bluuebunny/crossref_metadata_2025 huggingface dataset:
The Effect of Ti, N and V Content and Heat Treatment on Irradiation and Mechanical Property of SCRAM Steels
Authors: Yu Chi , Yang Feng , Suo Jinping
Published: November 2014

An Fe-Cr-W-V-Ti-N steel named SCRAM (super-clean reduced-activation martensitic) steel was designed for the first wall and blanket structure of fusion power plants. Compared with the Fe-Cr-W-V-Ta steel, TiN can precipitate first at 1650°C rather than TaC by the Thermal-cal Calculation. And we take vacuum induction melting (VIM) and electro-slag re-melting (ESR) together to manufacture the SCRAM steel, which can make the TiN fine and the steel pure. Mechanical properties and microstructures of SCRAM steels irradiated with single-beam (Fe) and sequential-beam (Fe plus He and Fe plus H) at 300°C were studied. The results show that, SCRAM steel can have better mechanical properties before and after irradiation while Ti was doped into the SCRAM steel. It has been reported that the precipitation formed in the steel has effect on the mechanical properties, irradiation properties and the ductile brittle transition temperature (DBTT). The effect of intermediate heat treatment on precipitation behavior and mechanical properties of SCRAM steel was investigated in order to obtain dispersed fine M23C6 carbides. The results indicated that MX carbonitrides precipitated first in the steel with intermediate heat treatment at 870°C rather than M23C6, which led to a decrease of carbon concentration in the supersaturated martensitic matrix and correspondingly a reduced volume fraction and mean size of M23C6. The intermediate heat treatment was beneficial to the mechanical properties, and proposed for reduction on the ductile brittle transition temperature (DBTT).
Analysis of RCS Depressurization Effects on Containment Temperature for Qinshan Phase II Nuclear Power Plant
Authors: Lili Liu , Ming Zhang , Jian Deng
Published: November 2014

A severe accident code was applied for modeling of a typical pressurized water reactor (PWR) nuclear power plant, and the effects of RCS depressurization on the gas temperature of the relief tank cell in the containment during a station blackout (SBO) induced accident was analyzed. The sensitivity calculation indicated that the hydrogen generation rate obviously increased due to RCS depressurization in a critical stage. The results show that RCS depressurization can play an important role in hydrogen generation rate and total accumulation, and the temperature of the containment atmosphere is highly influenced by hydrogen combustion. High temperature induced by hydrogen combustion may degrade the equipment and instruments capabilities. Based on this analysis, a feasible strategy of RCS depressurization for mitigating the accident consequence is provided for developing the capacity of the SBO treatment of Qinshan Phase Nuclear Power Plant (QSP-II NPP).
MARS-KS Code Analysis of the Pressure Wave Propagation Test 0 Performed at the PMK-2 Test Facility
Authors: Yu Xin-Guo , Choi Ki-Yong , Song Chul-Hwa , Trosztel Istvan , Toth Ivan , Ezsol Gyorgy
Published: November 2014

The pressure waves might be expected in the nuclear reactor systems due to sudden rupture of a pipe, quick opening or closure of a system valve. If generated, they can result in large mechanical loads on the RPV internal structures and pipelines, threating their integrity. This kind of phenomena is an important issue and a limiting accident case for the nuclear power plant safety, which requires extensive analysis to ensure the nuclear power plant safety. To study these phenomena, four PWP (Pressure Wave Propagation) tests have been performed in the PMK-2 test facility in MTA EK. In addition, these tests have been used to assess the capability of the MARS-KS code in simulating the PWP phenomena. Then, an input model representing the PMK-2 test facility was developed to simulate the tests. The MARS-KS simulation results are then compared with the test results. The comparison shows that the MARS code can well simulate the PWP frequencies and the initial pressure peaks as well. After the qualified assessment, the MARS-KS code is then deployed to conduct the sensitivity analysis on the effect of the break size, break time, coolant initial conditions on the PWP phenomena. The sensitivity analysis on the break sizes shows that the pressure wave amplitude is relevant to the break times: the shorter the break opening time is, the faster the pressure. The sensitivity analysis on the break sizes shows that the larger the break size is, the higher the pressure peak is.
Study on the Flow Characteristics for an Open Natural Circulation System
Authors: Guo Xueqing , Sun Zhongning , Wang Jianjun , Su Jiqiang
Published: November 2014

An open natural circulation system applied to the passive containment cooling system is proposed for the large dry double-wall concrete containment after severe accidents. In order to investigate the operating behaviors of the open loop, a numerical code is developed to simulate the transient flow characteristics from the startup to the long-term operation. The results show that the open system may experience such four stages as startup process, quasi-steady single-phase flow, transition stage and quasi-steady two-phase flow. Flashing is the dominated factor in the last two flow phases. Additionally, the effects of some systematic parameters on the mass flow rate are analyzed in detail. Both increasing the condensation heat transfer coefficient and the containment temperature can enhance the flow effectively whichever stage the system operates in. But the the initial fluid temperature can only affect the duration of the single-phase flow but has little impact on the mass flow rate. Moreover, greater riser height can lead to bigger single-phase flow rate, but nearly has no influence on the long-term two-phase flow and the flashing region.
Optimization of Chemical Composition in the Iron Phosphate Glass as the Matrix of High Level Waste Generated From Pyroprocessing
Authors: Kofuji Hirohide , Yano Tetsuji , Myochin Munetaka , Matsuyama Kanae , Okita Takeshi , Miyamoto Shinya
Published: November 2014

