bluuebunny/crossref_metadata_2025 huggingface dataset:
Generalized Sensitivity Analysis With Continuous-Energy Monte Carlo Code RMC
Authors: Qiu Yishu , Aufiero Manuele , Wang Kan , Fratoni Massimiliano
Published: October 2016

A new capability for computing sensitivity coefficients of bilinear response functions has been developed in the Reactor Monte Carlo code RMC based on the collision history-based method. Originally implemented in the Monte Carlo code SERPENT2 in the frame of Delta-tracking technique, this method computes the perturbation of particle weight based on the concept of accepted events and rejected events. The implementation of this method in RMC is based on ray-tracking technique. The new capability in RMC has been verified by comparing sensitivity coefficients of adjoint-weighted kinetic parameters including effective prompt lifetime and effective delayed neutron fraction from SERPENT2 as well as two deterministic codes based on Equivalent Generalized Perturbation Theory (EGPT), TSUNAMI-1D and SUSD3D, through two fast metallic systems, the Jezebel and flattop problems. Good agreement among RMC, SERPENT2, SUSD3D and TSUNAMI-1D (EGPT) is observed.
Program of Human-Machine Interface Verification and Validation of HTR-PM in China
Authors: Qianqian Jia , Guo Chao , Ronghong Qu , Liangju Zhang
Published: October 2016

The demonstration construction of the Chinese design of modular high temperature gas cooled reactor (MHTGR), named as high-temperature gas-cooled reactor-pebble bed module (HTR-PM) is now under construction. HTR-PM employs a layout of two nuclear steam supply system (NSSS) modules coupled to one steam turbine. The verification and validation (V&V) of HMIs is important for HTR-PM, as the first two-modular design. The V&V program of HMIs for HTR-PM is introduced in this paper. The V&V activities may include the static and dynamic verifications. The basic layout and static HMI usabilities will be verified on the static stage, and the task support based on operation procedures will be evaluated in the dynamic V&V phase. A verification platform is built to make verification and validation of the HMIs. The HMIs of HTR-PM will be improved according to the V&V result.
Analysis and Prospect of the Duplex Fuel Pellets of LOWI Type for Water-Cooled Reactors
Authors: Li Songyang , Wang Dingqu , Guo Wenli , Jiang Yueyuan
Published: October 2016

The duplex pellets under a “Low-Interact” (LOWI) nuclear fuel design, which consist of an outer enriched annulus and a depleted or natural core, can provide lower center temperature and reduced probability of pellet-clad mechanical interact (PCMI). Analysis and experiments were done in 1970s to examine the benefits and cost of LOWI design for water-cooled reactors. Results showed that the additional economic cost of this design should not be neglected in spite of the benefits. However, due to the improvement of nuclear fuel fabrication technology in the past 30 years, the benefits of LOWI design become more significant, especially when the potential of other methods to elevate the power density and overcome the constraints on ramp rates in power reactors is running out. In order to evaluate the feasibility of deploying the LOWI fuel in commercial and research reactors, neutronics and thermal calculations are made to figure out the performance of duplex UO2 pellets in particular reactors. It is shown that the center temperature of pellet has been greatly reduced without any change on assembly and core geometry, which means the opportunity of less fission gas production, higher power density and more adequate safety margin. A mechanical analysis of a typical LOWI design is also done. The challenges on duplex pellet manufacture are also discussed. Several fabrication techniques are presented to show the potential of cutting the cost of pellet production.
Evaluation of Passive Circulation Flow Characteristics and CHF Under In-Vessel Retention
Authors: Hu Qiang , Yan Xiao , Huang Shanfang , Huang Yanping , Yu JunChong
Published: October 2016

A method to evaluate the effect of geometric structure and thermal hydraulic parameters on the passive circulation flow rates in the annular flow channel between the insulation structure and the lower head of the reactor pressure vessel and critical heat flux on the equator of lower head outer surface under core degradation accidents has been proposed in this work. Based on the one-dimensional steady-state mass, momentum and energy conservation equation, a numerical calculation model has been established. The drift flux model is adopted in the numerical calculation model since velocity difference between the liquid and vapor phase in the induced upward two-phase flow could be significant. The calculation results are compared with the available experiment data collected from the literature under the same condition for the validation of the numerical calculation model proposed in this work. The effects of the heating power, loss coefficient, flooding levels, gap clearance and inlet area on the passive circulation flow are studied in this study. Furthermore, combined with the CHF correlation proposed for the determination of the CHF on the downward heating surface, the variation of the CHF at the equator of lower head outer surface with those geometric structure and thermal hydraulic parameters are analyzed utilizing the numerical calculation model for the better understanding of the limitation of the in-vessel retention through external reactor vessel cooling.
Fragility of a Flood Defense Structure Subjected to Multi-Hazard Scenario
Authors: Bodda Saran Srikanth , Sandhu Harleen Kaur , Gupta Abhinav
Published: October 2016

