surface of the vessel was heated to high temperature. On the other hand the temperature of water was lower than that of the deposition of the potassium chloride. As described above, the influence of the high dielectric constant solution on the flashing phenomenon by microwave heating was evaluated. And the mechanism of flashing phenomena was assumed by detailed observation of bubble nucleation. Eventually the mechanism of flashing phenomena was examined.
Investigation on Flow and Breakdown Characteristics Under Horizontal Shear of Water Film Falling Down Vertical Corrugated Plate Dryer
Authors: Wang Bo , Tian Rui-feng , Bowen Chen , Feng Mao
Published: October 2018

The corrugated plate dryer is a very important equipment for the steam-water separation in the steam generator. Its separation efficiency determines the economic indicators of nuclear power plants. The study of the complicated steam-water separation mechanism of the corrugated plate dryer is very helpful to improve its separation efficiency.

Flow and breakdown characteristics under horizontal shear of water film falling down vertical corrugated plate dryer is investigated. Air and water are used as the two phases in the research. This paper also derives a model calculating the critical airflow velocity of water film breakdown and completes the comparison of the experimental result and the result calculated by the model. The relationship between the water film thickness and critical airflow velocity of water film breakdown is investigated through the experiment. In addition, the water film thickness is measured according to CCD high-speed camera acquisition system and the Planar Laser Induced Fluorescence (PLIF) method is also used for the measurement.

The experimental result reveals that the critical airflow velocity of water film breakdown is related to the corrugated plate structural parameters, the properties of the two phases and the water film thickness. The critical airflow velocity of water film breakdown and the water film thickness are negatively correlated. The result calculated by the model is in good agreement with the experimental result as the water film thickness is large in a certain range.
Hydrogen Risk Reducing Technology in Small Modular Reactor
Authors: Chen YanLin , Lv XueFeng
Published: October 2018

With the development of small modular reactors, the hydrogen risk reducing technology cannot be ignored. Special safety facilities of small modular reactor (SMR) are investigated and studied, and a serious accident analysis program model for SMR is established. The combination of Pre-inerting and hydrogen recombination was used to control the hydrogen risk. The effectiveness of the hydrogen control system is analyzed by using the GASFLOW program. The results show that the volume fraction of hydrogen in the containment dome is higher than that in the other parts of the containment during the calculation. Because of the small size and tight internal structure, hydrogen accumulates in the narrow channel, which increases the hydrogen concentration in the local channel. Inerting reduces the concentration of oxygen in the containment and effectively controls the possibility of flame acceleration and blasting transition in high hydrogen concentration regions.
Propagation of Nuclear Data Uncertainties in PWR Pin-Cell Burnup Calculations via Stochastic Sampling
Authors: Mercatali Luigi , Alzaben Yousef , Sanchez Espinoza Victor Hugo
Published: October 2018

In recent years there has been an increasing demand from nuclear research, industry, safety, and regulation bodies for best estimate predictions of Light Water Reactors (LWR’s) performances to be provided with their confidence bounds. From a neutronic point of view, among the different sources of uncertainty the main challenge is represented by the one related to the accuracy of the nuclear data libraries used in the transport calculations. The assessment of nuclear data uncertainties and their impact on the main reactor parameters plays a fundamental role not only for criticality safety but also in burnup analyses. In facts, the accurate prediction of nuclear parameters in burnup calculations strongly affects the management of spent nuclear fuel, the core design, as well as the economy and safety of nuclear reactors. In this paper a study related to the impact of the nuclear data uncertainties on the evolution in time of the criticality and the nuclide concentrations in burnup calculations is presented. The analysis has been performed by using a statistical sampling methodology in which all the uncertain parameters are handled as random dependent variables by a sampling procedure. The probability distributions of the uncertain input parameters are used to generate random variations of these input quantities starting from a covariance library in a 56-group energy structure. Calculations have been performed by means of the SCALE 6.2.2 code and ENDF/B-VII.1 nuclear data. The method has been tested on a PWR pin cell model representative of the TMI-1 15 × 15 assembly as defined in an international benchmark exercise. In the first part of the paper the methodology and the neutronics modelling of the problem are presented.
Numerical Study of the Amplitude and the Convection Speed of Periodic Large-Scale Vortices in a Square Array of Cylinders Subjected to Axial Flow
Authors: De Moerloose Laurent , De Ridder Jeroen , Vierendeels Jan , Degroote Joris
Published: October 2018