As part of ongoing research and development of nuclear waste disposal techniques suitable for the pyrochemical processing system [1], iron-phosphate glass was examined as an alternative waste form for high level waste generated from the electro-refining process [2]. To enhance the waste element content in the glass matrix and improve the durability of the waste form, optimization experiments of the glass composition were performed, and the effects of other additional transition metal oxides were determined. From the surface analysis of iron phosphate glass, a leaching mechanism was assumed for various elements contained in the glass matrix. We have selected suitable a glass composition for the treatment of radioactive waste generated from the spent electrolytes of pyrochemical processing.
Design Improvement Studies for Future SFR
Authors: Babu V. Rajan , Balasubramaniyan V. , Sundararajan Raghupathy , Puthiyavinayagam P. , Perumal Chellapandi
Published: November 2014

Prototype Fast Breeder Reactor (PFBR), a 500 MWe, (U-Pu)O2 fuelled, sodium cooled, pool type fast reactor, is in advanced stage of construction at Kalpakkam, India. Based on the experience gained during the design, manufacture and erection of various reactor components of PFBR, it is planned to construct Sodium cooled Fast Reactors (SFR) by adopting twin unit (2×500 MWe reactors) concept. The future Fast Breeder Reactor (FBR) – 1 & 2 have three main heat transport circuits, namely primary sodium, secondary sodium and steam-water systems. All the reactor internals including core and primary heat transport circuit systems are contained in a single vessel called main vessel and it is closed with top shield. Reactor assembly forms the heart of the Nuclear Steam Supply System. A detailed and exhaustive design / optimisation exercise was initiated towards improving the economic competitiveness and enhancing the safety of future FBRs. It is observed that the overall dimensions of the reactor assembly contribute immensely to the capital cost. In this context, detailed studies were carried out towards optimizing the overall dimensions of the reactor assembly. Further, the reactor assembly design in particular has been engineered to favour manufacture of integrated assembly and erection of the same, as a single unit, in reactor vault to reduce construction time. Various activities undertaken towards technology development of critical components have enhanced the confidence level in the improved design concepts and reducing time for manufacture and erection.

In addition to the reactor assembly, specific improvements have been made in decay heat removal systems and sodium purification system. The layout incorporates a twin unit concept in which the ex-vessel fuel handling system and fuel storage building are shared. This paper discusses the basis for undertaking the review exercise and experience gained during construction of PFBR and highlights the design studies and technology development carried out for future SFR.
Results of a New Production of Nuclear-Grade 1.4970 ‘15-15Ti’ Stainless Steel Fuel Cladding Tubes for GEN IV Reactors
Authors: Delville Rémi , Stergar Erich , Verwerft Marc
Published: November 2014

Titanium stabilized 1.4970 ‘15-15Ti’ stainless steel cladding is the primary choice for fuel cladding of several current fast spectrum research reactor projects. The choice of cladding material is based on past experiences and the availability of material databases from similar steel grades that have proven their reliability in past sodium-cooled fast reactors programs. However the last production in Europe of nuclear-grade 15-15Ti was more than 20 years ago and it remained to be seen if the know-how to produce such steel with the strict specifications for nuclear fuel cladding was still available. Results of a new production of nuclear-grade 15-15Ti cladding tubes at Sandvik for SCK•CEN is presented in this paper. It is shown that materials properties are within the strict specifications similar to the ones used during past sodium-cooled fast reactors programs. Special attention is given to microstructural analysis of the newly produced steel which contains a large number of stabilizing Ti(C-N) precipitates known for their beneficial effect on in-pile material properties and thermal creep. Results from metallography, SEM and TEM investigations are presented.
Prediction of Gas-Holdup in Filtered Containment Venting System
Authors: Strömgren Tobias , Perzon Sven , Karlsson Martin , Lillberg Eric
Published: November 2014

A filter system to be used in order to reduce the radioactive pollution when reducing the pressure in the containment in case of a severe accident is developed. The radioactive pollution will be captured in the vessel that acts as a bubble column. To be able to maximize the water content in order to ensure that the radioactive pollution is captured it is of importance to predict the swelled water level in the vessel. In this study is three different methods to predict the gas holdup is presented and compared to experimental results; empirical correlations, thermal-hydraulic codes and Computational Fluid Dynamics (CFD). The advantages and disadvantages with the different methods are discussed.

