bluuebunny/crossref_metadata_2025 huggingface dataset:
Analyses of the Start-Up and Shut-Down of SCWR Fuel Qualification Test Loop
Authors: Vojacek Ales , Mazzini Guido
Published: November 2014

In the development of the SuperCritical Water Reactor (SCWR), it is planned to carry out a fuel qualification test (FQT) on a small scale fuel assembly in the research reactor LVR-15 located in Research Center Rez (CZ). The subject of the SCWR-FQT project is to design and license a facility containing the small scale fuel assembly comprised of 4 fuel pins with wire wraps, including the required loop with its safety and auxiliary systems. The present paper focuses on the transient analyses for start-up and shut-down of the FQT loop, particularly to explore the ability of system codes ATHLET 3.0A to simulate the transient between subcritical conditions and supercritical conditions. Simulations of heating up of the SCWR-FQT loop and achieving nominal supercritical conditions (Tiio=390 °C and Pio=25 MPa for primary circuit) were carried out. Further, analyses of shut-down of the SCWR-FQT loop from its supercritical conditions by shutting down the reactor and cooling down the loop until subcritical conditions were performed. Particular attention was dedicated for the models of heat transfer in supercritical condition.
Comparison of Nuclear Fitness for Service Codes: ASME Section XI and RSE-M AFCEN Codes
Authors: Faidy Claude
Published: November 2014

Two major Codes are used for Fitness for Service of Nuclear Power Plants: one is the ASME B&PV Code Section XI and the other one is the French RSE-M Code. Both of them are largely used in many countries, partially or totally.

The last 2013 RSE-M covers “Mechanical Components of Pressurized Water Reactors (PWRs):

- Pre-service and In-service inspection

- Surveillance in operation or during shutdown

- Flaw evaluation

- Repairs-Replacements parts for plant in operation

- Pressure tests

The last 2013 ASME Section XI covers “Mechanical components and containment of Light Water Reactors (LWRs)” and has a larger scope with similar topics: more types of plants (PWR and Boiling Water Reactor-BWR), other components like metallic and concrete containments…

The paper is a first comparison covering the scope, the jurisdiction, the general organization of each section, the major principles to develop In Service Inspection, Repair-Replacement activities, the flaw evaluation rules, the pressure test requirements, the surveillance procedures (monitoring…) and the connections with Design Codes…

These Codes are extremely important for In-service inspection programs in particular and essential tools to justify long term operation of Nuclear Power Plants.
Fatigue Design Rules in Pressure Vessel Codes Uncertainties and Remaining Open Points
Authors: Faidy Claude
Published: November 2014

During the past 30 years the main rules for fatigue analysis of pressure vessels were based on elastic approaches in order to evaluate a cyclic strain amplitude and compare with an S-N fatigue curve for the corresponding material. After review of some rules in different Nuclear and Non Nuclear Codes, like ASME Boiler & Pressure Vessel Code Section III, French RCC-M and RCC-MRx, European Standards EN 13445, the major conservatisms and uncertainties of different rules are discussed.

All these Codes propose simple rules to evaluate the strain amplitude based on elastic approaches and simplified correction factors (Ke and Kv), transient combination rules and damage cumulating procedure.

In the other hand, the material properties are based on standard fatigue tests done on the material associated to reduction factors to consider some particular effects like scatter, scale, surface roughness, mean stress or environmental effects to transfer them from small specimen to real structures.

Concerned components are mainly piping systems.

No existing Code covers all the aspects of fatigue with similar conservatisms that can affect the in-service inspection programs and the remaining life assessment of the corresponding components.

After the review of different rules, key factors that affect the results and predictions will be identified.

Some proposals will be issued to progress in the near future. Finally, a first set of recommendation on fatigue analysis will be presented to improve existing codes on an harmonized way, associated to material properties needed, as fatigue curves associated to reduction factors.
An Alternative Approach to Prediction of Critical Heat Flux: Projection Support Vector Regression
Authors: Jiang Botao , Zhao Fuyu
Published: November 2014

Critical heat flux (CHF) is one of the most crucial design criteria in other boiling systems such as evaporator, steam generators, fuel cooling system, boiler, etc. This paper presents an alternative CHF prediction method named projection support vector regression (PSVR), which is a combination of feature vector selection (FVS) method and support vector regression (SVR). In PSVR, the FVS method is first used to select a relevant subset (feature vectors, FVs) from the training data, and then both the training data and the test data are projected into the subspace constructed by FVs, and finally SVR is applied to estimate the projected data. An available CHF dataset taken from the literature is used in this paper. The CHF data are split into two subsets, the training set and the test set. The training set is used to train the PSVR model and the test set is then used to evaluate the trained model. The predicted results of PSVR are compared with those of artificial neural networks (ANNs). The parametric trends of CHF are also investigated using the PSVR model. It is found that the results of the proposed method not only fit the general understanding, but also agree well with the experimental data. Thus, PSVR can be used successfully for prediction of CHF in contrast to ANNs.
Two-Way Coupling Between the Reactor Dynamics Code DYN3D and the Fuel Performance Code TRANSURANUS at Assembly Level
Authors: Holt L. , Rohde U. , Seidl M. , Schubert A. , Van Uffelen P. , Macián-Juan R.
Published: November 2014

In the last two decades the reactor dynamics code DYN3D was coupled to thermal hydraulics system codes, a sub-channel thermal hydraulics code and CFD codes. These earlier developed code systems allow modeling of the thermal hydraulics phenomena occurring during reactor transients and accidents in greater detail. Still these code systems lack a sufficiently sophisticated fuel behavior model, which is able i.e. to take into account the fission gas behavior during normal operation, off-normal conditions and transients. To our knowledge a two-way coupling to a fuel performance code hasn’t so far been reported in the open literature for calculating a full core with detailed and well validated fuel behavior models.

