WorldWideScience

Sample records for supercritical water coolant

  1. Channel type reactors with supercritical water coolant. Russian experience

    International Nuclear Information System (INIS)

    Kuznetsov, Y.N.; Gabaraev, B.A.

    2003-01-01

    Transition to coolant of supercritical parameters allows for principle engineering-andeconomic characteristics of light-water nuclear power reactors to be substantially enhanced. Russian experience in development of channel-type reactors with supercritical water coolant has demonstrated advantages and practical feasibility of such reactors. (author)

  2. CANDU with supercritical water coolant: conceptual design features

    International Nuclear Information System (INIS)

    Spinks, N.

    1997-01-01

    An advanced CANDU reactor, with supercritical water as coolant, has many attractive design features. The pressure exceeds 22 MPa but coolant temperatures in excess of 370 degrees C can be reached without encountering the two-phase region with its associated fuel-dry-out and flow-instability problems. Increased coolant temperature leads to increased plant thermodynamic efficiency reducing unit energy cost through reduced specific capital cost and reduced fueling cost. Increased coolant temperature leads to reduced void reactivity via reduced coolant in-core density. Light water becomes a coolant option. To preserve neutron economy, an advanced fuel channel is needed and is described below. A supercritical-water-cooled CANDU can evolve as fuel capabilities evolve to withstand increasing coolant temperatures. (author)

  3. Predicted Variations of Water Chemistry in the Primary Coolant Circuit of a Supercritical Water Reactor

    International Nuclear Information System (INIS)

    Yeh, Tsung-Kuang; Wang, Mei-Ya; Liu, Hong-Ming; Lee, Min

    2012-09-01

    In response to the demand over a higher efficiency for a nuclear power plant, various types of Generation IV nuclear reactors have been proposed. One of the new generation reactors adopts supercritical light water as the reactor coolant. While current in-service light water reactors (LWRs) bear an average thermal efficiency of 33%, the thermal efficiency of a supercritical water reactor (SCWR) could generally reach more than 44%. For LWRs, the coolants are oxidizing due to the presence of hydrogen peroxide and oxygen, and the degradation of structural materials has mainly resulted from stress corrosion cracking. Since oxygen is completely soluble in supercritical water, similar or even worse degradation phenomena are expected to appear in the structural and core components of an SCWR. To ensure proper designs of the structural components and suitable selections of the materials to meet the requirements of operation safety, it would be of great importance for the design engineers of an SCWR to be fully aware of the state of water chemistry in the primary coolant circuit (PCC). Since SCWRs are still in the stage of conceptual design and no practical data are available, a computer model was therefore developed for analyzing water chemistry variation and corrosion behavior of metallic materials in the PCC of a conceptual SCWR. In this study, a U.S. designed SCWR with a rated thermal power of 3575 MW and a coolant flow rate of 1843 kg/s was selected for investigating the variations in redox species concentration in the PCC. Our analyses indicated that the [H 2 ] and [H 2 O 2 ] at the core channel were higher than those at the other regions in the PCC of this SCWR. Due to the self-decomposition of H 2 O 2 , the core channel exhibited a lower [O 2 ] than the upper plenum. Because the middle water rod region was in parallel with the core channel region with relatively high dose rates, the [H 2 ] and [H 2 O 2 ] in this region were higher than those in the other regions

  4. The effect of outflowing water coolant with supercritical parameters on a barrier

    Directory of Open Access Journals (Sweden)

    Alekseev Maksim

    2017-01-01

    Full Text Available The outflow of supercritical coolant with different initial parameters and its impact on the barrier have been numerically simulated. Spatial and axial distributions of pressure and steam quality are presented. The force acting on the barrier at different parameters of the outflow has been calculated.

  5. Prospects for development of an innovative water-cooled nuclear reactor for supercritical parameters of coolant

    Science.gov (United States)

    Kalyakin, S. G.; Kirillov, P. L.; Baranaev, Yu. D.; Glebov, A. P.; Bogoslovskaya, G. P.; Nikitenko, M. P.; Makhin, V. M.; Churkin, A. N.

    2014-08-01

    The state of nuclear power engineering as of February 1, 2014 and the accomplished elaborations of a supercritical-pressure water-cooled reactor are briefly reviewed, and the prospects of this new project are discussed based on this review. The new project rests on the experience gained from the development and operation of stationary water-cooled reactor plants, including VVERs, PWRs, BWRs, and RBMKs (their combined service life totals more than 15 000 reactor-years), and long-term experience gained around the world with operation of thermal power plants the turbines of which are driven by steam with supercritical and ultrasupercritical parameters. The advantages of such reactor are pointed out together with the scientific-technical problems that need to be solved during further development of such installations. The knowledge gained for the last decade makes it possible to refine the concept and to commence the work on designing an experimental small-capacity reactor.

  6. Prospects for and problems of using light-water supercritical-pressure coolant in nuclear reactors in order to increase the efficiency of the nuclear fuel cycle

    International Nuclear Information System (INIS)

    Alekseev, P. N.; Semchenkov, Yu. M.; Sedov, A. A.; Subbotin, S. A.; Chibinyaev, A. V.

    2011-01-01

    Trends in the development of the power sector of the Russian and world power industries both at present time and in the near future are analyzed. Trends in the rise of prices for reserves of fossil and nuclear fuels used for electricity production are compared. An analysis of the competitiveness of electricity production at nuclear power plants as compared to the competitiveness of electricity produced at coal-fired and natural-gas-fired thermal power plants is performed. The efficiency of the open nuclear fuel cycle and various versions of the closed nuclear fuel cycle is discussed. The requirements on light-water reactors under the scenario of dynamic development of the nuclear power industry in Russia are determined. Results of analyzing the efficiency of fuel utilization for various versions of vessel-type light-water reactors with supercritical coolant are given. Advantages and problems of reactors with supercritical-pressure water are listed.

  7. Supercritical Water Reactors

    International Nuclear Information System (INIS)

    Bouchter, J.C.; Dufour, P.; Guidez, J.; Latge, C.; Renault, C.; Rimpault, G.

    2014-01-01

    The supercritical water reactor (SCWR) is one of the 6 concepts selected for the 4. generation of nuclear reactors. SCWR is a new concept, it is an attempt to optimize boiling water reactors by using the main advantages of supercritical water: only liquid phase and a high calorific capacity. The SCWR requires very high temperatures (over 375 C degrees) and very high pressures (over 22.1 MPa) to operate which allows a high conversion yield (44% instead of 33% for a PWR). Low volumes of coolant are necessary which makes the neutron spectrum shift towards higher energies and it is then possible to consider fast reactors operating with supercritical water. The main drawbacks of supercritical water is the necessity to use very high pressures which has important constraints on the reactor design, its physical properties (density, calorific capacity) that vary strongly with temperatures and pressures and its very high corrosiveness. The feasibility of the concept is not yet assured in terms of adequate materials that resist to corrosion, reactor stability, reactor safety, and reactor behaviour in accidental situations. (A.C.)

  8. Solvation in supercritical water

    International Nuclear Information System (INIS)

    Cochran, H.D.; Cummings, P.T.; Karaborni, S.

    1991-01-01

    The aim of this work is to determine the solvation structure in supercritical water composed with that in ambient water and in simple supercritical solvents. Molecular dynamics studies have been undertaken of systems that model ionic sodium and chloride, atomic argon, and molecular methanol in supercritical aqueous solutions using the simple point charge model of Berendsen for water. Because of the strong interactions between water and ions, ionic solutes are strongly attractive in supercritical water, forming large clusters of water molecules around each ion. Methanol is found to be a weakly-attractive solute in supercritical water. The cluster of excess water molecules surrounding a dissolved ion or polar molecule in supercritical aqueous solutions is comparable to the solvent clusters surrounding attractive solutes in simple supercritical fluids. Likewise, the deficit of water molecules surrounding a dissolved argon atom in supercritical aqueous solutions is comparable to that surrounding repulsive solutes in simple supercritical fluids. The number of hydrogen bonds per water molecule in supercritical water was found to be about one third the number in ambient water. The number of hydrogen bonds per water molecule surrounding a central particle in supercritical water was only mildly affected by the identify of the central particle--atom, molecule, or ion. These results should be helpful in developing a qualitative understanding of important processes that occur in supercritical water. 29 refs., 6 figs

  9. The light water integral reactor with natural circulation of the coolant at supercritical pressure B-500 SKDI

    International Nuclear Information System (INIS)

    Silin, V.A.; Voznesensky, V.A.; Afrov, A.M.

    1993-01-01

    Pressure increase in the primary circuit over the critical value gives a possibility to construct the B-500SKDI (500 MWe) lightwater integral reactor with natural circulation of the coolant in the vessel with a diameter less than 5 m. The given reactor has a high safety level, simple operability, its specific capital cost and fuel expenditure being lower as compared to a conventional PWR. The development of the reactor is carried out taking into consideration verified technical decisions of current NPPs on the basis of Russian LWR technology. (orig.)

  10. European supercritical water cooled reactor

    International Nuclear Information System (INIS)

    Schulenberg, T.; Starflinger, J.; Marsault, P.; Bittermann, D.; Maraczy, C.; Laurien, E.; Lycklama a Nijeholt, J.A.; Anglart, H.; Andreani, M.; Ruzickova, M.; Toivonen, A.

    2011-01-01

    Highlights: → The HPLWR reactor design is an example of a supercritical water cooled reactor. → Cladding material tests have started but materials are not yet satisfactory. → Numerical heat transfer predictions are promising but need further validation. → The research project is most suited for nuclear education and training. - Abstract: The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 o C average core outlet temperature. It is designed and analyzed by a European consortium of 10 partners and 3 active supporters from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small fuel assemblies in boxes with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. Besides the scientific and technical progress, the HPLWR project turned out to be most successful in training the young generation of nuclear engineers

  11. Numerical experiment on different validation cases of water coolant flow in supercritical pressure test sections assisted by discriminated dimensional analysis part I: the dimensional analysis

    International Nuclear Information System (INIS)

    Kiss, A.; Aszodi, A.

    2011-01-01

    As recent studies prove in contrast to 'classical' dimensional analysis, whose application is widely described in heat transfer textbooks despite its poor results, the less well known and used discriminated dimensional analysis approach can provide a deeper insight into the physical problems involved and much better results in all cases where it is applied. As a first step of this ongoing research discriminated dimensional analysis has been performed on supercritical pressure water pipe flow heated through the pipe solid wall to identify the independent dimensionless groups (which play an independent role in the above mentioned thermal hydraulic phenomena) in order to serve a theoretical base to comparison between well known supercritical pressure water pipe heat transfer experiments and results of their validated CFD simulations. (author)

  12. Materials challenges for the supercritical water-cooled reactor (SCWR)

    International Nuclear Information System (INIS)

    Baindur, S.

    2008-01-01

    This paper discusses the materials requirements of the Supercritical Water-cooled Reactor (SCWR) which arise from its severe expected operating conditions: (i) Outlet Temperature (to 650 C); (ii) Pressure of 25 MPa for the coolant containment, (iii) Thermochemical stress in the presence of supercritical water, and (iv) Radiative damage (up to 150 dpa for the fast spectrum variant). These operating conditions are reviewed; the phenomenology of materials in the supercritical water environment that create the materials challenges is discussed; knowledge gaps are identified, and efforts to understand material behaviour under the operating conditions expected in the SCWR are described. (author)

  13. Advanced Thermal Storage for Central Receivers with Supercritical Coolants

    Energy Technology Data Exchange (ETDEWEB)

    Kelly, Bruce D.

    2010-06-15

    The principal objective of the study is to determine if supercritical heat transport fluids in a central receiver power plant, in combination with ceramic thermocline storage systems, offer a reduction in levelized energy cost over a baseline nitrate salt concept. The baseline concept uses a nitrate salt receiver, two-tank (hot and cold) nitrate salt thermal storage, and a subcritical Rankine cycle. A total of 6 plant designs were analyzed, as follows: Plant Designation Receiver Fluid Thermal Storage Rankine Cycle Subcritical nitrate salt Nitrate salt Two tank nitrate salt Subcritical Supercritical nitrate salt Nitrate salt Two tank nitrate salt Supercritical Low temperature H2O Supercritical H2O Two tank nitrate salt Supercritical High temperature H2O Supercritical H2O Packed bed thermocline Supercritical Low temperature CO2 Supercritical CO2 Two tank nitrate salt Supercritical High temperature CO2 Supercritical CO2 Packed bed thermocline Supercritical Several conclusions have been drawn from the results of the study, as follows: 1) The use of supercritical H2O as the heat transport fluid in a packed bed thermocline is likely not a practical approach. The specific heat of the fluid is a strong function of the temperatures at values near 400 °C, and the temperature profile in the bed during a charging cycle is markedly different than the profile during a discharging cycle. 2) The use of supercritical CO2 as the heat transport fluid in a packed bed thermocline is judged to be technically feasible. Nonetheless, the high operating pressures for the supercritical fluid require the use of pressure vessels to contain the storage inventory. The unit cost of the two-tank nitrate salt system is approximately $24/kWht, while the unit cost of the high pressure thermocline system is nominally 10 times as high. 3) For the supercritical fluids, the outer crown temperatures of the receiver tubes are in the range of 700 to 800 °C. At temperatures of 700 °C and above

  14. Numerical modeling of the waves evolution generated by the depressurization of the vessels containing a supercritical parameters coolant

    Science.gov (United States)

    Alekseev, Maksim V.; Vozhakov, Ivan S.; Lezhnin, Sergey I.; Pribaturin, Nikolay A.

    2017-10-01

    The development of power plants focuses on increasing the parameters of water coolants up to a supercritical level. Depressurization of the unit circuits with such a coolant leads to emergency situations. Their scenarios can change significantly with the variation of initial pressure and temperature before the start of depressurization. When the pressure drops from the supercritical single-phase region of the initial thermodynamic parameters of the coolant, either the liquid boils up, or the vapor is condensed. Because of the rapid pressure decrease, the phase transition can be non-equilibrium that must be taken into account in the simulation. In the present study, an axisymmetric problem of the outflow of a water coolant from the pipe butt-end is considered. The equations of continuity, momentum and energy for a two-phase homogeneous mixture are solved numerically. The vapor and liquid properties are calculated using the TTSE software package (The Tabular Taylor Series Expansion Method). On the basis of the computer complex LCPFCT (The Flux-Corrected Transport Algorithm) the program code was developed for solving numerous problems on the depressurization of vessels or pipelines, containing superheated water or gas under high pressure. Different variants of outflow in the external model atmosphere and generation of waves are analyzed. The calculated data on the interaction of pressure waves with a barrier are calculated. To describe phase transitions, an asymptotic relaxation model of nonequilibrium evaporation and condensation has been created and tested.

  15. ENGINEERING BULLETIN: SUPERCRITICAL WATER OXIDATION

    Science.gov (United States)

    This engineering bulletin presents a description and status of supercritical water oxidation technology, a summary of recent performance tests, and the current applicability of this emerging technology. This information is provided to assist remedial project managers, contractors...

  16. High temperature and high performance light water cooled reactors operating at supercritical pressure, research and development

    International Nuclear Information System (INIS)

    Oka, Y.; Koshizuka, S.; Katsumura, Y.; Yamada, K.; Shiga, S.; Moriya, K.; Yoshida, S.; Takahashi, H.

    2003-01-01

    The concept of supercritical-pressure, once-through coolant cycle nuclear power plant (SCR) was developed at the University of Tokyo. The research and development (R and D) started worldwide. This paper summarized the conceptual design and R and D in Japan. The big advantage of the SCR concept is that the temperatures of major components such as reactor pressure vessel, control rod drive mechanisms, containments, coolant pumps, main steam piping and turbines are within the temperatures of the components of LWR and supercritical fossil fired power plants (FPP) in spite of the high outlet coolant temperature. The experience of these components of LWR and supercritical fossil fired power plants will be fully utilized for SCR. The high temperature, supercritical-pressure light water reactor is the logical evolution of LWR. Boiling evolved from circular boilers, water tube boilers and once-through boilers. It is the reactor version of the once-through boiler. The development from LWR to SCR follows the history of boilers. The goal of the R and D should be the capital cost reduction that cannot be achieved by the improvement of LWR. The reactor can be used for hydrogen production either by catalysis and chemical decomposition of low quality hydrocarbons in supercritical water. The reactor is compatible with tight lattice fast core for breeders due to low outlet coolant density, small coolant flow rate and high head coolant pumps

  17. Supercritical Water Mixture (SCWM) Experiment

    Science.gov (United States)

    Hicks, Michael C.; Hegde, Uday G.

    2012-01-01

    The subject presentation, entitled, Supercritical Water Mixture (SCWM) Experiment, was presented at the International Space Station (ISS) Increment 33/34 Science Symposium. This presentation provides an overview of an international collaboration between NASA and CNES to study the behavior of a dilute aqueous solution of Na2SO4 (5% w) at near-critical conditions. The Supercritical Water Mixture (SCWM) investigation, serves as important precursor work for subsequent Supercritical Water Oxidation (SCWO) experiments. The SCWM investigation will be performed in DECLICs High Temperature Insert (HTI) for the purpose of studying critical fluid phenomena at high temperatures and pressures. The HTI includes a completely sealed and integrated test cell (i.e., Sample Cell Unit SCU) that will contain approximately 0.3 ml of the aqueous test solution. During the sequence of tests, scheduled to be performed in FY13, temperatures and pressures will be elevated to critical conditions (i.e., Tc = 374C and Pc = 22 MPa) in order to observe salt precipitation, precipitate agglomeration and precipitate transport in the presence of a temperature gradient without the influences of gravitational forces. This presentation provides an overview of the motivation for this work, a description of the DECLIC HTI hardware, the proposed test sequences, and a brief discussion of the scientific research objectives.

  18. European supercritical water cooled reactor (HPLWR Phase 2 project)

    International Nuclear Information System (INIS)

    Schulenberg, Thomas; Starflinger, Joerg; Marsault, Philippe; Bittermann, Dietmar; Maraczy, Czaba; Laurien, Eckart; Lycklama, Jan Aiso; Anglart, Henryk; Andreani, Michele; Ruzickova, Mariana; Heikinheimo, Liisa

    2010-01-01

    The High Performance Light Water Reactor (HPLWR), how the European Supercritical Water Cooled Reactor is called, is a pressure vessel type reactor operated with supercritical water at 25 MPa feedwater pressure and 500 deg C maximum core outlet temperature. It is designed and analyzed by a European consortium of 13 partners from 8 Euratom member states in the second phase of the HPLWR project. Most emphasis has been laid on a core with a thermal neutron spectrum, consisting of small, housed fuel assemblies with 40 fuel pins each and a central water box to improve the neutron moderation despite the low coolant density. Peak cladding temperatures of the fuel rods have been minimized by heating up the coolant in three steps with intermediate coolant mixing. The innovative core design with upward and downward flow through its assemblies has been studied with neutronic, thermal-hydraulic and stress analyses and has been reviewed carefully in a mid-term assessment. The containment design with its safety and residual heat removal systems is based on the latest boiling water reactor concept, but with different passive high pressure coolant injection systems to cause a forced convection through the core. The design concept of the steam cycle is indicating the envisaged efficiency increase to around 44%. Moreover, it provides the constraints to design the components of the balance of the plant. The project is accompanied by numerical studies of heat transfer of supercritical water in fuel assemblies and by material tests of candidate cladding alloys, performed by the consortium and supported by additional tests of the Joint Research Centre of the European Commission. An overview of results achieved up to now, given in this paper, is illustrating the latest scientific and technological advances. (author)

  19. Mathematical modelling of nonstationary processes in a regenerator with dissociating coolant at supercritical parameters

    International Nuclear Information System (INIS)

    Tashchilova, Eh.M.; Sharovarov, G.A.

    1985-01-01

    The mathematical model of nonstationary processes in heat exchangers with dissociating coolant at supercritical parameters is given. Its dimensionless criteria are deveped. The effect of NPP regenerator parameters on criteria variation is determined. The proceeding nonstationary processes are estimated qualitatively using the dimensionless parameters. Dynamics of the processes in heat exchangers is described by the energy, mass and moment-of-momentum equations for heating and heated medium taking into account heat accumulation in the heat-transfer wall and distribution of parameters along the length of a heat exchanger

  20. Feasibility analysis of the modified ATHLET code for supercritical water cooled systems

    Energy Technology Data Exchange (ETDEWEB)

    Zhou Chong, E-mail: ch.zhou@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai 200240 (China); Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Vincenz-Priessnitz-Str. 3, 76131 Karlsruhe (Germany); Yang Yanhua [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, 800 Dongchuan Road, Shanghai 200240 (China); Cheng Xu [Institute of Fusion and Reactor Technology, Karlsruhe Institute of Technology, Vincenz-Priessnitz-Str. 3, 76131 Karlsruhe (Germany)

    2012-09-15

    Highlights: Black-Right-Pointing-Pointer Modification of system code ATHLET for supercritical water application. Black-Right-Pointing-Pointer Development and assessment of a heat transfer package for supercritical water. Black-Right-Pointing-Pointer Validation of the modified code at supercritical pressures with the theoretical point-hydraulics model and the SASC code. Black-Right-Pointing-Pointer Application of the modified code to LOCA analysis of a supercritical water cooled in-pile fuel qualification test loop. - Abstract: Since the existing thermal-hydraulic computer codes for light water reactors are not applicable to supercritical water cooled reactors (SCWRs) owing to the limitation of physical models and numerical treatments, the development of a reliable thermal-hydraulic computer code is very important to design analysis and safety assessment of SCWRs. Based on earlier modification of ATHLET for SCWR, a general interface is implemented to the code, which serves as the platform for information exchange between ATHLET and the external independent physical modules. A heat transfer package containing five correlations for supercritical water is connected to the ATHLET code through the interface. The correlations are assessed with experimental data. To verify the modified ATHLET code, the Edwards-O'Brian blow-down test is simulated. As first validation at supercritical pressures, a simplified supercritical water cooled loop is modeled and its stability behavior is analyzed. Results are compared with that of the theoretical model and SASC code in the reference and show good agreement. To evaluate its feasibility, the modified ATHLET code is applied to a supercritical water cooled in-pile fuel qualification test loop. Loss of coolant accidents (LOCAs) due to break of coolant supply lines are calculated for the loop. Sensitivity analysis of some safety system parameters is performed to get further knowledge about their influence on the function of the

  1. Supercritical Water Oxidation Program (SCWOP)

    International Nuclear Information System (INIS)

    1994-02-01

    Purpose of SCWOP is to develop and demonstrate supercritical water oxidation as a viable technology for treating DOE hazardous and mixed wastes and to coordinate SCWO research, development, demonstration, testing, and evaluation activities. The process involves bringing together organic waste, water, and an oxidant (air, O 2 , etc.) to temperatures and pressures above water's critical point (374 C, 22.1 MPa); organic destruction is >99.99% efficient, and the resulting effluents (mostly water, CO 2 ) are relatively benign. Pilot-scale (300--500 gallons/day) SCWO units are to be constructed and demonstrated. Two phases will be conducted: hazardous waste pilot plant demonstration and mixed waste pilot demonstration. Contacts for further information and for getting involved are given

  2. Systems design of direct-cycle supercritical-water-cooled fast reactors

    International Nuclear Information System (INIS)

    Oka, Yoshiaki; Koshizuka, Seiichi; Jevremovic, Tatjana; Okano, Yashushi

    1995-01-01

    The system design of a direct-cycle supercritical-water-cooled fast reactor is presented. The supercritical water does not exhibit a change of phase. the recirculation system, steam separator, and dryer of a boiling water reactor (BWR) are unnecessary. Roughly speaking, the reactor pressure vessel and control rods are similar to those of a pressurized water reactor, the containment and emergency core cooling system are similar to a BWR, and the balance of plant is similar to a supercritical-pressure fossil-fired power plant (FPP). the electric power of the fast converter is 1,508 MW(electric). The number of coolant loops is only two because of the high coolant enthalpy. Containment volume is much reduced. The thermal efficiency is improved 24% over a BWR. The coolant void reactivity is negative by placing thin zirconium-hydride layers between seeds and blankets. The power costs would be much reduced compared with those of a light water reactor (LWR) and a liquid-metal fast breeder reactor. The concept is based on the huge amount of experience with the water coolant technology of LWRs and FPPs. The oxidation of stainless steel cladding is avoided by adopting a much lower coolant temperature than that of the FPP

  3. Canadian supercritical water reactor modeling using G4STORK

    International Nuclear Information System (INIS)

    Ford, W.; Buijs, A.