Both Computational Fluid Dynamics and the thermal hydraulic code (Gothic) show good predictions of the water level. Using empirical models it is crucial to have knowledge of the two-phase flow region in the bubble column.
Design of Experimental Loop With Supercritical Carbon Dioxide
Authors: Vesely Ladislav , Dostal Vaclav , Hajek Petr
Published: November 2014

Supercritical carbon dioxide (S-CO2) cycles are recently very perspective and are researched all around the world. For successful deployment of these cycles experimental research is necessary. This paper describes the design and research program of S-CO2 experimental loop that was performed in collaboration of the Czech Technical University in Prague (CTU) and the Research Centre Rez i.n.c.. The loop will be constructed at the premises of the Research Centre Rez i.n.c a part of the project SUSEN. This paper particularly focuses on the design of components, their thermotechnical calculations, construction and arrangement of the experimental loop. It also deals with the preparation of experimental measurements to be implemented. The loop is envisioned for the research of the heat transfer for various operating conditions, testing of the material in the environment of S-CO2 and sampling of gas from the loop.
Feasibility Study on Direct Waste Salt Disposition Through Measurement and Analysis of Salt Bead Dissolutions in Aqueous Solvents
Authors: Shaltry Michael , Phongikaroon Supathorn
Published: November 2014

Experimentation, modeling, and analysis of the dissolution of chloride salt beads in aqueous solvents (pure H2O, ERDA-6, and GSEEP) were conducted to provide fundamental information in support of evaluating the feasibility of directly disposing electrorefiner waste salt. Currently the waste salt is immobilized by a ceramic waste form and stored. Direct disposal of the waste salt could provide time and cost efficiencies. Dissolution tests involved LiCl-KCl eutectic, LiCl, SrCl2, CeCl3, LaCl3, PrCl3, and YCl3 as solutes at 20 °C. Bead diameter was measured and recorded at intervals while submerged in the aqueous solvent. The change in the diameter of the bead was found to be linear in time and the applied model exhibited a good fit with R2 > 0.94. Initial results indicated that dissolution was faster in water than in the brines. Further results will be presented and discussed.
The Gamma Radiation Emitted by a PWR Core Under Severe Accident Conditions
Authors: Brachem Carsten , Konheiser Jörg , Hampel Uwe
Published: November 2014

The gamma radiation emitted by a nuclear reactor core might contain information about the reactor state. This information may be used in a monitoring system for severe accidents. The Technische Universität Dresden and the Zittau/Görlitz University of Applied Sciences are currently carrying out feasibility studies for the development of such a system in a collaborative effort. As one part of such feasibility studies we performed Monte Carlo simulations on a simplified model of a generic pressurized water reactor. For a set of states which represent scenarios of a coolant level decrease and core melt, the gamma radiation distribution outside the reactor pressure vessel has been computed. The results are presented in this paper.

They indicate that different coolant levels yield different gamma radiation distributions, and that an accumulation of corium inside the lower head is detectable from the outside.
Neutron Exposure and Electrical Analysis of P-Type Boron-Doped Silicon Resistors and Transistors
Authors: Medina Hector E. , Hinderliter Brian
Published: November 2014

Neutron detectors are deployed at ports of entry across the world to monitor people and cargos for smuggled nuclear materials and are often incorporated in nuclear power plant designed to monitor power levels and ensure safe operations. With the supply of Helium-3 rapidly decreasing and due to increase of terrorism threats it is vital to U.S. national security that a viable alternative material be identified, and a new neutron detector design made available, especially for portal monitoring applications. The interest to study Boron-10 as an alternative to helium-3, due to the vast natural supply that the United States possesses and its nontoxic characteristics, is increasing in different research arenas. Our work consists on taking advantage of the near 20% of boron-10 present in naturally occurring boron which (just as phosphorus) is used to dope semiconductors. Boron doped semiconductor wafers were characterized using four-point probe techniques. Resistors and transistors made with various levels of boron concentration were exposed to a thermal flux at various fluencies at Los Alamos National Laboratory. The reaction, 10B+n → Li + α, caused by neutron irradiation, introduces impurities in the silicon lattice thus producing measurable differences in electronic parameters. These changes are likely to be proportional to the fluence of the source, and hence to the neutron flux. The results show that for irradiated resistors possessing very high values of boron concentration there is a significant reduction in resistivity. This trend is not seen for medium or low values of boron. Additionally, there was no observation of significant changes in other electronic parameters such as threshold voltage or trans-conductance, for the transistors exposed and tested.
Polycrystalline Diamond Films as Protection of Zircaloy Fuel Cladding
Authors: Škarohlíd Jan , Škoda Radek
Published: November 2014

Polycrystalline diamond coating is a promising possibility for prevention, or reduction of high temperature oxidation of zirconium alloys. Zirconium alloys are used as cladding material in almost all types of nuclear reactors, where creates a barrier between nuclear fuel and cooling water in the primary circuit. Hydrogen and considerable amount of heat is released during steam oxidation that may occur in an eventual accident. In this paper Zircaloy-2 alloy was covered by polycrystalline diamond layer using Plasma Enhanced Linear Antennas Microwave Chemical Vapor Deposition system reactor. X-Ray Diffraction and Raman spectroscopy measurements confirmed coverage of the surface area with crystalline and amorphous carbon layer. Characterizations were done for zirconium alloy covered with diamond layer before and after corrosion and irradiation tests - ion beam irradiation tests and high temperature steam exposure.
The Super Equivalence Method in Monte Carlo Based Homogenization
Authors: Li Mancang , Wang Kan , Yao Dong
Published: November 2014