A new two-way coupling approach between DYN3D and the fuel performance code TRANSURANUS is presented. In the coupling, DYN3D provides the time-dependent rod power and thermal hydraulics conditions to TRANSURANUS, which in turn transfers parameters like fuel temperature and cladding temperature back to DYN3D. The main part of the development is a general TRANSURANUS coupling interface that is applicable for linking of any other reactor dynamics codes, thermal hydraulics system codes and sub-channel codes to TRANSURANUS. Beside its generality, other features of this interface are the application at either fuel assembly or fuel rod level, one-way or two-way coupling, automatic switching from steady to transient conditions in TRANSURANUS (including update of the material properties etc.), writing of all TRANSURANUS output files and the possibility of manual pre- and post-calculations with TRANSURANUS in standalone mode. The TRANSURANUS code can be used in combination with this coupling interface in various scenarios: different fuel compositions in the reactor types BWR, PWR, VVER, HWR and FBR, considering time scales from milliseconds (i.e. RIA) over seconds/ minutes (i.e. LOCA) to years (i.e. normal operation) and thence different reactor states.

Results of DYN3D-TRANSURANUS are shown for a control rod ejection transient in a German PWR. In particular, it appears that for all burn-up levels the two-way coupling approach systematically calculates higher maximum values for the node fuel enthalpy (max. difference of 46 J/g) and node centerline fuel temperature (max. difference of 181 K), compared to DYN3D standalone in best estimate calculations. These differences can be completely explained by the more detailed TRANSURANUS modeling of fuel thermal conductivity, radial power density profile and heat transfer in the gap. As known from fuel performance codes, the modeling of the heat transfer in the gap is sensitive and causes also larger differences in case of low burn-up.

The numerical convergence for DYN3D-TRANSURANUS is quick and stable. The coupled code system can improve the assessment of safety criteria, at a reasonable computational cost with a CPU time of less than seven hours without parallelization.
HFE Safety Evaluation in Angra-1 NPP for the Preparation and Inclusion of a Chapter 18 in the Plant Final Safety Analysis Report
Authors: Alonso Pedro Trueba , Valdivia Juan Carlos , Illobre Luís Fernández , Hulsmans Mark
Published: November 2014

Angra-1 Nuclear Power Station (Westinghouse PWR-600 MW, 2 loops) started commercial operation in 1985, being property of Eletronuclear, subsidiary of Eletrobras in Brazil. Angra-1 has been preparing the necessary measures to renew the operating license and to apply for a lifetime extension up to 60 years.

Among the many activities to perform, there are some related to fulfilling the requirements of the Brazilian regulator, the CNEN. These include requirements related to Human Factors Engineering (HFE) that included the preparation of a Chapter 18 of HFE, to become part of the plant’s Final Safety Analysis Report (FSAR).

In the framework of the Instrument for Nuclear Safety Cooperation (INSC), created and funded by the European Union (EU) to enhance nuclear safety world-wide, cooperation activities between the EU and the Government of Brazil were set up in 2009. One of the INSC projects funded was to support the Brazilian nuclear operator of Angra-1 in the field of HFE. In 2010, the implementation of the project was awarded to a consortium lead by Tecnatom for performing a HFE Safety Evaluation to the plant and to provide support for preparing this Chapter 18.

For this Project a specific methodology was developed for the execution of the Safety Evaluation. The methodology has been developed for evaluating — from the HFE viewpoint — a plant in operation, from the beginning of commercial operation until nowadays, including the design modifications performed to date. The obtained results have been used for developing the aforementioned Chapter 18.

The main results of the Project Execution have been:

1. The developed methodology has made it possible to perform a comprehensive HFE evaluation of Angra-1, including the analysis of Post-TMI requirements, the design included in the current FSAR, the existing Angra-1 procedures and the verification of the current Main Control Room.

2. Technical support has been provided to Angra-1 for the preparation of Chapter 18 of the FSAR, following the structure of NUREG-0711, and using the results of the HFE Safety Evaluation.

