bluuebunny/crossref_metadata_2025 huggingface dataset:
Analyses of Severe Accident Sequences in the Spent Fuel Pool of the VVER-1000 Type of Reactor With MELCOR 1.8.6
Authors: Kotouč Miroslav
Published: November 2014

Since the Fukushima nuclear disaster in 2011, much attention has been paid to investigation of severe accidents (SA) progression in spent fuel pools (SFP) of various types of nuclear power plants (NPP). In Czech Republic, 4 VVER-440 and 2 VVER-1000 types of reactors (at the Dukovany and Temelin NPPs, respectively) are currently under operation. In order to enhance their safety, especially with respect to station black-out (SBO) events, numerical analyses have been carried out evaluating the risks associated with accidents occurring also in the SFP. The present paper deals with analyses of 2 postulated scenarios (loss of cooling and loss of coolant) and is mainly focused on the input deck preparation for the integral, lumped parameter (LP) code for SA analyses — MELCOR 1.8.6. Emphasis is put on description of correct implementation of the complex geometry of the SFP, consisting of 3 distinct pools separated by concrete walls (lined with steel plates) in which fuel assemblies (FA) are stored in an absorber grid (rack). In the description of the prepared numerical model, light is shed on the encountered modeling issues, and corresponding hints and solutions are proposed in order to provide guidance for preparing adequate models for various types of SFP designs. Finally, some of the most characteristic results are presented for both postulated scenarios.
Transient Analysis of CP300 Based on RELAP5/MOD3.4
Authors: Liu Jianchang , Nina Yue , Xie Xu , Ducheng Sun , Tian Wenxi , Su Guanghui , Qiu Suizheng
Published: November 2014

The thermal-hydraulic characteristics of Nuclear Power Plant (NPP) during Steam Generator Tube Rupture (SGTR) accident are of great concern. In this paper, the thermal-hydraulic characteristics of CP300 during SGTR accident with no operator actionsduring the first 30 min are investigated using the best estimate RELAP5/MOD3.4.

Modeling and nodalization of CP300 was executed, including the vessel, pumps, pressurizer, steam generations and necessary auxiliary systems. Some main transient parameters were obtained, such as Reactor Coolant Pump (RCP) coolant temperature and pressure, steam generator flow rate and pressure. The calculation results gives the sequences of the NPP during the SGTR as described below. As the tube rupture occurs, the primary pressure drops and secondary pressure increases. When the primary pressure drops down to the set-point of the scram, the control rods drop down and the power of the NPP begins to decrease, causing the primary coolant temperature to decrease. The primary pressure continues to drop. When it drops down to 10.78MPa, the High pressure Safety Injection System is put into operation and the accident is mitigated, even without operator’s actions.

In conclusion, the calculated results indicate that the key thermal hydraulic parameters of CP300 during the SGTR accident are in acceptable ranges and the accident is effectively controlled.
Simulation of a Gas Jet in the Secondary Side of a Steam Generator Bundle Following a Tube Rupture Sequence
Authors: Mukin Roman , Dehbi Abdel
Published: November 2014

The main objective of the following work is to simulate flow in the secondary side of a steam generator (SG) tube bundle during tube rupture sequence using approaches for turbulence simulation. The interest in this topic stems from experimental findings at PSI in the framework of the ARTIST program (Aerosol Trapping in a Steam Generator) which show that particles that are injected into a dry SG tube bundle through a tube breach are principally retained in the region close to breach (the so-called “break-stage”). This region is characterized by highly non-uniform flow, with very high velocities near the breach, and low-velocity recirculating flow away from the breach. Owing to complexity of the flow, 3D simulations with highly resolved computational mesh near the breach were done. First, the flow inside an isolated tube with a guillotine tube breach has been studied. The next part is devoted to simulation of a gas jet entering the tube bundle SG via the guillotine tube breach. Comparison of the calculated flow with the experimental data of axial velocity distribution at different vertical levels have been performed, and we have found good agreement of the obtained results with the experiments.
Effect of Pressure on Boiling Heat Transfer Mechanism by Using MEMS Technology
Authors: Oda Kei , Ohtake Hiroyasu , Hasegawa Koji
Published: November 2014