A square array of cylinders subjected to axial flow is commonly encountered in nuclear reactors and other heat exchangers. Large-scale vortices have been observed in the gaps between the cylinders, both experimentally and numerically. These periodic flow instabilities occur in tightly-spaced cylinder arrays and originate from the velocity difference between the gap and the subchannel regions. The pressure fluctuations caused by the coherent vortex structures are possibly a source of fretting and fatigue in the aforementioned applications. In order to quantify and comprehend this phenomenon, Large-Eddy Simulations are performed on an incompressible, Newtonian fluid flowing adiabatically through a numerical domain containing a single rigid cylinder with periodic boundary conditions, representative for a cylinder in an infinite square array. Subsequently, the temporal frequency spectrum of the wall pressure profile is calculated. The spatial autocorrelation function of this Fourier spectrum, the so-called Cross Spectral Density function, contains information regarding the amplitude and convection speed of the pressure fluctuations. It is shown that the flow instability is strongest for a pitch-over-diameter ratio of 1.03. Also, the simulations indicate that the convection speed is monotonously increasing with the pitch-over-diameter ratio. An updated model for this convection speed is proposed. Finally, it is shown that the single-cylinder approximation has its limitations, but provides valuable information with minimal computational cost.
Thermal-Hydraulic Analysis of TOPAZ-II With Modified RELAP5
Authors: Tang Simiao , Wang Chenglong , Su G. H. , Qiu Suizheng , Tian Wenxi
Published: October 2018

With the advantages of high reliability, high power density and long life, small nuclear power reactor has become one of the most excellent space power options in the space missions. TOPAZ-II is the most mature space nuclear power reactor based on thermionic conversion. In this paper, the thermo-physical and transport properties of NaK-78 and heat transfer correlations for liquid metals are implemented into the RELAP5 code. The modified RELAP5 has already been accessed to analyze the thermal-hydraulic characteristics of the space reactor cooled by NaK-78. A RELAP5 model including the core, TFEs, radiator, coolant loop and volume accumulator is developed. Temperature reactivity feedback, TFE emitter, TFE collector, moderator and the reactivity insertion effects of control drums and safety drums are modeled in the point reactor kinetics equations with six-group delayed neutrons. To V&V the integrated TOPAZ-II system model, the steady state is simulated and analyzed. The steady state calculated results are in good agreement with the designed values. On the basis of V&V, a hypothetical reactivity insertion accident is simulated and analyzed. During the accident, the automatic control system is assumed to be malfunctioned, 0.01$ positive reactivity is introduced for 500s and then control drums start to rotate inward. The maximum temperatures of fuel and emitter are below the melting temperature, respectively. The maximum temperature of coolant is 940K with 160K margin from boiling. With the rotating of control drums, the reactor reaches critical again.
Optimal Design and Performance Simulation of a Novel Semi-Active Vibration Absorber for Pipeline System of NPP
Authors: Wei Zhiguo , Gou Jinlan , Liao Mengran , Li Shaodan , Dai Lu
Published: October 2018