3. An Action Plan has been developed for identifying and addressing in the future all those deficiencies found during the HFE Safety Evaluation, as well as those activities that are the consequence of the new FSAR Chapter 18.
Comparative Analysis of High Void Fraction Regimes Using an Averaging Euler-Euler Multi-Fluid Approach and a Generalized Two-Phase Flow (GENTOP) Concept
Authors: Montoya Gustavo , Baglietto Emilio , Lucas Dirk , Krepper Eckhard , Hoehne Thomas
Published: November 2014

Complex multiphase gas-liquid flows, including boiling, are usually encountered in safety related nuclear applications. For CFD purposes, modeling the transition from low to high void fraction regimes represents a non-trivial challenge due to the increasing complexity of its interface. For example, churn-turbulent and slug flows, which are typically encountered for these gas volume fraction ranges, are dominated by highly deformable bubbles. Multiphase CFD has been so far relying on an averaged Euler-Euler simulation approach to model a wide range of two-phase applications. While this methodology has shown to date demonstrated reasonable results (Montoya et al., 2013), it is evidently highly dependable on the accuracy and validity of the mechanistic models for interfacial forces, which are necessary to recover information lost during the averaging process. Unfortunately existing closures, which have been derived from experimental as well as DNS data, are hardly applicable to high void fraction highly-deformable gas structures. An alternative approach for representing the physics behind the high void fraction phenomena, is to consider a multi-scale method. Based on the structure of the gas-liquid interfaces, different gaseous morphologies should be described by different CFD approaches, such as interface tracking methods for larger than the grid size interfacial-scales, or the averaged Euler-Euler approach for smaller than grid size scales, such as bubbly or droplet flow. A novel concept for considering flow regimes where both, dispersed and continuous interfacial structures, could occur has been developed in the past (Hänsch et al., 2012), and has been further advanced and validated for pipe flows under high void fraction regimes (Montoya et al., 2014) and other relevant cases, such as the dam-break with an obstacle (Hänsch et al., 2013). Still, various short-comings have been shown in this approach associated mostly to the descriptive models utilized to obtain the continuous gas morphology from within the averaged Eulerian simulations. This paper presents improvements on both concepts as well as direct comparison between the two approaches, based on newly obtained experimental data. Both models are based on the bubble populations balance approach known as the inhomogeneous MUltiple SIze Group or MUSIG (Krepper et al., 2008) in order to define an adequate number of bubble size groups with its own velocity fields. The numerical calculations have been performed with the commercially available ANSYS CFX 14.5 software, and the results have been validated using experimental data from the MT-Loop and TOPFLOW facilities from the Helmholtz-Zentrum Dresden-Rossendorf in Germany (Prasser et al., 2007).
Decommissioning Combustible Waste Treatment Using Oxygen-Enriched Incinerator
Authors: Min Byung Youn , Lee Yoon Ji , Yun Gyoung Su , Lee Ki Won , Moon Jae Kwon
Published: November 2014

The aim of the paper is current status of treatment for the decommissioning combustible waste in KAERI. For the purpose of the volume reduction and clearance for decommissioning combustible wastes generated by the decommissioning projects in Korea, the incineration technology has been selected for the treatment of combustible wastes. About 34 tons of decommissioning combustible waste has been treated using Oxygen Enriched incineration. Temperature, pressure of major components, stack gas concentration, i.e., SOx, NOx, CO, CO2 and HCl, and the residual oxygen were measured. Measured major parameters during normal operation were sustained on a stable status within a criteria operation condition. Oxygen enriched air, 22vol. % (dry basis) was used for stable incineration. The volume reduction ratio has achieved about 1/117.
Application of an Integrated Head Assembly Concept at Pressurized Water Reactor Commercial Nuclear Plants
Authors: Baliga Ravi , Watts Tom Neal , Kamath Harish
Published: November 2014

In a typical pressurized water reactor commercial nuclear plant, a number of components such as CRDMs, a lift rig to lift the Reactor Vessel Closure Head (RVCH), seismic restraints, missile shield, and a cooling system with large air ducts are installed on or directly over the RVCH. These components and systems are typically designed and installed individually to perform designated functions during plant operation.

During refueling outages the removal of the RVCH from the pressure vessel and its subsequent re-installation on the pressure vessel for fuel loading requires individual dismantling and reassembly of these components resulting in an expensive and time-consuming process. Prior to detensioning the RVCH from the vessel, a lengthy series of steps or detailed procedures must be followed to safely remove the head area components and to store them in their designated spaces inside containment. The procedure generally includes: removal and storage of the concrete missile shield; removal and storage of CRDM cooling ducts; removal of seismic restraints; removal of head area cables; installation of the tripod assembly over the service structure; disconnecting the vent and level indicator lines; and installation of temporary lead blankets around the RVCH. Once the refueling is complete, these procedural steps are repeated in reverse order.

Each procedure in the refueling process contributes significantly to the total cost associated with personnel time required to perform the refueling, power plant down time and consequent loss of electricity production, radiation exposure to personnel, and risks and costs associated with potential human errors. In addition, these components require a large amount of storage space inside containment raising the risk of having inadvertent contamination of work and storage areas.

To reduce the outage duration and the associated radiation exposure to the workers, the authors have designed an Integrated Head Assembly (IHA) for Callaway nuclear plant based on Mr. Baliga’s patented design as disclosed in U.S. Patents. The IHA is an assembly of all head area components integrally attached to the RVCH so that all these components can be lifted with the RVCH in one assembly (see Figure 1). The IHA also provides a forced air convection system that improves the efficiency of the CRDM cooling. The IHA reduces a significant amount of critical path time and radiation dosage during refueling outages.