    2015-01-01

    The Canadian Supercritical Water Reactor design was simulated using G4STORK. The results showed the expected trends but the determined Keff of 1.253±0.001 with a Coolant Void Reactivity (CVR) of -25mk differed greatly from the results achieved using MCNP of Keff=1.2914 and a CVR of -14mk. This discrepancy is partly due to the different data libraries used and the mixing of different temperature libraries in MCNP, but is also likely due to a difference in the physics methodology. Work is ongoing to further clarify reasons for discrepancies and improve the efficiency of the simulation. (author)

  4. Canadian supercritical water reactor modeling using G4STORK

    Energy Technology Data Exchange (ETDEWEB)

    Ford, W.; Buijs, A. [McMaster University, Hamilton, ON (Canada)

    2015-07-01

    The Canadian Supercritical Water Reactor design was simulated using G4STORK. The results showed the expected trends but the determined Keff of 1.253±0.001 with a Coolant Void Reactivity (CVR) of -25mk differed greatly from the results achieved using MCNP of Keff=1.2914 and a CVR of -14mk. This discrepancy is partly due to the different data libraries used and the mixing of different temperature libraries in MCNP, but is also likely due to a difference in the physics methodology. Work is ongoing to further clarify reasons for discrepancies and improve the efficiency of the simulation. (author)

  5. Supercritical-pressure, once-through cycle light water cooled reactor concept

    International Nuclear Information System (INIS)

    Oka, Yoshiaki; Koshizuka, Seiichi

    2001-01-01

    The purpose of the study is to develop new reactor concepts for the innovation of light water reactors (LWR) and fast reactors. Concept of the once-through coolant cycle, supercritical-pressure light water cooled reactor was developed. Major aspects of reactor design and safety were analysed by the computer codes which were developed by ourselves. It includes core design of thermal and fast reactors, plant system, safety criteria, accident and transient analysis, LOCA, PSA, plant control, start up and stability. High enthalpy rise as supercritical boiler was achieved by evaluating the cladding temperature directly during transients. Fundamental safety principle of the reactor is monitoring coolant flow rate instead of water level of LWR. The reactor system is compact and simple because of high specific enthalpy of supercritical water and the once-through cycle. The major components are similar to those of LWR and supercritical thermal plant. Their temperature are within the experiences in spite of the high outlet coolant temperature. The reactor is compatible with tight fuel lattice fast reactor because of the high head pumps and low coolant flow rate. The power rating of the fast reactor is higher than the that of thermal reactor because of the high power density. (author)

  6. Code Development in Coupled PARCS/RELAP5 for Supercritical Water Reactor

    Directory of Open Access Journals (Sweden)

    Po Hu

    2014-01-01

    Full Text Available The new capability is added to the existing coupled code package PARCS/RELAP5, in order to analyze SCWR design under supercritical pressure with the separated water coolant and moderator channels. This expansion is carried out on both codes. In PARCS, modification is focused on extending the water property tables to supercritical pressure, modifying the variable mapping input file and related code module for processing thermal-hydraulic information from separated coolant/moderator channels, and modifying neutronics feedback module to deal with the separated coolant/moderator channels. In RELAP5, modification is focused on incorporating more accurate water properties near SCWR operation/transient pressure and temperature in the code. Confirming tests of the modifications is presented and the major analyzing results from the extended codes package are summarized.

  7. Once-through cycle, supercritical-pressure light water cooled reactor concept

    Energy Technology Data Exchange (ETDEWEB)

    Oka, Y.; Koshizuka, S. [Tokyo Univ., Tokai, Ibaraki (Japan). Nuclear Engineering Research Lab

    2001-07-01

    Concept of once-through cycle, supercritical-pressure light water cooled reactors was developed. The research covered major aspects of conceptual design such as cores of thermal and fast reactors, plant system and heat balance, safety system and criteria, accident and transient analysis, LOCA, PSA, plant control and start-up. The advantages of the reactor lie in the compactness of the plant from high specific enthalpy of supercritical water, the simplicity of the once-through cycle and the experiences of major component technologies which are based on supercritical fossil-fired power plants and LWRs. The operating temperatures of the major components are within the experience in spite of high coolant outlet temperature. The once-through cycle is compatible with the tight fuel lattice fast reactor because of high head pumps and small coolant flow rate. (author)

  8. Once-through cycle, supercritical-pressure light water cooled reactor concept

    International Nuclear Information System (INIS)

    Oka, Y.; Koshizuka, S.

    2001-01-01

    Concept of once-through cycle, supercritical-pressure light water cooled reactors was developed. The research covered major aspects of conceptual design such as cores of thermal and fast reactors, plant system and heat balance, safety system and criteria, accident and transient analysis, LOCA, PSA, plant control and start-up. The advantages of the reactor lie in the compactness of the plant from high specific enthalpy of supercritical water, the simplicity of the once-through cycle and the experiences of major component technologies which are based on supercritical fossil-fired power plants and LWRs. The operating temperatures of the major components are within the experience in spite of high coolant outlet temperature. The once-through cycle is compatible with the tight fuel lattice fast reactor because of high head pumps and small coolant flow rate. (author)

  9. Coolant mixing in pressurized water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Hoehne, T; Grunwald, G

    1998-10-01

    The behavior of PWRs during cold water or boron dilution transients is strongly influenced by the distribution of coolant temperature and boron concentration at the core inlet. This distribution is the needed input to 3-dimensional neutron kinetics to calculate the power distribution in the core. It mainly depends on how the plugs of cold or unborated water formed in a single loop are mixed in the downcomer and in the lower plenum. To simulate such mixture phenomena requires the application of 3-dimensional CFD (computational fluid dynamics) codes. The results of the simulation have to be validated against mixture experiments at scaled facilities. Therefore, in the framework of a research project funded by BMBF, the institute creates a 1:5 mixture facility representing first the geometry of a German pressurized water reactor and later the European Pressurized Water Reactor (EPR) geometry. The calculations are based on the CFD Code CFX-4. (orig.)

  10. Efficiency of water coolant for DEMO divertor

    International Nuclear Information System (INIS)

    Fetzer, Renate; Igitkhanov, Yuri; Bazylev, Boris

    2015-01-01

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  11. Efficiency of water coolant for DEMO divertor

    Energy Technology Data Exchange (ETDEWEB)

    Fetzer, Renate, E-mail: renate.fetzer@kit.edu; Igitkhanov, Yuri; Bazylev, Boris

    2015-10-15

    Up to now, water-cooled divertor concepts have been developed for limited incident fluxes without taking into account transient power loadings. In this paper we analyzed the efficiency of water as a coolant for the particular PFC tungsten monoblock shield with a cooling tube made from Cu alloy (Cu OFHC) as a laminate adjacent to W and a low activation martensitic steel (Eurofer) as inner tube contacting the coolant. Thermal analysis is carried out by using the code MEMOS, which simulates W armour damage under the repetitive ELM heat loads. We consider cooling conditions which allow one to keep relatively high material temperatures (in the range 300–600 °C) thus minimizing Eurofer embrittlement under neutron irradiation. Expected DEMO I and DEMO II heat loads including type I ELMs are found to cause melting of the W surface during unmitigated ELMs. By mitigation of the ELMs melting of W is avoided. DEMO I operation under these conditions is save for cooling at water pressure 15.5 MPa and temperature 325 °C, while for DEMO II with mitigated ELMs the critical heat flux is exceeded and safe operation is not provided.

  12. Development Project of Supercritical-water Cooled Power Reactor

    International Nuclear Information System (INIS)

    Kataoka, K.; Shiga, S.; Moriya, K.; Oka, Y.; Yoshida, S.; Takahashi, H.

    2002-01-01

    A Supercritical-water Cooled Power Reactor (SCPR) development project (Feb. 2001- Mar. 2005) is being performed by a joint team consisting of Japanese universities and nuclear venders with a national fund. The main objective of this project is to provide technical information essential to demonstration of SCPR technologies through concentrating three sub-themes: 'plant conceptual design', 'thermohydraulics', and 'material and water chemistry'. The target of the 'plant conceptual design sub-theme' is simplify the whole plant systems compared with the conventional LWRs while achieving high thermal efficiency of more than 40 % without sacrificing the level of safety. Under the 'thermohydraulics sub-theme', heat transfer characteristics of supercritical-water as a coolant of the SCPR are examined experimentally and analytically focusing on 'heat transfer deterioration'. The experiments are being performed using fron-22 for water at a fossil boiler test facility. The experimental results are being incorporated in LWR analytical tools together with an extended steam/R22 table. Under the 'material and water chemistry sub-theme', material candidates for fuel claddings and internals of the SCPR are being screened mainly through mechanical tests, corrosion tests, and simulated irradiation tests under the SCPR condition considering water chemistry. In particular, stress corrosion cracking sensitivity is being investigated as well as uniform corrosion and swelling characteristics. Influences of water chemistry on the corrosion product characteristics are also being examined to find preferable water condition as well as to develop rational water chemistry controlling methods. (authors)

  13. Upgrading of bitumen using supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Kayukawa, T. [JGC Corp., Ibaraki (Japan)

    2009-07-01

    This presentation outlined the technical and economic aspects of thermal cracking by supercritical water. Supercritical water (SCW) is a commonly used method for upgrading heavy oil to produce pipeline-transportable oil from high-viscous bitumen. The process uses water and does not require hydrogen nor catalysts. Pre-heated bitumen and water enter a vertical reactor with flows of counter current at the supercritical point of water. The upgraded synthetic crude oil (SCO) and pitch are obtained from the top of the reactor when the bitumen is thermally cracked. Bench-scale studies have shown that Canadian oil sands bitumen can be converted to 80 volume per cent of SCO and 20 volume per cent of pitch. The SCO has satisfied Canadian pipeline specifications in terms of API gravity and kinetic viscosity. The kinetic viscosity of the pitch has also satisfied boiler fuel specifications. tabs., figs.

  14. Thermo hydraulic analysis of narrow channel effect in supercritical-pressure light water reactor

    International Nuclear Information System (INIS)

    Zhou Tao; Chen Juan; Cheng Wanxu

    2012-01-01

    Highlights: ► Detailed thermal analysis with different narrow gaps between fuel rods is given. ► Special characteristics of narrow channels effect on heat transfer in supercritical pressure are shown. ► Reasonable size selection of gaps between fuel rods is proposed for SCWR. - Abstract: The size of the gap between fuel rods has important effects on flow and heat transfer in a supercritical-pressure light water reactor. Based on thermal analysis at different coolant flow rates, the reasonable value range of gap size between fuel rods is obtained, for which the maximum cladding temperature safety limits and installation technology are comprehensively considered. Firstly, for a given design flow rate of coolant, thermal hydraulic analysis of supercritical pressure light water reactor with different gap sizes is provided by changing the fuel rod pitch only. The results show that, by means of reducing the gap size between fuel rods, the heat transfer coefficients between coolant and fuel rod, as well as the heat transfer coefficient between coolant and water rod, would both increase noticeably. Furthermore, the maximum cladding temperature will significantly decrease when the moderator temperature is decreased but coolant temperature remains essentially constant. Meanwhile, the reduction in the maximum cladding temperature in the inner assemblies is much larger than that in the outer assemblies. In addition, the maximum cladding temperature could be further reduced by means of increasing coolant flow rate for each gap size. Finally, the characteristics of narrow channels effect are proposed, and the maximum allowable gap between fuel rods is obtained by making full use of the enhancing narrow channels effect on heat transfer, and concurrently considering installation. This could provide a theoretical reference for supercritical-pressure light water reactor design optimization, in which the effects of gap size and flow rate on heat transfer are both considered.

  15. Successful treatment with supercritical water oxidation

    International Nuclear Information System (INIS)

    Jensen, R.

    1994-01-01

    Supercritical Water Oxidation (SCWO) operates in a totally enclosed system. It uses water at high temperatures and high pressure to chemically change wastes. Oily substances become soluble and complex hydrocarbons are converted into water and carbon dioxide. Research and development on SCWO is described

  16. Elements of Design Consideration of Once-Through Cycle, Supercritical-Pressure Light Water Cooled Reactor

    International Nuclear Information System (INIS)

    Yoshiaki Oka; Sei-ichi Koshizuka; Yuki Ishiwatari; Akifumi Yamaji

    2002-01-01

    The paper describes elements of design consideration of supercritical-pressure, light water cooled reactors as well as the status and prospects of the research and development. It summarizes the results of the conceptual design study at the University of Tokyo from 1989. The research and development started in Japan, Europe and USA. The major advantages of the reactors are 1. Compact reactor and turbines due to high specific enthalpy of supercritical water 2.Simple plant system because of the once-through coolant cycle 3.Use of the experience of LWR and fossil-fired power plants. The temperatures of the major components such as reactor pressure vessel, coolant pipes, pumps and turbines are within the experience, in spite of the high outlet coolant temperature. 4.Similarity to LWR safety design and criteria, but no burnout phenomenon 5.Potential cost reduction due to smaller material expenditure and short construction period 6.The smallest reactor not in power rating, but in plant sizes. 7.High-thermal efficiency and low coolant flow rate because of high enthalpy rise. 8.Water cooled reactors potentially free from SCC (stress corrosion cracking) problems. 9.Compatibility of tight-fuel-lattice fast reactor core due to small coolant flow rate, potentially easy shift to fast breeder reactor without changing coolant technology. 10.Potential of producing energy products such as hydrogen and high quality hydro carbons. (authors)

  17. Stability analysis of supercritical-pressure light water-cooled reactor in constant pressure operation

    International Nuclear Information System (INIS)

    Suhwan, JI; Shirahama, H.; Koshizuka, S.; Oka, Y.

    2001-01-01

    The purpose of this study is to evaluate the thermal-hydraulic and the thermal-nuclear coupled stabilities of a supercritical pressure light water-cooled reactor. A stability analysis code at supercritical pressure is developed. Using this code, stabilities of full and partial-power reactor operating at supercritical pressure are investigated by the frequency-domain analysis. Two types of SCRs are analyzed; a supercritical light water reactor (SCLWR) and a supercritical water-cooled fast reactor (SCFR). The same stability criteria as Boiling Water Reactor are applied. The thermal-hydraulic stability of SCLWR and SCFR satisfies the criteria with a reasonable orifice loss coefficient. The decay ratio of the thermal-nuclear coupled stability in SCFR is almost zero because of a small coolant density coefficient of the fast reactor. The evaluated decay ratio of the thermal-nuclear coupled stability is 3,41 ∼ 10 -V at 100% power in SCFR and 0,028 at 100% power in SCLWR. The sensitivity is investigated. It is found that the thermal-hydraulic stability is sensitive to the mass flow rate strongly and the thermal-nuclear coupled stability to the coolant density coefficient. The bottom power peak distribution makes the thermal-nuclear stability worse and the thermal-nuclear stability better. (author)

  18. Supercritical water decontamination of town gas soil

    International Nuclear Information System (INIS)

    Kocher, B.S.; Azzam, F.O.; Lee, S.

    1994-01-01

    Town gas sites represent a large environmental problem that exists in more than 2,000 sites across North America alone. The major contaminants in town gas sods are polycyclic aromatic hydrocarbons (PAHs). These are stable compounds that migrate deep into the soil and are traditionally very difficult to remove by conventional remediation processes. Supercritical fluids offer enhanced solvating properties along with reduced mass transfer resistances that make them ideal for removing compounds that are difficult or impossible to remove by conventional processes. Supercritical water is ideal for removing PAHs and other hydrocarbons from soil due to its high solvating power towards most hydrocarbon species. Supercritical water was investigated for its ability to remediate two different town gas sods containing from 3--20 wt% contamination. The sod was remediated in a 300-cc semi-continuous system to a more environmentally acceptable level

  19. Nuclear and thermal analyses of supercritical-water-cooled solid breeder blanket for fusion DEMO reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yanagi, Yoshihiko; Sato, Satoshi; Enoeda, Mikio; Hatano, Toshihisa; Kikuchi, Shigeto; Kuroda, Toshimasa; Kosaku, Yasuo; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment

    2001-11-01

    Within a design study of a fusion DEMO reactor aiming at demonstrating technologies of fusion power plant, supercritical water is applied as a coolant of solid breeder blanket to attain high thermal efficiency. The blanket has multi-layer composed of solid breeder pebbles (Li{sub 2}O) and neutron multiplier pebbles (Be) which are radially separated by cooling panels. The first wall and the breeding region are cooled by supercritical water below and above the pseudo-critical temperature, respectively. Temperature distribution and tritium breeding ratio (TBR) have been estimated by one-dimensional nuclear and thermal calculations. The local TBR as high as 1.47 has been obtained after optimization of temperature distribution in the breeder region under the following conditions: neutron wall loading of 5 MW/m{sup 2}, {sup 6}Li enrichment of 30% and coolant temperature at inlet of breeder region of 380degC. In the case of the higher coolant temperature 430degC of the breeder region the local TBR was reduced to be 1.40. This means that the net TBR higher than 1.0 could be expected with the supercritical-water-cooled blanket, whose temperature distribution in the breeder region would be optimized by following the coolant temperature, and where a coverage of the breeder region is assumed to be 70%. (author)

  20. Pulse radiolysis study of supercritical water-G-value measurement up to 450 degree C

    International Nuclear Information System (INIS)

    Katsumura, Y.

    2006-01-01

    It is widely recognized that the understanding of water radiolysis at elevated temperatures is inevitably important in the field of water chemistry in light water reactors because water radiolysis is closely related to many subjects such as hydrogen water chemistry (H 2 injection), SCC (stress corrosion cracking), dose accumulation and so on. This situation would also be applied to the future reactor using supercritical water (>374 C, 22.1MPa) as a coolant, so called supercritical water-cooled reactor (SCWR). Therefore, it is important to investigate water radiolysis of supercritical water. In 1989 Prof. Oka, University of Tokyo, proposed the SCWR as a future reactor and done much design study. This reactor has many advantages such as high energy efficiency, applicability of experience accumulated in light water reactors and supercritical fissile plant, and compact structure. In 2002 the Department of Energy in USA has selected the SCWR as one of the six Generation IV reactors and fundamental research has started in different countries as a national or an international project. In the present research G-values of water radiolysis have been measured by using a pulse radiolysis method up to 450 degree C to obtain the fundamental data relevant to the development of the SCWR. In supercritical water, the pressure controls the density of water easily and it was found that the G-values are strongly dependent not only on temperature but also on density in supercritical water. After presentation of experimental method and its difficulties, temperature and density dependent G-values of water decomposition products in supercritical water would be summarized. (authors)

  1. MIF-SCD computer code for thermal hydraulic calculation of supercritical water cooled reactor core

    International Nuclear Information System (INIS)

    Galina P Bogoslovskaia; Alexander A Karpenko; Pavel L Kirillov; Alexander P Sorokin

    2005-01-01

    Full text of publication follows: Supercritical pressure power plants constitute the basis of heat power engineering in many countries to day. Starting from a long-standing experience of their operation, it is proposed to develop a new type of fast breeder reactor cooled by supercritical water, which enables the economical indices of NPP to be substantially improved. In the Thermophysical Department of SSC RF-IPPE, an attempt is made to provide thermal-hydraulic validation of the reactor under discussion. The paper presents the results of analysis of the thermal-hydraulic characteristics of fuel subassemblies cooled by supercritical water based on subchannel analysis. Modification of subchannel code MIF - MIF-SCD Code - developed in the SSC RF IPPE is designed as block code and permits one to calculate the coolant temperature and velocity distributions in fuel subassembly channels, the temperature of fuel pin claddings and fuel subassembly wrapper under conditions of irregular geometry and non-uniform axial and radial power generation. The thermal hydraulics under supercritical pressure of water exhibits such peculiarities as abrupt variation of the thermal physical properties in the range of pseudo-critical temperature, the absence of such phenomenon as the critical heat flux which can lead to fuel element burnout in WWERs. As compared with subchannel code for light water, in order to take account of the variation of the coolant properties versus temperature in more detail, a block for evaluating the thermal physical properties of supercritical water versus the local coolant temperature in the fuel subassembly channels was added. The peculiarities of the geometry and power generation in the fuel subassembly of the supercritical reactor are considered as well in special blocks. The results of calculations have shown that considerable preheating of supercritical coolant (several hundreds degrees) can occur in the fuel subassembly. The test calculations according to

  2. Thermal-Hydraulic Analysis of a Supercritical Water Reactor (SCWR) Core

    International Nuclear Information System (INIS)

    Kucukboyaci, V.N.; Oriani, L.

    2004-01-01

    The supercritical water reactor (SCWR) has been the object of interest throughout the nuclear Generation IV community because of its high potential: a simple, direct cycle, compact configuration; elimination of many traditional LWR components, operation at coolant temperatures much higher than traditional LWRs and thus high thermal efficiency. It could be said that the SWR was viewed as the water counterpart to the high temperature gas reactor

  3. Destruction of Energetic Materials in Supercritical Water

    Science.gov (United States)

    2002-06-25

    controls and difficulties associated with controlling processes and obtaining permits can negate potential advantages . Supercritical water oxidation...for H2 and an Alltech CTR-1 column with a temperature ramp program from -10 °C to 180 °C was used for the other gases. A mass spectrometer (HP 5971

  4. Updated heat transfer correlations for supercritical water-cooled reactor applications

    International Nuclear Information System (INIS)

    Mokry, S.J.; Pioro, I.L.; Farah, A.; King, K.

    2011-01-01

    In support of the development of SuperCritical Water-cooled Reactors (SCWRs), research is currently being conducted for heat-transfer at supercritical conditions. Currently, there are no experimental datasets for heat transfer from power reactor fuel bundles to the fuel coolant (Water) available in open literature. Therefore, for preliminary calculations, heat-transfer correlations obtained with bare tube data can be used as a conservative approach. A large set of experimental data, for supercritical water was analyzed and an updated heat-transfer correlation for forced-convective heat-transfer, in the normal heat transfer regime, was developed. This experimental dataset was obtained within conditions similar to those for proposed SCWR concepts. Thus, this new correlation can be used for preliminary heat-transfer calculations in SCWR fuel channels. It has demonstrated a good fit for the analyzed dataset. Experiments with SuperCritical Water (SCW) are very expensive. Therefore, a number of experiments are performed in modeling fluids, such as carbon dioxide and refrigerants. However, there is no common opinion if SC modeling fluids' correlations can be applied to SCW and vice versa. Therefore, a correlation for supercritical carbon dioxide heat transfer was developed as a less expensive alternative to using supercritical water. The conducted analysis also meets the objective of improving our fundamental knowledge of the transport processes and handling of supercritical fluids. These correlations can be used for supercritical water heat exchangers linked to indirect-cycle concepts and the cogeneration of hydrogen, for future comparisons with other independent datasets, with bundle data, for the verification of computer codes for SCWR core thermalhydraulics and for the verification of scaling parameters between water and modeling fluids. (author)

  5. Oxidation behavior of steels and Alloy 800 in supercritical water

    International Nuclear Information System (INIS)

    Olmedo, A.M.; Bordoni, R.; Dominguez, G.; Alvarez, M.G.

    2011-01-01

    The oxidation behavior of a ferritic-martensitic steel T91 and a martensitic steel AISI 403 up to 750 h, and of AISI 316L and Alloy 800 up to 336 h in deaerated supercritical water, 450ºC-25 MPa, was investigated in this paper. After exposure up to 750 h, the weight gain data, for steels T91 and AISI 403, was fitted by ∆W=k t n , were n are similar for both steels and k is a little higher for T91. The oxide films grown in the steels were characterized using gravimetry, scanning electron microscopy/energy dispersive X-ray spectroscopy (SEM/EDS) and X-ray diffraction. The films were adherent and exhibited a low porosity. For this low oxygen content supercritical water exposure, the oxide scale exhibited a typical duplex structure, in which the scale is composed of an outer iron oxide layer of magnetite (Fe 3 O 4 ) and an inner iron/chromium oxide layer of a non-stoichiometric iron chromite (Fe,Cr) 3 O 4 . Preliminary results, with AISI 316L and Alloy 800, for two exposure periods (168 and 336 h), are also reported. The morphology shown for the oxide films grown on both materials up to 336 h of oxidation in supercritical water, resembles that of a duplex layer film like that shown by stainless steels and Alloy 800 oxide films grown in a in a high temperature and pressure (220-350ºC) of a primary or secondary coolant of a plant. (author) [es

  6. Supercritical-pressure light water cooled reactors

    CERN Document Server

    Oka, Yoshiaki

    2014-01-01

    This book focuses on the latest reactor concepts, single pass core and experimental findings in thermal hydraulics, materials, corrosion, and water chemistry. It highlights research on supercritical-pressure light water cooled reactors (SCWRs), one of the Generation IV reactors that are studied around the world. This book includes cladding material development and experimental findings on heat transfer, corrosion and water chemistry. The work presented here will help readers to understand the fundamental elements of reactor design and analysis methods, thermal hydraulics, materials and water

  7. Etching of glass microchips with supercritical water

    Czech Academy of Sciences Publication Activity Database

    Karásek, Pavel; Grym, Jakub; Roth, Michal; Planeta, Josef; Foret, František

    2015-01-01

    Roč. 15, č. 1 (2015), s. 311-318 ISSN 1473-0197 R&D Projects: GA ČR(CZ) GAP106/12/0522; GA ČR(CZ) GBP206/12/G014; GA MŠk(CZ) EE2.3.20.0182 Institutional support: RVO:68081715 Keywords : glass microchips * channel etching * supercritical water Subject RIV: CB - Analytical Chemistry, Separation Impact factor: 5.586, year: 2015

  8. Steady state and linear stability analysis of a supercritical water natural circulation loop

    International Nuclear Information System (INIS)

    Sharma, Manish; Pilkhwal, D.S.; Vijayan, P.K.; Saha, D.; Sinha, R.K.