Since boiling heat transfer has a high heat transfer coefficient, it has been used as a cooling technique for high-temperature bodies and it has been investigated for more than 70 years [1]. However, it has not yet been fully clarified: because the boiling phenomena are affected by many factors, such as the coalescence of bubbles, the fluid convection, the heat conduction and the physics on the contact of the gas, the liquid and the solid phase, boiling phenomena are considerably complicated. The present paper investigated an effect of pressure on boiling heat transfer mechanism by using the MEMS technology. And, boiling heat transfer enhancement in water under low pressure and low boiling temperature was examined experimentally. Steady state pool boiling experiments were conducted by using a copper thin-film and a silicon wafer for the test heater and pure water at atmospheric condition for the test liquid. The system pressure was 0.010, 0.10 and 0.15 MPa, respectively. The heaters were made of a printed circuit board and a commercial silicon wafer. The width was 7.5 mm, the length was 10 mm. The test sections were arranged for horizontal position facing upward. The test heaters had an artificial cylindrical-cavity of 0.010 or 0.040 mm in diameter; the cavities were fabricated by using the MEMS technology, i.e., wet etching technology and deep RIE. The test heaters were heated by Joule heating of d.c. current from a low-voltage high-current stabilizer. The heating rate of the heater was determined from supplied current and voltage. The temperature of the heater was obtained by referring to the measured electric resistance. The present experimental results showed the boiling bubble grew up to about 20 mm in diameter then the bubble released without coalesce of bubble under low pressure condition. Thus, the bubble coalesce was slight. From the experimental results, the gradient of boiling curve by using the copper thin-film was about 3: the heat transfer characteristic was dominant to nucleate boiling. On the other hand, the gradient of boiling curve by using the Silicon wafer (Non-cavity) was about unity: the heat transfer was dominant to heat convection of single-phase flow. According to the present observation of the boiling bubbles, the boiling heat transfer was dominant to latent heat: the ratio of the phase change and the convection was about 90 % and 10 %, respectively. The heat transfer ratio of the convection increased as the system pressure increased.
Investigation of Electron Beam Welding Characteristics Using AA6061-T6 for Nuclear Plate-Type Fuel Assembly
Authors: Kim Soo-sung , Jeong Yong-jin , Park Jong-man , Lee Yoon-sang
Published: November 2014

This study was carried out to establish an electron beam welding process for nuclear plate-type fuel assembly fabrication. A preliminary investigation for plate fuel fabrication was conducted with a consideration of the weld performance using AA6061-T6 aluminum alloy made through the EBW (Electron Beam Welding) process. The optimum welding parameters for the plate-type fuel assembly were obtained in terms of the accelerating voltage, beam current and welding time. The welds made by the optimum parameters showed slightly lower tensile strengths than those of the un-welded specimens. The integrity of the welds by the EBW process was confirmed by the results of the tensile test, an examination of the macro-cross sections and the fracture surfaces of the welded specimens.
Subcooled Flow Boiling Heat Transfer in Volumetrically Heated Packed Bed
Authors: Guangzhan Xu , Sun Zhongning , Meng Xianke , Zhang Xiaoning , Su Jiqiang , Zhou Yanmin
Published: November 2014

The volumetrically heated packed bed has been widely utilized in modern industry, however, due to the variability and randomness of packed bed channels, flow boiling heat transfer characteristics is complex. To study subcooled flow boiling heat transfer characteristics of volumetrically heated packed bed, here electromagnetic induction heating method was used to heat the oxidized carbon steel balls adopted to stack packed bed, while water was utilized as refrigerant in the experiment. The mass flux and heat flux effects on the subcooled flow boiling heat transfer were analyzed. A new correlation was developed to predict the subcooled flow boiling heat transfer coefficient in volumetrically heated packed bed.
Development of a Thermal-Hydraulic Analysis Code and Transient Analysis for a FHTR
Authors: Xiao Yao , Hu Lin-wen , Qiu Suizheng , Zhang Dalin , Su Guanghui , Tian Wenxi
Published: November 2014

The Fluoride-salt-cooled High-temperature Reactor (FHR) is an advanced reactor concept that uses high temperature TRISO fuel with a low-pressure liquid salt coolant. Design of Fluoride-salt-cooled High-temperature Test Reactor (FHTR) is a key step in the development of the FHR technology and is currently in progress both in China and the United States. An FHTR based on pebble bed core design with coolant temperature 600–700 °C is being planned for construction by the Chinese Academy of Sciences (CAS)’s Thorium Molten Salt Reactor (TMSR) Research Center, Shanghai Institute of Applied Physics (SINAP). This paper provides preliminary thermal hydraulic transient analyses of an FHTR using SINAP’s pebble core design as a reference case. A point kinetic model is calculated by developing a microcomputer code coupling with a simplified porous medium heat transfer model in the core. The founded models and developed code are applied to analyze the safety characteristics of the FHTR by simulating basic transient conditions including the unprotected loss of flow, unprotected overcooling, and unprotected transient overpower accidents. The results show that the SINAP’s pebble core design is an inherently safe reactor design.
Study of Decay Heat Removal Approach for Advanced Passive PWR During Station Blackout
Authors: Zou Jie , Tong Lili , Cao Xuewu
Published: November 2014

After Fukushima accident, decay heat removal in station blackout (SBO) accident is concerned for different NPP design. Advanced passive PWR relies on passive systems to cool reactor core and containment, such as the passive residual heat removal system (PRHR), passive injection system and passive containment cooling system (PCCS). Passive safety systems are considered more reliable than traditional active safety system under accident condition. However, in long-term SBO situation, possible failure of passive safety systems is noticed as active valves are needed in system actuation. Moreover, probability safety analysis results of advanced passive PWR show that system failure is possible without external event. Given different passive safety system failure assumptions, response of reactor coolant system and containment of advanced passive PWR is calculated in SBO accident, the integrity of core, reactor pressure vessel and containment is assessed, and decay heat removal approach is studied. The results show that containment failure is predicted with the failure of PCCS and PRHR, reactor vessel failure together with containment failure is predicted with the failure of PCCS, passive injection system and PRHR. Advices to deal with the risk of advanced passive PWR in SBO are given based on the study.
Design of 500 MWe Metal Fuel Demonstration Fast Reactor
Authors: Perumal Chellapandi , Balasubramaniyan V. , Puthiyavinayagam P. , Sundararajan Raghupathy , Kanakkil Madhusoodanan , Selvaraj P. , Rao P. R. Vasudeva
Published: November 2014