The March 2011 Fukushima Daiichi nuclear power plant disaster has highlighted the significance of maintaining the integrity of flood protection systems in the vicinity of a nuclear power plant. In the US, Oyster Creek nuclear plant was shut down when high storm surge during hurricane Sandy threatened its water intake and circulation systems. A gravity dam located upstream of a power plant can undergo seismic failure or flooding failure leading to flooding at the nuclear plant. In this paper, we present the results from a study on evaluating the fragilities for failure of a concrete gravity dam under both the flooding and the seismic events. Finite element analysis is used for modeling the seismic behavior as well as the seepage through foundation. A time-dependent analysis is considered to account for appropriate nonlinearities. Failure of dam foundation is characterized by rupture, and the failure of dam body is characterized by excessive deformation for the flooding and seismic loads respectively. The study presented in this paper has focused on a concrete gravity dam because of the need of validation of models which exist in prior studies only for concrete gravity dams. However, the concepts are directly applicable to any concrete flood defense structure.
First Experiments at the CIGMA Facility for Investigations of LWR Containment Thermal Hydraulics
Authors: Sibamoto Yasuteru , Abe Satoshi , Ishigaki Masahiro , Yonomoto Taisuke
Published: October 2016

There has been an extensive reorientation of the light water reactor (LWR) research in Japan since the Fukushima Dai-ichi nuclear power station (NPS) accident, which focuses on severe accidents and accident managements. The Japan Atomic Energy Agency (JAEA) initiated the ROSA-SA project in 2013 for the purpose of studying thermal hydraulics relevant to over-temperature containment damage, hydrogen risk, and fission product transport. For this purpose, the JAEA newly constructed the Containment InteGral Measurement Apparatus (CIGMA) in 2015 for the experiments addressing containment responses, separate effects, and accident managements. Recently, we successfully conducted first experiments using CIGMA to characterize the facility under typical experimental conditions investigating basic phenomena such as buildup of pressure by steam injection, containment cooling and depressurization by internal or external cooling, and density stratified layer mixing by impinging jet. This paper provides an overview of the research programs, the brief description of the facility specification and the outcomes obtained from the first experiments.
First-Principles Study of Atomic Hydrogen and Oxygen Adsorption on Doped-Iron Nanoclusters
Authors: Das Nishith K. , Shoji T.
Published: October 2016

Density functional theory calculations have been used to calculate the ground state structure and oxygen and hydrogen adsorption properties of the pure and doped-iron nanoclusters. Small atomic clusters containing two to six atoms have been considered and a single Fe atom has replaced by a minor element i.e. Zr, Ti, and Sc. Doping of a minor element increases the cluster stability and octahedron Fe5Zr is the most stable structure within this study. Zr- and Sc-doped clusters have the highest oxygen and hydrogen adsorption energy. The electronic structure shows a strong hybridization between the metal 3d and oxygen 2p orbitals with a small contribution from metal 4s and 3p orbitals. Additionally, H s and metal 4s states form a new peak below the Fermi energy and a small modification is observed for 3d orbitals near the Fermi level. A small amount of Zr- and Sc-doping into the Fe-based alloys might improve the oxide film adherence.
Coordinated Control Strategy of Multiple FACTS Devices in 750kV Power Grid
Authors: Yunwei Li , Zhang Yujun , Liu Hua
Published: October 2016

Aiming at a wide range and high-frequency reactive voltage fluctuation problem in 750kV power grid which due to large-scale wind power, a coordinated control strategy of a multiple FACTS (Flexible Alternative Current Transmission Systems) configuration is investigated in the paper. System-level and electromagnetic transient level control method of multiple FACTS devices are proposed in the small wind power fluctuation, big disturbance and off-grid level, and the priorities and the principles are determined. The strategy is applied to “Xinjiang and Northwest 750kV main grid interconnection second channel project”, and the correctness and effectiveness is proved.
iB1350: A Generation III.7 Reactor After the Fukushima Daiichi Accident
Authors: Sato Takashi , Matsumoto Keiji , Hosomi Kenji , Taguchi Keisuke
Published: October 2016