    2010-01-01

    Supercritical water (SCW) has excellent heat transfer characteristics as a coolant for nuclear reactors. Besides it results in high thermal efficiency of the plant. However, the flow can experience instabilities in supercritical water reactors, as the density change is very large for the supercritical fluids. A computer code SUCLIN using supercritical water properties has been developed to carry out the steady state and linear stability analysis of a SCW natural circulation loop. The conservation equations of mass, momentum and energy have been linearized by imposing small perturbation in flow rate, enthalpy, pressure and specific volume. The equations have been solved analytically to generate the characteristic equation. The roots of the equation determine the stability of the system. The code has been qualitatively assessed with published results and has been extensively used for studying the effect of diameter, height, heater inlet temperature, pressure and local loss coefficients on steady state and stability behavior of a Supercritical Water Natural Circulation Loop (SCWNCL). The present paper describes the linear stability analysis model and the results obtained in detail.

  9. Kinetics of Chemical Agents Destruction in Supercritical Water

    National Research Council Canada - National Science Library

    Tester, Jefferson

    2003-01-01

    .... An experimental study of methylphosphonic acid (MPA) oxidation has been completed that includes macroscopic modeling of the overall global rate law for MPA oxidation in supercritical water (SCW...

  10. A design study of high electric power for fast reactor cooled by supercritical light water

    International Nuclear Information System (INIS)

    Koshizuka, Seiichi

    2000-03-01

    In order to evaluate the possibility to achieve high electric power by a fast reactor with supercritical light water, the design study was carried out on a large fast reactor core with high coolant outlet temperature (SCFR-H). Since the reactor coolant circuit uses once-through direct cycle where all feedwater flows through the core to the turbine at supercritical pressure, it is possible to design much simpler and more compact reactor systems and to achieve higher thermal efficiency than those of current light water reactors. The once-through direct cycle system is employed in current fossil-fired power plants. In the present study, three types of core were designed. The first is SCFR-H with blankets cooled by ascending flow, the second is SCFR-H with blankets cooled by descending flow and the third is SCFR-H with high thermal power. Every core was designed to achieve the thermal efficiency over 43%, positive coolant density reactivity coefficient and electric power over 1600 MW. Core characteristics of SCFR-Hs were compared with those of SCLWR-H (electric power: 1212 MW), which is a thermal neutron spectrum reactor cooled and moderated by supercritical light water, with the same diameter of the reactor pressure vessel. It was shown that SCFR-H could increase the electric power about 1.7 times maximally. From the standpoint of the increase of a reactor thermal power, a fast reactor has advantages as compared with a thermal neutron reactor, because it can increase the power density by adopting tight fuel lattices and eliminating the moderator region. Thus, it was concluded that a reactor cooled by supercritical light water could further improve the cost competitiveness by using a fast neutron spectrum and achieving a higher thermal power. (author)

  11. Coolant circuit water chemistry of the Paks Nuclear Power Plant

    International Nuclear Information System (INIS)

    Tilky, Peter; Doma, Arpad

    1985-01-01

    The numerous advantages of the proper selection of water chemistry parameters including low corrosion rate of the structural materials, hence the low-level activity build-up, depositions, radiation doses were emphasized. Major characteristics of water chemistry applied to the primary coolant of pressurized water reactors including neutral, slightly basic and strong basic ones are discussed. Boric acid is widely used to control reactivity. Primary coolant water chemistry of WWER type reactors which is based on the addition of ammonia and potassium hydroxide to boric acid is compared with that of other reactors. The demineralization of the total condensate of the steam turbines became a general trend in the water chemistry of the secondary coolant circuits. (V.N.)

  12. Supercritical water oxidation treatment of textile sludge.

    Science.gov (United States)

    Zhang, Jie; Wang, Shuzhong; Li, Yanhui; Lu, Jinling; Chen, Senlin; Luo, XingQi

    2017-08-01

    In this work, we studied the supercritical water oxidation (SCWO) of the textile sludge, the hydrothermal conversion of typical textile compounds and the corrosion properties of stainless steel 316. Moreover, the influence mechanisms of NaOH during these related processes were explored. The results show that decomposition efficiency for organic matter in liquid phase of the textile sludge was improved with the increment of reaction temperature or oxidation coefficient. However, the organic substance in solid phase can be oxidized completely in supercritical water. Serious coking occurred during the high pressure water at 250-450°C for the Reactive Orange 7, while at 300 and 350°C for the polyvinyl alcohol. The addition of NaOH not only accelerated the destruction of organic contaminants in the SCWO reactor, but effectively inhibited the dehydration conversion of textile compounds during the preheating process, which was favorable for the treatment system of textile sludge. The corrosion experiment results indicate that the stainless steel 316 could be competent for the body materials of the reactor and the heat exchangers. Furthermore, there was prominent enhancement of sodium hydroxide for the corrosion resistance of 316 in subcritical water. On the contrary the effect was almost none during SCWO.

  13. Investigation in justification of innovation supercritical water-cooled reactor - WWER-SCP

    International Nuclear Information System (INIS)

    Kirillov, P.L.; Baranaev, Yu.D.; Bogoslovskaya, G.P.; Glebov, A.P.; Grabezhnaya, V.A.; Kartashov, K.V.; Klushin, A.V.; Popov, V.V.

    2014-01-01

    State-of-the-art, gathered experience and development prospects of water-cooled reactors of next generation are considered. It is pointed out that development of SCWR is more attractive from the viewpoint of the basis principle of infrastructure - NPP adaptation without excessive investments. The results of experimental and calculational study of reactor installations on supercritical parameters (SCP) of water and freon are given. Consideration is given to the data on heat transfer at SCP of coolant, optimization of thermodynamic cycle, codes for thermohydraulic calculations, processes of heat and mass transfer at SCP, mass transfer and corrosion in SCP water, fuel elements and martials [ru

  14. Thermodynamic analysis of a supercritical water reactor

    International Nuclear Information System (INIS)

    Edwards, M.

    2007-01-01

    A thermodynamic model has been developed for a hypothetical design of a Supercritical Water Reactor, with emphasis on Canadian design criteria. The model solves for cycle efficiency, mass flows and physical conditions throughout the plant based on input parameters of operating pressures and efficiencies of components. The model includes eight feedwater heaters, three feedwater pumps, a deaerator, a condenser, the core, three turbines and two reheaters. To perform the calculations, Microsoft Excel was used in conjunction with FLUIDCAL-IAPWS95 and VBA code. The calculations show that a thermal efficiency of 47.5% can be achieved with a core outlet temperature of 625 o C. (author)

  15. Design and analysis on super-critical water cooled power reactors

    International Nuclear Information System (INIS)

    Ishiwatari, Yuki

    2005-01-01

    The Super-Critical Water Cooled Power Reactors (SCPR) is cooled by 25 MPa supercritical water of 280degC at reactor inlet and greater than 500degC at reactor outlet and directly connected with turbine/generators with high energy conversion efficiency. This corresponds to the deletion of recirculation system and steam-water separation system of BWR type reactors or of pressurizer and steam generator of PWR type reactors. In addition to the design study of the university of Tokyo, technology development of the SCPR for practical use has started under the collaboration of industry and academia since 2000. Mockup single tube and bundle tests for heat transfer/fluid flow characteristics of the design have been conducted with 3D heat transfer analysis. Materials compatible with coolant conditions for fuel cans and reactor internals are also assessed. Overall evaluation of the reactor concept is under way. (T. Tanaka)

  16. Pressurized-water coolant nuclear reactor steam generator

    International Nuclear Information System (INIS)

    Mayer, H.; Schroder, H.J.

    1975-01-01

    A description is given of a pressurized-water coolant nuclear reactor steam generator having a vertical housing for the steam generating water and containing an upstanding heat exchanger to which the pressurized-water coolant passes and which is radially surrounded by a guide jacket supporting a water separator on its top. By thermosiphon action the steam generating water flows upward through and around the heat exchanger within the guide chamber to the latter's top from which it flows radially outwardly and downwardly through a down draft space formed between the outside of the jacket and the housing. The water separator discharges separated water downwardly. The housing has a feedwater inlet opening adjacent to the lower portion of the heat exchanger, providing preheating of the introduced feedwater. This preheated feedwater is conveyed by a duct upwardly to a location where it mixes with the water discharged from the water separator

  17. Oxidation of oily sludge in supercritical water

    International Nuclear Information System (INIS)

    Cui Baochen; Cui Fuyi; Jing Guolin; Xu Shengli; Huo Weijing; Liu Shuzhi

    2009-01-01

    The oxidation of oily sludge in supercritical water is performed in a batch reactor at reaction temperatures between 663 and 723 K, the reaction times between 1 and 10 min and pressure between 23 and 27 MPa. Effect of reaction parameters such as reaction time, temperature, pressure, O 2 excess and initial COD on oxidation of oily sludge is investigated. The results indicate that chemical oxygen demand (COD) removal rate of 92% can be reached in 10 min. COD removal rate increases as the reaction time, temperature and initial COD increase. Pressure and O 2 excess have no remarkable affect on reaction. By taking into account the dependence of reaction rate on COD concentration, a global power-law rate expression was regressed from experimental data. The resulting pre-exponential factor was 8.99 x 10 14 (mol L -1 ) -0.405 s -1 ; the activation energy was 213.13 ± 1.33 kJ/mol; and the reaction order for oily sludge (based on COD) is 1.405. It was concluded that supercritical water oxidation (SCWO) is a rapidly emerging oily sludge processing technology.

  18. Steam generator for a pressurized-water coolant nuclear reactor

    International Nuclear Information System (INIS)

    Schroeder, H.J.; Berger, W.

    1975-01-01

    A description is given of a steam generator which has a vertical cylindrical housing having a steam output outlet, a horizontal tube sheet closing the lower end of this housing, and an inverted U-shaped tube bundle inside of the housing and having vertical inlet and outlet legs with their ends mounted in the tube sheet. Beneath the tube sheet there are inlet and outlet manifolds for the respective ends of the tube bundle so that pressurized-water coolant from a pressurized-water coolant nuclear reactor can be circulated through the tube bundle

  19. Coolant mixing in pressurized water reactors. Proceedings

    International Nuclear Information System (INIS)

    Hoehne, T.; Grunwald, G.; Rohde, U.

    1998-10-01

    For the analysis of boron dilution transients and main steam like break scenarios the modelling of the coolant mixing inside the reactor vessel is important. The reactivity insertion due to overcooling or deboration depends strongly on the coolant temperature and boron concentration. The three-dimensional flow distribution in the downcomer and the lower plenum of PWR's was calculated with a computational fluid dynamics (CFD) code (CFX-4). Calculations were performed for the PWR's of SIEMENS KWU, Westinghouse and VVER-440 / V-230 type. The following important factors were identified: exact representation of the cold leg inlet region (bend radii etc.), extension of the downcomer below the inlet region at the PWR Konvoi, obstruction of the flow by the outlet nozzles penetrating the downcomer, etc. The k-ε turbulence model was used. Construction elements like perforated plates in the lower plenum have large influence on the velocity field. It is impossible to model all the orifices in the perforated plates. A porous region model was used to simulate perforated plates and the core. The porous medium is added with additional body forces to simulate the pressure drop through perforated plates in the VVER-440. For the PWR Konvoi the whole core was modelled with porous media parameters. The velocity fields of the PWR Konvoi calculated for the case of operation of all four main circulation pumps show a good agreement with experimental results. The CFD-calculation especially confirms the back flow areas below the inlet nozzles. The downcomer flow of the Russian VVER-440 has no recirculation areas under normal operation conditions. By CFD calculations for the downcomer and the lower plenum an analytical mixing model used in the reactor dynamic code DYN3D was verified. The measurements, the analytical model and the CFD-calculations provided very well agreeing results particularly for the inlet region. The difficulties of analytical solutions and the uncertainties of turbulence

  20. Supercritical Water Oxidation Total Organic Carbon (TOC) Analysis

    Science.gov (United States)

    The work presented here is the evaluation of the modified wet‐oxidation method described as Supercritical Water Oxidation (SCWO) for the analysis of total organic carbon (TOC) in very difficult oil/gas produced water sample matrices.

  1. Corrosion and stress corrosion cracking in supercritical water

    Science.gov (United States)

    Was, G. S.; Ampornrat, P.; Gupta, G.; Teysseyre, S.; West, E. A.; Allen, T. R.; Sridharan, K.; Tan, L.; Chen, Y.; Ren, X.; Pister, C.

    2007-09-01

    Supercritical water (SCW) has attracted increasing attention since SCW boiler power plants were implemented to increase the efficiency of fossil-based power plants. The SCW reactor (SCWR) design has been selected as one of the Generation IV reactor concepts because of its higher thermal efficiency and plant simplification as compared to current light water reactors (LWRs). Reactor operating conditions call for a core coolant temperature between 280 °C and 620 °C at a pressure of 25 MPa and maximum expected neutron damage levels to any replaceable or permanent core component of 15 dpa (thermal reactor design) and 100 dpa (fast reactor design). Irradiation-induced changes in microstructure (swelling, radiation-induced segregation (RIS), hardening, phase stability) and mechanical properties (strength, thermal and irradiation-induced creep, fatigue) are also major concerns. Throughout the core, corrosion, stress corrosion cracking, and the effect of irradiation on these degradation modes are critical issues. This paper reviews the current understanding of the response of candidate materials for SCWR systems, focusing on the corrosion and stress corrosion cracking response, and highlights the design trade-offs associated with certain alloy systems. Ferritic-martensitic steels generally have the best resistance to stress corrosion cracking, but suffer from the worst oxidation. Austenitic stainless steels and Ni-base alloys have better oxidation resistance but are more susceptible to stress corrosion cracking. The promise of grain boundary engineering and surface modification in addressing corrosion and stress corrosion cracking performance is discussed.

  2. Fuel cladding interaction with water coolant in power reactors

    International Nuclear Information System (INIS)

    1985-11-01

    Water coolant chemistry and corrosion processes are important factors in reliable operation of NPP's, as at elevated temperatures water is aggressive towards structural materials. Water regimes for commercial Pressurized Water Reactors and Boiling Water Reactors were developed and proved to be satisfactory. Nevertheless, studies of operation experience continue and an amount of new Research and Development work is being conducted for further improvements of technology and better understanding of the physicochemical nature of those processes. In this report information is presented on the IAEA programme on fuel element cladding interaction with water coolant. Some results of this survey and recommendations made by the group of consultants who participated in this work are given as well as recommendations for continuation of this study. Separate abstracts were prepared for 6 papers of this report

  3. The installation welding of pressure water reactor coolant piping

    International Nuclear Information System (INIS)

    Deng Feng

    2010-01-01

    Large pressure water reactor nuclear power plants are constructing in our country. There are three symmetry standard loops in reactor coolant system. Each loop possesses a steam generator and a primary poop, in which one of the loops is equipped with a pressurizer. These components are connected with reactor pressure vessel by installation welding of the coolant piping. The integrity of reactor coolant pressure boundary is the second barrier to protect the radioactive substance from release to outside, so the safe operation of nuclear power plant is closely related to the quality of coolant piping installation welding. The heavy tube with super low carbon content austenitic stainless steel is selected for coolant piping. This kind of material has good welding behavior, but the poor thermal conductivity, the big liner expansion coefficient and the big welding deformation will cause bigger welding stress. To reduce the welding deformation, to control the dimension precision, to reduce the residual stress and to ensure the welding quality the installation sequence should be properly designed and the welding technology should be properly controlled. (authors)

  4. Muonium in sub- and supercritical water

    International Nuclear Information System (INIS)

    Percival, P.W.; Brodovitch, J.-C.; Ghandi, K.; Addison-Jones, B.; Schuth, J.; Bartels, D.M.

    1999-01-01

    Muonium has been studied in muon-irradiated water over a wide range of conditions, from standard temperature and pressure (STP) up to 350 bar and up to 420 o C, corresponding to water densities from 1.0 down to 0.1 g cm -3 . This is the first report of muonium in supercritical water. Muonium was unambiguously identified from its spin precession frequencies in small transverse magnetic fields. The hyperfine constant was determined and found to be similar to the published values for muonium in water at STP and in vacuum. Muonium was found to be long-lived over the whole range of conditions studied. The fraction of muons which form muonium was found to vary markedly over the density range studied. Correlation of the muonium fraction with the ionic product of water suggests a common cause, such as the rate of proton transfer between molecules involved in the radiolysis of water and the formation of MuOH, which competes with muonium formation

  5. Supercritical water natural circulation flow stability experiment research

    Energy Technology Data Exchange (ETDEWEB)

    Ma, Dongliang; Zhou, Tao; Li, Bing [North China Electric Power Univ., Beijing (China). School of Nuclear Science and Engineering; North China Electric Power Univ., Beijing (China). Inst. of Nuclear Thermalhydraulic Safety and Standardization; North China Electric Power Univ., Beijing (China). Beijing Key Lab. of Passive Safety Technology for Nuclear Energy; Huang, Yanping [Nuclear Power Institute of China, Chengdu (China). Science and Technology on Reactor System Design Technology Lab.

    2017-12-15

    The Thermal hydraulic characteristics of supercritical water natural circulation plays an important role in the safety of the Generation-IV supercritical water-cooled reactors. Hence it is crucial to conduct the natural circulation heat transfer experiment of supercritical water. The heat transfer characteristics have been studied under different system pressures in the natural circulation systems. Results show that the fluctuations in the subcritical flow rate (for natural circulation) is relatively small, as compared to the supercritical flow rate. By increasing the heating power, it is observed that the amplitude (and time period) of the fluctuation tends to become larger for the natural circulation of supercritical water. This tends to show the presence of flow instability in the supercritical water. It is possible to observe the flow instability phenomenon when the system pressure is suddenly reduced from the supercritical pressure state to the subcritical state. At the test outlet section, the temperature is prone to increase suddenly, whereas the blocking effect may be observed in the inlet section of the experiment.

  6. Contaminants in light water reactor coolants

    International Nuclear Information System (INIS)

    Michael, I.; Bechtold, G.

    1975-01-01

    At a lower oxygen content of the pressurized water a reduced metal loss by about 10% was detected. The state of oxidation for incoloy resulting from surface examination was 2,3 +- 0,3 which corresponds to Fe 3 O 4 and a smaller fraction of iron hydroxide. (orig.) [de

  7. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-01-01

    The use of supercritical temperature and pressure light water as the coolant in a direct-cycle nuclear reactor offers potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to 46%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type recirculation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed. If a tight fuel rod lattice is adopted, it is possible to significantly reduce the neutron moderation and attain fast neutron energy spectrum conditions. In this project a supercritical water reactor concept with a simple, blanket-free, pancake-shaped core will be developed. This type of core can make use of either fertile or fertile-free fuel and retain the hard spectrum to effectively burn plutonium and minor actinides from LWR spent fuel while efficiently generating electricity.

  8. Supercritical CO2 Brayton power cycles for DEMO (demonstration power plant) fusion reactor based on dual coolant lithium lead blanket

    International Nuclear Information System (INIS)

    Linares, José Ignacio; Cantizano, Alexis; Moratilla, Beatriz Yolanda; Martín-Palacios, Víctor; Batet, Lluis

    2016-01-01

    This paper presents an exploratory analysis of the suitability of supercritical CO 2 Brayton power cycles as alternative energy conversion systems for a future fusion reactor based on a DCLL (dual coolant lithium-lead) blanket, as prescribed by EUROfusion. The main issue dealt is the optimization of the integration of the different thermal sources with the power cycle in order to achieve the highest electricity production. The analysis includes the assessment of the pumping consumption in the heating and cooling loops, taking into account additional considerations as control issues and integration of thermal energy storage systems. An exergy analysis has been performed in order to understand the behavior of each layout. Up to ten scenarios have been analyzed assessing different locations for thermal sources heat exchangers. Neglecting the worst four scenarios, it is observed less than 2% of variation among the other six ones. One of the best six scenarios clearly stands out over the others due to the location of the thermal sources in a unique island, being this scenario compatible with the control criteria. In this proposal 34.6% of electric efficiency (before the self-consumptions of the reactor but including pumping consumptions and generator efficiency) is achieved. - Highlights: • Supercritical CO 2 Brayton cycles have been proposed for BoP of DCLL fusion reactor. • Integration of different available thermal sources has been analyzed considering ten scenarios. • Neglecting the four worst scenarios the electricity production varies less than 2%. • Control and energy storage integration issues have been considered in the analysis. • Discarding the vacuum vessel and joining the other sources in an island is proposed.

  9. Muonium kinetics in sub- and supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Ghandi, K.; Addison-Jones, B.; Brodovitch, J.C.; Kecman, S.; McKenzie, I.; Percival, P.W

    2003-02-01

    Muonium is long-lived in pure water and has been studied over a very wide range of temperatures and pressures, from 5 deg. C to over 400 deg. C and from 1 to 400 bar. We have determined rate constants for representative reactions of muonium in aqueous solution; equivalent data on H atom kinetics is sparse and stops well short of the maximum temperature and pressure attained in our experiments. The results show remarkable deviations from the predictions of standard reaction theories. In particular, rate constants pass through a maximum with temperature well below the critical point. This seems to be a general phenomenon, since we have observed it for spin-exchange and chemical reactions that are diffusion limited at low temperatures, as well as for activated reactions. We believe that a key factor in the drop of rate constants at high temperature is the cage effect, in particular the number of collisions between a pair of reactants over the duration of their encounter. Whatever the reason, the implications are profound for both the efficiency of supercritical water oxidation reactors and for the modelling of radiation chemistry in pressurized water nuclear reactors.

  10. Muonium kinetics in sub- and supercritical water

    International Nuclear Information System (INIS)

    Ghandi, K.; Addison-Jones, B.; Brodovitch, J.C.; Kecman, S.; McKenzie, I.; Percival, P.W.

    2003-01-01

    Muonium is long-lived in pure water and has been studied over a very wide range of temperatures and pressures, from 5 deg. C to over 400 deg. C and from 1 to 400 bar. We have determined rate constants for representative reactions of muonium in aqueous solution; equivalent data on H atom kinetics is sparse and stops well short of the maximum temperature and pressure attained in our experiments. The results show remarkable deviations from the predictions of standard reaction theories. In particular, rate constants pass through a maximum with temperature well below the critical point. This seems to be a general phenomenon, since we have observed it for spin-exchange and chemical reactions that are diffusion limited at low temperatures, as well as for activated reactions. We believe that a key factor in the drop of rate constants at high temperature is the cage effect, in particular the number of collisions between a pair of reactants over the duration of their encounter. Whatever the reason, the implications are profound for both the efficiency of supercritical water oxidation reactors and for the modelling of radiation chemistry in pressurized water nuclear reactors

  11. Risk analysis for a radiolysis gas detonation in an in-pile loop with supercritical water

    International Nuclear Information System (INIS)

    Zeiger, T.; Raque, M.; Kuznetsov, M.; Redlinger, R.; Schulenberg, T.

    2012-01-01

    The SCWR (supercritical water reactor) -FQT project is a cooperation between European and Chinese partners aimed to test the fuel SCWR elements under reactor conditions. In the frame of this work the risk of radiolysis gas production in the active range of the test track was assessed. The radiolysis gas could accumulate in an emergency cooling system with stagnating coolant. The ignition of this radiolysis gas could cause pressure peaks that are able to damage the primary coolant circuit. Pressure increase and deformations in case of ignition of accumulated gas were investigated. As piping material the Ti stabilized austenitic steel 08Ch18N10T was assumed, the simulation was performed using the ANSYS code. The results show that pipes without significant wall thickness enhancement cannot withstand the radiolysis gas detonation.

  12. Destruction of polyphasic systems in supercritical water reaction media

    International Nuclear Information System (INIS)

    Leybros, A.