The vibration of pipeline system induced by pump is one of the most important factors affecting the safety and stability of nuclear power plant (NPP) operation, and is also the most prominent vibration and noise transmission channel in the plant area. Analysis results show that various support structures such as pipe clamps, hangers and brackets are the main transmission routes of pipeline vibration. However, it is usually difficult to obtain a satisfactory absorption effect by using traditional passive vibration isolation device, especially for a variety of low-frequency vibration. Besides, the effective isolation frequency is fixed which cannot adapt to the variation of the operation condition excitation source. Therefore, it is of great practical significance to propose a method and scheme which can effectively restrain the low-frequency vibration of pipeline system of NPP. In order to meet the above requirements, an electromagnetic semi-active vibration absorber is designed in this paper. Its electromagnetic stiffness can be adjusted by changing the excitation current, so as to achieve the purpose of changing the natural frequency online and improving the vibration attenuation effect. In order to verify the rationality of its electromagnetic structure, the magnetic field distribution of vibration absorber is simulated and analyzed. Results show that it is close to magnetic saturation inside without leakage. On this basis, a platform of simulated condensate water pipeline system for vibration experiment is set up, on which the frequency response together with time & frequency-domain vibration characteristics of typical support points are tested, and the stiffness, damping and other parameters of the vibration absorber are optimized according to the test results. Finally, the vibration of pipeline system is simulated and compared. The results show that under the condition of the optimal combination of parameters, the vibration spectrum of the pipeline is obviously reduced after the vibration absorber is used, which indicates a significant vibration control effect.
Development of Hydrogen Treatment System in Severe Accident: Part 4 — Study of Fission Products and Steam Effect on Hydrogen Treatment Characteristics
Authors: Yamada Akira , Yagyu Motoshige , Iwaki Chikako , Sugita Tsukasa , Haruguchi Yoshiko , Tanabe Masashi
Published: October 2018

A large amount of hydrogen is generated by the metal water-reaction in the Primary containment vessel (PCV) of light water reactors in the severe accident (SA). In the present accident management for boiling water reactor (BWR), vent of mixing gas with filtered vent is regarded as the most likely method that prevents the PCV overpressure. However, it is difficult to vent in early stage of SA because of high radioactive dose. Then we have been developing the hydrogen treatment system to prevent excessive pressure without PCV vent. In focusing on the oxidation-reduction reaction of metal oxides (MOs) with high reaction rate, we have been studying hydrogen treatment system using MOs as effective device under oxygen deficit conditions like PCV of BWR. In the previous studies, we evaluated the hydrogen treatment rate using a couple of MOs, and confirmed that CuO, Co3O4, and MnO2 were effective for the hydrogen oxidation under the oxygen-free condition. We also found that granules of these three MOs could achieve the goal of hydrogen treatment rate with reactor of hydrogen treatment system. We predicted that the performance of MOs decreased with exposure to steam and fission products (FPs) in the PCV during the hydrogen treatment, and investigated their influence. The objective of the present research is to investigate how the steam and FPs, which is supposed to be a reaction-inhibiting-factor, influence hydrogen treatment rate. Then, we conducted hydrogen treatment experiments using a fixed bed reactor with MOs layer. As the results, we confirmed that the hydrogen treatment rate of MnO2 decrease from 70 g/s/m3 to 15 g/s/m3 when partial pressure of vapor went above 0.1 MPa-abs, though, that of CuO didn’t depend on the partial pressure of vapor and sustain the same rate about 40 g/s/m3. We also confirmed that the hydrogen treatment rate was decreased with the consumption of granulated MOs faster than our expectation estimated with unreacted-core model*. We also estimated that CsI selected as typical FPs could not affect the hydrogen treatment rate of CuO. From these results, we have evaluated the reaction rate equation including the steam influence in CuO, which could estimate the hydrogen treatment rate of reactor unit.

*Gas reacts only on the surface of solid and generates shell of products around reactants core. The core shrinks with reaction.
Vaporisation and Condensation in the Feed-Water System in the Turbine Building: How the Phenomenon Arises and How it Can Be Avoided
Authors: Probert Thomas Arne
Published: October 2018

When restarting Oskarshamn Unit 3 after the re-fuelling outage 2016 a turbine trip occurred at low reactor power. Inspections of the feed-water system in the turbine building revealed a damaged support. A water-hammer caused by an open valve that should have been closed was suspected to be the cause of the damage. Measurement data from the event showed that the pressure in the piping near the damaged support had decreased slowly to be followed by rapid pressure peaks with high amplitudes. The damaged support was reinforced to regain its function and the plant was restarted but an investigation was initiated to analyse the event.