Indian nuclear power programme is being implemented in three stages taking in to consideration limited uranium resources and vast thorium resources in the country. The first stage consists of investing natural uranium in Pressurized Heavy Water Reactors (PHWR). This stage has the potential of 10 GWe. The second stage involves large scale deployment of Fast Breeder Reactors (FBR) with co-located fuel cycle facilities to utilize the plutonium and depleted uranium extracted from the PHWR spent fuel. This stage has a potential of about 300 GWe. In the third stage, effective utilization of the vast thorium resources is planned. Indira Gandhi Centre for Atomic Research (IGCAR) instituted in 1971 at Kalpakkam, is involved in the mission of developing the technology of FBR. A host of multidisciplinary laboratories are established in the centre around the central facility of the 40 MWt Fast Breeder Test Reactor (FBTR). Presently, the construction of indigenously designed MOX fueled 500 MWe Prototype Fast Breeder Reactor (PFBR) that started in 2003 is in advanced stage and commissioning activities are underway. The design of PFBR incorporates several state-of-art features and is foreseen as an industrial scale techno-economic viability demonstrator for the FBR program. Beyond PFBR, the proposal is to build one twin unit having two reactors, with each of improved design compared to PFBR, to be commissioned by 2025. Subsequently, towards rapid realization of nuclear power, the department is planning a series of metal fueled FBRs starting with a 500 MWe Metal fuel Demonstration Fast Breeder Reactor (MDFR-500) to be followed by industrial scale 1000 MWe metal fueled reactors. The paper discusses in detail the above aspects and highlights the activities carried out towards designing MDFR.
Effect of Thermal Stratification on Full-Cone Spray Performance in Reactor Containment for a Scaled Scenario
Authors: Paranjape Sidharth , Mignot Guillaume , Paladino Domenico
Published: November 2014

The results of an experimental study on the nuclear reactor containment spray system are presented. Depending on the initial conditions, the spray nozzle configuration and flow rates, the spray may cause higher hydrogen concentration during depressurization due to steam condensation, or it may erode the hydrogen stratification by enhanced mixing. To investigate these phenomena, the tests are performed using a full-cone spray nozzle in PANDA facility at Paul Scherrer Institut, Switzerland. Temporal evolution and spatial distribution of the fluid temperature and the fluid concentrations are measured using thermocouples and mass spectrometers. Two tests are performed with initial vessel wall temperatures of 105°C and 135°C, which create condensing and non-condensing environments respectively. The different initial conditions lead to different density stratifications. The effect of these different density stratification on the flow patterns and mixing of gases in the vessels due to the action of the spray is revealed by these tests.
CFD Analysis of the Flow in Crud-Coated Nuclear Rod Bundles
Authors: Cinosi Nicolas , Walker Simon P. , Bluck Mike , Issa Raad I.
Published: November 2014

This paper presents a methodology to assess the effects of crud surface roughnesses on the normal operation of a PWR reactor.

A CFD methodology based on the use of RANS models and the implementation of roughness embedded in STAR-CCM+ CFD code has been developed and applied.

A typical PWR reactor case is investigated in details.

Results show how the presence of crudded rough rods induces a modification of the flow distribution as compared to the ideal case with clean smooth rods.

Notably, these modifications reflect a reduction of the mass flow rate, which in turn has a deteriorating consequence on the efficiency of the fuel-to-coolant heat transfer.
Best Estimate Analysis of Maanshan PWR LBLOCA by TRACE Coupling With DAKOTA
Authors: Yang Jung-Hua , Wang Jong-Rong , Lin Hao-Tzu , Shih Chunkuan
Published: November 2014

This research is focused on the Large Break Loss of Coolant Accident (LBLOCA) analysis of the Maanshan power plant by TRACE-DAKOTA code. In the acceptance criteria for Loss of Coolant Accidents (LOCAs), there are two accepted analysis methods: conservative methodology and best estimate methodology. Compared with conservative methodology, the best estimate and realistic input data with uncertainties to quantify the limiting values i.e., Peak Cladding Temperature (PCT) for LOCAs analysis. By the conservative methodology, the PCTCM (PCT calculated by conservative methodology) of Maanshan power plant LBLOCA calculated is 1422K. On the other hand, there are six key parameters taken into account in the uncertainty analysis in this study. In PCT95/95 (PCT of 95/95 confidence level and probability) calculation, the PCT95/95 is 1369K lower than the PCTCM (1422K). In addition, the partial rank correlation coefficients between input parameters and PCT indicate that accumulator pressure is the most sensitive parameter in this study.
Development of a Spatial Neutronics Model for Control-Oriented Dynamics Simulation
Authors: Lorenzi Stefano , Cammi Antonio , Luzzi Lelio , Ponciroli Roberto
Published: November 2014