Mr. Baliga’s invention has been implemented at several commercial nuclear plants in the USA (Turkey Point Units 3 and 4; Salem Units 1 and 2; DC Cook Units 1 and 2; Diablo Canyon Units 1 and 2; Davis Besse Unit 1; and Callaway Unit 1). This paper provides details of the IHA design implemented at Callaway nuclear plant in the USA.
An Experimental Study on the Critical Heat Flux for Low Flow of Water in Bilaterally Heated Vertical Annuli Under Low Pressure Conditions
Authors: Zhou Lei , Liu Guangxu , Zan Yuanfeng , Yan Xiao
Published: November 2014

Critical heat flux (CHF) has been widely studied in the past decades because of its importance for nuclear power plant design. But most of the studies are based on flow under normal operating conditions for light water reactors. CHF under low flow and low pressure is of significance when considering operating transients and accidents. In this study, experimental study has been carried out on CHF for low flow rate and low pressure water flow in vertical bilaterally heated annuli. Parameter trends on CHF is discussed and a new predictive correlation was fitted based on the CHF data points. This study is meaningful for concerned nuclear engineering and similar experiment design.
Eulerian Two-Phase Flow Modeling of Steam Direct Contact Condensation for the Fukushima Accident Investigation
Authors: Pellegrini Marco , Agostinelli Giulia , Okada Hidetoshi , Naitoh Masanori
Published: November 2014

Steam condensation is characterized by a relatively large interfacial region between gas and liquid which, in computational fluid dynamic (CFD) analyses, allows the creation of a discretized domain whose average cell size is larger than the interface itself. For this reason generally one fluid model with interface tracking (e.g. volume of fluid method, VOF) is employed for its solution in CFD, since the solution of the interface requires a reasonable amount of cells, reducing the modeling efforts. However, for some particular condensation applications, requiring the computation of long transients or the steam ejected through a large number of holes, one-fluid model becomes computationally too expensive for providing engineering information, and a two-fluid model (i.e. Eulerian two-phase flow) is preferable. Eulerian two-phase flow requires the introduction of closure terms representing the interactions between the two fluids in particular, in the condensation case, drag and heat transfer. Both terms involve the description of the interaction area whose definition is different from the typical one adopted in the boiling analyses. In the present work a simple but effective formulation for the interaction area is given based on the volume fraction gradient and then applied to a validation test case of steam bubbling in various subcooling conditions. It has been shown that this method gives realistic values of bubble detachment time, bubble penetration for the cases of interest in the nuclear application and in the particular application to the Fukushima Daiichi accident.
An Analysis of a Three-Region Conduction Controlled Rewetting of Hot Surface by Variational Method
Authors: Agrawal Manish K. , Garg Krati , Singh Aditya , Sahu Santosh K.
Published: November 2014

When the liquid comes in contact with a sufficiently hot surface, a stable vapor blanket is formed between solid–liquid interface, this significantly reduces the heat transfer rate between liquid and solid interface because of the poor conduction through the vapour layer. As the process continues the surface cools off and the vapor blanket collapses, consequently, the liquid wets the hot surface and this phenomenon is termed as rewetting. The phenomena of rewetting has been observed in many industrial and scientific applications such as emergency core cooling of fuel pins during loss of coolant accident (LOCA) in nuclear reactor, mettalurgical processing and cryogenic chill down during refueling of space transfer vehicles.
Effect of Physical Properties on Gas Entrainment Rate From Free Surface by Vortex (2nd Report)
Authors: Koizumi Yasuo , Ohte Naosuke , Kamide Hideki , Ohno Shuji , Ito Kei
Published: November 2014

A sodium-cooled fast breeder reactor is now at the developing stage in Japan. One concern for safety is cover gas entrainment into the sodium coolant. The gas entrainment rate into liquid by the vortex formed on the free surface was examined experimentally. Four kinds of test fluid were used; water at 25 °C, water at 60 °C, 20 cSt silicone oil and kerosene. Gas was air. The flow state of gas entrainment was visually observed by using a high speed video camera. The gas entrainment rate into liquid was measured. Following conclusions were obtained. When exit velocity was low, bubble-type gas entrainment occurred. As the exit velocity increased, the gas entrainment type turned from the bubble type to a vortex type and gas entrainment rate considerably increased. The relation between gas entrainment rate and liquid velocity was mainly affected by the viscosity of liquid. As viscosity became large, higher exit velocity was required to get the same gas entrainment rate. The effect of surface tension on the gas entertainment rate was minor or little. No systematic trend by the surface tension was noticed in the gas entrainment rate. Present results of the onset of the bubble type gas entrainment are well expressed with both the modified Baum and the modified Takahashi et al. correlation although the modified Takahashi et al. correlation provides a little better prediction than the modified Baum correlation. A flow state at the outlet piping has significant effect on the gas entrainment rate. The dimension of the outlet piping may become important to consider the gas entrainment rate in the vortex type region.
Development of Inherently Safe Technologies for Large Scale BWRs: (3) Infinite-Time Air-Cooling System
Authors: Tamura Akinori , Kawamura Toshinori , Ishida Naoyuki , Kitou Kazuaki
Published: November 2014