iB1350 stands for an innovative, intelligent and inexpensive boiling water reactor 1350. It is the first Generation III.7 reactor after the Fukushima Daiichi accident. It has incorporated lessons learned from the Fukushima Daiichi accident and Western European Nuclear Regulation Association safety objectives. It has innovative safety to cope with devastating natural disasters including a giant earthquake, a large tsunami and a monster hurricane. The iB1350 can survive passively such devastation and a very prolonged station blackout without any support from the outside of a site up to 7 days even preventing core melt. It, however, is based on the well-established proven Advance Boiling Water Reactor (ABWR) design. The nuclear steam supply system is exactly the same as that of the current ABWR. As for safety design it has a double cylinder reinforced concrete containment vessel (Mark W containment) and an in-depth hybrid safety system (IDHS). The Mark W containment has double fission product confinement barriers and the in-containment filtered venting system (IFVS) that enable passively no emergency evacuation outside the immediate vicinity of the plant for a severe accident (SA). It has a large volume to hold hydrogen, a core catcher, a passive flooding system and an innovative passive containment cooling system (iPCCS) establishing passively practical elimination of containment failure even in a long term. The IDHS consists of 4 division active safety systems for a design basis accident, 2 division active safety systems for a SA and built-in passive safety systems (BiPSS) consisting of an isolation condenser (IC) and the iPCCS for a SA. The IC/PCCS pools have enough capacity for 7-day grace period. The IC/PCCS heat exchangers, core and spent fuel pool are enclosed inside the containment vessel (CV) building and protected against a large airplane crash. The iB1350 can survive a large airplane crash only by the CV building and the built-in passive safety systems therein. The dome of the CV building consists of a single wall made of steel and concrete composite. This single dome structure facilitates a short-term construction period and cost saving. The CV diameter is smaller than that of most PWR resulting in a smaller R/B. Each active safety division includes only one emergency core cooling system (ECCS) pump and one emergency diesel generator (EDG). Therefore, a single failure of the EDG never causes multiple failures of ECCS pumps in a safety division. The iB1350 is based on the proven ABWR technology and ready for construction. No new technology is incorporated but design concept and philosophy are initiative and innovative.
Pool Nucleate Boiling on Heat Transfer Surface With Deposited Sea Salts
Authors: Uesawa Shinichiro , Koizumi Yasuo , Shibata Mitsuhiko , Yoshida Hiroyuki
Published: October 2016

Pool nucleate boiling heat transfer experiments of the 3.5 - 10wt% NaCl solution, the real seawater and the 3.5 - 10wt% artificial seawater solution as well as distilled water for the basis of comparison were performed to examine the effect of salts on boiling heat transfer. Seawater was injected into the reactor cores in the accident at the Fukushima Daiichi Nuclear Power Station of Tokyo Electric Power Company. This study intended to provide base data to consider reactor core cooling by seawater. Boiling curves of the 3.5 - 10wt% NaCl solution, the real seawater and the 3.5 - 10wt% artificial seawater solutions as well as distilled water were well predicted with the Rohsenow pool nucleate boiling heat transfer correlation although the curves were a little shifted to the higher wall superheat region. The formation of secondary coalescent large bubble was suppressed in the experiments of the NaCl solutions, real seawater and the artificial seawater solutions, and small primary bubbles detached directly from the heat transfer surface. Sea salt deposition was observed only in the experiments of the 7.0wt% and 10wt% artificial seawater solutions. The deposited salt was calcium sulfate. Slow heat transfer surface temperature excursion occurred in the experiments of the 7.0wt% and 10wt% artificial seawater solutions after the heat flux was raised to 600 kW/m2 and 120 kW/m2, respectively. The critical heat flux of the 7.0wt% and 10wt% artificial seawater solutions were 600 kW/m2 and 120 kW/m2, respectively if the occurrence of the slow heat transfer surface temperature excursion was defined as the critical heat flux condition. The heat transfer surface temperature excursion might be caused by the growth of the deposited salt layer.
RELAP5 Code Uncertainty Analysis of LSTF Small Break LOCA Tests With Steam Generator Intentional Depressurization as an Accident Management Procedure
Authors: Kinoshita Ikuo , Torige Toshihide
Published: October 2016

The Best Estimate Plus Uncertainty (BEPU) method is applied to analysis of the “intentional depressurization of steam generator secondary side” which is an accident management procedure in a small break loss-of-coolant accident with high pressure injection system failure. In the present study, experimental analyses using the RELAP5/MOD3.2 code were carried out for the ROSA/Large Scale Test Facility (LSTF) secondary-side depressurization tests. The two test cases were selected with different break sizes and different depressurization conditions to ensure the reliability for the accident scenario analyses. The input parameter uncertainty propagation analyses were performed to get 95%/95% tolerance limit values of the output parameters. It was confirmed that the code predicted well the major event progressions of the accident for both test cases and the 95%/95% uncertainty bounds of the peak cladding temperatures included the measured values. On the other hand, the same ranges of some input uncertainty parameters could lead to different influences on the output uncertainties between the test cases. The dominating input uncertainty parameters could be different depending on the break sizes and depressurization conditions of the accident scenario.
Effects of External Grid Disturbances to Nuclear Power Plants
Authors: Arians Robert , Arnold Simone , Mueller Christian , Quester Claudia , Sommer Dagmar
Published: October 2016

The reliability of the auxiliary power supply of a nuclear power plant (NPP) is of high importance for safe operation. The loss of the electrical power supply is one of the major contributions to the calculated core damage frequency in probabilistic safety assessments. Among others, the events in Forsmark in 2006 [1] and 2012 [2] as well as in Byron in 2012 [3] illustrate that disturbances in the external power grid can propagate into the NPP and have an impact on the safety important electrical equipment. Therefore, the grid reliability contributes considerably to the reliability of the auxiliary power supply.