In control-oriented simulators, the neutronics is usually modeled by implementing the point-wise kinetics. In the framework of the study of the control strategy for innovative reactor concepts, such a simplified description is less effective since prevents the possibility of exploiting the capabilities of advanced control schemes. In order to overcome these limitations, a spatial neutronics model based on Modal Method has been considered. This approach allows separating the spatial and time dependence of the neutron flux, which can be represented as sum of the eigenfunctions of the multi-group neutron diffusion equation weighted by time-dependent coefficients. In this way, the system dynamic behavior is reduced to the study of these time-dependent coefficients and can be represented by a set of Ordinary Differential Equations (ODEs), reducing the computational burden. In this paper, a test case involving three fuel pins of an innovative Lead-cooled Fast Reactor has been set up and investigated. Once obtained the eigenfunctions, the set of ODEs for studying the time-dependent coefficients has been implemented in the MATLAB environment [1]. Finally, in order to assess the performance of the developed model, the outcomes have been compared with the results obtained from the neutron diffusion partial differential equation, achieving a satisfactory agreement.
High Efficiency Gamma-Beta Blind Alpha Spectrometry for Nuclear Energy Applications
Authors: Webster Jeffrey A. , Hagen Alexander , Archambault Brian C. , Hume Nicholas , Taleyarkhan Rusi
Published: November 2014

A novel, Centrifugally Tensioned Metastable Fluid Detector (CTMFD) sensor technology has been developed over the last decade to demonstrate high selective sensitivity and detection efficiency to various forms of radiation for wide-ranging conditions (e.g., power level, safeguards, security, and health physics) relevant to the nuclear energy industry. The CTMFD operates by tensioning a liquid with centrifugal force to weaken the bonds in the liquid to the point whereby even a femto-scale nuclear particle interactions can break the fluid and cause a detectable vaporization cascade. The operating principle has only peripheral similarity to the superheated bubble chamber based superheated droplet detectors (SDDs); instead, CTMFDs utilize mechanical “tension pressure” instead of thermal superheat offering a lot of practical advantages. CTMFDs have been used to detect a variety of alpha and neutron emitting sources in near real-time. The CTMFD is selectively blind to gamma photons and betas allowing for detection of alphas and neutrons in extreme gamma/beta background environments such as spent fuel reprocessing plants or under full power conditions within an operating nuclear reactor itself. The selective sensitivity allows for differentiation between alpha emitters including the isotopes of Plutonium. Mixtures of Plutonium isotopes have been measured in ratios of 1:1, 2:1, and 3:1 Pu-238:Pu-239 with successful differentiation. Due to the lack of gamma-beta background interference, the CTMFD’s LLD can be effectively reduced to zero and hence, is inherently more sensitive than scintillation based alpha spectrometers or SDDs and has been proven capable to detect below femtogram quantities of Plutonium-238. Plutonium is also easily distinguishable from Neptunium making it easy to measure the Plutonium concentration in the NPEX stream of a UREX reprocessing facility. The CTMFD has been calibrated for alphas from Americium (5.5 MeV) and Curium (∼6 MeV) as well. The CTMFD has furthermore, recently also been used to detect spontaneous and induced fission events which can be differentiated from alpha decay allowing for detection of fissionable material in a mixture of isotopes. This paper discusses these transformational developments which are also being entered for real-world commercial use.
Sensitivity Analyses of a Simulation Model for Estimating Fiber-Induced Sump Screen and Core Failure Rates
Authors: Morton David , Letellier Bruce , Tejada Jeremy , Johnson David , Mohaghegh Zahra , Kee Ernie , Moiseytseva Vera , Reihani Seyed , Zolan Alexander
Published: November 2014

Output from a high-order simulation model with random inputs may be difficult to fully evaluate absent an understanding of sensitivity to the inputs. We describe, and apply, a sensitivity analysis procedure to a large-scale computer simulation model of the processes associated with Nuclear Regulatory Commission (NRC) Generic Safety Issue (GSI) 191. Our GSI-191 simulation model has a number of distinguishing features: (i) The model is large in scale in that it has a high-dimensional vector of inputs; (ii) some model inputs are governed by probability distributions; (iii) a key model output is the probability of system failure — a rare event; (iv) the model’s outputs require estimation by Monte Carlo sampling, including the use of variance reduction techniques; (v) it is computationally expensive to obtain precise estimates of the failure probability; (vi) we seek to propagate key uncertainties on model inputs to obtain distributional characteristics of the model’s outputs; and, (vii) the overall model involves a loose coupling between a physics-based stochastic simulation sub-model and a logic-based Probabilistic Risk Assessment (PRA) sub-model via multiple initiating events. Our proposal is guided by the need to have a practical approach to sensitivity analysis for a computer simulation model with these characteristics. We use common random numbers to reduce variability and smooth output analysis; we assess differences between two model configurations; and, we properly characterize both sampling error and the effect of uncertainties on input parameters. We show selected results of studies for sensitivities to parameters used in the South Texas Project Electric Generating Station (STP) GSI-191 risk-informed resolution project.
Analysis of Melt Progression and Debris Cooling at the Fukushima Daiichi Unit 3 by the SAMPSON code
Authors: Pellegrini Marco , Mizouchi Hideo , Suzuki Hiroaki , Naitoh Masanori
Published: November 2014