The general equivalence theory (GET) and the superhomogenization method (SPH) are widely used for equivalence in the standard two-step reactor physics calculation. GET has behaved well in light water reactor calculation via nodal reactor analysis methods. The SPH was brought up again lately to satisfy the need of accurate pin-by-pin core calculations. However, both of the classical methods have their limitations. The super equivalence method (SPE) is proposed in the paper as an attempt to preserve the surface current, the reaction rates and the reactivity. It enhances the good property of the SPH method through reaction rates based normalization. The concept of pin discontinuity factors are utilized to preserve the surface current, which is the basic idea in the GET technique. However, the pin discontinuity factors are merged into the homogenized cross sections and diffusion coefficients, thus no additional homogenization parameters are needed in the succedent reactor core calculation. The eigenvalue preservation is performed after the reaction rate and surface current have been preserved, resulting in reduced errors of reactivity. The SPE has been implemented into the Monte Carlo method based homogenization code MCMC, as part of RMC Program, under developed in Tsinghua University. The C5G7 benchmark problem have been carried out to test the SPE. The results show that the SPE method not only suits for the equivalence in Monte Carlo based homogenization but also provides improved accuracy compared to the traditional GET or SPH method.
The Study and Development of Excitation Source for Sensors of the AMB System in HTGR
Authors: Liu Xiaohan , Mo Ni
Published: November 2014

The High Temperature Gas-Cooled Reactor (HTGR) is developed widespread in the world nowadays. The helium turbine of the 10 MW high temperature gas-cooled test module reactor with the core made of spherical fuel elements is under research at the Institute of Nuclear and New Energy Technology of Tsinghua University in China. Magnetic bearings are widely applied in HTGR, where the rotating machineries are running under highly pure helium environment. Compared with conventional bearings, Active Magnetic Bearings (AMBs) possess several attractive advantages, such as contact-free, no-lubricating and active damping vibration. The sensors are the key components of the AMBs system and are responsible for measuring the position and movement of the rotor, the operating temperature and so on. In order to operate the sensors successfully, a stable and efficient excitation source is needed. The conventional excitation source, which is equivalent to the class AB power amplifier circuit, is inefficient and unstable in the previous experiment, and therefore a new source using class D power amplifier is under research and development. The paper will present the design, simulation and test of the new excitation source for the sensors of the AMBs system.
Characterization of Incandescent Particles Emitted by a Cut-Off Grinder During Decommissioning Operations for Evaluating Filters Degradation
Authors: Marchal Pierre , Porcheron Emmanuel , Grehan Gérard , Lafanechere Luc , Walter Jonathan , Gelain Thomas
Published: November 2014

The use of mechanical or thermal cutting tools in decommissioning of nuclear facilities generates a lot of incandescent particles. Those particles may represent a deterioration risk of the containment barriers associated with a potential risk of fire starting.

The aim of this study is to characterize the incandescent particles emitted by a wheel grinder (in terms of temperature, diameter and velocity) and to follow those parameters all along their path from their emission point to their impact on the air filter. The characteristics of particles correlated with a possible loss of filter efficiency should highlight the destructing particles for the filter. All the measurement techniques used to characterize experimentally the incandescent particles are presented in this article. Particles are characterized in terms of diameter by microscope visualizations. The particle velocity is measured with a high speed camera using Particle Tracking Velocimetry (PTV) technique. An adaptation of a commercial monochromatic pyrometer is achieved to measure the in-flight particles temperature in our specific configuration. All of these techniques have been implemented on an experimental facility which reproduces representative conditions of the cutting processes realized during dismantling operations. Concerning the investigation of the filter, a global and a local approaches about filter degradation are used. The decontamination factor of High Efficiency Particle Air (HEPA) filter is measured, and detailed visualizations of the filter fibers deteriorations are obtained using Scanning Electrons Microscope (SEM).
Estimation of Loss Coefficient for T-Junction by an Entropy Production Approach
Authors: Ji Y. , Zhang H.-C. , Zhang Yi.-N. , Li Y. , Yan L.-M.
Published: November 2014

A T-junction structure is common in nuclear reactor systems, and plays an essential role in safety operation. An accurate estimation of the resistance characteristics of a T-junction is important for both energy conservation and engineering design. In the current study, the concept of the head loss coefficient K for a local structure was proposed, and its thermodynamic definition is extended to laminar flow within a T-junction by an entropy production approach, which was initially for turbulent flows. Specific numbers of K was determined by integration of the entropy generation rate obtained from a numerical simulation. In addition, the correlation between K and Re of combining and dividing flows was also derived.
News Media Framing of Nuclear Power in China From 2004 to 2013
Authors: Wang Yongxiang , Li Jingping , Li Nan , Chen Ronghua
Published: November 2014