In the research work presented in this paper the international operating experience has been evaluated concerning events which include disturbance in the external grid to discover those types of grid disturbances which may have influence on the safe operation of the NPPs. The identified events have then been categorized within a developed classification scheme to determine those with the highest relevance. Based on this scheme representative scenarios of grid disturbances have been developed.

The investigation of the impact of the developed scenarios on the electrical equipment of NPPs will be performed using a grid analysis, planning and optimization tool which also allows executing dynamic simulations of electrical grids [4]. Therefore, a generalized auxiliary power supply of a pressurized water reactor was modeled according to German NPPs of the type Konvoi.

In this paper, an overview of the developed scenarios of grid disturbances and the actual status of the simulation of the auxiliary power supply of NPPs is presented.
Development of Prediction Technology of Two-Phase Flow Dynamics Under Earthquake Acceleration: (17) Effect of Structure Vibration on a Rising Single Bubble
Authors: Miyazaki Akifumi , Kato Yuki , Kanagawa Tetsuya , Kaneko Akiko , Monji Hideaki , Yoshida Hiroyuki , Abe Yutaka
Published: October 2016

Earthquake is one of the most serious phenomena for safety of a nuclear reactor in Japan. Structural safety of nuclear reactors has been studied and nuclear reactors were contracted with structural safety for a big earthquake. However, it is not enough for safe operation of nuclear reactors because thermal-fluid safety is not confirmed under the earthquake. For instance, behavior of gas-liquid two-phase flow is unknown under the earthquake conditions. Especially, fluctuation of void fraction is an important factor for the safe operation of the nuclear reactor. In the previous work, fluctuation of void fraction in bubbly flow was studied experimentally and numerically. In case of the earthquake, the fluctuation is not only the flow rate, but also body force on the two-phase flow and shear stress through the pipe wall. Interactions of gas and liquid through their interface also act on the behavior of the two-phase flow. The fluctuation of the void fraction is not clear for such complicated situation under the earthquake.

Our study has investigated the behavior of gas-liquid two-phase flow experimentally and numerically. In this paper effects of vibration on bubbly flow in the components and construct experimental database for validation and performs visualization experiments of a rising single bubble in a rectangular water tank on which sinusoidal vibration was applied. In this paper, results of visualized experiment evaluated by the visualization techniques, including positions of a bubble, a shape of the bubble and the bubble tilt angle were shown. In the results, bubble behavior were affected by the table oscillation. The bubble tilt angle is also almost same value of the bubble movement angle. It is implied that higher table oscillation frequency than 20 Hz quite weakly affects on fluctuation of bubble tilt angle frequency.
Experimental and Numerical Simulation on Hydraulic Behavior of Molten Flow in the Lower Part in Reactor Core
Authors: Matsuura Kota , Monji Hideaki , Yamashita Susumu , Yoshida Hiroyuki
Published: October 2016

In the decommissioning work of nuclear power plants, it is important to grasp the sedimentation place of molten materials. However, the technique to grasp exactly sedimentation place is not established now. Therefore, the detailed and phenomenological numerical simulation code named JUPITER for predicting the molten core behavior is developed. In the study, visualization experiment and numerical simulation were performed to validate the applicability of the JUPITER to the hydraulic relocation behavior in core internals. The test section used in this experiment simulated the structure of the core internals, such as a control rod and a fuel support piece, simply. The working fluid is water under the atmospheric pressure. The experiment uses a high-speed video camera to visualize the flow behavior. The behavior and the speed of the liquid film in a narrow flow channel is obtained. For the numerical analysis carried out prior to the experiment, the behavior of flow down liquid was shown. The typical behavior was also observed that the tip of a liquid film flowing down splits into.
Modeling and Analyzing of Simple Combined Cycle of Modular High Temperature Gas Cooled Reactor
Authors: Qu Xinhe , Yang Xiaoyong , Wang Jie
Published: October 2016

High temperature gas cooled reactor (HTGR) which is one of generation IV reactor has been widely given attention in many countries since the sixties of the last century because of its inherent safety and high efficiency. Currently, the HTGR commonly uses regenerative Brayton cycle. However, as reactor outlet temperature (ROT) rising, regenerative Brayton cycle has a higher reactor inlet temperature (RIT) than 500°C and is limited by reactor materials. Combined cycle of HTGR not only can solve the problem of high RIT, but also can get a higher cycle efficiency than 50%. In this paper an accurate model of combined cycle consisting of topping Brayton cycle, bottoming Rankine cycle and heat recovery steam generator (HRSG) was established. In terms of new model of combined cycle, this paper analyzed the main properties of simple combined cycle. And put forward two optimization schemes improving the cycle efficiency of combined cycle.
Parametric Study of the Burnup Credit Criticality Analysis in BWR Fuels
Authors: Wu Shang-Chien , Chao Der-Sheng , Liang Jenq-Horng
Published: October 2016