    2009-12-01

    Spent ion exchange resins (IER) are, hence, radioactive process wastes for which there is no satisfactory industrial treatment. Supercritical water oxidation offers a viable alternative treatment to destroy the organic structure of resins by using supercritical water properties. The reactor used in Supercritical Fluids and Membranes Laboratory is a double shell stirred reactor. Total Organic Carbon reduction rates higher than 99% were obtained thanks to POSCEA2 experimental set-up when using a co-fuel, isopropyl alcohol. Influence of operating parameters was studied. A detailed reactional mechanism for cationic and anionic resins is created. For the solubilization of the particles in supercritical water, a mechanism has been created with the identified rate determining species and implemented into Fluent software through the EDC approach. Experimental temperature profiles are well represented by EDC model. Reaction rates are hence controlled by the chemical species mixing. (author)

  13. Fundamental Aspects of Water Coolant Radiolysis

    International Nuclear Information System (INIS)

    Christensen, Hilbert

    2006-04-01

    The current state of knowledge of radiolysis in Light Water Reactors (LWR) is presented in this report. High-temperature data for rate constants and primary radiolysis yields have been collected and are shown in tables. Data from different sources have been compared and based on this recommended values have been selected. There is generally a good agreement between g-values for gamma-radiation at ambient temperature from different sources. There are larger discrepancies between results for primary yields from fast neutrons and also for g-values at reactor temperatures. Complete reaction mechanisms, including rate constants at reactor temperatures, from different sources are discussed and shown in tables. Experimentally determined activation energies are also shown, including the temperature range within which they have been determined. In normal cases rate constants at high temperature have been calculated from the rate constant at ambient temperature and the activation energy. Exceptions from this rule are shown and uncertainties have been discussed. The results of a number of radiolysis calculations, carried out for reactor temperatures, are also shown. The results of some sensitivity analyses are discussed. It has been shown that results from radiolysis calculations are rather sensitive to the rate constant ratio k(OH + H 2 )/(k(OH + H 2 O 2 ). The first reaction leads to recombination, whereas the last reaction leads to decomposition. In some cases reactions which are unimportant at ambient temperature may play a role at reactor temperatures. This may be the case for reactions with a low rate constant at ambient temperature in combination with a high activation energy

  14. Effects of Gravity on Supercritical Water Oxidation (SCWO) Processes

    Science.gov (United States)

    Hegde, Uday; Hicks, Michael

    2013-01-01

    The effects of gravity on the fluid mechanics of supercritical water jets are being studied at NASA to develop a better understanding of flow behaviors for purposes of advancing supercritical water oxidation (SCWO) technologies for applications in reduced gravity environments. These studies provide guidance for the development of future SCWO experiments in new experimental platforms that will extend the current operational range of the DECLIC (Device for the Study of Critical Liquids and Crystallization) Facility on board the International Space Station (ISS). The hydrodynamics of supercritical fluid jets is one of the basic unit processes of a SCWO reactor. These hydrodynamics are often complicated by significant changes in the thermo-physical properties that govern flow behavior (e.g., viscosity, thermal conductivity, specific heat, compressibility, etc), particularly when fluids transition from sub-critical to supercritical conditions. Experiments were conducted in a 150 ml reactor cell under constant pressure with water injections at various flow rates. Flow configurations included supercritical jets injected into either sub-critical or supercritical water. Profound gravitational influences were observed, particularly in the transition to turbulence, for the flow conditions under study. These results will be presented and the parameters of the flow that control jet behavior will be examined and discussed.

  15. Research and development of the supercritical-pressure light water cooled reactor

    International Nuclear Information System (INIS)

    Oka, Yoshiaki

    2003-01-01

    The concept of high temperature reactor cooled by light water (SCR) has been developed at the University of Tokyo since 1989. Major elements of reactor conceptual design and safety were studied. It includes fuel rod design, core design of thermal and fast reactors, plant heat balance, safety design, accident and transient analysis, LOCA, PSA, plant control, start-up and stability. The big advantage of the SCR concept is that the temperatures of major components such as reactor pressure vessel, control rod drive mechanisms, containments, coolant pumps, main steam piping and turbines are within the temperatures of the components of LWR and supercritical FPP in spite of the high outlet coolant temperature. The experience of these components of LWR and supercritical fossil Fired Power Plants (FPP) will be fully utilized for SCR. Although the concept was developed at the University of Tokyo mostly with our own funds and resources, four funding was/is provided for the research in Japan so far. Those are TEPCO studies with Japanese vendors in 1994 and 1995. JSPS (Monbusho) funding of pulse radiolysis of supercritical water to the University of Tokyo, Japanese-NERI program of METI to Toshiba team on thermal hydraulics, corrosion and plant system and Japanese-NERI program of MEXT on water chemistry to the University of Tokyo. The concept was taken as the reference of HPLWR study in Europe with funding of EU in 2000 and 2001. The concept was evaluated in the Generation 4 reactor program in USA. It was selected as only one water-cooled Generation 4 reactor. This paper describes the overview of the conceptual design at the University of Tokyo and R and D in the world

  16. Fast reactor cooled by supercritical light water

    Energy Technology Data Exchange (ETDEWEB)

    Ishiwatari, Yuki; Mukouhara, Tami; Koshizuka, Seiichi; Oka, Yoshiaki [Tokyo Univ., Nuclear Engineering Research Lab., Tokai, Ibaraki (Japan)

    2001-09-01

    This report introduces the result of a feasibility study of a fast reactor cooled by supercritical light water (SCFR) with once-through cooling system. It is characterized by (1) no need of steam separator, recirculation system, or steam generator, (2) 1/7 of core flow rate compared with BWR or PWR, (3) high temperature and high pressure permits small turbine and high efficiency exceeding 44%, (4) structure and operation of major components are already experienced by LWRs or thermal power plants. Modification such as reducing blanket fuels and increasing seed fuels are made to achieve highly economic utilization of Pu and high power (2 GWe). The following restrictions were satisfied. (1) Maximum linear heat rate 39 kW/m, (2) Maximum surface temperature of Inconel cladding 620degC, (3) Negative void reactivity coefficient, (4) Fast neutron irradiation rate at the inner surface of pressure vessel less than 2.0x10{sup 19} n/cm{sup 2}. Thus the high power density of 167 MW/m{sup 3} including blanket is thought to contributes economy. The high conversion is attained to be 0.99 Pu fission residual rate by the outer radius of fuel rod of 0.88 mm. The breeding of 1.034 by Pu fission residual rate can be achieved by using briquette (tube-in-shell) type fuel structure. (K. Tsuchihashi)

  17. Supercritical water gasification of sewage sludge

    Energy Technology Data Exchange (ETDEWEB)

    Aye, L.; Yamaguchi, D. [Melbourne Univ. International Technologies Centre, Melbourne, Victoria (Australia). Dept. of Civil and Environmental Engineering

    2006-07-01

    Supercritical water gasification (SCWG) is an attractive technology for producing fuels from biomass and waste materials. As a result of greenhouse gas emissions and issues related to local air pollutants, hydrogen production from these renewable energy resources has been gaining in popularity. Disposal of sewage sludge is another environmental problem that have led to severe regulations. Incineration has been one of the most commonly used means of sewage sludge disposal. Thermal gasification produces gaseous fuel, making it a better option over incineration. However, due to its high moisture content, this process is not feasible to make use of sewage sludge directly. In order to analyze SCWG of sewage sludge, it has been determined that equilibrium analysis is most suitable since the maximum achievable amount of hydrogen in a given reacting condition can be estimated. The equilibrium model can be divided into two types of models, namely stoichiometric and non-stoichiometric. This paper presented the results of a study that used a computer program to develop a nonstoichiometric model with the direct Gibbs free energy minimization technique. In addition, various biomass were simulated for comparisons in order to identify if sewage sludge is a potential feedstock for hydrogen production. Last, the effects of reaction pressure and temperature on product distribution were also examined. It was shown that the proposed model is capable of estimating the product distribution at equilibrium. 33 refs., 4 tabs., 6 figs.

  18. Modular Porous Plate Sublimator /MPPS/ requires only water supply for coolant

    Science.gov (United States)

    Rathbun, R. J.

    1966-01-01

    Modular porous plate sublimators, provided for each location where heat must be dissipated, conserve the battery power of a space vehicle by eliminating the coolant pump. The sublimator requires only a water supply for coolant.

  19. Development of an Accelerated Methodology to Study Degradation of Materials in Supercritical Water for Application in High Temperature Power Plants

    Science.gov (United States)

    Rodriguez, David

    The decreasing supply of fossil fuel sources, coupled with the increasing concentration of green house gases has placed enormous pressure to maximize the efficiency of power generation. Increasing the outlet temperature of these power plants will result in an increase in operating efficiency. By employing supercritical water as the coolant in thermal power plants (nuclear reactors and coal power plants), the plant efficiency can be increased to 50%, compared to traditional reactors which currently operate at 33%. The goal of this dissertation is to establish techniques to characterize the mechanical properties and corrosion behavior of materials exposed to supercritical water. Traditionally, these tests have been long term exposure tests spanning months. The specific goal of this dissertation is to develop a methodology for accelerated estimation of corrosion rates in supercritical water that can be sued as a screening tool to select materials for long term testing. In this study, traditional methods were used to understand the degradation of materials in supercritical water and establish a point of comparison to the first electrochemical studies performed in supercritical water. Materials studied included austenitic steels (stainless steel 304, stainless steel 316 and Nitronic 50) and nickel based alloys (Inconel 625 and 718). Surface chemistry of the oxide layer was characterized using scanning electron microscopy, X-ray diffraction, FT-IR, Raman and X-ray photoelectron spectroscopies. Stainless steel 304 was subjected to constant tensile load creep tests in water at a pressure of 27 MPa and at temperatures of 200 °C, 315 °C and supercritical water at 450 °C for 24 hours. It was determined that the creep rate for stainless steel 304 exposed to supercritical water would be unacceptable for use in service. It was observed that the formation of hematite was favored in subcritical temperatures, while magnetite was formed in the supercritical region. Corrosion of

  20. Mechanism study of c.f.c Fe-Ni-Cr alloy corrosion in supercritical water

    International Nuclear Information System (INIS)

    Payet, M.

    2011-01-01

    Supercritical water can be use as a high pressure coolant in order to improve the thermodynamic efficiency of power plants. For nuclear concept, lifetime is an important safety parameter for materials. Thus materials selection criteria concern high temperature yield stress, creep resistance, resistance to irradiation embrittlement and also to both uniform corrosion and stress corrosion cracking.This study aims for supplying a new insight on uniform corrosion mechanism of Fe-Ni-Cr f.c.c. alloys in deaerated supercritical water at 600 C and 25 MPa. Corrosion tests were performed on 316L and 690 alloys as sample autoclaves taking into account the effect of surface finishes. Morphologies, compositions and crystallographic structure of the oxides were determined using FEG scanning electron microscopy, glow discharge spectroscopy and X-ray diffraction. If supercritical water is expected to have a gas-like behaviour in the test conditions, the results show a significant dissolution of the alloy species. Thus the corrosion in supercritical water can be considered similar to corrosion in under-critical water assuming the higher temperature and its effect on the solid state diffusion. For alloy 690, the protective oxide layer formed on polished surface consists of a chromia film topped with an iron and nickel mixed chromite or spinel. The double oxide layer formed on 316L steel seems less protective with an outer porous layer of magnetite and an inhomogeneous Cr-rich inner layer. For each alloy, the study of the inner protective scale growth mechanisms by marker or tracer experiments reveals that diffusion in the oxide scale is governed by an anionic process. However, surface finishes impact deeply the growth mechanisms. Comparisons between the results for the steel suggest that there is a competition between the oxidation of iron and chromium in supercritical water. Sufficient available chromium is required in order to form a thin oxide layer. Highly deformed or ultra fine

  1. Core design of a high breeding fast reactor cooled by supercritical pressure light water

    Energy Technology Data Exchange (ETDEWEB)

    Someya, Takayuki, E-mail: russell@ruri.waseda.jp; Yamaji, Akifumi

    2016-01-15

    Highlights: • Core design concept of supercritical light water cooled fast breeding reactor is developed. • Compound system doubling time (CSDT) is applied for considering an appropriate target of breeding performance. • Breeding performance is improved by reducing fuel rod diameter of the seed assembly. • Core pressure loss is reduced by enlarging the coolant channel area of the seed assembly. - Abstract: A high breeding fast reactor core concept, cooled by supercritical pressure light water has been developed with fully-coupled neutronics and thermal-hydraulics core calculations, which takes into account the influence of core pressure loss to the core neutronics characteristics. Design target of the breeding performance has been determined to be compound system doubling time (CSDT) of less than 50 years, by referring to the relationship of energy consumption and economic growth rate of advanced countries such as the G7 member countries. Based on the past design study of supercritical water cooled fast breeder reactor (Super FBR) with the concept of tightly packed fuel assembly (TPFA), further improvement of breeding performance and reduction of core pressure loss are investigated by considering different fuel rod diameters and coolant channel geometries. The sensitivities of CSDT and the core pressure loss with respect to major core design parameters have been clarified. The developed Super FBR design concept achieves fissile plutonium surviving ratio (FPSR) of 1.028, compound system doubling time (CSDT) of 38 years and pressure loss of 1.02 MPa with positive density reactivity (negative void reactivity). The short CSDT indicates high breeding performance, which may enable installation of the reactors at a rate comparable to energy growth rate of developed countries such as G7 member countries.

  2. Reaction kinetics of cellulose hydrolysis in subcritical and supercritical water

    Science.gov (United States)

    Olanrewaju, Kazeem Bode

    The uncertainties in the continuous supply of fossil fuels from the crisis-ridden oil-rich region of the world is fast shifting focus on the need to utilize cellulosic biomass and develop more efficient technologies for its conversion to fuels and chemicals. One such technology is the rapid degradation of cellulose in supercritical water without the need for an enzyme or inorganic catalyst such as acid. This project focused on the study of reaction kinetics of cellulose hydrolysis in subcritical and supercritical water. Cellulose reactions at hydrothermal conditions can proceed via the homogeneous route involving dissolution and hydrolysis or the heterogeneous path of surface hydrolysis. The work is divided into three main parts. First, the detailed kinetic analysis of cellulose reactions in micro- and tubular reactors was conducted. Reaction kinetics models were applied, and kinetics parameters at both subcritical and supercritical conditions were evaluated. The second major task was the evaluation of yields of water soluble hydrolysates obtained from the hydrolysis of cellulose and starch in hydrothermal reactors. Lastly, changes in molecular weight distribution due to hydrothermolytic degradation of cellulose were investigated. These changes were also simulated based on different modes of scission, and the pattern generated from simulation was compared with the distribution pattern from experiments. For a better understanding of the reaction kinetics of cellulose in subcritical and supercritical water, a series of reactions was conducted in the microreactor. Hydrolysis of cellulose was performed at subcritical temperatures ranging from 270 to 340 °C (tau = 0.40--0.88 s). For the dissolution of cellulose, the reaction was conducted at supercritical temperatures ranging from 375 to 395 °C (tau = 0.27--0.44 s). The operating pressure for the reactions at both subcritical and supercritical conditions was 5000 psig. The results show that the rate-limiting step in

  3. Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

    International Nuclear Information System (INIS)

    2014-08-01

    The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in

  4. Heat Transfer Behaviour and Thermohydraulics Code Testing for Supercritical Water Cooled Reactors (SCWRs)

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    2014-08-15

    The supercritical water cooled reactor (SCWR) is an innovative water cooled reactor concept which uses water pressurized above its thermodynamic critical pressure as the reactor coolant. This concept offers high thermal efficiencies and a simplified reactor system, and is hence expected to help to improve economic competitiveness. Various kinds of SCWR concepts have been developed, with varying combinations of reactor type (pressure vessel or pressure tube) and core spectrum (thermal, fast or mixed). There is great interest in both developing and developed countries in the research and development (R&D) and conceptual design of SCWRs. Considering the high interest shown in a number of Member States, the IAEA established in 2008 the Coordinated Research Project (CRP) on Heat Transfer Behaviour and Thermo-hydraulics Code Testing for SCWRs. The aim was to foster international collaboration in the R&D of SCWRs in support of Member States’ efforts and under the auspices of the IAEA Nuclear Energy Department’s Technical Working Groups on Advanced Technologies for Light Water Reactors (TWG-LWR) and Heavy Water Reactors (TWG-HWR). The two key objectives of the CRP were to establish accurate databases on the thermohydraulics of supercritical pressure fluids and to test analysis methods for SCWR thermohydraulic behaviour to identify code development needs. In total, 16 institutes from nine Member States and two international organizations were involved in the CRP. The thermohydraulics phenomena investigated in the CRP included heat transfer and pressure loss characteristics of supercritical pressure fluids, development of new heat transfer prediction methods, critical flow during depressurization from supercritical conditions, flow stability and natural circulation in supercritical pressure systems. Two code testing benchmark exercises were performed for steady state heat transfer and flow stability in a heated channel. The CRP was completed with the planned outputs in

  5. Destruction of energetic materials by supercritical water oxidation

    International Nuclear Information System (INIS)

    Beulow, S.J.; Dyer, R.B.; Harradine, D.M.; Robinson, J.M.; Oldenborg, R.C.; Funk, K.A.; McInroy, R.E.; Sanchez, J.A.; Spontarelli, T.

    1993-01-01

    Supercritical water oxidation is a relatively low-temperature process that can give high destruction efficiencies for a variety of hazardous chemical wastes. Results are presented examining the destruction of high explosives and propellants in supercritical water and the use of low temperature, low pressure hydrolysis as a pretreatment process. Reactions of cyclotrimethylene trinitramine (RDX), cyclotetramethylene tetranitramine (HMX), nitroguanidine (NQ), pentaerythritol tetranitrate (PETN), and 2,4,6-trinitrotoluene (TNT) are examined in a flow reactor operated at temperatures between 400 degrees C and 650 degrees C. Explosives are introduced into the reactor at concentrations below the solubility limits. For each of the compounds, over 99.9% is destroyed in less than 30 seconds at temperatures above 600 degrees C. The reactions produce primarily N 2 , N 2 O,CO 2 , and some nitrate and nitrite ions. The distribution of reaction products depends on reactor pressure, temperature, and oxidizer concentration. Kinetics studies of the reactions of nitrate and nitrite ions with various reducing reagents in supercritical water show that they can be rapidly and completely destroyed at temperatures above 525 degrees C. The use of slurries and hydrolysis to introduce high concentrations of explosives into a supercritical water reactor is examined. For some compounds the rate of reaction depends on particle size. The hydrolysis of explosives at low temperatures (<100 degrees C) and low pressures (<1 atm) under basic conditions produces water soluble, non-explosive products which are easily destroyed by supercritical water oxidation. Large pieces of explosives (13 cm diameter) have been successfully hydrolyzed. The rate, extent, and products of the hydrolysis depend on the type and concentration of base. Results from the base hydrolysis of triple base propellant M31A1E1 and the subsequent supercritical water oxidation of the hydrolysis products are presented

  6. Water vapor as a perspective coolant for fast reactors

    International Nuclear Information System (INIS)

    Kalafati, D.D.; Petrov, S.I.

    1978-01-01

    Based on analysis of foreign projects of nuclear power plants with steam-cooled fast reactors, it is shown that low breeding ratio and large doubling time were caused by using nickel alloys, high vapor pressure and small volume heat release. The possibility is shown of obtaining doubling time in the necessary limits of T 2 =10-12 years when the above reasons for steam-cooled reactors are eliminated. Favourable combination of thermophysical and thermodynamic properties of water vapor makes it perspective coolant for power fast reactors

  7. Loss of coolant accident at boiling water reactors

    International Nuclear Information System (INIS)

    Ramirez G, R.

    1975-01-01

    A revision is made with regard to the methods of thermohydraulic analysis which are used at present in order to determine the efficiency of the safety systems against loss of coolant at boiling water reactors. The object is to establish a program of work in the INEN so that the personnel in charge of the safety of the nuclear plants in Mexico, be able to make in a near future, independent valuations of the safety systems which mitigate the consequences of the above mentioned accident. (author)

  8. Method and apparatus for suppressing water-solid overpressurization of coolant in nuclear reactor power apparatus

    International Nuclear Information System (INIS)

    Aanstad, O.J.; Sklencar, A.M.

    1983-01-01

    A reactor-coolant relief valve is opened for increase in mass influx if the rate of change of coolant pressure exceeds a setpoint during a predetermined interval, if, during this interval, the coolant temperature is less than a setpoint and if the level of the fluid in the pressurizer is above a predetermined setpoint (water-solid state). (author)

  9. A Conceptual Supercritical Water Cooled Reactor Design Using a Cruciform Solid Moderator

    Energy Technology Data Exchange (ETDEWEB)

    Joo, Hyung Kook; Bae, Kang Mok; Yoo, Jae Woon; Lee, Hyun Chul; Noh, Jae Man; Bae, Yoon Yong

    2006-01-15

    A Super Critical Water-Cooled Reactor(SCWR) concept proposed by Gen-IV has an advantage of a high thermal efficiency. However, there are some difficulties in neutronic core design for a SCWR due to lower moderator density resulting from the high operating temperature over the pseudo-critical temperature. In this report, the design concepts for the fuel assembly and the core for a SCWR were described as a feasibility study on the SCWR core design. HELIOS lattice code which will be used for group constants generation was verified for the application to the low coolant density condition of a SCWR. The TAF module for a thermal hydraulic feedback in MASTER was modified to consider high pressure and temperature of the supercritical coolant with single-phase fluid. A cruciform ZrH{sub 2} solid moderator was proposed for the SCWR fuel assembly design to compensate the lower coolant density. The axial zoning concept with three different enrichments for a fuel rod was used for the axial power shape control. Gadolinia burnable poison rods were used to reduce excess reactivity. Control rod system was grouped into 6 banks to control the excess reactivity of the core during normal operation. An orifice concept for each assembly was applied to control a coolant flow rate individually. As a result of the neutronic analysis for the equilibrium SCWR core, the maximum linear heat generation rete limit was satisfied and the maximum coolant temperature of the core outlet was {approx}590 .deg. C which is lower than 620 .deg. C of the maximum clad temperature limit.

  10. Conceptual design of solid breeder blanket system cooled by supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Enoeda, Mikio; Akiba, Masato [Japan Atomic Energy Research Inst., Naka, Ibaraki (Japan). Naka Fusion Research Establishment; Ohara, Yoshihiro [Japan Atomic Energy Research Inst., Takasaki, Gunma (Japan). Takasaki Radiation Chemistry Research Establishment] [and others

    2001-12-01

    This report is a summary of the design works, which was discussed in the design workshop held in 2000 for the demonstration (DEMO) blanket aimed to strengthen the commercial competitiveness and technical feasibility simultaneously. The discussion of the Fusion Council in 1999 updated the assessment of the mission of DEMO blanket. Updated mission of the DEMO blanket is to be the prototype of the commercially competitive power plant. The DEMO blanket must supply the feasibility and experience of the total design of the power plant and the materials. From such standing point, the conceptual design study was performed to determine the updated strategy and goal of the R and D of the DEMO blanket which applies the supercritical water cooling proposed in A-SSTR, taking into account the recent progress of the plasma research and reactor engineering technology. The DEMO blanket applies the solid breeder materials and supercritical water cooling. The product tritium is purged out by helium gas stream in the breeder region. In the breeder region, the pebble bed concept was applied to withstand instable cracking of the breeder and multiplier materials in high neutron irradiation and high temperature operation. Inlet temperature of the coolant is planned to be 280degC and final outlet temperature is 510degC to obtain high energy conversion efficiency up to 43%. Reduced activation ferritic steel, F82H and ODS ferritic steel were selected as the structural material. Lithium ceramics, Li{sub 2}TiO{sub 3} or Li{sub 2}O were selected as the breeder materials. Beryllium or its inter-metallic compound Be12Ti was selected as the neutron multiplier materials. Basic module structure was selected as the box type structure which enables the remote handling replacement of the module from in-vessel access. Dimension of the box is limited to 2 m x 2 m, or smaller, due to the dimension of the replacement port. In the supercritical water cooling, the high coolant temperature is the merit for

  11. Conceptual design of solid breeder blanket system cooled by supercritical water

    International Nuclear Information System (INIS)

    Enoeda, Mikio; Akiba, Masato; Ohara, Yoshihiro

    2001-12-01

    This report is a summary of the design works, which was discussed in the design workshop held in 2000 for the demonstration (DEMO) blanket aimed to strengthen the commercial competitiveness and technical feasibility simultaneously. The discussion of the Fusion Council in 1999 updated the assessment of the mission of DEMO blanket. Updated mission of the DEMO blanket is to be the prototype of the commercially competitive power plant. The DEMO blanket must supply the feasibility and experience of the total design of the power plant and the materials. From such standing point, the conceptual design study was performed to determine the updated strategy and goal of the R and D of the DEMO blanket which applies the supercritical water cooling proposed in A-SSTR, taking into account the recent progress of the plasma research and reactor engineering technology. The DEMO blanket applies the solid breeder materials and supercritical water cooling. The product tritium is purged out by helium gas stream in the breeder region. In the breeder region, the pebble bed concept was applied to withstand instable cracking of the breeder and multiplier materials in high neutron irradiation and high temperature operation. Inlet temperature of the coolant is planned to be 280degC and final outlet temperature is 510degC to obtain high energy conversion efficiency up to 43%. Reduced activation ferritic steel, F82H and ODS ferritic steel were selected as the structural material. Lithium ceramics, Li 2 TiO 3 or Li 2 O were selected as the breeder materials. Beryllium or its inter-metallic compound Be12Ti was selected as the neutron multiplier materials. Basic module structure was selected as the box type structure which enables the remote handling replacement of the module from in-vessel access. Dimension of the box is limited to 2 m x 2 m, or smaller, due to the dimension of the replacement port. In the supercritical water cooling, the high coolant temperature is the merit for the energy

  12. Kinetics and mechanism of methane oxidation in supercritical water

    International Nuclear Information System (INIS)

    Rofer, C.K.; Streit, G.E.