China has vigorously developed nuclear power in the past several years whereas the nuclear power expansion has met with considerable objections, principally due to low widespread public support. News media frames can influence public perception. We conducted content analysis of articles related to nuclear power in the People’s Daily from 2004 to 2013. Analysis results showed that articles in the newspaper mainly reflected pro-nuclear or balanced arguments whereas little anti-nuclear opinion was published. Pro-nuclear arguments most often presented environmental benefits and stakeholder support, and informational text primarily reflected nuclear power information, nuclear technology, and regulatory processes. The results of content analysis of the newspaper are highly consistent with the Chinese government’s pro-nuclear decision.
Experimental Study on Burning Process and Characteristic of Sodium Columnar Fire
Authors: Tang Ye-xin , Zhang Zhi-gang , Guo Ming , Sun Shu-bin
Published: November 2014

A variety of sodium fire generated by the leakage of liquid sodium in the FBR is common. This paper focuses on the burning process and characteristics of sodium fire in a columnar flow. About 290°C liquid sodium was injected into a 7.9 m3 stainless steel cylindrical combustion space to shape the sodium columnar fire by 0.2 MPa high pressure nitrogen. The data of temperature field for the study of burning characteristic of sodium columnar fire have been collected by the temperature acquisition system located in the combustion space. The sodium flow maintains the columnar shape at first, and disperses by hydrodynamic effects on its way down. About 64s after the initiating time of sodium ejection for this experiment, the maximum temperature of the area close to the ejection center reaches over 1200°C. And the maximum temperature appears at the space of 1–1.5m from the plate. But the high temperature lasts for a short time and reduces rapidly. The radial temperature of the area far from the sodium flow is relatively low and generally about 200°C, and maximally about 350 °C. This study is helpful to evaluate the combustion characteristics and burning process of the sodium fire in the sodium-related facilities.
Two-Phase CFD of Channel Boiling for Boiling Transition Problems
Authors: Povolny Antonin , Cuhra Martin
Published: November 2014

In order to ensure safety of nuclear installations, thermohydraulics has developed many ways how to predict the behavior of coolant in a heated boiling channel. Accuracy of these predictions can be improved using three-dimensional Computational Fluid Dynamics (CFD) method, which is based on first principles of fluid mechanics. Even though when using CFD, there is a struggle between the accuracy and low computation costs, in many cases CFD can provide feasible improvement of accuracy compared to more traditional approaches. In this research, the focus is set on channel boiling problems, especially those associated with boiling transitions. The phenomenon of critical heat flux (CHF) is investigated using two-phase CFD computation and is compared to experimental data. There is also comparison with other computation methods. When experiment provides some set of data, CFD calculation provides description of the whole flow behavior that provides significantly more information and is of great value during the design process when it gives the understanding of undergoing effects. Besides CHF, general ability of CFD to predict changes in boiling patterns in two-phase channel boiling flows is discussed.
Modeling and Application of TRACE in the Kuosheng BWR/6 NPP Dry-Storage System
Authors: Chiang Yu , Lin Hao-Tzu , Wang Jong-Rong , Chen Shao-Wen , Tseng Yung-Shin , Shih Chunkuan
Published: November 2014

The dry-storage systems of nuclear power plants (NPPs) in Taiwan have become one of the major safety concerns. There are two steps considered in this study. The first step is the verification of the TRACE by using VSC-17 experimental data. By using VSC-17 data, the VSC-17 dry-storage system was established by TRACE. Then, the analysis results of TRACE were compared with the VSC-17 data. The results of TRACE were similar to the VSC-17 data. It indicates that TRACE has the respectable accuracy in the simulation and analysis of the dry-storage systems. The next step is the application of TRACE in the dry-storage system of Kuosheng nuclear power plant (BWR/6). Kuosheng NPP is the second BWR NPP of Taiwan Power Company. In order to solve the storage of the spent fuels, Taiwan Power Company developed the new dry-storage system for Kuosheng NPP. In this step, the dry-storage system model of Kuosheng NPP was established by TRACE. Then, the steady state simulation of this model was performed and the results of TRACE were compared with the Kuosheng NPP data. The results of TRACE in the steady state calculation were consistent with the Kuosheng NPP data. Finally, this model was used to perform the safety analysis of Kuosheng NPP dry-storage system.
Integrated Functional Modeling Method for Configuring NPP Plant DiD Risk Monitor and its Application for AP1000
Authors: Yoshikawa Hidekazu , Yang Ming , Lind Morten , Matsuoka Takeshi
Published: November 2014

A new risk monitor system has been under developing, in order not only to prevent severe accident in daily operation but also even to serve as to mitigate the radiological hazard just after severe accident happens and long term management of post-severe accident consequences. In the proposed risk monitor, the range of risk is not limited to core melt accidents but includes all kinds of negative outcome events, i.e., not only precursor troubles and incident but also any types of hazard states resulting from a severe accident. The whole system of the proposed risk monitor system is configured the by plant Defense-in Depth (DiD) risk monitor and reliability monitor, and in this paper the result of how to configure the Plant Defense in-Depth Risk Monitor by functional modeling approach is first presented. Then, a preliminary study is conducted on applying the integrated functional modeling for Plant Defense-in-depth risk monitor for passive safety system of AP1000.
Development of Inherently Safe Technologies for Large Scale BWRs: (2) Passive Water-Cooling System
Authors: Ishida Naoyuki , Watahiki Naohisa , Fujimoto Kiyoshi , Hosoi Hideaki , Kitou Kazuaki
Published: November 2014

Working from the lessons of the Fukushima Daiichi nuclear accident, we have been developing the following various safe technologies for boiling water reactors (BWRs), a passive water-cooling system, an infinite-time air-cooling system, a hydrogen explosion prevention system, and an operation support system for reactor accidents. The objective of the study reported here was development of the passive water-cooling system. The above technologies are referred to as ‘Inherently Safe Technology’.