To address long-term station black outs, which occurred at the Fukushima Nuclear Power Station, we have been developing the infinite-time air-cooling system which operates without electricity by a natural circulation loop. The air-cooling heat exchanger, which is located outside the primary containment vessel of a reactor, transfers the decay heat to the atmosphere by natural circulation resulting from the density difference of the air. Improvement in the heat-transfer performance of air-cooling is a key technology in the development of the infinite-time air-cooling system. In this paper, we developed the air-cooling enhancing technology for the infinite-time air-cooling system by using a micro-fabrication surface, turbulence-enhancing structures, and heat-transfer fins. To evaluate the performance of this air-cooling enhancing technology, we conducted a heat exchange test using an element test apparatus. A single tube of the air-cooling heat exchanger, which includes a sheath heater and thermo-couples, was used. The air flow outside the tube and the heat quantity were respectively controlled using an air-compressor and the sheath heater. The heat-transfer performance was calculated from the heat-quantity and temperature difference measured using thermo-couples. The developed air-cooling enhancing technology demonstrated superior heat-transfer performance in this test. The heat-transfer performance increased approximately 100 % with this technology compared with a bare pipe. From these experimental results, we confirmed good feasibility for implementing the infinite-time air-cooling system.
Safety Assessment and Accident Analysis of VVER-1000/466B With Active and Passive Safety Systems for Belene NPP
Authors: Sabotinov Luben , Dimitrov Borislav , Bruna Giovanni B.
Published: November 2014

The paper presents the methodology adopted to assess the Interim Safety Analysis Report (ISAR) of the Belene NPP in the framework of the contract between the Bulgarian Nuclear Regulatory Authority (BNRA) and RISKAUDIT (IRSN&GRS). It stresses the in-depth analysis carried-out for several relevant-to-safety issues and illustrates in some detail the investigation of the Large Break Loss of Coolant Accident (LB LOCA) with loss of power and failure of the active part of the Emergency Core Cooling System (High Pressure and Low Pressure Safety Injection pumps), performed with the French best estimate thermal-hydraulic code CATHARE. The role, problems and efficiency of the passive and active safety systems during the accident scenarios are discussed. Finally, the main conclusions of the safety evaluation of the Belene NPP project are summarized.
Lead Fast Reactor Sustainability
Authors: Ciotti Marco , Manzano Jorge L. , Grasso Giacomo , Mansani Luigi , Petrovich Carlo
Published: November 2014

The electricity production systems, especially those based on nuclear fission, are increasingly facing more tight constraints and are subjected to more deep analyses based on the three aspects of economical sustainability, environmental sustainability and social sustainability. Nuclear Reactors future development has been outlined in the framework of the GIF (Generation IV International Forum), where the Lead Fast Reactor (LFR) is placed among the most promising innovative solutions. Many aspects of LFR offer a huge improvement from different points of view. The non pressurization of the system and the absence of sources of hazardous chemical potential energy enhances consistently its safety aspects, improving the perception of inherent safety of the Generation IV (G4) reactors in the public opinion. At the moment, due to the abundance of the new fossil resources, the competitiveness of Nuclear Power Plants is severely challenged, this aspect representing the most difficult to manage, besides the public acceptability. Moreover, for G4 reactors, an additional “cost premium” associated with the innovative technological concept has to be taken into account. Conversely, looking at the mid-term future, the real economical comparison has to be performed considering as competing sources, according to the IPCC recommendations and constraints enacted by the European Community, only CO2 free sources. In this context, economical competitiveness could be regained depending on the “cost premium” to be added to fossil fuels to become CO2 free, through the improvement of the carbon separation and storage techniques. The intrinsic lead properties (e.g.: low absorption cross section) permit to easily design LFR flexible cores, optimized with respect to a number of possible goals, as a long-lived core with minimal reactivity swing intended for battery concepts, or what is called an “adiabatic” core, where the entire Pu and MA inventory in the spent fuel can be indefinitely reused in a closed fuel cycle. The latter option allows to limit the waste throughput to the fission products only (along with the — unavoidable — losses from fuel reprocessing), and to benefit of natural resources minimization. These are both specific Generation IV goals envisioned to reach nuclear energy sustainability. An overall fuel cycle balance in a scenario with a step by step introduction of LFR reactors fleet grown in a specific geographical area, is in details analyzed in [1] and presented in this conference.
In-Service Inspection and On-Line Monitoring of Inaccessible Components in Nuclear Power Plants Using Guided Wave Technology
Authors: Turcu Florin , Cappelli Mauro , Mazzini Davide , Pistelli Sergio , Raugi Marco
Published: November 2014

One of the most challenging problems in the on-line monitoring of critical parameters of nuclear plants is the inspection of components that result inaccessible or difficult to reach. In this context, there is an increasing interest of the scientific community and industry for the use of Ultrasonic Guided Waves (UGW) for addressing this issue.