A study of measurement data from the plant showed that this was not a single event. Similar events had occurred earlier. The event turned out to be independent of the valve being open or closed. It was also observed that main pumps that were not in operation at the time in the feed-water system had rotated in the reverse direction during the pressure decrease (normally two out of three pumps are in operation). This event could explain a lot of the support damage in the feed-water system that had occurred since the plant was licensed and that was considered to be caused by operational displacements and vibrations.

A fruitful and intense collaboration started between Engineering, Operations and Maintenance to find the root cause of the event. This resulted in a number of observations that led to a general understanding of the nature of the event. The event was caused by depressurisation in the feed-water tank caused by a turbine trip at low power. The gradual pressure drop caused the water in the piping lower down to evaporate (flash), causing a mass-flow upwards. When the evaporation stopped the flow changed direction and fell downwards causing condensation and large pressure peaks.

The event could be simulated (by RElAP5) with results that qualitatively compared well to measurements. Simulations were then used to show how the event could be mitigated and ultimately avoided.

Hopefully future events will be avoided by changing operation procedures. This has been verified when the plant had to shut-down in November 2017 due to a fuel failure. During the shut-down process no flashing event was detected.
Investigation of Flammability of Hydrogen Gases With Diluent Gases Under Severe Accident Conditions Using CNFT Model
Authors: Jeon Joongoo , Kim Nam Kyung , Choi Wonjun , Kim Taeseok , Kim Sung Joong
Published: October 2018

After the Fukushima Daiichi accident, predicting lower flammability limits (LFL) as a part of hydrogen risk analysis has become an ever important task. Although many experimental studies have been conducted extensively, the LFL results for mixtures abided by the severe accident conditions are still lacking. The objective of this study is to develop a calculated non-adiabatic flame temperature (CNFT) model, which facilitates to predict the LFL of hydrogen mixtures. This model considers heat loss due to radiative heat transfer from flame to ambient environment during flame propagation. The model shows better agreement with experimental results for various mixtures than previous model, which predicts the LFL through a calculated adiabatic flame temperature. Especially, prediction accuracy for H2-air-steam mixture and mixtures at elevated initial temperature is improved substantially. Thus it is worth to evaluate the applicability of the CNFT model in the hydrogen risk analysis during severe accident. The postulated hydrogen risk in the current Optimized Power Reactor 1000 MWe (OPR1000) under Station Blackout (SBO) scenario was investigated with MELCOR 1.8.6 code. As a result, it was observed that uncertainty of hydrogen risk calculated with the MELCOR default model can be reduced by the CNFT model. This study suggests that the developed CNFT model can enhance reliability of severe accident analysis related to the flammability of hydrogen mixtures.
The Necessity of Independent Configuration of the DC Power Supply System Used for Power Supply and Auxiliary Power Supply in Nuclear Power Plant
Authors: Zhichao Zhang
Published: October 2018

At present, in the typical design of China’s nuclear power plant, main generation system is connected to the power grid by 500 kV system. 500 kV system as a priority power source, 220kV system as an auxiliary power source. Independent operation of 500kV and 220kV system, improved the reliability of power supply of nuclear power plant. However, the DC 220V power system used to control the 500kV and 220kV system in the switch station of partial nuclear power project is not independently configured, and the design form of one set of DC system is used in the transformer station. In recent years, there are many accidents that AC power enters into the DC power system, resulted in the loss of power source in the transformer station. The loss of external power source in the whole plant is very significant. In this paper, the influence of AC power entering into DC power system on relay protection device is analyzed, the measures to prevent the AC power into DC power system are discussed, the necessity of independent configuration of DC control power system for the 500kV priority power system and 220kV auxiliary power supply system is analyzed.
Research of Fast Modeling and Simulating Platform for Nuclear Power Plant Secondary Loop
Authors: Gong Meijie , Peng Minjun , Zhu Haishan
Published: October 2018