The progression of core melting in a nuclear power plant is characterized by various complications due mainly to chemical reactions and interactions occurring between core materials, which modify the main core temperatures, affecting the time when melt is effectively beginning. On top of it the geometry of a Boiling Water Reactor (BWR) is introducing additional challenges to code modelers. Challenges exist because of the necessity to adopt approximations in the core discretization, which might result in incorrect predictions of the actual phenomena. In particular large uncertainties exist for a clear definition of the fuel and control rod interfaces, discretization of the lower core region as well as control rod guide tubes and monitor penetrations in the lower head. The severe accident code SAMPSON, with an advanced characterization of the core and lower plenum region, is able to recreate a more realistic representation of the phenomena occurring during the melt progression. In this work, SAMPSON description of molten fuel relocation and debris spreading/cooling are presented. State of the art simulation by the SAMPSON code of Fukushima Daiichi unit 3 has shown that core melting progression develops until zirconium is contained within the core, triggering large heat release during oxidation. Thereafter core conditions almost stabilize and 35% of the core falls into the lower plenum, where pipe penetration ejection is predicted releasing molten material within the pedestal.
Study on a Small Power Reactor With Compact Pressure Vessel and Natural Circulation
Authors: Takiwaki Kenya , Sakurai Shungo , Takeuchi Yutaka , Yamamoto Yasushi
Published: November 2014

There is movement which is developing the small reactor for the small electricity grid in place of a big power reactor which requires the high capital cost. This paper introduces a small power reactor whose purpose is to achieve high economic competitiveness and advanced safety. In order to attain high economic competitiveness, it is designed to be small and simple and uses natural circulation and high pressure. A steam generator is integrated into the reactor pressure vessel (RPV), thus dispensing with a primary system and preventing radiation leakage from the reactor core. The small core is designed to have a high power density (100 MW/m3, almost twice that of a conventional boiling water reactor). The concept of a 300 MWt (100 MWe) core design is established by introducing a boiling heat transfer system. By boiling cooling water, the cooling-water circulating flow quantity in a reactor core is enlarged.

By increasing a flow, the minimum critical power ratio is improved, which is an important core characteristic. Furthermore, using a burnable poison (Gd2O3), the excess reactivity of a reactor core is reduced and excess reactivity is controlled only by the control rod.

Moreover, the maximum linear power density is improved and the critical power ratio is minimized by optimizing the burnable poison arrangement and the control rod pattern.

In order to attain high safety, our small reactor has an advanced decay heat removal system that can cool the core without external support. This decay heat removal system is part of the secondary cooling system and combined with a cooling tower. As a result, the quantity of cooling water stored in the decay heat removal system is reduced, and longtime decay heat removal is possible by small equipment.
Validation DualSPHysics Code for Liquid Sloshing Phenomena
Authors: Buruchenko Sergei K. , Crespo Alejandro J. C.
Published: November 2014

The DualSPHysics code is proposed as a numerical tool for the simulation of liquid sloshing phenomena. A particular type of sloshing motion can occur during the core meltdown of a liquid metal cooled reactor (LMR) and can lead to a compaction of the fuel in the center of the core possibly resulting in energetic nuclear power excursions. This phenomenon was studied in series of “centralized sloshing” experiments with a central water column collapsing inside the surrounding cylindrical tank. These experiments provide data for a benchmark exercise for accident analysis codes. To simulate “centralized sloshing” phenomena, a numerical method should be capable to predict the motion of the free surface of a liquid, wave propagation and reflection from the walls. The DualSPHysics code based on the smoothed particle hydrodynamics method was applied to the simulation of “centralized sloshing” experiments. Simulation results are compared with the experimental results. In a series of numerical calculations it is shown that overall motion of the liquid is in a good agreement with experimental observations. Dependence on the initial and geometrical symmetry is studied and compared with experimental data.
Pseudo-Transient Simulation of Turbulent Mixing in a Rectangular Channel
Authors: Otic Ivan , Chai Xiang , Cheng Xu
Published: November 2014

Fast and robust numerical approach for turbulent heat transfer in case of high gradients of physical properties and unsteady heat transfer in case of strong temperature fluctuations is developed. This pseudo transient Large Eddy Simulation approach is applied to perform transient calculations of turbulent mixing between helium and air in a rectangular mixing channel for Atwood numbers of 0.04 and 0.6. Comparisons of these numerical results with the experimental results by Banerjee, Kraft, Andrews (2010) show a good agreement. The results confirmed applicability of the pseudo transient approach also for higher Atwood numbers while wider time stepping and higher Courant numbers are used. Also reported are simulation results using Reynolds averaged (RANS) method where standard k–ε and k–ω SST models are applied.
Advancement of Reactor Coolant Pump (RCP) Performance Verification Test in KAERI
Authors: Cho Yun-Je , Kim Yeon-Sik , Cho Seok , Kim Seok , Bae Byoung-Uhn , Chung Heung-June , Youn Young-Jung , Park Jong-Kuk , Choi Hae-Seob , Jeon Woo-Jin , Kim Bok-Deuk , Kwon Tae-Soon , Song Chul-Hwa
Published: November 2014

Korea Atomic Energy Research Institute (KAERI) has designed and constructed a test facility for reactor coolant pumps (RCPs). The RCP Test Facility (RCPTF) has the capability to test a RCP under the operation condition of an Advanced Power Reactor 1400 MW (APR1400). The design values of the facility are 17.2 MPa, 343 °C, 11.7 m3/s, and 13 MW in maximum pressure, temperature, flow rate, and electrical power, respectively. In the facility, it is possible to perform a type test for a newly-developed RCP as well as a production test for a RCP before its installation in a nuclear power plant. For the production test, H-Q curves under the cold and hot conditions are acquired. For the type test, various transient tests are additionally performed including four types of seal transient tests, a thrust bearing transient test, a cost down test, and so on.