In this work, the problem of the applicability of the UGW technique with magnetostrictive sensors to NPP structures is described, together with the outline of the related advantages as well as the main technical concerns that may arise from such applications. This methodology has been tested on experimental activities concerning high temperature applications. Results show the effectiveness of such an approach.
The ARCADIA Project in Support of the ALFRED Concept
Authors: Diaconu Daniela , Constantin Marin , Grasso Giacomo , Glinatsis Georgios , Di Gabriele Fosca , Alemberti Alessandro , De Bruyn Didier , Urbonavicius Egidijus
Published: November 2014

The FP7 Euratom project ARCADIA (Assessment of Regional CApabilities for new reactors Development through an Integrated Approach), launched in November 2013, has the aim to assess the regional capabilities of the EU New Member States (NMS) to effectively contribute in bringing closer the European research to ESNII (European Sustainable Nuclear Industrial Initiative) and NUGENIA (NUclearGEneration II&III Association) objectives. In particular, for the ESNII, the project will work in support to the ALFRED concept towards its realization in Romania, and for NUGENIA it will approach safety aspects of Gen III/III+.

The project focuses on the identification of the primary needs for the ALFRED project, mainly to what concerns engineering and technology, supporting infrastructures, competences and regulatory aspects. At the same time, it investigates the existing national and regional supporting structures — with a particular attention to the ones in Romania and in all the participating NMS — for defining a map of competences potentially eligible to satisfy the identified needs. Since the most of the existing competences and experiences rely on present generation systems, the review will also provide a frame for potential support to NUGENIA on Gen III/III+ systems, as well as the starting point for building the Gen IV culture and future capabilities.

On the other hand, ARCADIA takes the profit of the large participation to the project itself to carry out an advising activity to the ALFRED Consortium (FALCON), addressing recommendations regarding the national and international best practices and guidelines for the advancement of the ALFRED project.

Investigation of the existing Research Reactors, among the Project Partner, for their potential involvement in future (potential) experimental activities in support of both the ALFRED project realization and the Gen III/III+ safety activities will be also considered. Networking, Cooperation and Dissemination activities are also included in the project in order to provide the connection with the international scientific community, with the European institutional organizations and with the general public. This work gives an overview of the project objectives, structure and contents, with focus on the international effort towards the GenIV.
Design and R&D Progress of China Lead-Based Reactor
Authors: Wu Yican , Bai Yunqing , Song Yong , Huang Qunying , Zhao Zhumin , Song Gang , Hu Liqin , Jiang Jieqiong
Published: November 2014

Liquid lead or lead-based alloy is a potential candidate coolant for fast reactors and ADS subcritical reactors because of its many unique nuclear, thermal-physical and chemical attributes. Chinese Academy of Sciences (CAS) had launched an engineering project to develop ADS system and lead-based reactors. Series CLEAR reactor conceptual designs were finished, and the preliminary engineering design for the China Lead-based research reactor (CLEAR-I) was underway. The key components prototypes for engineering validation including the control rod drive system, refueling system, fuel assembly have been constructed, the validation experiment are carrying out. KYLIN series PbBi experimental loops has already been built to perform structure material corrosion experiment, thermal-hydraulics experiment and safety experiment for CLEAR series reactors. The highly intensified neutron generator HINEG will be constructed to take the benchmark experiment of neutronics simulation codes. In this paper, the design and R&D progress are presented.
Modeling of a Reactor Trip System Using Markov Chains: Case Study
Authors: Butenko Valentina
Published: November 2014

Markov models (MM) are widely used in dependability assessment of complex safety-critical systems, such as NPP I&C system. The main computational difficulties in case of using MM are model size and stiffness, which pose a problems in its construction, storage and solution. Selection of the solution approach and method, based on analysis of such MM features as stiffness and complexity, increases the assessment accuracy. Result of such analysis helps in making decision between direct and indirect research techniques and set of software packages (SP) to provide the high accurate assessment results.

This paper presents the case study for safety assessment of NPP I&Cs. This is a two-channel FPGA-based Reactor Trip System with three parallel tracks on voting logic “2-out-of-3” in each channel. Several solution techniques and SPs were used to analyze and describe the ways in which main modelling risks can be avoid. Analysis of case study results using different SP allows to formulate few application problems: importance of usability-oriented SP selection in case of solving complex MM; achieving an accurate result for stiff MM; support the results verification to ensure the needed level of confidence.
A 3D, Interactive Virtual Instruction Laboratory and a Virtual Model of a Reactor Control Room
Authors: Li Ye , Haddish Imran , Zhu Xuefeng , Satoh Yoshinori , Uddin Rizwan
Published: November 2014

Efficient and effective education and training of nuclear engineering students, nuclear reactor operators, their supervisors, and other personnel are critical for the safe operation and maintenance of any nuclear reactor, whether for research or commercial power generation. Radiation and reactor laboratories are a very important part of such training. Recent increase in the student population in nuclear engineering programs has put strains on laboratory resources. This increase in student population, constraints on resources and qualitative improvements in gaming technology have led researchers in the field of radiological and nuclear engineering to explore virtual, game-like models to provide the needed experience [1–3]. Though virtual lab experience may never completely replace an actual physical lab experience in educational institutions, in some ways virtual labs may provide a better experience than limited cookbook style executions in a physical lab or reactor operator training course.