Secondary loop system is an essential part in nuclear power plant (NPP). The traditional methods for the secondary loop cannot reveal the universality for the most of the NPPs. When there is the need to build another NPP secondary loop simulation program or project, the simulation model always need to modify or adapt to the new design parameters and operation parameters. In this work a fast modeling simulation platform for secondary loop in NPP is developed. In this simulation platform there is the modular, parametric simulation model in the device library and graphical human-computer interaction. There is the modeling procedure for the NPP secondary loop system simulating and modeling. In the device library there are different simulation models of secondary loop system. The secondary loop simulation model can be made up by the device simulation model with reasonable data interface. Each device simulation model could represent the practical device steady and dynamic performance with accurate calculation. The simulation results calculated in the simulation platform can represent the accuracy steady calculation and reasonable dynamic tendency of main parameters for the secondary loop system built. From the simulation platform the NPP secondary loop system simulation model can be built conveniently and fast. In the simulation platform there is the appropriate data input form and output form. The simulation platform can be used in the different purposes of simulation for NPP secondary loop as training, evaluating, operation and characteristic analysis.
Preliminary Analysis of Hydrogen Behavior Using GASFLOW-MPI in the OPR1000 Containment Under a Severe Accident Condition
Authors: Go JongWook , Hong TaeHyub , Seo MiRo
Published: October 2018

Hydrogen can be generated by the oxidation of fuel claddings with steam at a high temperature in a nuclear reactor. When the generated hydrogen is discharged from the reactor vessel into the containment, the integrity of the containment can be challenged by the high pressure and shock wave resulting from a hydrogen explosion similar to those during the Fukushima accident and the TMI accident. Thus, an investigation of the hydrogen behavior is required in order to assess the threat of a hydrogen explosion. The objective of this research is to conduct a preliminary 3D computational fluid dynamics (CFD) analysis to examine the hydrogen behavior in the containment. To achieve this objective, GASFLOW-MPI, a computational fluid dynamics tool, was selected given that it is specialized for analysis of hydrogen behavior in nuclear power plants. A large-break loss-of-coolant accident (LBLOCA) scenario was selected for the assessment. A guillotine break was assumed at a cold leg for the LBLOCA. The boundary condition, specifically the guillotine break, and the initial conditions were set based on calculation results from MAAP 5.03. Mitigation measures, including a spray system, passive autocatalytic recombiners (PARs), and hydrogen ignitors installed in the containment of a NPP were considered in the assessment. In this case, the mitigation measure models inserted by GASFLOW-MPI were used. Specifically, the Korean PAR was modeled by GASFLOW-MPI for the PAR installed in the OPR1000. The results of the preliminary analysis show that GASFLOW-MPI has the capacity to assess hydrogen behavior. The operation of the spray system increased the hydrogen volume fraction and halved the steam volume fraction compared to when it did not operate due to condensation.
Conceptual Design of the Water Cooled Breeder Blanket for Both Phases of CFETR
Authors: Liu Songlin , Ma Xuebin , Jiang Kecheng , Li Min , Zhang Xiaokang
Published: October 2018