The passive water-cooling system works without electricity for the first 10 days after an event to remove a relatively large mount of decay heat from the core. The system consists of a condenser and a steam turbine-driven pump for transferring water from a suppression pool to the reactor. Steam from the reactor pressure vessel is condensed in the condensation tubes of the condenser, and the condensate flows out into the suppression pool in the primary containment vessel (PCV). The water temperature at the condensation tube outlet is lowered to less than the saturated temperature at the partial steam pressure of the maximum PCV design pressure to prevent the PCV failure. The condenser is located at a lower level, e.g., underground, for easier access and for supplying cooling water to a condenser pool without electricity during an event. The lower level condenser pool has an advantage that it can be seismically designed.

To evaluate our concept of the water-cooling system, heat transfer tests were conducted using full-scale U-shaped single tubes with three diameter sizes under a wide range of pressure and inlet steam velocity conditions. The heat transfer data were obtained at system pressures of 0.2 to 3.0 MPa (absolute) and inlet steam velocities of 5 to 56 m/s. The heat transfer data with this wide range of pressure and inlet velocity conditions include thermal hydraulics conditions for a passive containment cooling system (PCCS) and some of the data can be extrapolated to isolation condenser (IC) conditions. We also confirmed thermal hydraulics conditions to determine the practicality of our new concept.
Multi-Scale 2D Sensitivity Analysis of the TALL-3D Experiment
Authors: Geffray Clotaire , Macián-Juan Rafael
Published: November 2014

The TALL-3D facility built by KTH in the scope of the European THINS project aims at investigating the behavior of lead-bismuth eutectic under forced and natural circulation. The design additionally presents complex 3D flowing patterns in a pool.

GRS developed a multi-scale thermal-hydraulics computational framework based on a coupling between a CFD code and a system code. The coupled program enables the simulation of the entire primary side of the TALL-3D facility at affordable cost but still provides 3D information where complex flow phenomena need to be accounted for.

TUM performs Uncertainty and Sensitivity Analysis of the TALL-3D experiment using GRS’ computational framework. In the present paper a Sensitivity Analysis is performed. The influence of modeling uncertainty on the model output is investigated. This analysis is brought to a detailed level with the monitoring of local effects within the pool.
Heat Transfer in Gas-Liquid Flow in a Triangular Arrranged Rod Bundle
Authors: Kashinsky O. N. , Lobanov P. D. , Kurdyumov A. S. , Pribaturin N. A. , Volkov S. E.
Published: November 2014

An experimental study of gas-liquid flow in a vertical bundle of rods in a hexagonal arrangement with spacer grids was performed. The test section was a bundle consisted of 37 rods. Experiments were performed in system water – air at room temperature. New information about characteristics of two-phase flow near a spacer grid is presented. A comparison of distribution of local heat transfer from the central rod surface and distribution of local void fraction around this rod for two ways of gas phase addition was performed. A non uniform void profile around the central rod of the bundle was obtained for both cases. Just downstream the spacer grid an increase of void fraction was obtained Gas addition to the flow results in heat transfer enhancement. This is especially clear for the case of gas supply by a single nozzle. In this case high void fraction and high heat transfer coefficient were observed in one of the subchannels. The data obtained allow to estimate the transfer of gas bubbles in rod bundles and heat transfer augmentation by the addition of gas phase into the coolant flow.
A Neutronic Evaluation of the HTR-10 Using Scale, MCNPX and MCNP5 Nuclear Codes
Authors: Sousa Rômulo V. , Silva Clarysson A. M. , Fortini Ângela , Pereira Cláubia , Veloso Maria Auxiliadora F. , Costa Antonella L.
Published: November 2014

The HTR-10 (High Temperature Gas-cooled Test Reactor) is a 10 MW modular pebble bed type reactor, built by the Institute of Nuclear Energy Technology (INET), Tsinghua University, China. As an advanced reactor, it has good passive safety characteristics: capacity of retaining all fission products inside the coated particles (up to 1,600° C), passive decay heat removal, large heat capacity of the core to mitigate temperature transition, large fuel temperature margin and negative temperature reactivity coefficient sufficient to accommodate reactivity insertion and small amount of excess reactivity in the core. This reactor, which core is filled with 27,000 spherical fuel elements, e.g. TRISO coated particles, is used to test and develop fuel, verify PBR safety features, demonstrate combined electricity production and cogeneration of heat, and provide experience in PBR design, operation and construction. Using the SCALE 6.0 (Standardized Computer Analysis for Licensing Evaluation), the MCNPX 2.6.0 (Monte Carlo N-Particle eXtended) and the MCNP 5 (Monte Carlo N-Particle) nuclear codes, the HTR-10 first critical core described in the Evaluation of The Initial Critical Configuration of The HTR-10 Pebble-Bed Reactor was modeled and analyzed. A three-dimension model was simulated and the keff was obtained and compared with the reference. The result presents good agreement with experimental value. The goal is to validate the DEN/UFMG model to be applied in transmutation studies changing the fuel.
Evaluation of Fatigue Life of Pressurized Water Reactor Internals Considering Light-Water Reactor Coolant Environmental Effect for Aging Management Program
Authors: Yang Jianfeng , O’Brien Paul
Published: November 2014