We have earlier reported our initial efforts toward the development of a generic virtual and interactive laboratory environment [3]. This virtual lab presents a fully immersive learning experience. We here report the specifics of a radiation lab in which half-life and shielding experiments can be conducted, and simulation-based real-physics data can be gathered.
Plant Event Signatures on Neutron Noise Data
Authors: Yang Jianfeng
Published: November 2014

Neutron noise monitoring and analysis has been a valuable tool for in-service monitoring of the structural integrity of reactor internals. Several nuclear power plants have experienced structural degradation in their reactor internals. Significant signatures were noticed in the neutron noise data when the degradations occurred. This article briefly summarizes the findings based on these experiences. A significant amount of neutron noise time-history data has been obtained based on the continuous neutron noise data monitoring and analysis for pressurized water reactors (PWRs). Over the past decade, power plants have made many major component repairs and replacements. This article correlates the plant operating history with its corresponding signatures on the neutron noise data. This information is an extremely valuable reference for diagnosing the condition of the reactor internals through neutron noise data monitoring and analyses.
Characteristics of Pressure Buildup From Local Fuel-Coolant Interactions in a Simulated Molten Fuel Pool
Authors: Cheng Songbai , Matsuba Ken-ichi , Isozaki Mikio , Kamiyama Kenji , Suzuki Tohru , Tobita Yoshiharu
Published: November 2014

Studies on local fuel-coolant interactions (FCI) in a molten pool are important for severe accident analyses of sodium-cooled fast reactors (SFRs). To clarify the mechanisms underlying this interaction, in this study a series of experiments was conducted by delivering a given quantity of water into a simulated molten fuel pool (formed with a low-melting-point alloy). Based on the experimental data obtained from a variety of conditions, including difference in water volume, melt temperature and water subcooling, the characteristics of pressure-buildup during local FCIs was investigated. It is found that under our experimental conditions the water volume and melt temperature have remarkable impact on the interaction, while the role of water subcooling seems to be less prominent. The performed analyses also suggest that the pressurization from local FCIs should be intrinsically limited, due to a suppressing role caused by the increasing of coolant volume entrapped within the pool as well as the transition of boiling mode. Current work, which gives a palette of favorable data for a better understanding and an improved estimation of severe accidents in SFRs, is expected to benefit future analyses and verifications of computer models developed in advanced fast reactor safety analysis codes.
Atomic layer deposition of magnesium fluoride via bis(ethylcyclopentadienyl)magnesium and anhydrous hydrogen fluoride
Authors: Hennessy John , Jewell April D. , Greer Frank , Lee Michael C. , Nikzad Shouleh
Published: November 2014

A new process has been developed to deposit magnesium fluoride (MgF2) thin films via atomic layer deposition (ALD) for use as optical coatings in the ultraviolet. MgF2 was deposited in a showerhead style ALD reactor using bis(ethylcyclopentadienyl)magnesium and anhydrous hydrogen fluoride (HF) as precursors at substrate temperatures from 100 to 250 °C. The use of HF was observed to result in improved morphology and reduced impurity content compared to other reported MgF2 ALD approaches that use metal fluoride precursors as the fluorine-containing chemistry. Characterization of these films has been performed using spectroscopic ellipsometry, atomic force microscopy, and x-ray photoelectron spectroscopy for material deposited on silicon substrates. Films at all substrate temperatures were transparent at wavelengths down to 190 nm and the low deposition temperature combined with low surface roughness makes these coatings good candidates for a variety of optical applications in the far ultraviolet.
Right ventricular strain analysis from three‐dimensional echocardiography by using temporally diffeomorphic motion estimation
Authors: Zhang Zhijun , Zhu Meihua , Ashraf Muhammad , Broberg Craig S. , Sahn David J. , Song Xubo
Published: November 2014

Purpose:

Quantitative analysis of right ventricle (RV) motion is important for study of the mechanism of congenital and acquired diseases. Unlike left ventricle (LV), motion estimation of RV is more difficult because of its complex shape and thin myocardium. Although attempts of finite element models on MR images and speckle tracking on echocardiography have shown promising results on RV strain analysis, these methods can be improved since the temporal smoothness of the motion is not considered.

Methods:

The authors have proposed a temporally diffeomorphic motion estimation method in which a spatiotemporal transformation is estimated by optimization of a registration energy functional of the velocity field in their earlier work. The proposed motion estimation method is a fully automatic process for general image sequences. The authors apply the method by combining with a semiautomatic myocardium segmentation method to the RV strain analysis of three‐dimensional (3D) echocardiographic sequences of five open‐chest pigs under different steady states.

Results:

The authors compare the peak two‐point strains derived by their method with those estimated from the sonomicrometry, the results show that they have high correlation. The motion of the right ventricular free wall is studied by using segmental strains. The baseline sequence results show that the segmental strains in their methods are consistent with results obtained by other image modalities such as MRI. The image sequences of pacing steady states show that segments with the largest strain variation coincide with the pacing sites.