The purpose of this study is to investigate the effects of operation parameters on BWR fuel reactivity and nuclide inventory. The three high-priority parameters that are suggested by NRC/ORNL [1], that is, axial burnup profile, axial moderator density profile, and control blade insertion during operation, were examined in order to evaluate their impact on BWR burnup credit. All of the calculations were performed using SCALE6.1 based on the ENDF/B-VII nuclear data library and involved 238 energy groups.

The results indicate that the axial burnup profile is the most critical factor in reactivity when the burnup value is greater than 10 GWd/MTU. The insertion of control blades during operation is also more important for the reactivity of spent fuel as compared to the influence of the axial moderator density profile. At a burnup of 60 GWd/MTU, these three operation parameters (i.e., axial burnup profile, axial moderator density profile, and control blade insertion during operation) in turn lead to multiplication factor differences of 18850, 7600, and 1980 pcm, respectively, for set2 nuclides. The influence between different parameters was also investigated in this study. The results reveal that multiple operation parameters affect the reactivity reduction tendency. In addition to the three parameters investigated in this study, other important factors related to burnup credit analysis are also currently under investigation.
Dynamic Processes During Pulsed Gas Injection Into Liquid Column
Authors: Lobanov P. D. , Kashinsky O. N. , Kurdyumov A. S. , Pribaturin N. A.
Published: October 2016

An experimental study of dynamic processes during pulsed gas injection into quiescent liquids was performed. Both water and low melting temperature metal alloy were used as test liquids. Air and argon were used as gas phase. The test sections were vertical cylindrical columns 25 and 68 mm inner diameter. Measurements of flow parameters during gas injection were performed. Water – air experiments were performed at room temperature, the temperature of liquid metal alloy was 135 deg C. Time records of pressure in the liquid and in gas phase above the liquid were obtained. Measurements of liquid temperature and level of liquid surface were performed. It was shown that at pulse gas injection into liquid metal high amplitude pressure fluctuation may arise. Also the fluctuation variation of the free surface of the liquid may appear which are connected with the oscillations of the gas volume. Experimental data obtained may be used for verification & validation of modern CFD codes.
Activation Analysis of the Reactor Pressure Vessel of HTR-PM Reactor Using FLUKA Code
Authors: Fang Sheng , Li Wenqian , Sun Sida , Li Hong
Published: October 2016

Activation products are the primary radiation source in the maintenance of the reactor pressure vessel (RPV) of High temperature reactor pebble-bed module (HTR-PM). In order to properly plan the maintenance and reduce the related occupational exposure, it is important to correctly evaluate the activation of the impurities in the metal material of HTR-PM’s RPV. In this study, activation of the material of HTR-PM’s RPV is performed using both FLUKA program and experimental formula. Based on the impurity control limit in the technical specification, various impurity nuclides are taken into account. The specific activities of activated radionuclides calculated by the two methods after 40 years irradiation and at the shutdown time are compared. It is demonstrated that FLUKA and experimental formula are in agreement. Also primary contributory activation products at 30 days, 1 year, 5 years and 10 years after the shutdown time are listed.
Jet Breakup and Droplet Formation in Immiscible Liquid-Liquid System
Authors: Saito Shimpei , Iwasawa Yuzuru , Abe Yutaka , Kaneko Akiko , Kanagawa Tetsuya , Nariai Hideki , Koyama Kazuya
Published: October 2016

Mitigative measures against the event of a core disruptive accident (CDA) are of the importance from the viewpoint of safety of a sodium-cooled fast reactor (SFR). If the CDA occurs, the so-called post-accident heat removal must be surely achieved. The present study focuses on the scenario that the molten materials are injected into the lower plenum as jets. The jet breakup behavior during the CDA will be very complicated. Therefore, a specialized study on the fundamental process during the jet breakup is believed to be an effective approach.