The Chinese Fusion Engineering Testing Reactor (CFETR) will be operated in two phases. Phase I focuses on fusion power Pfusion = 200 MW, fusion power gain Qplasma = 1 – 5, tritium breeding ratio TBR>1.0, neutron DPA requirement ∼10 dpa. Phase II emphasizes DEMO validation, which means Qplasma > 10, Pfusion > 1 GW, e.g. 1.5 GW. It is required that one blanket design can cover the operation of both phases of CFETR from the viewpoint of saving construction cost and reducing waste inventory. However, fusion power in Phase-II is 4–6.5 times larger than those in Phase-I, and this also causes the great challenge facing the thermal-hydraulics design of the blanket. A new version of water cooled ceramic breeder (WCCB) blanket for both phases is proposed for CFETR, based on a trade-off considering on TBR, release tritium temperature in breeder zone, and removal heat capability of coolant. This design continues to employ the mixed breeder of Li2TiO3 and Be12Ti as tritium breeder and primary neutron multiplier, and a few Be as supplement of multiplying neutrons, Reduced Activation Ferritic/Martensitic steel as structural material, tungsten as plasma facing material. Pressurized water of 15.5 MPa is chosen as coolant with 285 °C inlet/325 °C outlet temperature. Main design change is that it employs two independent coolant systems in the blanket cooling components. For Phase I, one coolant system is only used and hoped to improve the breeder zone temperature higher than tritium release temperature. For Phase II, all of two coolant systems are put into using to ensure the material temperature less than the allowable limit. In this paper, the WCCB blanket design work is presented and its feasibility is investigated from the aspect of neutronics and thermo-hydraulics.
ALFRED Steam Generator Assessment: Design and Pre-Test Analysis of HERO Experiment
Authors: Lorusso Pierdomenico , Pesetti Alessio , Tarantino Mariano
Published: October 2018

In the framework of the ALFRED design (Advanced Lead Fast Reactor European Demonstrator) for DEMO-LFR, a new concept of steam generator (SG) has been proposed consisting in a double wall bayonet tube bundle which improves the plant safety reducing the possibility of water-lead interaction thanks to a double physical separation between them, and allowing an easier control of eventual leakages from the coolant by pressurizing the separation region with inert gas.

In order to support the development of this innovative SG configuration, the ENEA has designed and realized the HERO (Heavy liquid mEtal pRessurized water cOoled tubes) test section, a mock-up (1:1 in length) which represents the ALFRED SG. This test section, implemented in the CIRCE pool facility, aims to investigate on the thermal-hydraulic features of the system, providing a database for STH codes validation. The experimental campaign consists of high pressure tests at about 180 bar carried out in the framework of the HORIZON2020 SESAME project (Simulations and Experiments for the Safety Assessment of MEtal cooled reactors).

The secondary loop has been realized for the HERO SG feeding, consisting in an open loop circuit fed by demineralized water. The system is equipped with a volumetric pump and a heater in order to reach the water nominal working conditions of 335°C at the SG inlet and about 180 bar at the outlet.

A preliminary test analysis is carried out by RELAP5-3D© thermal-hydraulic system code. A numerical 1-D model of the HERO SG and the secondary loop has been realized in order to test the loop layout and to characterize the main components from a thermal-hydraulic point of view, defining the start-up procedures for the achievement of the working conditions of the water for the high pressure tests. Furthermore, several simulations are carried out to investigate on the secondary system behavior both for steady states and transients.
Experimental Investigation on Critical Heat Flux From Downward-Facing Flat Plate for Different Orientation Angles
Authors: Deng K. H. , Zhang Y. , Wang C. L. , Zhang Y. P. , Tian W. X. , Qiu S. Z. , Su G. H. , Wang Y.
Published: October 2018

After the severe accident inside a nuclear reactor, the IVR (In-vessel retention) management strategy is an effective way to keep the integrity of pressure vessel and reduce risk of radioactive leakage by holding the damaged core materials through External Reactor Vessel Cooling (ERVS). The damaged core materials aggregate in the lower head of pressure vessel and releasing heat to the lower head. Therefore, it is very important to remove heat timely to keep the integrity of pressure vessel by ERVS. The shape of lower head is hemispherical and the local Critical Heat Flux (CHF) of different parts changed with latitude. In this paper, influence of orientation angles, area and length-width ratio on CHF of plate heating surface for saturated pool boiling is investigate experimentally. The results show that CHF increases with increasing orientation angles and decreasing area, meanwhile, length-width ratio has a significantly effect on CHF.