Conclusions:

The high correlation of the peak two‐point strains of their method and sonomicrometry under different steady states demonstrates that their RV motion estimation has high accuracy. The closeness of the segmental strain of their method to those from MRI shows the feasibility of their method in the study of RV function by using 3D echocardiography. The strain analysis of the pacing steady states shows the potential utility of their method in study on RV diseases.
Pseudomonas Strains Naturally Associated with Potato Plants Produce Volatiles with High Potential for Inhibition of Phytophthora infestans
Authors: Hunziker Lukas , Bönisch Denise , Groenhagen Ulrike , Bailly Aurélien , Schulz Stefan , Weisskopf Laure
Published: November 2014

Bacteria emit volatile organic compounds with a wide range of effects on bacteria, fungi, plants, and animals. The antifungal potential of bacterial volatiles has been investigated with a broad span of phytopathogenic organisms, yet the reaction of oomycetes to these volatile signals is largely unknown. For instance, the response of the late blight-causing agent and most devastating oomycete pathogen worldwide,
Phytophthora infestans
, to bacterial volatiles has not been assessed so far. In this work, we analyzed this response and compared it to that of selected fungal and bacterial potato pathogens, using newly isolated, potato-associated bacterial strains as volatile emitters.
P. infestans
was highly susceptible to bacterial volatiles, while fungal and bacterial pathogens were less sensitive. Cyanogenic
Pseudomonas
strains were the most active, leading to complete growth inhibition, yet noncyanogenic ones also produced antioomycete volatiles. Headspace analysis of the emitted volatiles revealed 1-undecene as a compound produced by strains inducing volatile-mediated
P. infestans
growth inhibition. Supplying pure 1-undecene to
P. infestans
significantly reduced mycelial growth, sporangium formation, germination, and zoospore release in a dose-dependent manner. This work demonstrates the high sensitivity of
P. infestans
to bacterial volatiles and opens new perspectives for sustainable control of this devastating pathogen.
Propionyl Coenzyme A (Propionyl-CoA) Carboxylase in Haloferax mediterranei: Indispensability for Propionyl-CoA Assimilation and Impacts on Global Metabolism
Authors: Hou Jing , Xiang Hua , Han Jing
Published: November 2014

Propionyl coenzyme A (propionyl-CoA) is an important intermediate during the biosynthesis and catabolism of intracellular carbon storage of poly(3-hydroxybutyrate-
co
-3-hydroxyvalerate) (PHBV) in haloarchaea. However, the haloarchaeal propionyl-CoA carboxylase (PCC) and its physiological significance remain unclear. In this study, we identified a PCC that catalyzed propionyl-CoA carboxylation with an acetyl-CoA carboxylation side activity in
Haloferax mediterranei
. Gene knockout/complementation demonstrated that the PCC enzyme consisted of a fusion protein of a biotin carboxylase and a biotin-carboxyl carrier protein (PccA [HFX_2490]), a carboxyltransferase component (PccB [HFX_2478]), and an essential small subunit (PccX [HFX_2479]). Knockout of
pccBX
led to an inability to utilize propionate and a higher intracellular propionyl-CoA level, indicating that the PCC enzyme is indispensable for propionyl-CoA utilization. Interestingly,
H. mediterranei
DBX (
pccBX
-deleted strain) displayed multiple phenotypic changes, including retarded cell growth, decreased glucose consumption, impaired PHBV biosynthesis, and wrinkled cells. A propionyl-CoA concentration equivalent to the concentration that accumulated in DBX cells was demonstrated to inhibit succinyl-CoA synthetase of the tricarboxylic acid cycle
in vitro
. Genome-wide microarray analysis showed that many genes for glycolysis, pyruvate oxidation, PHBV accumulation, electron transport, and stress responses were affected in DBX. This study not only identified the haloarchaeal PCC for the metabolism of propionyl-CoA, an important intermediate in haloarchaea, but also demonstrated that impaired propionyl-CoA metabolism affected global metabolism in
H. mediterranei
.
Novel Alginate‐Gelatin Hybrid Nanoparticle for Drug Delivery and Tissue Engineering Applications
Authors: Lee Eun Mi , Singh Deepti , Singh Dolly , Choi Soon Mo , Zo Sun Mi , Park Seon Joo , Han Sung Soo
Published: November 2014

Novel alginate‐gelatin hybrid nanoparticles were fabricated using single oil in water (O/W) emulsification techniques. Physicochemical property of the particle was characterized using scanning electron microscopy and Fourier’s transmission infrared spectroscopy. Particle size was determined using zeta potential metastasize analyzer and was found to be in range of 400–600 nm. AGNPs were used for culturing human keratinocytes for two weeks to check biocompatibility of synthesized AGNPs. 3‐(4,5‐Dimethylthiazol‐2‐yl)‐2,5‐diphenyltetrazolium bromide (MTT) assay showed increased metabolic activity of cells cultured on AGNPs in comparison to two‐dimensional (2D) system (control). Cellular attachment on nanoparticle was further confirmed using SEM and 4′,6‐diamidino‐2‐phenylindole staining. The drug release profile shows possible electrostatic bond between alginate and gelatin resulting in controlled release of drug from AGNPs. For the first time alginate‐gelatin hybrid nanosystem has been fabricated and all results showed it can be used as potential system for delivery of drug and therapeutical agents to cells and can also be used for regenerative medicine applications.