    1988-10-01

    This project, is a Hazardous Waste Remedial Actions Program (HAZWRAP) Research and Development task being carried out by the Los Alamos National Laboratory. Its objective is to achieve an understanding of the technology for use in scaling up and applying oxidation in supercritical water as a viable process for treating a variety of Department of Energy Defense Programs (DOE-DP) waste streams. This report presents experimental results for the kinetics of the oxidation of methane and methanol in supercritical water and computer modeling results for the oxidation of carbonmonoxide and methane in supercritical water. The experimental and modeling results obtained to date on these one-carbon model compounds indicate that the mechanism of oxidation in supercritical water can be represented by free-radical reactions with appropriate modifications for high pressure and the high water concentration. If these current trends are sustained, a large body of existing literature data on the kinetics of elementary reactions can be utilized to predict the behavior of other compounds and their mixtures. 7 refs., 4 figs., 3 tabs

  13. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production, 3rd Quarterly Report

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth

    2002-06-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed.

  14. Stability analysis of a heated channel cooled by supercritical water

    International Nuclear Information System (INIS)

    Magni, M. C.; Delmastro, D. F; Marcel, C. P

    2009-01-01

    A simple model to study thermal-hydraulic stability of a heated cannel under supercritical conditions is presented. Single cannel stability analysis for the SCWR (Supercritical Water Cooled Reactor) design was performed. The drastic change of fluid density in the reactor core of a SCWR may induce DWO (Density Wave Oscillations) similar to those observed in BWRs. Due to the similarities between subcritical and supercritical systems we may treat the supercritical fluid as a pseudo two-phase system. Thus, we may extend the modeling approach often used for boiling flow stability analysis to supercritical pressure operation conditions. The model developed in this work take into account three regions: a heavy fluid region, similar to an incompressible liquid; a zone where a heavy fluid and a light fluid coexist, similar to two-phase mixture; and a light fluid region which behaves like superheated steam. It was used the homogeneous equilibrium model (HEM) for the pseudo boiling zone, and the ideal gas model for the pseudo superheated steam zone. System stability maps were obtained using linear stability analysis in the frequency domain. Two possible instability mechanisms are observed: DWO and excursive Ledinegg instabilities. Also, a sensitivity analysis showed that frictions in pseudo superheated steam zone, together with acceleration effect, are the most destabilizing effects. On the other hand, frictions in pseudo liquid zone are the most important stabilizing effect. [es

  15. Supercritical water oxidation data acquisition testing. Final report, Volume I

    International Nuclear Information System (INIS)

    1996-11-01

    This report discusses the phase one testing of a data acquisition system for a supercritical water waste oxidation system. The system is designed to destroy a wide range of organic materials in mixed wastes. The design and testing of the MODAR Oxidizer is discussed. An analysis of the optimized runs is included

  16. Design and technology development of solid breeder blanket cooled by supercritical water in Japan

    Science.gov (United States)

    Enoeda, M.; Kosaku, Y.; Hatano, T.; Kuroda, T.; Miki, N.; Honma, T.; Akiba, M.; Konishi, S.; Nakamura, H.; Kawamura, Y.; Sato, S.; Furuya, K.; Asaoka, Y.; Okano, K.

    2003-12-01

    This paper presents results of conceptual design activities and associated R&D of a solid breeder blanket system for demonstration of power generation fusion reactors (DEMO blanket) cooled by supercritical water. The Fusion Council of Japan developed the long-term research and development programme of the blanket in 1999. To make the fusion DEMO reactor more attractive, a higher thermal efficiency of more than 40% was strongly recommended. To meet this requirement, the design of the DEMO fusion reactor was carried out. In conjunction with the reactor design, a new concept of a solid breeder blanket cooled by supercritical water was proposed and design and technology development of a solid breeder blanket cooled by supercritical water was performed. By thermo-mechanical analyses of the first wall, the tresca stress was evaluated to be 428 MPa, which clears the 3Sm value of F82H. By thermal and nuclear analyses of the breeder layers, it was shown that a net TBR of more than 1.05 can be achieved. By thermal analysis of the supercritical water power plant, it was shown that a thermal efficiency of more than 41% is achievable. The design work included design of the coolant flow pattern for blanket modules, module structure design, thermo-mechanical analysis and neutronics analysis of the blanket module, and analyses of the tritium inventory and permeation. Preliminary integration of the design of a solid breeder blanket cooled by supercritical water was achieved in this study. In parallel with the design activities, engineering R&D was conducted covering all necessary issues, such as development of structural materials, tritium breeding materials, and neutron multiplier materials; neutronics experiments and analyses; and development of the blanket module fabrication technology. Upon developing the fabrication technology for the first wall and box structure, a hot isostatic pressing bonded F82H first wall mock-up with embedded rectangular cooling channels was

  17. Heat transfer in vertical pipe flow at supercritical pressures of water

    International Nuclear Information System (INIS)

    Loewenberg, M.F.

    2007-05-01

    A new reactor concept with light water at supercritical conditions is investigated in the framework of the European project ''High Performance Light Water Reactor'' (HPLWR). Characteristics of this reactor are the system pressure and the coolant outlet temperature above the critical point of water. Water is regarded as a single phase fluid under these conditions with a high energy density. This high energy density should be utilized in a technical application. Therefore in comparison with up to date nuclear power plants some constructive savings are possible. For instance, steam dryers or steam separators can be avoided in contrast to boiling water reactors. A thermal efficiency of about 44% can be accomplished at a system pressure of 25MPa through a water heat-up from 280 C to 510 C. To ensure this heat-up within the core reliable predictions of the heat transfer are necessary. Water as the working fluid changes its fluid properties dramatically during the heat up in the core. As such; the density in the core varies by the factor of seven. The motivation to develop a look-up table for heat transfer predications in supercritical water is due to the significant temperature dependence of the fluid properties of water. A systematic consolidation of experimental data was performed. Together with further developments of the methods to derive a look-up table made it possible to develop a look-up table for heat transfer in supercritical water in vertical flows. A look-up table predicts the heat transfer for different boundary conditions (e.g. pressure or heat flux) with tabulated data. The tabulated wall temperatures for fully developed turbulent flows can be utilized for different geometries by applying hydraulic diameters. With the developed look-up table the difficulty of choosing one of the many published correlations can be avoided. In general, the correlations have problems with strong fluid property variations. Strong property variations combined with high heat

  18. IAEA coordinated research project on thermal-hydraulics of Supercritical Water-Cooled Reactors (SCWRs)

    International Nuclear Information System (INIS)

    Yamada, K.; Aksan, S. N.

    2012-01-01

    The Supercritical Water-Cooled Reactor (SCWR) is an innovative water-cooled reactor concept, which uses supercritical pressure water as reactor coolant. It has been attracting interest of many researchers in various countries mainly due to its benefits of high thermal efficiency and simple primary systems, resulting in low capital cost. The IAEA started in 2008 a Coordinated Research Project (CRP) on Thermal-Hydraulics of SCWRs as a forum to foster the exchange of technical information and international collaboration in research and development. This paper summarizes the activities and current status of the CRP, as well as major progress achieved to date. At present, 15 institutions closely collaborate in several tasks. Some organizations have been conducting thermal-hydraulics experiments and analysing the data, and others have been participating in code-to-test and/or code-to-code benchmark exercises. The expected outputs of the CRP are also discussed. Finally, the paper introduces several IAEA activities relating to or arising from the CRP. (authors)

  19. Safety system consideration of a supercritical-water cooled fast reactor with simplified PSA

    International Nuclear Information System (INIS)

    Lee, J.H.; Oka, Y.; Koshizuka, S.

    1999-01-01

    The probabilistic safety of the supercritical-water cooled fast reactor (SCFR) is evaluated with the simplified probabilistic safety assessment (PSA) methodology. SCFR has a once-through direct cycle where all feedwater flows through the core to the turbine at supercritical pressure. There are no recirculation loops in the once-through direct cycle system, which is the most important difference from the current light water reactor (LWR). The main objective of the present study is to assess the effect of this difference on the safety in the stage of conceptual design study. A safety system configuration similar to the advanced boiling water reactor (ABWR) is employed. At loss of flow events, no natural recirculation occurs. Thus, emergency core flow should be quickly supplied before the completion of the feedwater pump coastdown at a loss of flow accident. The motor-driven high pressure coolant injection (MD-HPCI) system cannot be used for the quick core cooling due to the delay of the emergency diesel generator (D/G) start-up. Accordingly, an MD-HPCI system in an ABWR is substituted by a turbine-driven (TD-) HPCI system for the SCFR. The calculated core damage frequency (CDF) is a little higher than that of the Japanese ABWR and a little lower than that of the Japanese BWR when Japanese data are employed for initiating event frequencies. Four alternatives to the safety system configurations are also examined as a sensitivity analysis. This shows that the balance of the safety systems designed here is adequate. Consequently, though the SCFR has a once-through coolant system, the CDF is not high due to the diversity of feedwater systems as the direct cycle characteristics

  20. Design of a supercritical water-cooled reactor. Pressure vessel and internals

    Energy Technology Data Exchange (ETDEWEB)

    Fischer, Kai

    2008-08-15

    The High Performance Light Water Reactor (HPLWR) is a light water reactor with supercritical steam conditions which has been investigated within the 5th Framework Program of the European Commission. Due to the supercritical pressure of 25 MPa, water, used as moderator and as coolant, flows as a single phase through the core and can be directly fed to the turbine. Using the technology of coal fired power plants with supercritical steam conditions, the heat-up in the core is done in several steps to achieve the targeted high steam outlet temperature of 500.C without exceeding available cladding material limits. Based on a first design of a fuel assembly cluster for a HPLWR with a single pass core, the surrounding internals and the reactor pressure vessel (RPV) are dimensioned for the first time, following the safety standards of the nuclear safety standards commission in Germany. Furthermore, this design is extended to the incorporation of core arrangements with two and three passes. The design of the internals and the RPV are verified using mechanical or, in the case of large thermal deformations, combined mechanical and thermal stress analyses. Additionally, a passive safety component for the feedwater inlet of the RPV of the HPLWR is designed. Its purpose is the reduction of the mass flow rate in case of a LOCA for a feedwater line break until further steps are executed. Starting with a simple vortex diode, several steps are executed to enhance the performance of the diode and adapt it to this application. Then, this first design is further optimized using combined 1D and 3D flow analyses. Parametric studies determine the performance and characteristic for changing mass flow rates for this backflow limiter. (orig.)

  1. Heat Transfer Phenomena in Supercritical Water Nuclear Reactors

    International Nuclear Information System (INIS)

    Mark H. Anderson; MichaelL. Corradini; Riccardo Bonazza; Jeremy R. Licht

    2007-01-01

    A supercritical water heat transfer facility has been built at the University of Wisconsin to study heat transfer in a circular and square annular flow channel. A series of integral heat transfer measurements has been carried out over a wide range of heat flux, mass velocity and bulk water temperatures at a pressure of 25 MPa. The circular annular test section geometry is a 1.07 cm diameter heater rod within a 4.29 diameter flow channel

  2. Heat Transfer Phenomena in Supercritical Water Nuclear Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Mark H. Anderson; MichaelL. Corradini; Riccardo Bonazza; Jeremy R. Licht

    2007-10-03

    A supercritical water heat transfer facility has been built at the University of Wisconsin to study heat transfer in ancircular and square annular flow channel. A series of integral heat transfer measurements has been carried out over a wide range of heat flux, mas velocity and bulk water temperatures at a pressure of 25 MPa. The circular annular test section geometry is a 1.07 cm diameter heater rod within a 4.29 diameter flow channel.

  3. Prediction and analysis of onset of turbulent convective heat transfer deterioration in supercritical water flows

    International Nuclear Information System (INIS)

    Anglart, H.; Gallaway, T.; Antal, St.P.; Podowski, M.Z.

    2007-01-01

    Supercritical water is considered as a coolant in one of the six systems defined as Generation IV reactors. Such reactor will operate at pressures higher than the thermodynamic critical point of water (374 C degrees and 22.1 MPa), allowing for a significant increase of the system thermal efficiency. During normal operation no boiling crisis will occur, thereby sudden temperature excursions will be avoided. However, since the physical properties of supercritical fluids change rapidly with temperature in the pseudo critical region, the local heat transfer coefficient may still show unusual behaviour depending upon the heat flux. It can be either enhanced or deteriorated, depending on flow conditions and heat flux. It has been shown that the complexity of the phenomena involved makes it very difficult to develop acceptable predictive capabilities solely based on phenomenological models and correlations. It has also been shown that a multidimensional approach based on CFD (computational fluid dynamics) concepts is capable of properly capturing local effects that may lead to either heat transfer deterioration or enhancement

  4. Reactor water chemistry relevant to coolant-cladding interaction

    International Nuclear Information System (INIS)

    1987-09-01

    The report is a summary of the work performed in a frame of a Coordinated Research Program organized by the IAEA and carried out from 1981 till 1986. It consists of a survey on our knowledge on coolant-cladding interaction: the basic phenomena, the relevant parameters, their control and the modelling techniques implemented for their assessment. Based upon the results of this Coordinated Research Program, the following topics are reviewed on the report: role of water chemistry in reliable operation of nuclear power plants; water chemistry specifications and their control; behaviour of fuel cladding materials; corrosion product behaviour and crud build-up in reactor circuits; modelling of corrosion product behaviour. This report should be of interest to water chemistry supervisors at the power plants, to experts in utility engineering departments, to fuel designers, to R and D institutes active in the field and to the consultants of these organizations. A separate abstract was prepared for each of the 3 papers included in the Annex of this document. Refs, figs, tabs

  5. Generic supercritical water technology; Generic technology to shite no chorinkaisui riyo gijutsu

    Energy Technology Data Exchange (ETDEWEB)

    Arai, K; Ajiri, M; Inomata, H; Smith, R; Hakuta, Y [Tohoku University, Sendai (Japan). Faculty of Engineering; Yokoyama, C [Tohoku University, Sendai (Japan). The Institute forChemical Reaction Science; Chin, L [New Energy and Industrial Technology Development Organization, Tokyo, (Japan)

    1997-02-01

    This paper describes the measurement and analysis for clarifying solution structure of supercritical water and exhibition mechanism of solvent functions. It also describes the development of new processes using supercritical water as reaction solvent. The PVT measurements were conducted in the supercritical region using pure water and NaCl aqueous solution, to confirm the reduction of molar volume of the electrolyte solution. The hydration structure was examined in the supercritical aqueous solution by the molecular dynamic simulation. As a result, presence of hydrogen bond structure, where the contribution of two branching hydrogen bond can not be ignored, was suggested under the supercritical condition. Characteristics of supercritical aqueous solutions are analyzed through in-situ Raman and scattered X-ray spectral measurements. Moreover, this paper introduces developments of some processes in the supercritical water, such as decomposition of wasted polymers, recovery of chemical materials, reforming of heavy hydrocarbons by contact hydrogenation, and synthesis of fine powders of metal oxide by reaction crystallization.

  6. Solubility of 1:1 Alkali Nitrates and Chlorides in Near-Critical and Supercritical Water : 1 Alkali Nitrates and Chlorides in Near-Critical and Supercritical Water

    NARCIS (Netherlands)

    Leusbrock, Ingo; Metz, Sybrand J.; Rexwinkel, Glenn; Versteeg, Geert F.

    2009-01-01

    To increase the available data oil systems containing supercritical water and inorganic compounds, all experimental setup was designed to investigate the solubilities of inorganic compounds Ill supercritical water, In this work, three alkali chloride salts (LiCl, NaCl, KCl) and three alkali nitrate

  7. FILM-30: A Heat Transfer Properties Code for Water Coolant

    International Nuclear Information System (INIS)

    MARSHALL, THERON D.

    2001-01-01

    A FORTRAN computer code has been written to calculate the heat transfer properties at the wetted perimeter of a coolant channel when provided the bulk water conditions. This computer code is titled FILM-30 and the code calculates its heat transfer properties by using the following correlations: (1) Sieder-Tate: forced convection, (2) Bergles-Rohsenow: onset to nucleate boiling, (3) Bergles-Rohsenow: partially developed nucleate boiling, (4) Araki: fully developed nucleate boiling, (5) Tong-75: critical heat flux (CHF), and (6) Marshall-98: transition boiling. FILM-30 produces output files that provide the heat flux and heat transfer coefficient at the wetted perimeter as a function of temperature. To validate FILM-30, the calculated heat transfer properties were used in finite element analyses to predict internal temperatures for a water-cooled copper mockup under one-sided heating from a rastered electron beam. These predicted temperatures were compared with the measured temperatures from the author's 1994 and 1998 heat transfer experiments. There was excellent agreement between the predicted and experimentally measured temperatures, which confirmed the accuracy of FILM-30 within the experimental range of the tests. FILM-30 can accurately predict the CHF and transition boiling regimes, which is an important advantage over current heat transfer codes. Consequently, FILM-30 is ideal for predicting heat transfer properties for applications that feature high heat fluxes produced by one-sided heating

  8. Candidate Materials Evaluation for Supercritical Water-Cooled Reactor

    International Nuclear Information System (INIS)

    Allen, T.R.; Was, G.S.

    2008-01-01

    Final technical report on the corrosion, stress corrosion cracking, and radiation response of candidate materials for the supercritical water-cooled reactor concept. The objective of the proposed research was to investigate degradation of materials in the supercritical water environment (SCW). First, representative alloys from the important classes of candidate materials were studied for their corrosion and stress-corrosion cracking (SCC) resistance in supercritical water. These included ferritic/martensitic (F/M) steels, austenitic stainless steels, and Ni-base alloys. Corrosion and SCC tests were conducted at various temperatures and exposure times, as well as in various water chemistries. Second, emerging plasma surface modification and grain boundary engineering technologies were applied to modify the near surface chemistry, microstructure, and stress-state of the alloys prior to corrosion testing. Third, the effect of irradiation on corrosion and SCC of alloys in the as-received and modified/engineered conditions were examined by irradiating samples using high-energy protons and then exposing them to SCW

  9. Supercritical water gasification of Victorian brown coal: Experimental characterisation

    Energy Technology Data Exchange (ETDEWEB)

    Yamaguchi, Doki; Aye, Lu [Department of Civil and Environmental Engineering, The University of Melbourne, Vic 3010 (Australia); Sanderson, P. John; Lim, Seng [CSIRO Minerals, Clayton, Vic 3168 (Australia)

    2009-05-15

    Supercritical water gasification is an innovative thermochemical conversion method for converting wet feedstocks into hydrogen-rich gaseous products. The non-catalytic gasification characteristics of Victorian brown coal were investigated in supercritical water by using a novel immersion technique with quartz batch reactors. Various operating parameters such as temperature, feed concentration and reaction time were varied to investigate their effect on the gasification behaviour. Gas yields, carbon gasification efficiency and the total gasification efficiency increased with increasing temperature and reaction time, and decreasing feed concentration. The mole fraction of hydrogen in the product gases was lowest at 600 C, and increased to over 30 % at a temperature of 800 C. Varying parameters, especially reaction time, did not improve the coal utilisation for gas production significantly and the measured data showed a large deviation from the equilibrium level. (author)

  10. Supercritical water oxidation data acquisition testing. Final report, Volume II

    International Nuclear Information System (INIS)

    1996-11-01

    Supercritical Water Oxidation (SCWO) technology holds great promise for treating mixed wastes, in an environmentally safe and efficient manner. In the spring of 1994 the US Department of Energy (DOE), Idaho Operations Office awarded Stone ampersand Webster Engineering Corporation, of Boston Massachusetts and its sub-contractor MODAR, Inc. of Natick Massachusetts a Supercritical Water Oxidation Data Acquisition Testing (SCWODAT) program. The SCWODAT program was contracted through a Cooperative Agreement that was co-funded by the US Department of Energy and the Strategic Environmental Research and Development Program. The SCWODAT testing scope outlined by the DOE in the original Cooperative Agreement and amendments thereto was initiated in June 1994 and successfully completed in December 1995. The SCWODAT program provided further information and operational data on the effectiveness of treating both simulated mixed waste and typical Navy hazardous waste using the MODAR SCWO technology

  11. Corrosion properties of modified PNC1520 austenitic stainless steel in supercritical water as a fuel cladding candidate material for supercritical water reactor

    International Nuclear Information System (INIS)

    Nakazono, Yoshihisa; Iwai, Takeo; Abe, Hiroaki

    2009-01-01

    The supercritical water-cooled reactor (SCWR) has been designed and investigated because of its high thermal efficiency and plant simplification. There are some advantages including the use of a single phase coolant with high enthalpy. Supercritical Water (SCW) has never been used in nuclear power applications. There are numerous potential problems, particularly with materials. As the operating temperature of SCWR will be between 553 K and 893 K with a pressure of 25 MPa, the selection of materials is difficult and important. The PNC1520 austenitic stainless steel has been developed by Japan Atomic Energy Agency (JAEA) as a nuclear fuel cladding material for a Na-cooled fast breeder reactor. Austenitic Fe-base steels were selected for possible use in supercritical water systems because of their corrosion resistance and radiation resistance. The PNC1520 austenitic stainless steel was selected for possible use in supercritical water systems. The corrosion data of PNC1520 in SCW is required but does not exist. The purpose of the present study is to research the corrosion properties for PNC1520 austenitic stainless steel in SCW. The SCW corrosion test was performed for the standard PNC1520 (1520S) and the Ti-additional type of PNC1520 (1520T) by using a SCW autoclave. The 1520S and 1520T are the first trial production materials of SCWR cladding candidate material in our group. Corrosion and compatibility tests on the austenitic 1520S and 1520T steels in supercritical water were performed at 673, 773 and 600degC with exposures up to 1000 h. We have evaluated the amount of weight gain, weight loss and weight of scale after the corrosion test in SCW for 1520S and 1520T austenitic steels. After 1000 h corrosion test performed, the weight gains of both austenitic stainless steels were less than 2 g/m 2 at 400degC and 500degC. But 1520T weight increases more and weight loss than 1520S at 600degC. The SEM observation result of the surface after 1000 h corrosion of an test

  12. Problems of hydrogen - water vapor - inert gas mixture use in heavy liquid metal coolant technology

    International Nuclear Information System (INIS)

    Ul'yanov, V.V.; Martynov, P.N.; Gulevskij, V.A.; Teplyakov, Yu.A.; Fomin, A.S.