Most of the current operating nuclear power plants in the United States were designed using the American Society of Mechanical Engineers (ASME) Boiler and Pressure Vessel Code, Section III, for fatigue design curves. These design curves were developed in the late 1960s and early 1970s. They were often referred to as “air curves” because they were based on tests conducted in laboratory air environments at ambient temperatures. In recent years, laboratory fatigue test data showed that the light-water reactor environment could have significant impact on the fatigue life of carbon and low-alloy steels, austenitic stainless steel, and nickel-chromium-iron (Ni-Cr-Fe) alloys. United States Nuclear Regulatory Commission, Regulatory Guide 1.207 provides a guideline for evaluating fatigue analyses incorporating the life reduction of metal components due to the effects of the light-water reactor environment for new reactors. It recommend following the method developed in NUREG/CR-6909 [3] when designing reactor coolant pressure boundary components. The industry has invested a lot of effort in developing methods and rules for applying environmental fatigue evaluations for ASME Class 1 components and piping. However, the industry experience in applying the environmental fatigue evaluation for reactor core support structures and internal structures has been very limited. During the recent aging management programs, reactor internal component environmental fatigue evaluations for several pressurized water reactors were evaluated. The analyses calculated the cumulative fatigue usage using the recorded plant-specific transient cycles and the projected cycles for 60 years of plant life. The study concludes that the actual fatigue usages of the components are substantially lower than the specified original design conditions. Even assuming the most severe light-water reactor coolant environmental effects, fatigue will not be a concern for 60 years of plant life. The experiences with environmental fatigue evaluation for reactor internals are still very limited. This study shall provide the industry with beneficial information to develop the approaches and rules addressing the environmental effect on the fatigue life of reactor internals.
Verification for Ray Effects Elimination Module of Radiation Shielding Code ARES by Kobayashi Benchmarks
Authors: Chen Mengteng , Zhang Bin , Chen Yixue
Published: November 2014

ARES is a multi-group of anisotropic scattering transport shielding code based on discrete ordinates method. The 3D radiation transport benchmark problems proposed by Kobayashi were calculated by ARES with sub-module ARES_RayEffect which using first collision method for ray effects mitigation. ARES_RayEffect calculates uncollided flux and first collision source moments for ARES. The uncollided flux is obtained by a ray tracing calculation between a source point and a target mesh center. In addition, ARES_RayEffect has a modifying factor function to improve the quality of uncollided flux calculation. For verification, the results of MCNP code are used as reference solution and the results of TORT with FNSUNCL3 are compared. ARES_RayEffect introduced the modifying factor to reduce the relative difference of meshes near the source region. For example, at the position (15,15,15) in Problem 1 case i, the relative difference of the result of ARES with ARES_RayEffect is −2.34%, while relative difference of the result of TORT with FNSUNCL3 is −11.92%. The calculated total neutron fluxes show good agreement with the MCNP solutions. For the pure absorber cases, the maximum differences are less than 3%. For the half scattering cases, the maximum differences are less than 11%. Numerical results demonstrate that ray effects can be effectively mitigated.
Can Safety Culture Be This Important?
Authors: Conger Dorian S.
Published: November 2014

Organizational culture has been extensively studied since the 1950’s. The research consistently demonstrates that an organization’s culture has a direct and immediate impact on the behaviors of the people working within the organization. For many years, the culture of an organization was not a part of the evaluation process when performance deficiencies or incidents were evaluated. In some instances, organizations were even told specifically that such ‘soft’ issues were not to be considered. Now, it seems that the pendulum has swung completely in the opposite direction. Organizations are encouraged and sometimes even required to consider safety culture contributions to performance problems and accidents/incidents. Few systematic methods exist to evaluate the contributions of safety culture to incident and accidents as part of a root cause analysis. This paper explores several questions related to the importance of safety culture and how it can be evaluated and changed for the betterment of the organization(s) involved. Some of the critical questions are: 1) How is it possible that safety culture has become so important in evaluating the performance of an organization? 2) Whether in terms of deficiencies or accident/incidents — can safety culture be reliably measured, particularly during a root cause analysis? 3) If it can be measured, how can it be changed? 4) Does organizational culture change have to take years to accomplish?
Variants of the Perspective Closed Fuel Cycle, Based on REMIX-Technology
Authors: Bobrov Evgeniy , Teplov Pavel , Alekseev Pavel , Chibinyaev Alexander , Dudnikox Anatoliy
Published: November 2014

In the traditional closed fuel cycle, based on REMIX-technology (REgenerated MIXture of U and Pu oxides) the fuel composition is produced on the basis of a uranium and plutonium mixture from spent Light Water Reactor (LWR) fuel and additional natural uranium. In this case, there is some saving in the amount of natural uranium used.