The aim of this paper is to understand the fundamental process of hydrodynamic interaction of jet breakup and droplet formation Using the immiscible liquid-liquid system, water and silicon oil as the test fluids, visualization via high-speed videography was performed. From the visualization results, the breakup length and droplet diameter were measured by image processing. The experimental data were scaled with ambient Weber number. When the Weber number was smaller than 1, the droplet diameter was close to the nozzle diameter, and distribution of droplet size was not observed. When the Weber number exceeded 1, the breakup length became longer and the generated droplet diameter possessed a distribution with two peaks due to satellite droplet formation. In both cases, the droplet formed at the leading edge of jet. In case that Weber number is around 100, the droplets were formed by entrainment of interfacial wave at jet side. From the mass median diameter data, we can see that the increase of the Weber number caused the decrease of median diameter and the increase of the width of the distribution.
Design and Construction of a Test Rig to Analyze Thermal-Hydraulic Phenomena in Two-Phase Flow
Authors: Lanz Thomas , Alobaid Falah , Epple Bernd
Published: October 2016

Heat transfer, pressure drop, and instability phenomena in water-steam two-phase flows are important research areas with regard to thermal power plants. Here, demineralized water is vaporized to operate steam turbines that drive generators, converting the energy from fuel that is burned into electricity. To design the evaporation process, it is important to know the heat transfer coefficients of water-steam-mixtures and the pressure drop due to friction, elevation and momentum in terms of steam quality. Moreover, occurring instabilities need to be understood to prevent unstable and unsafe operation conditions. The aim of this research work is to design and construct a test rig to analyze these thermal-hydraulic phenomena. Furthermore, the obtained results will be applied to validate the heterogeneous flow model of the dynamic simulation software APROS (Advanced Process Simulation Software) from VTT (Technical Research Center of Finland).

In contrast to previous studies in this field, the fluid used for the experiments is demineralized water — instead of usually used refrigerants — to eliminate possible side effects related to different fluid properties. Furthermore, the heated test section of the test rig is a thermal-hydraulic scale down from a thermal power plant’s low-pressure evaporator tube. The focus lies on a huge parameter variability in terms of operation conditions (sub-cooling, inlet pressure, mass flow, heating power, upstream compressible volume) and test rig layouts (vertical and horizontal test section, different lengths and inner diameters, single and parallel tubes) to analyze their impact on the different instability phenomena (density-wave oscillations, pressure-drop oscillations). Literature dealing with experimental investigation of two-phase flow in horizontal evaporator tubes exposed to natural circulation at low pressure is rare. This work aims to contribute especially to this field of research.

Currently, the designed test rig is under construction. Once the experiments are performed, the results will be analyzed by means of stability maps and fast Fourier transformation (FFT) amongst others. The results will then be used to validate the heterogeneous six-equation model of the dynamic simulation software APROS with regard to heat transfer, adiabatic pressure drop and flow instabilities. The six-equation flow model (also known as Euler-Euler approach) implies a considerably higher accuracy than the usually regarded homogeneous three-equation model.
Preliminary Benchmark Validation of a Lattice Physics Code COSLATC
Authors: Yang Tongrui , Liu Zhanquan , Yan Yuhang , Li Shuo , Yu Hui , Chen Yixue
Published: October 2016

COre and System INtegrated Engine for design and analysis (COSINE) is developed by State Nuclear Power Software Development Center (SNPSDC), which is an integrated nuclear engineering code package. A lattice physics code named COSLATC is an essential part included in COSINE. COSLATC is a multi-group two-dimensional transport code. Pin cells and assemblies of Pressurized Water Reactor (PWR) can be calculated by COSLATC. It is used to calculate few group constants and nuclei densities for core simulator.

In order to make sure the quality of the COSLATC, a strategy of verification & validation (V&V) are discussed and applied to the COSLATC. Firstly a V&V requirement analysis is performed. A test matrix considering the variety of fuel enrichments, materials, geometric and working conditions is generated. Then the corresponding benchmark is collected and classified. Finally, the numerical results of COSLATC and the reference values are compared and analyzed. According to the validation strategy discussed above, the preliminary validation is carried out.

The benchmark provides consistent and comprehensive tests for high burnup (approx. 70GWd/t) fuels of PWR. By comparison with k-infinity and isotopic composition in the UO2 and MOX fuels pin cell problems or assembly problems benchmark, the result shows that the calculated results from COSLATC code agree well with reference results. COSLATC will be an important lattice calculation code in the scientific calculation and engineering application.
The Powerful MOC Module in Advanced Neutronics Lattice Code GALAXY
Authors: Chai Xiaoming , Tu Xiaolan , Lu Wei
Published: October 2016

Due to powerful geometry treatment capability, Method Of Characteristics (MOC) becomes the most popular method to solve neutron transport equation. However, boundary conditions always restrict the MOC method’s widely application. Most of the current neutronics lattice codes based on MOC can only be used to solve one or two specific geometrical shapes. In this paper, we developed a powerful MOC module, which can treat different geometrical shapes with two methods. For special geometrical shapes, such as rectangle, 1/8 of square, hexagon, 1/3 of hexagon, 1/6 of hexagon, the MOC module adopts special trajectory layout and angle quadrature set, which can reduce the computation time. For other general geometrical shapes, the MOC module use ray prolongation method, which can treat arbitrary geometry shapes and boundary conditions but need much computation time. This MOC module was incorporated into advanced neutronics lattice code GALAXY, which developed by Nuclear Power Institute of China. The numerical results show that the GALAXY code can be used to calculate 2D neutronics problems with rectangle, hexagon, and other complicated geometry shapes accurately. In future, the GALAXY code will gradually become the main neutroncis lattice code in NPIC.
Advanced Cooling Tube Mobile Heat Exchanger in the Outer Shape of a Fuel Assembly for Spent Fuel Pool Cooling
Authors: Reuter Matthias , Ornot Leo , Maier Torsten , Fuchs Thomas , Gauter Kerstin , Reck Markus , Redmond Joseph
Published: October 2016

After the Fukushima accident the development and implementation of additional safety features has become more and more important.