To acquire H-Q curves of a RCP, the flow rate should be controlled by varying the flow resistance in the test loop. The RCPTF uses a Variable Restriction Orifice (VRO) whose flow area can be controlled by moving the two orifice plates installed in-parallel. The need for flow control valves and bypass lines was eliminated using the VRO such that the flow disturbance was minimized. The flow rate in the main loop of the RCPTF is measured by a standard venture flow meter. The flow rate in the RCPTF is very high and thus the venture flow meter could not be calibrated in the entire range of Reynolds number corresponding to the operating condition in the APR1400. The calibration was conducted at the Colorado Experiment Engineering Station Inc. (CEESI) in the USA where natural gas is used for a working fluid. If a discharge coefficient calibrated with the gas is applied in the test results performed using the water as a working fluid, a discrepancy can occur due to the static hole error. Therefore, the static hole error was compensated in the test results and the result shows the improvement.

The effect of the temperature on the pressure pulsation amplitude was also evaluated. During a cold performance test and heat-up phase to the condition of a hot performance test, an abnormal increase in the pressure pulsation amplitude was observed near the specific temperature range. This is acoustic resonance phenomena that occur when a blade passing frequency of the RCP is proportional to the harmonic resonance frequency of the RCPTF.
Loss of Representativeness of the Fixed Incore Instrumentation of the EPR™ Reactor
Authors: Jardiné Brice , Bougeant Olivier , Pfeiffer Maxime
Published: November 2014

The EPR™ reactor features a fixed incore instrumentation, composed of 72 Self Powered Neutron Detectors (SPND), that provides the online reconstruction of the core maximum Linear Power Density (LPD) and minimum Departure from Nucleate Boiling Ratio (DNBR). The Instrumentation and Control (I&C) systems of the EPR™ reactor use this online reconstruction in surveillance and protection functions. The onsite thresholds of those I&C functions have to take into account all the uncertainties affecting the online reconstruction of core power distribution measured by SPNDs. One of these uncertainties is the so-called Loss Of Representativeness (LOR). This uncertainty is defined as the difference between the LPD (respectively DNBR) physical value and the LPD (respectively DNBR) computed value using SPND signals. The LOR parameter is mostly linked to the difference between the core power distribution at the time where SPNDs are calibrated and the core power distribution at the time where their signals are used. For the DNBR, LOR also takes into account the use of a simplified on-line DNBR calculation algorithm. A statistical approach is used in order to define this uncertainty. The analysis is based on the evaluation of different sets of core power distributions generated thanks to random drawings of the plant state parameters (including power level, core inlet temperature, pressure, control rod insertion and xenon distribution). The sets of core configurations representative of normal plant operation are used to define the surveillance thresholds. The sets representative of accidental transients (for which the LPD and DNBR protections are claimed) are used to define the protection thresholds. The analysis of LOR values provides an envelop probability law covering a minimum of 95% of LOR values. In order to derive the on-site threshold for LPD and DNBR, a Monte Carlo method is used to propagate the LOR probability law and the other uncertainties. Sensitivity calculations have been performed in order to cover a large spectrum of fuel loading patterns and to take into account SPND failures. In conclusion, this approach allows defining an optimized and robust set of thresholds for the on-line surveillance and protection system of EPR™ reactor.
Design of Advanced Controllers for Pressure Control in Nuclear Reactors: A General Approach
Authors: Cappelli M. , Castillo-Toledo B. , Di Gennaro S.
Published: November 2014

When a failure occurs in a nuclear plant, a lack in the response of the controller could lead to serious consequences. The fundamental property to be ensured by the controller is the plant stability, possibly proved formally, and at least in the range of validity of the process model. In this work, using a mathematical model for the primary circuit of a PWR, accurate enough to catch the nonlinear, time–varying, and switching nature of the system, and suitable for the control purposes, the reactor power controller, the inventory controller for the primary circuit, and the pressurizer pressure controllers are designed. These controllers do not use direct measurements of the pressurizer pressure or temperature, but they use instead pressurizer wall temperature measurements. Disturbances and parameter variations are compensated by the use of sliding–mode terms, which guarantee further robustness to the control scheme. The switching nature of the controller, reflecting the switching nature of the pressurizer dynamics, and the nonlinear terms implemented in the controllers, along with classical PI actions, ensure better transient behaviors. Hence, they represent an evolution and an improvement with respect to classical PID controllers, usually implemented in standard control actions.
Decommissioning of Obsolete Nuclear Facilities in the UJV Rez, a. s.
Authors: Tous Milan , Podlaha Josef
Published: November 2014

More than 55 years of activities in the company UJV Rez, a. s. (Nuclear Research Institute Rez a.s. in the past) which is a leading institution in all areas of nuclear R&D in the Czech Republic and had a dominant position in the nuclear program since it was established (1955), there are several obsolete nuclear facilities that shall be decommissioned.

The total amount of radioactive waste (RAW) resulting from decommissioning for the next processing will be ∼ 1500 m3 and the expected amount RAW for releasing into the environment is 240 tons after the decontamination. For the RAW processing several decontamination methods such as high press water jetting, chemical treating in ultrasonic bath, dry ice blasting and abrasive blasting were performed.