    2014-01-01

    The reasons of slag deposit formation in circulation circuits with heavy liquid metal coolants, which can cause reactor core blockage, are considered. To prevent formation of deposits hydrogen purification of coolant and surfaces of circulation circuit is used. It consists in introduction of gaseous mixtures hydrogen - water vapor - rare gas (argon or helium) directly into coolant flow. The principle scheme of hydrogen purification and the processes occurring during it are under consideration. Measures which make it completely impossible to overlap of the flow cross section of reactor core, steam generators, pumps and other equipment by lead oxides in reactor facilities with heavy liquid metal coolants are listed [ru

  13. Supercritical water: On a road from CFD to NPP simulations

    International Nuclear Information System (INIS)

    Rintala, Lauri; Danielyan, Davit; Salomaa, Rainer

    2010-01-01

    The Fission and Radiation Physics Group at the Aalto University is contributing to the Finnish SCWR activities within the GEN4FIN-network. Our research involves reactor core thermal hydraulics, and in particular, heat transfer phenomena in supercritical water including both theoretical studies and simulations with APROS and OpenFOAM. APROS is a software applicable to full-scale power plant simulations and OpenFOAM an open source CFD code. The complicated heat transfer in the supercritical region is a very challenging problem for the design of SCWRs and their safety assessment. The steam tables of APROS have been extended to the supercritical region and their functionality has been tested with, e.g. blowdown simulations where the transient is rapid, hence mainly challenging for numerical stability whereas heat transfer has negligible effects. Numerous different heat correlations for supercritical water have been suggested , but simulations of benchmark experiments have shown that for instance fuel clad temperatures generally cannot be described sufficiently accurately. This discrepancy has been encountered in several process simulation codes. The largest errors occur near the pseudo critical line, during the heat transfer deterioration. It turns out that the physics in supercritical water is clearly more intricate than in ordinary boiling heat transfer where rather satisfactory heat transfer correlations are available. Full 3D CFD calculations allow a better description of various aspects of heat transfer in the supercritical region, i.e., effects arising from turbulence , buoyancy , varying material properties etc. On the other hand, CFD calculations are not feasible for plant-scale simulations. We have selected some simplified geometries and parameter ranges to study SCW heat transfer in a reactor. Old experiments have been calculated with satisfactory results with OpenFOAM to check its validity. A steady state case of heat transfer in a circular pipe with upward

  14. The corrosion products in the coolant circuits of pressurized water nuclear power plants

    International Nuclear Information System (INIS)

    Darras, R.

    1983-01-01

    The characteristics of the corrosion products formed in the primary and secondary coolant circuits of light-water pressurized reactors are reviewed. The problem induced by the pollution of coolants and metallic surface are examined. Then, the recommendations to follow to minimize the disturbing effects of this pollution by the corrosion products are indicated [fr

  15. Reactions of nitrate salts with ammonia in supercritical water

    International Nuclear Information System (INIS)

    Dell'Orco, P.C.; Gloyna, E.F.; Buelow, S.J.

    1997-01-01

    Reactions involving nitrate salts and ammonia were investigated in supercritical water at temperatures from 450 to 530 C and pressures near 300 bar. Reaction products included nitrite, nitrogen gas, and nitrous oxide. Observed reaction rates and product distributions provided evidence for a free-radical reaction mechanism with NO 2 , NO, and NH 2 · as the primary reactive species at supercritical conditions. In the proposed elementary mechanism, the rate-limiting reaction step was determined to be the hydrolysis of MNO 3 species, which resulted in the formation of nitric acid and subsequently NO 2 . A simple second-order reaction model was used to represent the data. In developing an empirical kinetic model, nitrate and nitrate were lumped as an NO x - reactant. Empirical kinetic parameters were developed for four MNO x /NH 3 reacting systems, assuming first orders in both NH 3 and NO x - . Observed MNO x /NH 3 reaction rates and mechanisms suggest immediately a practical significance of these reactions for nitrogen control strategies in supercritical water oxidation processes

  16. Method and apparatus for waste destruction using supercritical water oxidation

    Science.gov (United States)

    Haroldsen, Brent Lowell; Wu, Benjamin Chiau-pin

    2000-01-01

    The invention relates to an improved apparatus and method for initiating and sustaining an oxidation reaction. A hazardous waste, is introduced into a reaction zone within a pressurized containment vessel. An oxidizer, preferably hydrogen peroxide, is mixed with a carrier fluid, preferably water, and the mixture is heated until the fluid achieves supercritical conditions of temperature and pressure. The heating means comprise cartridge heaters placed in closed-end tubes extending into the center region of the pressure vessel along the reactor longitudinal axis. A cooling jacket surrounds the pressure vessel to remove excess heat at the walls. Heating and cooling the fluid mixture in this manner creates a limited reaction zone near the center of the pressure vessel by establishing a steady state density gradient in the fluid mixture which gradually forces the fluid to circulate internally. This circulation allows the fluid mixture to oscillate between supercritical and subcritical states as it is heated and cooled.

  17. An Investigation into Water Chemistry in Primary Coolant Circuit of an Advanced Boiling Water Reactor

    International Nuclear Information System (INIS)

    Wu, Bing-Jhen; Yeh, Tsung-Kuang; Wang, Mei-Ya; Sheu, Rong-Jiun

    2012-09-01

    To ensure operation safety, an optimization on the coolant chemistry in the primary coolant circuit of a nuclear reactor is essential no matter what type or generation the reactor belongs to. For a better understanding toward the water chemistry in an advanced boiling water reactor (ABWR), such as the one being constructed in the northern part of Taiwan, and for a safer operation of this ABWR, we conducted a proactive, thorough water chemistry analysis prior to the completion of this reactor in this study. A numerical simulation model for water chemistry analyses in ABWRs has been developed, based upon the core technology we established in the past. This core technology for water chemistry modeling is basically an integration of water radiolysis, thermal-hydraulics, and reactor physics. The model, by the name of DEMACE - ABWR, is an improved version of the original DEMACE model and was used for radiolysis and water chemistry prediction in the Longmen ABWR in Taiwan. Predicted results pertinent to the water chemistry variation and the corrosion behavior of structure materials in the primary coolant circuit of this ABWR under rated-power operation were reported in this paper. (authors)

  18. Supercritical Water Reactor Cycle for Medium Power Applications

    International Nuclear Information System (INIS)

    BD Middleton; J Buongiorno

    2007-01-01

    Scoping studies for a power conversion system based on a direct-cycle supercritical water reactor have been conducted. The electric power range of interest is 5-30 MWe with a design point of 20 MWe. The overall design objective is to develop a system that has minimized physical size and performs satisfactorily over a broad range of operating conditions. The design constraints are as follows: Net cycle thermal efficiency (ge)20%; Steam turbine outlet quality (ge)90%; and Pumping power (le)2500 kW (at nominal conditions). Three basic cycle configurations were analyzed. Listed in order of increased plant complexity, they are: (1) Simple supercritical Rankine cycle; (2) All-supercritical Brayton cycle; and (3) Supercritical Rankine cycle with feedwater preheating. The sensitivity of these three configurations to various parameters, such as reactor exit temperature, reactor pressure, condenser pressure, etc., was assessed. The Thermoflex software package was used for this task. The results are as follows: (a) The simple supercritical Rankine cycle offers the greatest hardware simplification, but its high reactor temperature rise and reactor outlet temperature may pose serious problems from the viewpoint of thermal stresses, stability and materials in the core. (b) The all-supercritical Brayton cycle is not a contender, due to its poor thermal efficiency. (c) The supercritical Rankine cycle with feedwater preheating affords acceptable thermal efficiency with lower reactor temperature rise and outlet temperature. (d) The use of a moisture separator improves the performance of the supercritical Rankine cycle with feedwater preheating and allows for a further reduction of the reactor outlet temperature, thus it was selected for the next step. Preliminary engineering design of the supercritical Rankine cycle with feedwater preheating and moisture separation was performed. All major components including the turbine, feedwater heater, feedwater pump, condenser, condenser pump

  19. Corrosion in the SCWR: insights from molecular dynamics simulations of the supercritical water - iron hydroxide interface

    Energy Technology Data Exchange (ETDEWEB)

    Kallikragas, D.; Plugatyr, A.; Svishchev, I.M., E-mail: dimitrioskallikragas@trentu.ca [Trent University, Peterborough, Ontario (Canada)

    2013-07-01

    The adsorption properties of supercritical water confined between parallel iron (II) hydroxide surfaces were determined through molecular dynamics simulations. Simulations were conducted at temperatures and water densities typically found in the heat transport system of the supercritical water cooled nuclear reactor (SCWR). Surface water layer densities were compared to those of the bulk water. Adsorption coverage was calculated as a function of the number of waters per surface OH group. Images of the water molecules configurations are provided along with the density profile of the adsorption layer. The observed localized adsorption and surface clustering of supercritical water, would likely produce more localized corrosion phenomena in the water bearing components of the SCWR. (author)

  20. NOMAGE4 activities 2011, Part II, Supercritical water loop

    DEFF Research Database (Denmark)

    Vierstraete, Pierre; Van Nieuwenhove, Rudi; Lauritzen, Bent

    The supercritical water reactor (SCWR) is one of the six different reactor technologies selected for research and development under the Generation IV program. Several countries have shown interest to this concept but up to now, there exist no in-pile facilities to perform the required material...... and fuel tests. Working on this direction, the Halden Reactor Project has started an activity in collaboration with Risoe-DTU (with Mr. Rudi Van Nieuwenhove as the project leader) to study the feasibility of a SCW loop in the Halden Reactor, which is a Heavy Boiling Water Reactor (HBWR). The ultimate goal...

  1. Investigation on flow stability of supercritical water cooled systems

    International Nuclear Information System (INIS)

    Cheng, X.; Kuang, B.

    2006-01-01

    Research activities are ongoing worldwide to develop nuclear power plants with supercritical water cooled reactor (SCWR) with the purpose to achieve a high thermal efficiency and to improve their economical competitiveness. However, the strong variation of the thermal-physical properties of water in the vicinity of the pseudo-critical line results in challenging tasks in various fields, e.g. thermal-hydraulic design of a SCWR. One of the challenging tasks is to understand and to predict the dynamic behavior of supercritical water cooled systems. Although many thermal-hydraulic research activities were carried out worldwide in the past as well as in the near present, studies on dynamic behavior and flow stability of SC water cooled systems are scare. Due to the strong density variation, flow stability is expected to be one of the key items which need to be taken into account in the design of a SCWR. In the present work, the dynamic behavior and flow stability of SC water cooled systems are investigated using both numerical and theoretical approaches. For this purpose a new computer code SASC was developed, which can be applied to analysis the dynamic behavior of systems cooled by supercritical fluids. In addition, based on the assumptions of a simplified system, a theoretical model was derived for the prediction of the onset of flow instability. A comparison was made between the results obtained using the theoretical model and those from the SASC code. A good agreement was achieved. This gives the first evidence of the reliability of both the SASC code and the theoretical model

  2. The effect of low-concentration inorganic materials on the behaviour of supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Imre, A.R., E-mail: imre@aeki.kfki.h [KFKI Atomic Energy Research Institute, POB 49, Budapest (Hungary); Hazi, G.; Horvath, A.; Maraczy, Cs. [KFKI Atomic Energy Research Institute, POB 49, Budapest (Hungary); Mazur, V.; Artemenko, S. [Odessa State Academy of Refrigeration, 1/3 Dvoryanslaya Str., 65026, Odessa (Ukraine)

    2011-01-15

    Research highlights: Small amount of inorganic materials (like corrosion products) can be dissolved in the supercritical water. Pseudo-critical temperature and other properties will be changed. Thermal and hydraulic behaviours of the SCW with small amount of contaminants differ in great extent from the behaviour of pure SCW. - Abstract: Supercritical water is a promising working fluid in the new Generation IV nuclear power plants. Due to the presence of the pseudo-critical line, the thermo-hydraulics (thermal and flow properties) and the physical chemistry of the supercritical water differ significantly from the pressurized hot water used in pressurized water reactors. In this study we would like to analyse the effect of small amount of inorganic material on the thermo-hydraulics of the supercritical water cooled nuclear reactors and other, non-nuclear supercritical water loops.

  3. Coupled 3D neutron kinetics and thermalhydraulic characteristics of the Canadian supercritical water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Hummel, David William, E-mail: hummeld@mcmaster.ca; Novog, David Raymond

    2016-03-15

    Highlights: • A coupled spatial kinetics and thermalhydraulics model of the PT-SCWR was created. • Positive power excursions were demonstrated during accident-like transients. • The reactor will inherently self-shutdown in such transients with some delay. • A fast-acting shutdown system would limit the consequences of the power pulse. - Abstract: The Canadian Supercritical Water-cooled Reactor concept, as an evolution of the CANada Deuterium Uranium (CANDU) reactor, includes both pressure tubes and a low temperature heavy water moderator. The current Pressure Tube type SCWR (PT-SCWR) concept features 64-element fuel assemblies placed within High Efficiency Re-entrant Channels (HERCs) that connect to core inlet and outlet plena. Among current SCWR concepts the PT-SCWR is unique in that the HERC separates multiple coolant and moderator regions, giving rise to coupled neutronic-thermalhydraulic feedbacks beyond those present in CANDU or contemporary Light Water Reactors. The objective of this work was thus to model the coupled neutronic-thermal hydraulic properties of the PT-SCWR to establish the impact of these multiple regions on the core's transient behavior. To that end, the features of the PT-SCWR were first modeled with the neutron transport code DRAGON to create a database of homogenized and condensed cross-sections and thermalhydraulic feedback coefficients. These were used as input to a core-level neutron diffusion model created with the code DONJON. The behavior of the primary heat transport system was modeled with the thermalhydraulic system code CATHENA. A procedure was developed to couple the outputs of DONJON and CATHENA, facilitating three-dimensional spatial neutron kinetics and coupled thermalhydraulic analysis of the PT-SCWR core. Several postulated transients were initiated within the coupled model by changing the core inlet and outlet boundary conditions. Decreasing coolant density around the fuel was demonstrated to produce positive

  4. Fundamental R and D program on water chemistry of supercritical pressure water under radiation field

    International Nuclear Information System (INIS)

    Katsumura, Yosuke; Kiuchi, Kiyoshi; Wada, Yoichi; Yotsuyanagi, Tadasu

    2003-01-01

    In a supercritical water-cooled reactor, property of water changes significantly around the critical point. It is expected that irradiation and change of water property will affect the chemistry and material corrosion. Deep understanding of interactions between supercritical water and materials under irradiation is important. However, comprehensive data on radiolysis, kinetics, corrosion and thermodynamics have not been obtained due to the severe experimental condition. To get such data by experiments and computer simulations, a national program funded by Ministry of Education, Culture, Sports, Science and Technology (MEXT) has been started since December 2002. (author)

  5. Hydrogen production from high moisture content biomass in supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Antal, M.J. Jr.; Xu, X. [Univ. of Hawaii, Honolulu, HI (United States). Hawaii Natural Energy Inst.

    1998-08-01

    By mixing wood sawdust with a corn starch gel, a viscous paste can be produced that is easily delivered to a supercritical flow reactor by means of a cement pump. Mixtures of about 10 wt% wood sawdust with 3.65 wt% starch are employed in this work, which the authors estimate to cost about $0.043 per lb. Significant reductions in feed cost can be achieved by increasing the wood sawdust loading, but such an increase may require a more complex pump. When this feed is rapidly heated in a tubular flow reactor at pressures above the critical pressure of water (22 MPa), the sawdust paste vaporizes without the formation of char. A packed bed of carbon catalyst in the reactor operating at about 650 C causes the tarry vapors to react with water, producing hydrogen, carbon dioxide, and some methane with a trace of carbon monoxide. The temperature and history of the reactor`s wall influence the hydrogen-methane product equilibrium by catalyzing the methane steam reforming reaction. The water effluent from the reactor is clean. Other biomass feedstocks, such as the waste product of biodiesel production, behave similarly. Unfortunately, sewage sludge does not evidence favorable gasification characteristics and is not a promising feedstock for supercritical water gasification.

  6. Flow analysis in a supercritical water oxidation reactor

    International Nuclear Information System (INIS)

    Oh, C.H.; Kochan, R.J.; Beller, J.M.

    1996-01-01

    Supercritical water oxidation (SCWO), also known as hydrothermal oxidation (HTO), involves the oxidation of hazardous waste at conditions of elevated temperature and pressure (e.g., 500 C--600 C and 234.4 bar) in the presence of approximately 90% of water and a 10% to 20% excess amount of oxidant over the stoichiometric requirement. Under these conditions, organic compounds are completely miscible with supercritical water, oxygen and nitrogen, and are rapidly oxidized to carbon dioxide and water. The essential part of the process is the reactor. Many reactor designs such as tubular, vertical vessel, and transpiring wall type have been proposed, patented, and tested at both bench and pilot scales. These designs and performances need to be scaled up to a waste throughput 10--100 times that currently being tested. Scaling of this magnitude will be done by creating a numerical thermal-hydraulic model of the smaller reactor for which test data is available, validating the model against the available data, and then using the validated model to investigate the larger reactor performance. This paper presents a flow analysis of the MODAR bench scale reactor (vertical vessel type). These results will help in the design of the reactor in an efficient manner because the flow mixing coupled with chemical kinetics eventually affects the process destruction efficiency

  7. Water quality control device and water quality control method for reactor primary coolant system

    International Nuclear Information System (INIS)

    Wada, Yoichi; Ibe, Eishi; Watanabe, Atsushi.

    1995-01-01

    The present invention is suitable for preventing defects due to corrosion of structural materials in a primary coolant system of a BWR type reactor. Namely, a concentration measuring means measures the concentration of oxidative ingredients contained in a reactor water. A reducing electrode is disposed along a reactor water flow channel in the primary coolant system and reduces the oxidative ingredients. A reducing counter electrode is disposed along the reactor water flow channel in the primary coolant system, and electrically connected to the reducing electrode. The reactor structural materials are used as a reference electrode providing a reference potential to the reducing electrode and the reducing counter electrode. A potential control means controls the potential of the reducing electrode relative to the reference potential based on the signals from the concentration measuring means. A stable reference potential in a region where an effective oxygen concentration is stable can be obtained irrespective of the change of operation conditions by using the reactor structural materials disposed to a boiling region in the reactor core as a reference electrode. As a result, the water quality can be controlled at high accuracy. (I.S.)

  8. Lewis-acid catalyzed depolymerization of Protobind lignin in supercritical water and ethanol

    NARCIS (Netherlands)

    Güvenatam, B.; Heeres, E.H.J.; Pidko, E.A.; Hensen, E.J.M.

    2014-01-01

    The use of metal acetates, metal chlorides and metal triflates as Lewis acid catalysts for the depolymerization of soda lignin under supercritical conditions was investigated. The reactions were carried out at 400°C in water and ethanol. Lignin conversion in supercritical water led to formation of

  9. Lewis-acid catalyzed depolymerization of Protobind lignin in supercritical water and ethanol

    NARCIS (Netherlands)

    Guvenatam, Burcu; Heeres, Erik H.J.; Pidko, Evgeny A.; Hensen, Ernie J. M.

    2016-01-01

    The use of metal acetates, metal chlorides and metal triflates as Lewis acid catalysts for the depolymerization of soda lignin under supercritical conditions was investigated. The reactions were carried out at 400 degrees C in water and ethanol. Lignin conversion in supercritical water led to

  10. Supercritical water oxidation test bed effluent treatment study

    International Nuclear Information System (INIS)

    Barnes, C.M.

    1994-04-01

    This report presents effluent treatment options for a 50 h Supercritical Water Test Unit. Effluent compositions are calculated for eight simulated waste streams, using different assumed cases. Variations in effluent composition with different reactor designs and operating schemes are discussed. Requirements for final effluent compositions are briefly reviewed. A comparison is made of two general schemes. The first is one in which the effluent is cooled and effluent treatment is primarily done in the liquid phase. In the second scheme, most treatment is performed with the effluent in the gas phase. Several unit operations are also discussed, including neutralization, mercury removal, and evaporation

  11. Supercritical water gasification with decoupled pressure and heat transfer modules

    KAUST Repository

    Dibble, Robert

    2017-09-14

    The present invention discloses a system and method for supercritical water gasification (SCWG) of biomass materials wherein the system includes a SCWG reactor and a plurality of heat exchangers located within a shared pressurized vessel, which decouples the function of containing high pressure from the high temperature function. The present invention allows the heat transfer function to be conducted independently from the pressure transfer function such that the system equipment can be designed and fabricated in manner that would support commercial scaled-up SCWG operations. By using heat exchangers coupled to the reactor in a series configuration, significant efficiencies are achieved by the present invention SCWG system over prior known SCWG systems.

  12. Assessment of heat transfer correlations for supercritical water in the frame of best-estimate code validation

    International Nuclear Information System (INIS)

    Jaeger, Wadim; Espinoza, Victor H. Sanchez; Schneider, Niko; Hurtado, Antonio

    2009-01-01

    Within the frame of the Generation IV international forum six innovative reactor concepts are the subject of comprehensive investigations. In some projects supercritical water will be considered as coolant, moderator (as for the High Performance Light Water Reactor) or secondary working fluid (one possible option for Liquid Metal-cooled Fast Reactors). Supercritical water is characterized by a pronounced change of the thermo-physical properties when crossing the pseudo-critical line, which goes hand in hand with a change in the heat transfer (HT) behavior. Hence, it is essential to estimate, in a proper way, the heat-transfer coefficient and subsequently the wall temperature. The scope of this paper is to present and discuss the activities at the Institute for Reactor Safety (IRS) related to the implementation of correlations for wall-to-fluid HT at supercritical conditions in Best-Estimate codes like TRACE as well as its validation. It is important to validate TRACE before applying it to safety analyses of HPLWR or of other reactor systems. In the past 3 decades various experiments have been performed all over the world to reveal the peculiarities of wall-to-fluid HT at supercritical conditions. Several different heat transfer phenomena such as HT enhancement (due to higher Prandtl numbers in the vicinity of the pseudo-critical point) or HT deterioration (due to strong property variations) were observed. Since TRACE is a component based system code with a finite volume method the resolution capabilities are limited and not all physical phenomena can be modeled properly. But Best -Estimate system codes are nowadays the preferred option for safety related investigations of full plants or other integral systems. Thus, the increase of the confidence in such codes is of high priority. In this paper, the post-test analysis of experiments with supercritical parameters will be presented. For that reason various correlations for the HT, which considers the characteristics

  13. Exhaust temperature analysis of four stroke diesel engine by using MWCNT/Water nanofluids as coolant

    Science.gov (United States)

    Muruganandam, M.; Mukesh Kumar, P. C.

    2017-10-01

    There has been a continuous improvement in designing of cooling system and in quality of internal combustion engine coolants. The liquid engine coolant used in early days faced many difficulties such as low boiling, freezing points and inherently poor thermal conductivity. Moreover, the conventional coolants have reached their limitations of heat dissipating capacity. New heat transfer fluids have been developed and named as nanofluids to try to replace traditional coolants. Moreover, many works are going on the application of nanofluids to avail the benefits of them. In this experimental investigation, 0.1, 0.3 and 0.5% volume concentrations of multi walled carbon nanotube (MWCNT)/water nanofluids have been prepared by two step method with surfactant and is used as a coolant in four stroke single cylinder diesel engine to assess the exhaust temperature of the engine. The nanofluid prepared is characterized with scanning electron microscope (SEM) to confirm uniform dispersion and stability of nanotube with zeta potential analyzer. Experimental tests are performed by various mass flow rate such as 270 300 330 LPH (litre per hour) of coolant nanofluids and by changing the load in the range of 0 to 2000 W and by keeping the engine speed constant. It is found that the exhaust temperature decreases by 10-20% when compared to water as coolant at the same condition.