The basic features of the WWER-1000 fuel loadings with a new variant REMIX-fuel during multiple recycle in the closed nuclear fuel cycle are described in this paper. Such fuel compositions are produced on a basis of a uranium and plutonium mixture allocated at processing the spent fuel after irradiation in the WWER-1000 core, depleted uranium and fission material such as: 235U as a part of high-enriched uranium from the warheads superfluous for defense.

Also here variants are considered of the perspective closed fuel cycle in which fissile feed materials for fuel manufacture is produced in the blankets of fast breeder reactors. The fissile material is 233U or Pu. The raw material is depleted uranium from the stocks of enrichment factories, or thorium. Natural uranium is not used in this case. The minimum feed material required for the REMIX technology in a closed fuel cycle was determined through calculations of different types of fissile and raw materials, with different cycle lengths and fuel-water ratios.
Influence of Silicone Carbide on the Reactivity of Nuclear Fuels Using Cerium Dioxide as a Surrogate Material
Authors: Foral Štěpán , Salamon David , Katovský Karel , Varmuža Jan , Roleček Jakub
Published: November 2014

The most researched material in nuclear power industry is uranium dioxide however due to strict safety and sanitary restrictions this material can be researched only in specialized research institutes and universities which have sufficient technological background.

For this reason it can be suitable to find material which would show physical properties similar to UO2 but would not suffer by the strict limitations in storage and handling. In this case much more workplaces could be incorporated in the material research and the list of investigated problems could be significantly enlarged.

One of the possible substitutional materials is the cerium dioxide (CeO2) which shows similar chemical and physical properties like UO2 and in some cases shows also similar neutronic properties. The laboratory research was focused on comparing of the basic neutronic properties. For the comparison of nuclear properties the JANIS software [1] was used as it contains cross section libraries of both materials.

It will be shown that similarity of both materials is significant and in several cases application of cerium dioxide as alternative material is possible with sufficient accuracy. As an example of the use of CeO2, research of influence of the SiC content on the reactivity of nuclear fuels is presented in this work.
Supporting Qualified Database for V&V and Uncertainty Evaluation of Best-Estimate System Codes
Authors: Petruzzi A. , D’Auria F.
Published: November 2014

Uncertainty evaluation constitutes a key feature of BEPU (Best Estimate Plus Uncertainty) process. The uncertainty can be the result of a Monte Carlo type analysis involving input uncertainty parameters or the outcome of a process involving the use of experimental data and connected code calculations. Those uncertainty methods are discussed in several papers and guidelines (IAEA-SRS-52, OECD/NEA BEMUSE reports).

The present paper aims at discussing the role and the depth of the analysis required for merging from one side suitable experimental data and on the other side qualified code calculation results. This aspect is mostly connected with the second approach for uncertainty mentioned above, but it can be used also in the framework of the first approach.

Namely, the paper discusses the features and structure of the database that includes the following kinds of documents:

1. The” RDS-facility” (Reference Data Set for the selected facility): this includes the description of the facility, the geometrical characterization of any component of the facility, the instrumentations, the data acquisition system, the evaluation of pressure losses, the physical properties of the material and the characterization of pumps, valves and heat losses;

2. The “RDS-test” (Reference Data Set for the selected test of the facility): this includes the description of the main phenomena investigated during the test, the configuration of the facility for the selected test (possible new evaluation of pressure and heat losses if needed) and the specific boundary and initial conditions;

3. The “QP” (Qualification Report) of the code calculation results: this includes the description of the nodalization developed following a set of homogeneous techniques, the achievement of the steady state conditions and the qualitative and quantitative analysis of the transient with the characterization of the Relevant Thermal-Hydraulics Aspects (RTA);

4. The EH (Engineering Handbook) of the input nodalization: this includes the rationale adopted for each part of the nodalization, the user choices, and the systematic derivation and justification of any value present in the code input respect to the values as indicated in the RDS-facility and in the RDS-test.
Innovative Robot Technologies for Nuclear Power Plants Inspection and Maintenance
Authors: Kim Seungho , Jung Kyung Min , Lee Sung Uk , Shin Hocheol , Kim Chang Hoi , Seo Yong Chil , Bae Yeong Geol , Na Hyeonseok
Published: November 2014

Nuclear energy has become the main energy source in Korea, but the safety issues are being debated since the Fukusima accident. In order to guarantee the safety and reliability of nuclear power plants, the uncertainty of human errors are being minimized by utilizing innovative technologies for inspection and maintenance. KAERI has developed robotic systems to upgrade and ensure the safety of nuclear facilities, to detect unusual conditions of facilities through remote inspection and to prevent human workers from high dose radiation with efficient plant maintenance.