Decommissioning started in 2003 and will be finished in 2016. This decommissioning of nuclear facilities in UJV is the only ongoing decommissioning project in the Czech Republic.
Effects of Aging on the Thermal Conductivity of the AP1000® Containment Vessel Inorganic Zinc Coating
Authors: Kindred Thomas A. , Wright Richard F.
Published: November 2014

During a loss-of-coolant accident (LOCA) event the AP1000® passive safety features actuate to provide emergency core cooling (ECC) to the reactor core using passive features that do not rely on electrical power being available. The core makeup tanks (CMTs), and accumulators (ACCs) actuate to quench the fuel rods and refill the reactor vessel. After the CMTs and ACCs empty, the in-containment refueling water storage tank (IRWST) utilizing gravity injects a large volume of sub-cooled fluid into the reactor vessel. This floods the vessel and the lower region of containment (containment sump) initiating gravity induced long-term recirculation cooling. The discharge of high energy fluid during the blowdown, re-flood, and re-fill phases is assumed to condense on the colder structures inside the containment including the containment vessel shell. Heat is transferred through the shell to the film of water from the Passive Containment Cooling System (PCS) applied to the outside of the containment vessel shell. This results in evaporative heat transfer on the outside of the containment vessel. Due to the large heat transfer coefficients on the inside and outside of the shell the heat conduction through the shell is very important to the heat rejection capability of the PCS, and plays a large part in ensuring the containment vessel pressure is not exceeded during design basis events. The AP1000® containment vessel is forged from a high strength carbon steel alloy that is coated with an inorganic zinc coating which protects the containment vessel from corrosion during its design life. The coating acts as a sacrificial anodic layer which corrodes in lieu of corrosion of the substrate beneath it. The corrosion of the coating can potentially lead to degradation in thermal conductivity of the coating due to metallic oxides typically having a lower thermal conductivity than that of the non-oxidized state. A reduction in thermal conductivity of the protective coating will impact the overall heat transfer through the containment vessel during PCS operation. The purpose of this work is to develop a mechanistic model demonstrated against empirical validation for assessing the effects of oxidation on the thermal conductivity of the protective inorganic zinc coating (IOZ) on the AP1000® containment vessel.
Security Informed Safety Assessment of NPP I&C Systems: GAP-IMECA Technique
Authors: Kharchenko Vyacheslav S. , Illiashenko Oleg A. , Kovalenko Andriy A. , Sklyar Vladimir V. , Boyarchuk Artem V.
Published: November 2014

The application of complex electronic components such as systems-on-chips (including systems-on-programmable-chips using field programmable gate arrays (FPGAs)) in industrial instrumentation and control systems (I&Cs) causes risks for ensuring of safety. Nuclear power plant (NPP) I&C projects on FPGA are complex solutions which include both software and hardware components.

Information security (system’s ability to protect the information and data from unauthorized access and modification) is a subordinate property with respect to safety of many I&Cs, primarily to the NPP reactor trip systems. Such hierarchy may be taken into account by implementation of security informed safety (SIS) approach.

Recent events like Stuxnet or Duqu showed vulnerabilities in industrial embedded IT-Systems. In order to remove or reduce security risks, which could increase overall safety risk, the holistic analytical technique are necessary.

The goal of the paper is to present the technique for of SIS-based assessment of the NPP I&C systems. The proposed SIS-oriented method of NPP I&C systems assessment includes the models and techniques.

To decrease the risk of manual errors, the tool for the SIS-oriented assessment automation is described. The tool is based on joint use of abovementioned models and techniques, is proposed.

The tool allows conducting the joint use of the following analysis techniques: GAP and IMECA.
Validation of In-Core Measurements
Authors: Machek Jindřich
Published: November 2014

Paper describes original methods of signal validation, specially developed for in-core measurements, namely for Self Powered Neutron Flux Detectors (SPND) and for core exit coolant temperature measurements (thermocouples measurements). Some interesting examples of NPP Temelin data validation (SPND) and NPP Dukovany (thermocouple measurements) are presented; precision and sensitivity of methods used are discussed.
The Application of Probabilistic Safety Assessment in the Preliminary Reactor Design Stage: Challenges and Insights
Authors: Alrammah Ibrahim A.
Published: November 2014

This paper discusses some technical issues related to applying Probabilistic Safety Assessment (PSA) to a novel Nuclear Power Plant (NPP) design. These aspects include: initiating events, passive systems modeling, reliability and common-cause failure (CCF) data, modeling of novel design features, modeling of preventive maintenance and technical specifications, human reliability analysis (HRA), modeling of instrumentation and control (I&C), external hazards, PSA supporting studies and interpretation of PSA results for new plants.
Infrared and thermoelectric power generation in thin atomic layer deposited Nb-doped TiO2 films
Authors: Mann Harkirat S. , Lang Brian N. , Schwab Yosyp , Niemelä Janne-Petteri , Karppinen Maarit , Scarel Giovanna
Published: November 2014