In March 2014 AREVA filed a patent for the invention of a mobile heat exchanger module in the outer shape similar to the outer geometry of a fuel assembly called Advanced Cooling Tube (ACT) [1]. Since January 2015 there is a full scale test facility in the German headquarters of AREVA in Erlangen to test the performance of such a mobile heat exchanger. In spring 2015 there were investigations running to implement the ACT into the spent fuel pool of a nuclear power plant in Switzerland. In the meantime, these investigations were finished and a contract has been signed.

This paper gives an overview about the main facts of the Advanced Cooling Tube as a mobile heat exchanger in the shape of a fuel assembly for spent fuel pools, how it can be implemented in any nuclear power station and what the advantages of this product are. Further information about the performance and requirements of this special heat exchanger will be given.
Research on Natural Circulation Flow Oscillation Experiment in Narrow Rectangular Channel
Authors: Zhou Tao , Fang Xiaolu , Liu Liang , Wang Yaoxin , Xiao Zejun
Published: October 2016

A narrow rectangular natural circulation experimental system is used to research flow pulsation occurred during the experiment. During the experiment, the flow pulsation phenomenon is found in narrow rectangular channels. When compared with density wave instability, it is found that the flow pulsation has certain characteristics of density wave instability, but at the same time, there will be a transient low flow phenomenon, which is a unique flow oscillation phenomenon in the narrow rectangular channel. The mechanism of flow oscillation phenomenon is analyzed, on this basis, a forecast equation of mass velocity is put forward. Forecast equation of mass velocity is used to calculate the natural circulation mass flow rate. Calculation results are in good agreement with the experimental results, which proves that the forecast equation of mass velocity is correct, provides control and prevention theory basis for natural circulation flow instability in narrow rectangular channel.
Experimental Investigation of an Application of Two-Phase Flow Differential Pressure Measurement in Vertical Square Channel
Authors: Burian Ondrej , Dostal Vaclav
Published: October 2016

The experimental work described in this paper deals with an experimental research of two-phase flow focused on the study of thermal hydraulics of steam generating facilities for NPPs, like BWR reactors and steam generators. The results of this research can be used for development of new methods for the control and measurement of the operating parameters of these facilities at normal and abnormal operational conditions. Moreover the results are also applicable for a development of a method for a determination of two-phase flow regimes by pressure fluctuation analysis. The main idea of those methods is based on the measurement of pressure and pressure difference in steam-water mixture and their further analysis by mathematical methods of signal processing. Based on the results of these analysis, represented by statistical and frequency parameters of pressure signal, it is possible to determine other parameters of two-phase flow such as void fraction, slip ratio and flow regime of two-phase flow. This method may be applied to the steam generator, where it can be used for a measurement of a generated steam flow rate by the measurement of a differential pressure between two points located at the outer shell of the steam generator at steam-water mixture level. This method has several advantages, over other methods and might improve the total efficiency of the power plant. In this paper the validation method at low pressure is presented. For this experimental work an experimental facility was made consisting of 1.5 m tall square channel with the dimensions of 200 × 200 mm. This channel was equipped with electrical heaters with power of 9 kW in the bottom part and they serve for production of steam-water mixture. This facility was used for a measurement of an absolute pressure and a pressure difference of two-phase flow for various conditions — power and geometries of grid elements. A set of pressure data for every measurement was analyzed by basic statistical methods and results of those analyzes for various conditions were compared. This comparison was focused on the determination of dependencies among a parameters of analyzed pressure signal from a heater power and geometry of a grid which have significant influence on void fraction. The results of this comparison are presented and discussed in this paper as well as the method used for the statistical analysis. Further possibilities and limitation of this method are described, mainly the performance of the measurement instrumentation that has a significant influence on the final result. In the conclusion the applicability of this method for real power facilities and next direction of this research are discussed.
Heat Transfer of Subcooled Water Flow Boiling Under Flow Pulsation
Authors: Yuan Hongsheng , Liu Xin , Feng Li , Tan Sichao , Zhuang Nailiang , Yu Zhiting
Published: October 2016