  14. Assessment of Loss-of-Coolant Effect on Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    Kim, Won Young; Park, Joo Hwan; Kim, Bong Ghi

    2009-01-01

    A CANDU reactor is a heavy-water-moderated, natural uranium fuelled reactor with a pressure tube. The reactor contains a horizontal cylindrical vessel (calandria) and each pressure tube is isolated from the heavy-water moderator in a calandria. This allows the moderator system to be operated of a high-pressure and of a high-temperature coolant in pressure tube. This causes the pressurized liquid coolant in the channel to void and therefore give rise to a reactivity transient in the event of a break or fault in the coolant circuit. In particular, all CANDU reactors are well known to have a positive void reactivity coefficient and thus this phenomenon may lead to a positive feedback, which can cause a large power pulse. We assess the loss-of-coolant effect by coolant void reactivity versus fuel burnup, four factor parameters for fresh fuel and equilibrium fuel, reactivity change due to the change of coolant density and reactivity change in the case of half- and full-core coolant

  15. Present status of study on super-critical water cooled reactor

    International Nuclear Information System (INIS)

    Ookawa, Masahiro; Shiga, Shigenori; Moriya, Kumiaki; Oka, Yoshiaki; Yoshida, Suguru; Takahashi, Heishichiro

    2003-01-01

    Reactor structure design, the core design and coolant flow in sub-channel of fuel assembly are evaluated in the subtitle of plant concepts of the 2002 fiscal year. High temperature parts and high pressure parts are separated on the reactor structure design. Reactor pressure vessel (RPV) is designed under the condition of low temperature and high pressure, while, apparatuses and instruments in the reactor core are designed under the condition of high temperature and low pressure. Design of control rods for cold shut down of the reactor are estimated by using monte carlo computation code (MCNP). It reveals that the number of 16 control rods (0.7 cm in dia) per a fuel assembly is needed for getting control rod worth of conventional light water reactor. Radial power peaking factor reduces to 1.27 by using a load pattern of fuel assembly, number and load position of fuel elements with burnable poison and control rod pattern. Distributions of coolant flow rate in the fuel assembly are studied by sub-channel analysis code, SILFEED, for BWR. The fuel assembly with 1.0 mm gaps between fuel rod and water keeps an uniform flow distribution in which no sub-channel below 90% of flow rate appears in the fuel assembly. Heat transfer experiments for a single test fuel are carried out in the subtitle of heat transfer. The heat transfer data obtained by the experiments are fitted well to Watts' formula. Slow strain rate tests (SSRT) for SUS 304 and SUS 316L steels in the subtitle of materials are carried out for studying stress corrosion cracking (SCC) of the materials under the super-critical pressure water environment. Intergranular stress corrosion cracking (IGSCC) takes place in SUS 304, but doesn't take place in SUS 316L. (M. Suetake)

  16. Selective Synthesis of Manganese/Silicon Complexes in Supercritical Water

    Directory of Open Access Journals (Sweden)

    Jiancheng Wang

    2014-01-01

    Full Text Available A series of manganese salts (Mn(NO32, MnCl2, MnSO4, and Mn(Ac2 and silicon materials (silica sand, silica sol, and tetraethyl orthosilicate were used to synthesize Mn/Si complexes in supercritical water using a tube reactor. X-ray diffraction (XRD, X-ray photoelectron spectrometer (XPS, transmission electron microscopy (TEM, and scanning electron microscopy (SEM were employed to characterize the structure and morphology of the solid products. It was found that MnO2, Mn2O3, and Mn2SiO4 could be obtained in supercritical water at 673 K in 5 minutes. The roles of both anions of manganese salts and silicon species in the formation of manganese silicon complexes were discussed. The inorganic manganese salt with the oxyacid radical could be easily decomposed to produce MnO2/SiO2 and Mn2O3/SiO2. It is interesting to found that Mn(Ac2 can react with various types of silicon to produce Mn2SiO4. The hydroxyl groups of the SiO2 surface from different silicon sources enhance the reactivity of SiO2.

  17. Hydrogen production by supercritical water gasification of alkaline black liquor

    Energy Technology Data Exchange (ETDEWEB)

    Cao, Changqing; Guo, Liejin; Chen, Yunan; Lu, Youjun [Xi' an Jiatong Univ. (China)

    2010-07-01

    Black liquor was gasified continuously in supercritical water successfully and the main gaseous products were H{sub 2}, CO{sub 2} and CH{sub 4} with little amount of CO, C{sub 2}H{sub 4} and C{sub 2}H{sub 6}. The increase of the temperature and the decrease of the flow rate and black liquor concentration enhanced SCWG of black liquor. The change of the system pressure had limited influence on the gasification effect. The maximal COD removal efficiency of 88.69 % was obtained at the temperature of 600 C. The pH values of the aqueous residue were all decreased to the range of 6.4{proportional_to}8 while the pH value of cooling effluence below 360 C increased to about 11 and the sodium content was much higher than that in the aqueous residue. The reaction rate for COD degradation in supercritical water was obtained by assuming pseudo first order reaction. And the activation energy and pre-exponential for COD removal in SCWG were 74.38kJ/mol and 1.11 x 10{sup 4} s{sup -1} respectively. (orig.)

  18. NOMAGE4 activities 2011. Part II, Supercritical water loop

    Energy Technology Data Exchange (ETDEWEB)

    Vierstraete, P. (Ecole Nationale Superieure des mines, Paris (France)); Van Nieuwenhove, R. (Institutt for Energiteknikk, OECD Halden Reactor Project (HRP), Kjeller (Norway)); Lauritzen, B. (Technical Univ. of Denmark, Risoe National Lab. for Sustainable Energy, Roskilde (Denmark))

    2012-01-15

    The supercritical water reactor (SCWR) is one of the six different reactor technologies selected for research and development under the Generation IV program. Several countries have shown interest to this concept but up to now, there exist no in-pile facilities to perform the required material and fuel tests. Working on this direction, the Halden Reactor Project has started an activity in collaboration with Risoe-DTU (with Mr. Rudi Van Nieuwenhove as the project leader) to study the feasibility of a SCW loop in the Halden Reactor, which is a Heavy Boiling Water Reactor (HBWR). The ultimate goal of the project is to design a loop allowing material and fuel test studies at significant mass flow with in-core instrumentation and chemistry control possibilities. The present report focusses on the main heat exchanger required for such a loop in the Halden Reactor. The goal of this heat exchanger is to assure a supercritical flow state inside the test section (the core side) and a subcritical flow state inside the pump section. The objective is to design the heat exchanger in order to optimize the efficiency of the heat transfer and to respect several requirements as the room available inside the reactor hall, the maximal total pressure drop allowed and so on. (Author)

  19. Supercritical water oxidation of ion exchange resins: Degradation mechanisms

    Energy Technology Data Exchange (ETDEWEB)

    Leybros, A.; Roubaud, A. [CEA Marcoule, DEN DTCD SPDE LFSM, F-30207 Bagnols Sur Ceze (France); Guichardon, P. [Ecole Cent Marseille, F-13451 Marseille 20 (France); Boutin, O. [Aix Marseille Univ, UMR CNRS 6181, F-13545 Aix En Provence 4 (France)

    2010-07-01

    Spent ion exchange resins are radioactive process wastes for which there is no satisfactory industrial treatment. Supercritical water oxidation could offer a viable treatment alternative to destroy the organic structure of resins and contain radioactivity. IER degradation experiments were carried out in a continuous supercritical water reactor. Total organic carbon degradation rates in the range of 95-98% were obtained depending on operating conditions. GC-MS chromatography analyses were carried out to determine intermediate products formed during the reaction. Around 50 species were identified for cationic and anionic resins. Degradation of poly-styrenic structure leads to the formation of low molecular weight compounds. Benzoic acid, phenol and acetic acid are the main compounds. However, other products are detected in appreciable yields such as phenolic species or heterocycles, for anionic IERs degradation. Intermediates produced by intramolecular rearrangements are also obtained. A radical degradation mechanism is proposed for each resin. In this overall mechanism, several hypotheses are foreseen, according to HOO center dot radical attack sites. (authors)

  20. Practical Suggestions for Calculating Supercritical Water-Steam Properties

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Seongil; Choi, Sangmin [Korea Advanced Institute of Science and Technology, Daejeon (Korea, Republic of)

    2016-12-15

    A standard procedure for determining water-steam properties has been established through an international collaboration in addition to a domestic effort. The current accepted international standard for industrial application is based on the IAPWS-IF97 (International Association for the Properties of Water and Steam-Industrial Formation 97). Based on this standard, the ASME (American Society of Mechanical Engineers)/NIST (National Institute of Standard and Technology) developed the REPROP program in the USA, and the JSME (Japan Society of Mechanical Engineers) developed the steam table and calculation code. Upon applying this standard procedure, modified procedures were proposed for computational convenience, particularly in the supercritical pressure region where non-smooth variations of water-steam properties were distinctively observed. In this paper, the internationally adopted procedures and the progress of related activities are briefly summarized. Some practical considerations are presented for the efficient execution of computational code.

  1. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    Energy Technology Data Exchange (ETDEWEB)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-04-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications.

  2. Integrity of the reactor coolant boundary of the European pressurized water reactor (EPR)

    International Nuclear Information System (INIS)

    Goetsch, D.; Bieniussa, K.; Schulz, H.; Jalouneix, J.

    1997-01-01

    This paper is an abstract of the work performed in the frame of the development of the IPSN/GRS approach in view of the EPR conceptual safety features. EPR is a pressurized water reactor which will be based on the experience gained by utilities and designers in France and in Germany. The reactor coolant boundary of a PWR includes the reactor pressure vessel (RPV), those parts of the steam generators (SGs) which contain primary coolant, the pressurizer (PSR), the reactor coolant pumps (RCPs), the main coolant lines (MCLs) with their branches as well as the other connecting pipes and all branching pipes including the second isolation valves. The present work covering the integrity of the reactor coolant boundary is mainly restricted to the integrity of the main coolant lines (MCLs) and reflects the design requirements for the main components of the reactor coolant boundary. In the following the conceptual aspects, i.e. design, manufacture, construction and operation, will be assessed. A main aspect is the definition of break postulates regarding overall safety implications

  3. Feasibility Study of Supercritical Light Water Cooled Fast Reactors for Actinide Burning and Electric Power Production Progress Report for Year 1, Quarter 2 (January - March 2002)

    Energy Technology Data Exchange (ETDEWEB)

    Mac Donald, Philip Elsworth; Buongiorno, Jacopo; Davis, Cliff Bybee; Weaver, Kevan Dean

    2002-03-01

    The use of light water at supercritical pressures as the coolant in a nuclear reactor offers the potential for considerable plant simplification and consequent capital and O&M cost reduction compared with current light water reactor (LWR) designs. Also, given the thermodynamic conditions of the coolant at the core outlet (i.e. temperature and pressure beyond the water critical point), very high thermal efficiencies of the power conversion cycle are possible (i.e. up to about 45%). Because no change of phase occurs in the core, the need for steam separators and dryers as well as for BWR-type re-circulation pumps is eliminated, which, for a given reactor power, results in a substantially shorter reactor vessel and smaller containment building than the current BWRs. Furthermore, in a direct cycle the steam generators are not needed.

  4. Numerical study on coolant flow distribution at the core inlet for an integral pressurized water reactor

    Energy Technology Data Exchange (ETDEWEB)

    Sun, Lin; Peng, Min Jun; Xia, Genglei; Lv, Xing; Li, Ren [Fundamental Science on Nuclear Safety and Simulation Technology Laboratory, Harbin Engineering University, Harbin (China)

    2017-02-15

    When an integral pressurized water reactor is operated under low power conditions, once-through steam generator group operation strategy is applied. However, group operation strategy will cause nonuniform coolant flow distribution at the core inlet and lower plenum. To help coolant flow mix more uniformly, a flow mixing chamber (FMC) has been designed. In this paper, computational fluid dynamics methods have been used to investigate the coolant distribution by the effect of FMC. Velocity and temperature characteristics under different low power conditions and optimized FMC configuration have been analyzed. The results illustrate that the FMC can help improve the nonuniform coolant temperature distribution at the core inlet effectively; at the same time, the FMC will induce more resistance in the downcomer and lower plenum.

  5. High-frequency dynamics of liquid and supercritical water

    International Nuclear Information System (INIS)

    Bencivenga, F.; Cunsolo, A.; Krisch, M.; Monaco, G.; Sette, F.; Ruocco, G.

    2007-01-01

    The dynamic structure factor S(Q,ω) of water has been determined by high-resolution inelastic x-ray scattering (IXS) in a momentum (Q) and energy (E) transfer range extending from 2 to 4 nm -1 and from ±40 meV. IXS spectra have been recorded along an isobaric path (400 bar) in a temperature (T) interval ranging from ambient up to supercritical (T>647 K) conditions. The experimental data have been described in the frame of the generalized hydrodynamic theory, utilizing a model based on the memory function approach. This model allows identifying the active relaxation processes which affect the time decay of density fluctuations, as well as a direct determination of the Q, T, and density (ρ) dependencies of the involved transport parameters. The experimental spectra are well described by considering three different relaxation processes: the thermal, the structural, and the instantaneous one. On approaching supercritical conditions, we observe that the microscopic mechanism responsible for the structural relaxation is no longer related to the making and breaking of intermolecular bonds, but to binary intermolecular collisions

  6. Thermophysical properties of supercritical water and bond flexibility.

    Science.gov (United States)

    Shvab, I; Sadus, Richard J

    2015-07-01

    Molecular dynamics results are reported for the thermodynamic properties of supercritical water using examples of both rigid (TIP4P/2005) and flexible (TIP4P/2005f) transferable interaction potentials. Data are reported for pressure, isochoric and isobaric heat capacities, the thermal expansion coefficient, isothermal and adiabatic compressibilities, Joule-Thomson coefficient, speed of sound, self-diffusion coefficient, viscosities, and thermal conductivity. Many of these properties have unusual behavior in the supercritical phase such as maximum and minimum values. The effectiveness of bond flexibility on predicting these properties is determined by comparing the results to experimental data. The influence of the intermolecular potential on these properties is both variable and state point dependent. In the vicinity of the critical density, the rigid and flexible potentials yield very different values for the compressibilities, heat capacities, and thermal expansion coefficient, whereas the self-diffusion coefficient, viscosities, and thermal conductivities are much less potential dependent. Although the introduction of bond flexibility is a computationally expedient way to improve the accuracy of an intermolecular potential, it can be counterproductive in some cases and it is not an adequate replacement for incorporating the effects of polarization.

  7. Supercritical fluid extraction of selected pharmaceuticals from water and serum.

    Science.gov (United States)

    Simmons, B R; Stewart, J T

    1997-01-24

    Selected drugs from benzodiazepine, anabolic agent and non-steroidal anti-inflammatory drug (NSAID) therapeutic classes were extracted from water and serum using a supercritical CO2 mobile phase. The samples were extracted at a pump pressure of 329 MPa, an extraction chamber temperature of 45 degrees C, and a restrictor temperature of 60 degrees C. The static extraction time for all samples was 2.5 min and the dynamic extraction time ranged from 5 to 20 min. The analytes were collected in appropriate solvent traps and assayed by modified literature HPLC procedures. Analyte recoveries were calculated based on peak height measurements of extracted vs. unextracted analyte. The recovery of the benzodiazepines ranged from 80 to 98% in water and from 75 to 94% in serum. Anabolic drug recoveries from water and serum ranged from 67 to 100% and 70 to 100%, respectively. The NSAIDs were recovered from water in the 76 to 97% range and in the 76 to 100% range from serum. Accuracy, precision and endogenous peak interference, if any, were determined for blank and spiked serum extractions and compared with classical sample preparation techniques of liquid-liquid and solid-phase extraction reported in the literature. For the benzodiazepines, accuracy and precision for supercritical fluid extraction (SFE) ranged from 1.95 to 3.31 and 0.57 to 1.25%, respectively (n = 3). The SFE accuracy and precision data for the anabolic agents ranged from 4.03 to 7.84 and 0.66 to 2.78%, respectively (n = 3). The accuracy and precision data reported for the SFE of the NSAIDs ranged from 2.79 to 3.79 and 0.33 to 1.27%, respectively (n = 3). The precision of the SFE method from serum was shown to be comparable to the precision obtained with other classical preparation techniques.

  8. Conceptual designing of reduced-moderation water reactor with heavy water coolant

    Energy Technology Data Exchange (ETDEWEB)

    Hibi, Kohki; Shimada, Shoichiro; Okubo, Tsutomu E-mail: okubo@hems.jaeri.go.jp; Iwamura, Takamichi; Wada, Shigeyuki

    2001-12-01

    The conceptual designing of reduced-moderation water reactors, i.e. advanced water-cooled reactors using plutonium mixed-oxide fuel with high conversion ratios more than 1.0 and negative void reactivity coefficients, has been carried out. The core is designed on the concept of a pressurized water reactor with a heavy water coolant and a triangular tight lattice fuel pin arrangement. The seed fuel assembly has an internal blanket region inside the seed fuel region as well as upper and lower blanket regions (i.e. an axial heterogeneous core). The radial blanket fuel assemblies are introduced in a checkerboard pattern among the seed fuel assemblies (i.e. a radial heterogeneous core). The radial blanket region is shorter than the seed fuel region. This study shows that the heavy water moderated core can achieve negative void reactivity coefficients and conversion ratios of 1.06-1.11.

  9. High converter pressurized water reactor with heavy water as a coolant

    International Nuclear Information System (INIS)

    Ronen, Y.; Reyev, D.

    1983-01-01

    There is an increasing interest in water breeder and high converter reactors. The increase in the conversion ratio of these reactors is obtained by hardening the neutron spectrum achieved by tightening the reactor's lattice. Another way of hardening the neutron spectrum is to replace the light water with heavy water. Two pressurized water reactor fuel cycles that use heavy water as a coolant are considered. The first fuel cycle is based on plutonium and depleted uranium, and the second cycle is based on plutonium and enriched uranium. The uranium ore and separative work unit (SWU) requirements are calculated as well as the fuel cycle cost. The savings in uranium ore are about40 and 60% and about40% in SWU for both fuel cycles considered

  10. Analysis of a water-coolant leak into a very high-temperature vitrification chamber

    International Nuclear Information System (INIS)

    Felicione, F. S.

    1998-01-01

    A coolant-leakage incident occurred during non-radioactive operation of the Plasma Hearth Process waste-vitrification development system at Argonne National Laboratory when a stray electric arc ruptured az water-cooling jacket. Rapid evaporation of the coolant that entered the very high-temperature chamber pressurized the normally sub-atmospheric system above ambient pressure for over 13 minutes. Any positive pressurization, and particularly a lengthy one, is a safety concern since this can cause leakage of contaminants from the system. A model of the thermal phenomena that describe coolant/hot-material interactions was developed to better understand the characteristics of this type of incident. The model is described and results for a variety of hypothetical coolant-leak incidents are presented. It is shown that coolant leak rates above a certain threshold will cause coolant to accumulate in the chamber, and evaporation from this pool can maintain positive pressure in the system long after the leak has been stopped. Application of the model resulted in reasonably good agreement with the duration of the pressure measured during the incident. A closed-form analytic solution is shown to be applicable to the initial leak period in which the peak pressures are generated, and is presented and discussed

  11. Chemistry control challenges in a supercritical water-cooled reactor

    International Nuclear Information System (INIS)

    Guzonas, David; Tremaine, Peter; Jay-Gerin, Jean-Paul

    2009-01-01

    The long-term viability of a supercritical water-cooled reactor (SCWR) will depend on the ability of designers to predict and control water chemistry to minimize corrosion and the transport of corrosion products and radionuclides. Meeting this goal requires an enhanced understanding of water chemistry as the temperature and pressure are raised beyond the critical point. A key aspect of SCWR water chemistry control will be mitigation of the effects of water radiolysis; preliminary studies suggest markedly different behavior than that predicted from simple extrapolations from conventional water-cooled reactor behavior. The commonly used strategy of adding excess hydrogen at concentrations sufficient to suppress the net radiolytic production of primary oxidizing species may not be effective in an SCWR. The behavior of low concentrations of impurities such as transition metal corrosion products, chemistry control agents, anions introduced via make-up water or from ion-exchange resins, and radionuclides (e.g., 60 Co) needs to be understood. The formation of neutral complexes increases with temperature, and can become important under near-critical and supercritical conditions; the most important region is from 300-450 C, where the properties of water change dramatically, and solvent compressibility effects exert a huge influence on solvation. The potential for increased transport and deposition of corrosion products (active and inactive), leading to (a) increased deposition on fuel cladding surfaces, and (b) increased out-of-core radiation fields and worker dose, must be assessed. There are also significant challenges associated with chemistry sampling and monitoring in an SCWR. The typical methods used in current reactor designs (grab samples, on-line monitors at the end of a cooled, depressurized sample line) will be inadequate, and in-situ measurements of key parameters will be required. This paper describes current Canadian activities in SCWR chemistry and chemistry

  12. Effect of supercritical water shell on cavitation bubble dynamics

    International Nuclear Information System (INIS)

    Shao Wei-Hang; Chen Wei-Zhong

    2015-01-01

    Based on reported experimental data, a new model for single cavitation bubble dynamics is proposed considering a supercritical water (SCW) shell surrounding the bubble. Theoretical investigations show that the SCW shell apparently slows down the oscillation of the bubble and cools the gas temperature inside the collapsing bubble. Furthermore, the model is simplified to a Rayleigh–Plesset-like equation for a thin SCW shell. The dependence of the bubble dynamics on the thickness and density of the SCW shell is studied. The results show the bubble dynamics depends on the thickness but is insensitive to the density of the SCW shell. The thicker the SCW shell is, the smaller are the wall velocity and the gas temperature in the bubble. In the authors’ opinion, the SCW shell works as a buffering agent. In collapsing, it is compressed to absorb a good deal of the work transformed into the bubble internal energy during bubble collapse so that it weakens the bubble oscillations. (paper)

  13. Supercritical water oxidation benchscale testing metallurgical analysis report

    International Nuclear Information System (INIS)

    Norby, B.C.

    1993-02-01

    This report describes metallurgical evaluation of witness wires from a series of tests using supercritical water oxidation (SCWO) to process cutting oil containing a simulated radionuclide. The goal of the tests was to evaluate the technology's ability to process a highly chlorinated waste representative of many mixed waste streams generated in the DOE complex. The testing was conducted with a bench-scale SCWO system developed by the Modell Development Corporation. Significant test objectives included process optimization for adequate destruction efficiency, tracking the radionuclide simulant and certain metals in the effluent streams, and assessment of reactor material degradation resulting from processing a highly chlorinated waste. The metallurgical evaluation described herein includes results of metallographic analysis and Scanning Electron Microscopy analysis of witness wires exposed to the SCWO environment for one test series

  14. Measurement of delayed neutron-emitting fission products in nuclear reactor coolant water during reactor operation

    International Nuclear Information System (INIS)

    Anon.

    1981-01-01

    The method covers the detection and measurement of delayed neutron-emitting fission products contained in nuclear reactor coolant water while the reactor is operating. The method is limited to the measurement of the delayed neutron-emitting bromine isotope of mass 87 and the delayed neutron-emitting iodine isotope of mass 137. The other delayed neutron-emitting fission products cannot be accurately distinguished from nitrogen 17, which is formed under some reactor conditions by neutron irradiation of the coolant water molecules. The method includes a description of significance, measurement variables, interferences, apparatus, sampling, calibration, standardization, sample measurement procedures, system efficiency determination, calculations, and precision

  15. Integrated main coolant pumps for pressurized-water reactors

    International Nuclear Information System (INIS)

    Wieser, R.

    1975-01-01

    The efficiency of an integrated main coolant pump for PWR's is increased. For this purpose, the pump is installed eccentric relative to the vertical axis of the U-type steam generator in the three-section HP chamber in such a way that its impeller wheel and the shell of the latter penetrate into the outlet chamber. The axis of the pump lies in the vertical plane of symmetry of the outlet chamber of the steam generator. The suction tube is arranged in the outlet chamber. To allow it to be installed, it is manufactured out of several parts. The diffusor tube, which is also made of several components, is attached to the horizontal separation plate between the outlet chamber and the pressure chamber so as to penetrate into it. To improve the outflow conditions at the diffusor tube, a plowshare-shaped baffle shield is installed between the diffusor tube and the HP chamber. Moreover, in order to improve the outflow conditions from the pump and from the pressure chamber, the outflow opening of the pressure chamber is put into the cylindrical shell of the HP chamber. In this way, the tensioning anchor is located between the pump and the outlet opening. (DG/RF) [de

  16. SCRELA, LOCA Analysis of Super-Critical Light-Water Reactors

    International Nuclear Information System (INIS)

    Lee, J.H.; Koshizuka, S.; Oka, Y.

    2001-01-01

    Description of program or function: LOCA Analysis Code for the Supercritical-Water Cooled Reactor. - Blowdown Module: Calculation of the Blowdown Phase and Refill Phase. - Reflood Module: Calculation of the Reflood Phase

  17. Research on physical and chemical parameters of coolant in Light-Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Reis, Isabela C.; Mesquita, Amir Z., E-mail: icr@cdtn.br, E-mail: amir@cdtn.br [Centro de Desenvolvimento da Tecnologia Nuclear (CDTN/CNEM-MG), Belo Horizonte, MG (Brazil)

    2015-07-01

    The coolant radiochemical monitoring of light-water reactors, both power reactor as research reactors is one most important tasks of the system safe operation. The last years have increased the interest in the coolant chemical studying to optimize the process, to minimize the corrosion, to ensure the primary system materials integrity, and to reduce the workers exposure radiation. This paper has the objective to present the development project in Nuclear Technology Development Center (CDTN), which aims to simulate the primary water physical-chemical parameters of light-water-reactors (LWR). Among these parameters may be cited: the temperature, the pressure, the pH, the electric conductivity, and the boron concentration. It is also being studied the adverse effects that these parameters can result in the reactor integrity. The project also aims the mounting of a system to control and monitoring of temperature, electric conductivity, and pH of water in the Installation of Test in Accident Conditions (ITCA), located in the Thermal-Hydraulic Laboratory at CDTN. This facility was widely used in the years 80/90 for commissioning of several components that were installed in Angra 2 containment. In the test, the coolant must reproduce the physical and chemical conditions of the primary. It is therefore fundamental knowledge of the main control parameters of the primary cooling water from PWR reactors. Therefore, this work is contributing, with the knowledge and the reproduction with larger faithfulness of the reactors coolant in the experimental circuits. (author)

  18. Investigation of coolant mixture in pressurized water reactors at the Rossendorf mixing test facility ROCOM

    International Nuclear Information System (INIS)

    Grunwald, G.; Hoehne, T.; Prasser, H.M.; Richter, K.; Weiss, F.P.