Infrared radiation is used to radiatively transfer heat to a nanometric power generator (NPG) device with a thermoelectric Nb-doped TiO2 film deposited by atomic layer deposition (ALD) as the active element, onto a borosilicate glass substrate. The linear rise of the produced voltage with respect to the temperature difference between the “hot” and “cold” junctions, typical of the Seebeck effect, is missing. The discovery of the violation of the Seebeck effect in NPG devices combined with the ability of ALD to tune thermoelectric thin film properties could be exploited to increase the efficiency of these devices for energy harvesting purposes.
Nearly Complete Genome Sequence of Curionopolis Virus, a
Culicoides
-Related Rhabdovirus Isolated in the Brazilian Amazon Region
Authors: Medeiros Daniele Barbosa de Almeida , Diniz Júnior José Antonio P. , Cardoso Jedson F. , Silva Sandro Patroca , da Silva Daisy Elaine Andrade , de Oliveira Layanna F. , é Vasconcelos Janaina M. , Chiang Jannifer O. , Dias Amarilis Aragão , Nunes Márcio R. T. , Vianez Júnior João Lídio da S. G. , Vasconcelos Pedro F. C.
Published: November 2014

We report here the first nearly complete genome sequence related to curionopolis virus (CURV), that of strain AR440009, isolated from a pool of
Culicoides
sp. midges in Serra Norte, Pará State, northern Brazil. All genes showed similarities to those belonging to members of the family
Rhabdoviridae
.
Draft Genome Sequence of
Microvirga
sp. Strain BSC39, Isolated from Biological Soil Crust of Moab, Utah
Authors: Bailey Alexis C. , Kellom Matthew , Poret-Peterson Amisha T. , Noonan Kathryn , Hartnett Hilairy E. , Raymond Jason
Published: November 2014

Microvirga
sp. BSC39 was isolated from a biological soil crust near Moab, Utah. The strain appears to be capable of chemotaxis and exopolysaccharide synthesis for biofilm adhesion. The BSC39 genome contains iron siderophore uptake and hydrolysis enzymes; however, it lacks siderophore synthesis pathways, suggesting the uptake of siderophores produced by neighboring microbes.
Draft Genome Sequence of
Massilia
sp. Strain BSC265, Isolated from Biological Soil Crust of Moab, Utah
Authors: Bailey Alexis C. , Kellom Matthew , Poret-Peterson Amisha T. , Noonan Kathryn , Hartnett Hilairy E. , Raymond Jason
Published: November 2014

Massilia
sp. BSC265 was isolated from a biological soil crust near Moab, Utah. The strain appears to be capable of chemotaxis and exopolysaccharide synthesis for biofilm adhesion. The BSC265 genome contains a complete dissimilatory nitrate reduction pathway as well as a TCA cycle, making it a facultative anaerobe.
Cardiovascular evaluation of middle-aged individuals engaged in high-intensity sport activities: implications for workload, yield and economic costs
Authors: Menafoglio Andrea , Di Valentino Marcello , Porretta Alessandra Pia , Foglia Pietro , Segatto Jeanne-Marie , Siragusa Patrick , Pezzoli Reto , Maggi Mattia , Romano Gian Antonio , Moschovitis Giorgio , Gallino Augusto
Published: November 2014

Background

The European Association of Cardiovascular Prevention and Rehabilitation (EACPR) recommends cardiovascular evaluation of middle-aged individuals engaged in sport activities. However, very few data exist concerning the impact of such position stand. We assessed the implications on workload, yield and economic costs of this preventive strategy.

Methods

Individuals aged 35–65 years engaged in high-intensity sports were examined following the EACPR protocol. Athletes with abnormal findings or considered at high-cardiovascular risk underwent additional examinations. The costs of the overall evaluation until diagnosis were calculated according to Swiss medical rates.

Results

785 athletes (73% males, 46.8±7.3 years) were enrolled over a 13-month period. Among them, 14.3% required additional examinations: 5.1% because of abnormal ECG, 4.7% due to physical examination, 4.1% because of high-cardiovascular risk and 1.6% due to medical history. A new cardiovascular abnormality was established in 2.8% of athletes, severe hypercholesterolaemia in 1% and type 2 diabetes in 0.1%. Three (0.4%) athletes were considered ineligible for high-intensity sports, all of them discovered through an abnormal ECG. No athlete was diagnosed with significant coronary artery disease on the basis of a high-risk profile or an exercise ECG. The cost was US$199 per athlete and US$5052 per new finding.

Conclusions

Cardiovascular evaluation of middle-aged athletes detected a new cardiovascular abnormality in about 3% of participants and a high-cardiovascular risk profile in about 4%. Some of these warranted exclusion of the athlete from high-intensity sport. The overall evaluation seems to be feasible at reasonable costs.
The new car following model considering vehicle dynamics influence and numerical simulation
Authors: Sun Dihua , Liu Hui , Zhang Geng , Zhao Min
Published: November 2014

In this paper, the car following model is investigated by considering the vehicle dynamics in a cyber physical view. In fact, that driving is a typical cyber physical process which couples the cyber aspect of the vehicles' information and driving decision tightly with the dynamics and physics of the vehicles and traffic environment. However, the influence from the physical (vehicle) view was been ignored in the previous car following models. In order to describe the car following behavior more reasonably in real traffic, a new car following model by considering vehicle dynamics (for short, D-CFM) is proposed. In this paper, we take the full velocity difference (FVD) car following model as a case. The stability condition is given on the base of the control theory. The analytical method and numerical simulation results show that the new models can describe the evolution of traffic congestion. The simulations also show vehicles with a more actual acceleration of starting process than early models.