In thermohydraulic analysis, unsteady subcooled flow boiling is of vital importance to both natural circulation systems where flow instabilities are frequently encountered and offshore nuclear power systems which operate under the influence of ocean waves. An experimental study was conducted here to investigate how an imposed periodic flow oscillation affects the subcooled flow boiling heat transfer of water in a vertical tube. The average heat transfer characteristics and variations of the transient parameters are investigated. The result shows that there is a wall temperature overshoot as a consequence of boiling onset and the wall temperature downstream of boiling front could even drop below the saturation temperature under the high inlet subcooling of 75 °C. Under flow pulsation, intermittent flow boiling appears, when the imposed heat flux, q, is close to the boiling onset heat flux of steady flow, qs,onset. As a result of intermittent flow boiling, the average wall temperature of pulsating flow is lower than the wall temperature in steady flow when the qqs,onset. Moreover, during intermittent flow boiling, the boiling induced a decrease of the total pressure drop and can cause a large pressure fluctuation. In addition, two types variation of outlet fluid temperature fluctuations were observed and the wall temperatures present an extra local maximum and minimum values beside the extrema corresponding to the extremal mass flux.
Study on Differences in Floor Response Spectra due to the Seismic Input Approach
Authors: Bochert Jana Sue , Schau Henry , Schmitt Timo
Published: October 2016

In this paper the differences of floor response spectra (FRS) resulting from different ground response spectra are discussed. These spectra include the site effects which are quantified via site response analysis. This response is generated by wave propagation from the base rock through the overlying soil layers to the surface. The influences of the different layers and the corresponding dynamic soil properties are considered by using wave propagation analysis.

The paper then discusses the results obtained from seismic input at different depths conditions. Similar results might be expected, because the depth of the input spectra is adjusted for each layer. However, in comparing the floor response spectra of these calculations, significant differences are observed and therefore interpreted.

The paper is completed with the explanation of these significant differences and also with comparable floor response spectra.
Enhancing the Plant’s Capability for Design Extension Conditions Based on Best Estimate Context Evaluation of Human Performance in ATWS Events
Authors: Petkov Gueorgui
Published: October 2016

The experience of severe accidents shows that reliable determination of technological process parameters is necessary but not always sufficient to avoid catastrophic consequences. The accident measures should be considered in a broader context that includes the human factor, organization of the nuclear technology, external influences and safety culture.

The anticipated transient without scram (ATWS) events were not considered in the original WWER (Russian PWR) design basis accidents. The design extension conditions scenarios progress in a context which is very uncertain and highly stressful for the operators. If a specific scenario requires some operators’ actions as measures to mitigate, delay or distribute the accident consequences then the dynamics of accident context seem of primary importance for ‘best estimate’ evaluations and enhancing the plant’s capability.

The paper presents the capacities of the Performance Evaluation of Teamwork (PET) procedure for enhancing plant’s capability for design extension conditions based on ’‘best estimate’ context evaluation of human performance in ATWS events. The PET procedure is based on a thorough description of symptoms of various timelines and their context quantification. It is exemplified for different ATWS scenarios of the NPP with WWER-1000 based on thermal-hydraulic simulations with RELAP5/MOD3.2 code and models.
Assessment of SIMMER-III Code Based on Steam Generator Tube Rupture Experiments in LIFUS5/Mod2 Facility
Authors: Pesetti Alessio , Del Nevo Alessandro , Forgione Nicola
Published: October 2016

An experimental campaign investigating the postulated Steam Generator Tube Rupture (SGTR) event, in relevant configurations for Heavy Liquid Metal Reactors (HLMRs), was carried out in the separate-effect facility LIFU5/Mod2, at ENEA CR Brasimone. Ten tests were performed injecting pressurized subcooled water into the reaction tank partially filled by Lead-Bismuth Eutectic alloy (LBE) at 400°C with a cover gas of argon at about 2 bar. Fast pressure transducers, thermocouples and strain gages provided high-quality measurement data for improving the phenomena understanding and supporting the development and validation phase of computer codes for SGTR numerical simulation.

The experimental campaign is composed by two series of tests, characterized by different water pressure: 40 and 16 bar. The first two tests belonging to the low pressure experiments are presented, highlighting the pressurization time trends of the water injection tank, injection line and reaction vessel. The injected water mass flow rate and temperature trends in the reaction vessel were measured. The former test is the reference one and the latter was carried out for investigating the injection of water with higher sub-cooling.

A post-test analysis of the two mentioned tests was carried out by SIMMER-III code. The pressure profile in the water injection tank was set as boundary condition of the calculation. The numerical analysis provided injection line and reaction tank pressurization in agreement with the experimental data. The lower water temperature test provided a better accordance with the measured data, due to the lower evaporation along the injection line. The SIMMER-III analysis also studied the water-LBE interaction from the volume fraction point of view and the energy released in the total reaction tank and in its cover gas.