    1999-01-01

    During the so-called boron dilution or cold water transients at pressurized water reactors too weakly borated water or too cold water, respectively, might enter the reactor core. This results in the insertion of positive reactivity and possibly leads to a power excursion. If the source of unborated or subcooled water is not located in all coolant loops but in selected ones only, the amount of reactivity insertion depends on the coolant mixing in the downcomer and lower plenum of the reactor pressure vessel (RPV). Such asymmetric disturbances of the coolant temperature or boron concentration might e.g. be the result of a failure of the chemical and volume control system (CVCS) or of a main steam line break (MSLB) that does only affect selected steam generators (SG). For the analysis of boron dilution or MSLB accidents coupled neutron kinetics/thermo-hydraulic system codes have been used. To take into account coolant mixing phenomena in these codes in a realistic manner, analytical mixing models might be included. These models must be simple and fast running on the one hand, but must well describe the real mixing conditions on the other hand. (orig.)

  19. Radionuclide buildup in BWR [boiling water reactor] reactor coolant recirculation piping

    International Nuclear Information System (INIS)

    Duce, S.W.; Marley, A.W.; Freeman, A.L.

    1989-12-01

    Since the spring of 1985, thermoluminescent dosimeter, dose rate, and gamma spectral data have been acquired on the contamination of boiling water reactor primary coolant recirculation systems as part of a Nuclear Regulatory Commission funded study. Data have been gathered for twelve facilities by taking direct measurements and/or obtaining plant and vendor data. The project titled, ''Effectiveness and Safety Aspects of Selected Decontamination Processes'' (October 1983) initially reviewed the application of chemical decontamination processes on primary coolant recirculation system piping. Recontamination of the system following pipe replacement or chemical decontamination was studied as a second thrust of this program. During the course of this study, recontamination measurements were made at eight different commercial boiling water reactors. At four of the reactors the primary coolant recirculation system piping was chemically decontaminated. At the other four the piping was replaced. Vendor data were obtained from two boiling water reactors that had replaced the primary coolant recirculation system piping. Contamination measurements were made at two newly operating boiling water reactors. This report discusses the results of these measurements as they apply to contamination and recontamination of boiling water reactor recirculation piping. 16 refs., 29 figs., 9 tabs

  20. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

    Energy Technology Data Exchange (ETDEWEB)

    Kryk, Holger, E-mail: h.kryk@hzdr.de [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Hoffmann, Wolfgang [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany); Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany)

    2014-12-15

    Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products.

  1. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Physicochemical effects

    International Nuclear Information System (INIS)

    Kryk, Holger; Hoffmann, Wolfgang; Kästner, Wolfgang; Alt, Sören; Seeliger, André; Renger, Stefan

    2014-01-01

    Highlights: • Physicochemical effects due to post-LOCA zinc corrosion in PWR were elucidated. • Decreasing solubility of corrosion products with increasing temperature was found. • Solid corrosion products may be deposited on hot surfaces and/or within hot-spots. • Corrosion products precipitating from coolant were identified as zinc borates. • Depending on coolant temperature, different types of zinc borate are formed. - Abstract: Within the framework of the reactor safety research, generic experimental investigations were carried out aiming at the physicochemical background of possible zinc corrosion product formation, which may occur inside the reactor pressure vessel during the sump circulation operation after loss-of-coolant accidents in pressurized water reactors. The contact of the boric acid containing coolant with hot-dip galvanized steel containment internals causes corrosion of the corresponding materials resulting in dissolution of the zinc coat. A retrograde solubility of zinc corrosion products with increasing temperature was observed during batch experiments of zinc corrosion in boric acid containing coolants. Thus, the formation and deposition of solid corrosion products cannot be ruled out if the coolant containing dissolved zinc is heated up during its recirculation into hot regions within the emergency cooling circuit (e.g. hot-spots in the core). Corrosion experiments at a lab-scale test facility, which included formation of corrosion products at a single heated cladding tube, proved that dissolved zinc, formed at low temperatures in boric acid solution by zinc corrosion, turns into solid deposits of zinc borates when contacting heated zircaloy surfaces during the heating of the coolant. Moreover, the temperature of formation influences the chemical composition of the zinc borates and thus the deposition and mobilization behavior of the products

  2. Efficiency of water removal from water/ethanol mixtures using supercritical carbon dioxide

    Directory of Open Access Journals (Sweden)

    M. A. Rodrigues

    2006-06-01

    Full Text Available Techniques involving supercritical carbon dioxide have been successfully used for the formation of drug particles with controlled size distributions. However, these processes show some limitations, particularly in processing aqueous solutions. A diagram walking algorithm based on available experimental data was developed to evaluate the effect of ethanol on the efficiency of water removal processes under different process conditions. Ethanol feeding was the key parameter resulting in a tenfold increase in the efficiency of water extraction.

  3. Two-phase coolant pump model of pressurized light water nuclear reactors

    International Nuclear Information System (INIS)

    Santos, G.A. dos; Freitas, R.L.

    1990-01-01

    The two-phase coolant pump model of pressurized light water nuclear reactors is an important point for the loss of primary coolant accident analysis. The homologous curves set up the complete performance of the pump and are input for accidents analysis thermal-hydraulic codes. This work propose a mathematical model able to predict the two-phase homologous curves where it was incorporated geometric and operational pump condition. The results were compared with the experimental tests data from literature and it has showed a good agreement. (author)

  4. Analysis of thermo-hydraulic behavior of coolant during discharge of pressurized high-temperature water

    International Nuclear Information System (INIS)

    Suzuki, Mitsuhiro; Sobajima, Makoto; Sasaki, Shinobu; Onishi, Nobuaki; Shiba, Masayoshi

    1978-01-01

    The present report describes results of the analysis of the LOFT semiscale experiment No. 1011 using remodeled RELAP-3 code, performed at the Idaho National Engineering Laboratory to simulate a postulated loss-of-coolant accident in a pressurized water reactor. It was clarified through the analysis that coolant behavior during blowdown was influenced variously by the system components in the primary loop, comparing with coolant discharge from a pressure vessel. Good agreement was obtained between experimental and analytical results when phase separation was assumed in upper plenum and downcomer, since experimental data indicated existence of liquid level in those parts. It was also found that the use of the Wilson's equation to calculate bubble rise velocity and the use of discharge coefficient as the function of fluid quality at break location to calculate discharge flow rate resulted in good agreement with experimental data. (auth.)

  5. Results of studying of turbulent heat transfer deterioration and their application for development of engineering methods of calculation of heat transfer and pressure drop in supercritical-pressure coolant flow

    International Nuclear Information System (INIS)

    Vladimir A Kurganov; Yuri A Zeigarnik

    2005-01-01

    Full text of publication follows: Using of the supercritical-pressure (SCP) water as a working medium is an apparent way to increase specific capacity and economic efficiency of nuclear power installations. Nevertheless, to provide safe operation of SCP nuclear power units, it is necessary to considerably improve reliability and accuracy of calculations of pressure drop and heat transfer in the SCP working media and coolants flows and the methods of forecasting such a dangerous phenomenon as deterioration of the turbulent heat transfer at a certain level of heat flux density. A value of the latter changes within a very large range depending on the specific conditions of the process under consideration. In the paper, the main results of the experimental study of heat transfer, pressure drop, and velocity and temperature fields in both upward and downward flows of the SCP CO 2 in tubes are considered. This study was conducted at OIVT RAN under conditions of heat input and embraced the regimes of normal and deteriorated heat transfer as well. On the basis of this data, the concept regarding to physical mechanism of incipience of the regimes of deteriorated heat transfer was developed. Classification of different modes of heat transfer deterioration in vertical channels is proposed. A degree of a danger of certain regimes is assessed. It is shown that the above phenomenon is caused by transformation of the structure of nonisothermal flow of SCP fluid due to changes in proportions between the forces acting upon a flow, specifically, because of an increase in the inertia forces due to thermal acceleration of a flow and/or in Archimedes' (buoyancy) forces up to the level comparable or higher than that of friction forces. The efficiency of the most thorough correlations for calculating normal and deteriorated heat transfer in flows of SCP water and CO 2 is analyzed. Reliability of existed recommendations to determine boundaries of normal heat transfer regimes is considered

  6. Simulation of Thermal Hydraulic at Supercritical Pressures with APROS

    Energy Technology Data Exchange (ETDEWEB)

    Kurki, Joona [VTT Technical Research Centre of Finland, P.O. Box 1000, FI02044 VTT (Finland)

    2008-07-01

    The proposed concepts for the fourth generation of nuclear reactors include a reactor operating with water at thermodynamically supercritical state, the Supercritical Water Reactor (SCWR). For the design and safety demonstrations of such a reactor, the possibility to accurately simulate the thermal hydraulics of the supercritical coolant is an absolute prerequisite. For this purpose, the one-dimensional two-phase thermal hydraulics solution of APROS process simulation software was developed to function at the supercritical pressure region. Software modifications included the redefinition of some parameters that have physical significance only at the subcritical pressures, improvement of the steam tables, and addition of heat transfer and friction correlations suitable for the supercritical pressure region. (author)

  7. Water quality estimation method for primary coolant circuit

    International Nuclear Information System (INIS)

    Wada, Yoichi; Ibe, Hidefumi.

    1994-01-01

    The present invention is suitable to water quality diagnosis at each of the portions in a reactor upon hydrogen injection for preventing stress corrosion crackings (SCC) of a BWR type reactor. That is, a plurality of simulations are conducted how the water quality at each of the portions in the reactor is changed when hydrogen injection amount is changed depending on the design and operation conditions of the plant. The result of the calculation is stored in a memory device. A water quality distribution in a pressure vessel having a solution which agrees with a value actually measured by a water quality measuring device disposed at the outside of a reactor core is retrieved from the results of the calculation. If no agreeing solution can be found, water quality distribution containing the actually measured value is determined based on the result of the calculation by using interpolation. In the present invention, the result of the calculation obtained by the simulation and the actually measured value at the outside of the reactor core can be utilized, to map the distribution of reactor water ingredients on a screen, which can accurately estimate the water quality at the periphery of the reactor core on real time. As a result, an operational efficiency of a reactor which can control water quality upon hydrogen injection at an optimum condition. (I.S.)

  8. Supercritical Water Nuclear Steam Supply System: Innovations In Materials, Neutronics and Thermal-Hydraulics

    International Nuclear Information System (INIS)

    Anderson, Mark; Corradini, M.L.; Sridharan, K.; Wilson, P.; Cho, D.; Kim, T.K.; Lomperski, S.

    2004-01-01

    In the 1990's supercritical light-water reactors were considered in conceptual designs. A nuclear reactor cooled by supercritical waster would have a much higher thermal efficiency with a once-through direct power cycle, and could be based on standardized water reactor components (light water or heavy water). The theoretical efficiency could be improved by more than 33% over that of other water reactors and could be simplified with higher reliability; e.g., a boiling water reactor without steam separators or dryers

  9. Development status and application prospect of supercritical-pressure light water cooled reactor

    International Nuclear Information System (INIS)

    Li Manchang; Wang Mingli

    2006-01-01

    The Supercritical-pressure Light Water Cooled Reactor (SCWR) is selected by the Generation IV International Forum (GIF) as one of the six Generation IV nuclear systems that will be developed in the future, and it is an innovative design based on the existing technologies used in LWR and supercritical coal-fired plants. Technically, SCWR may be based on the design, construction and operation experiences in existing PWR and supercritical coal-fired plants, which means that there is no insolvable technology difficulties. Since PWR technology will be adopted in the near term and medium term projects in China, and considering the sustainable development of the technology, it is an inevitable choice to research and develop the nuclear system of supercritical light water cooled reactor. (authors)

  10. Analysis of loss-of-coolant accidents in pressurized water reactors

    International Nuclear Information System (INIS)

    Moldaschl, H.

    1982-01-01

    Analysis of loss-of-coolant accidents in pressurized water reactors -Quantification of the influence of leak size, control assembly worth, boron concentration and initial power by a dynamic operations criterion. Neutronic and thermohydraulic behaviour of a pressurized water reactor during a loss-of-coolant accident (LOCA) is mainly influenced by -change of fuel temperature, -void in the primary coolant. They cause a local stabilization of power density, that means that also in the case of small leaks local void is the main stabilization effect. As a consequence the increase of fuel temperature remains very small even under extremely hypothetical assumptions: small leak, positive reactivity feedback (positive coolant temperature coefficient, negative density coefficient) at the beginning of the accident and all control assemblies getting stuck. Restrictions which have been valid up to now for permitted start-up conditions to fulfill inherent safety requirements can be lossened substantially by a dynamic operations criterion. Burnable poisons for compensation of reactivity theorefore can be omitted. (orig.)

  11. Simulation of small break loss of coolant accident in pressurized water reactor (PWR)

    International Nuclear Information System (INIS)

    Abass, N. M. N.

    2012-02-01

    A major safety concern in pressurized-water-reactor (PWR) design is the loss-of-coolant accident (LOCA),in which a break in the primary coolant circuit leads to depressurization, boiling of the coolant, consequent reduced cooling of the reactor core, and , unless remedial measures are taken, overheating of the fuel rods. This concern has led to the development of several simulators for safety analysis. This study demonstrates how the passive and active safety systems in conventional and advanced PWR behave during the small break loss of Coolant Accident (SBLOCA). The consequences of SBOLOCA have been simulated using IAEA Generic pressurized Water Reactor Simulator (GPWRS) and personal Computer Transient analyzer (PCTRAN) . The results were presented and discussed. The study has confirmed the major safety advantage of passive plants versus conventional PWRs is that the passive safety systems provide long-term core cooling and decay heat removal without the need for operator actions and without reliance on active safety-related system. (Author)

  12. Thermal aspects of mixed oxide fuel in application to supercritical water-cooled nuclear reactors

    Energy Technology Data Exchange (ETDEWEB)

    Grande, L.; Peiman, W.; Rodriguez-Prado, A.; Villamere, B.; Mikhael, S.; Allison, L.; Pioro, I., E-mail: lisa.grande@mycampus.uoit.ca, E-mail: igor.pioro@uoit.ca [Univ. of Ontario Inst. of Tech., Faculty of Energy Systems and Nuclear Science, Oshawa, Ontario (Canada)

    2010-07-01

    SuperCritical Water-cooled nuclear Reactors (SCWRs) are a renewed technology being developed as one of the Generation IV reactor concepts. This reactor type uses a light water coolant at temperatures and pressures above its critical point. These elevated operating conditions will improve Nuclear Power Plant (NPP) thermal efficiencies by 10 - 15% compared to those of current NPPs. Also, SCWRs will have the ability to utilize a direct cycle, thus decreasing NPP capital and operational costs. The SCWR core has 2 configurations: 1) Pressure Vessel (PV) -type enclosing a fuel assembly and 2) Pressure Tube (PT) -type consisting of individual pressurized channels containing fuel bundles. Canada and Russia are developing PT-type SCWRs. In particular, the Canadian SCWR reactor has an output of 1200 MW{sub el} and will operate at a pressure of 25 MPa with inlet and outlet fuel-channel temperatures of 350 and 625°C, respectively. These extreme operating conditions require alternative fuels and materials to be investigated. Current CANadian Deuterium Uranium (CANDU) nuclear reactor fuel-channel design is based on the use of uranium dioxide (UO{sub 2}) fuel; zirconium alloy sheath (clad) bundle, pressure and calandria tubes. Alternative fuels should be considered to supplement depleting world uranium reserves. This paper studies general thermal aspects of using Mixed OXide (MOX) fuel in an Inconel-600 sheath in a generic PT-type SCWR. The bulk fluid, sheath and fuel centerline temperatures along with the Heat Transfer Coefficient (HTC) profiles were calculated at uniform and non-uniform Axial Heat Flux Profiles (AHFPs). (author)

  13. Corrosion behavior of porous chromium carbide in supercritical water

    International Nuclear Information System (INIS)

    Dong Ziqiang; Chen Weixing; Zheng Wenyue; Guzonas, Dave

    2012-01-01

    Highlights: ► Corrosion behavior of porous Cr 3 C 2 in various SCW conditions was investigated. ► Cr 3 C 2 is stable in SCW at temperature below 420–430 °C. ► Cracks and disintegration were observed at elevated testing temperatures. ► Degradation of Cr 3 C 2 is related to the intermediate product CrOOH. - Abstract: The corrosion behavior of highly porous chromium carbide (Cr 3 C 2 ) prepared by a reactive sintering process was characterized at temperatures ranging from 375 °C to 625 °C in a supercritical water environment with a pressure of 25–30 MPa. The test results show that porous chromium carbide is stable in SCW environments at temperatures under 425 °C, above which disintegration occurred. The porous carbide was also tested under hydrothermal conditions of pressures between 12 MPa and 50 MPa at constant temperatures of 400 °C and 415 °C, respectively. The pressure showed little effect on the stability of chromium carbide in the tests at those temperatures. The mechanism of disintegration of chromium carbide in SCW environments is discussed.

  14. Design requirements for the supercritical water oxidation test bed

    International Nuclear Information System (INIS)

    Svoboda, J.M.; Valentich, D.J.

    1994-05-01

    This report describes the design requirements for the supercritical water oxidation (SCWO) test bed that will be located at the Idaho National Engineering Laboratory (INEL). The test bed will process a maximum of 50 gph of waste plus the required volume of cooling water. The test bed will evaluate the performance of a number of SCWO reactor designs. The goal of the project is to select a reactor that can be scaled up for use in a full-size waste treatment facility to process US Department of Energy mixed wastes. EG ampersand G Idaho, Inc. will design and construct the SCWO test bed at the Water Reactor Research Test Facility (WRRTF), located in the northern region of the INEL. Private industry partners will develop and provide SCWO reactors to interface with the test bed. A number of reactor designs will be tested, including a transpiring wall, tube, and vessel-type reactor. The initial SCWO reactor evaluated will be a transpiring wall design. This design requirements report identifies parameters needed to proceed with preliminary and final design work for the SCWO test bed. A flow sheet and Process and Instrumentation Diagrams define the overall process and conditions of service and delineate equipment, piping, and instrumentation sizes and configuration Codes and standards that govern the safe engineering and design of systems and guidance that locates and interfaces test bed hardware are provided. Detailed technical requirements are addressed for design of piping, valves, instrumentation and control, vessels, tanks, pumps, electrical systems, and structural steel. The approach for conducting the preliminary and final designs and environmental and quality issues influencing the design are provided

  15. Power flattening and reactivity suppression strategies for the Canadian supercritical water reactor concept

    International Nuclear Information System (INIS)

    McDonald, M.; Colton, A.; Pencer, J.

    2015-01-01

    The Canadian supercritical water-cooled reactor (SCWR) is a conceptual heavy water moderated, supercritical light water cooled pressure tube reactor. In contrast to current heavy water power reactors, the Canadian SCWR will be a batch fuelled reactor. Associated with batch fuelling is a large beginning-of-cycle excess reactivity. Furthermore, radial power peaking arising as a consequence of batch refuelling must be mitigated in some way. In this paper, burnable neutron absorber (BNA) added to fuel and absorbing rods inserted into the core are considered for reactivity management and power flattening. A combination of approaches appears adequate to reduce the core radial power peaking, while also providing reactivity suppression. (author)

  16. Cerenkov Detectors for Fission Product Monitoring in Reactor Coolant Water

    Energy Technology Data Exchange (ETDEWEB)

    Strindehag, O

    1967-09-15

    The expected properties of Cerenkov detectors when used for fission product monitoring in water cooled reactors and test loops are discussed from the point of view of the knowledge of the sensitivity of these detectors to some beta emitting isotopes. The basic theory for calculation of the detector response is presented, taking the optical transmission in the sample container and the properties of the photomultiplier tube into account. Special attention is paid to the energy resolution of this type of Cerenkov detector. For the design of practical detectors the results from several investigations of various window and reflector materials are given, and the selection of photomultiplier tubes is briefly discussed. In the case of optical reflectors and photomultiplier tubes reference is made to two previous reports by the author. The influence of the size and geometry of the sample container on the energy resolution follows from a separate investigation, as well as the relative merits of sample containers with transparent inner walls. Provided that the energy resolution of the Cerenkov detector is sufficiently high, there are several reasons for using this detector type for failed-fuel-element detection. It seems possible to attain the desired energy resolution by careful detector design.

  17. Independent modification on water lubrication loop of radial-axial bearing of Russian reactor coolant pump

    International Nuclear Information System (INIS)

    Gu Yingbin

    2012-01-01

    Water lubrication was used for radial-axial bearings of 1391M reactor coolant pumps at both units of Tianwan Nuclear Power Plant Phase I Project, which was the first trial on large commercial pressurized water reactors in the world. As a prototype, there were inherent deficiencies leading to a series of operational events. Jiangsu Nuclear Power Corporation conducted the independent innovative technical modification to cope with the defects, and succeeded in reducing heat removal rate of the radial-axial bearings of the reactor coolant pumps, mitigating or preventing the cavitation abrasion of the bearings and improving the cooling effects. This paper illustrates the reasons of the innovative modification, the design and implementation preparation of modification program, the implementation process and evaluation of modification effect, including detailed follow-up work program. (author)

  18. Study on primary coolant system depressurization effect factor in pressurized water reactor

    International Nuclear Information System (INIS)

    Ji Duan; Cao Xuewu

    2006-01-01

    The progression of high-pressure core melting severe accident induced by very small break loss of coolant accident plus the loss of main feed water and auxiliary feed water failure is studied, and the entry condition and modes of primary cooling system depressurization during the severe accident are also estimated. The results show that the temperature below 650 degree C is preferable depressurization input temperature allowing recovery of core cooling, and the available and effective way to depressurize reactor cooling system and to arrest very small break loss of coolant accident sequences is activating pressurizer relief valves initially, then restoring the auxiliary feedwater and opening the steam generator relief valves. It can adequately reduce the primary pressure and keep the capacity loop of long-term core cooling. (authors)

  19. Exergy analysis of a system using a chemical heat pump to link a supercritical water-cooled nuclear reactor and a thermochemical water splitting cycle

    International Nuclear Information System (INIS)

    Granovskii, M.; Dincer, I.; Rosen, M. A.; Pioro, I

    2007-01-01

    The power generation efficiency of nuclear plants is mainly determined by the permissible temperatures and pressures of the nuclear reactor fuel and coolants. These parameters are limited by materials properties and corrosion rates and their effect on nuclear reactor safety. The advanced materials for the next generation of CANDU reactors, which employ steam as a coolant and heat carrier, permit the increased steam parameters (outlet temperature up to 625 degree C and pressure of about 25 MPa). Supercritical water-cooled (SCW) nuclear power plants are expected to increase the power generation efficiency from 35 to 45%. Supercritical water-cooled nuclear reactors can be linked to thermochemical water splitting cycles for hydrogen production. An increased steam temperature from the nuclear reactor makes it also possible to utilize its energy in thermochemical water splitting cycles. These cycles are considered by many as one of the most efficient ways to produce hydrogen from water and to have advantages over traditional low-temperature water electrolysis. However, even lower temperature water splitting cycles (Cu-Cl, UT-3, etc.) require a heat supply at the temperatures over 550-600 degree C. A sufficient increase in the heat transfer from the nuclear reactor to a thermochemical water splitting cycle, without jeopardizing nuclear reactor safety, might be effectively achieved by application of a heat pump which increases the temperature the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. A high temperature chemical heat pump which employs the reversible catalytic methane conversion reaction is proposed. The reaction shift from exothermic to endothermic and back is achieved by a change of the steam concentration in the reaction mixture. This heat pump, coupled with a SCW nuclear plant on one side and thermochemical water splitting cycle on the other, increases the temperature level of the 'nuclear' heat and, thus, the intensity of

  20. Temperature feedback effects in a supercritical water reactor concept with multiple heat-up steps

    Energy Technology Data Exchange (ETDEWEB)

    Barragan-Martinez, A.M., E-mail: albrm29@yahoo.com [Universidad Nacional Autonoma de Mexico, Departamento de Sistemas Energeticos, Facultad de Ingenieria, Jiutepec, Mor (Mexico); Espinosa-Paredes, G.; Vazquez-Rodriguez, A., E-mail: gepe@xanum.uam.mx, E-mail: vara@xanum.uam.mx [Universidad Autonoma Metropolitana-Iztapalapa, Area de Ingenieria en Rescursos Energeticos, Col. Vicentina (Mexico); Martin-del-Campo, C.; Francois, J.L., E-mail: cecilia.martin.del.campo@gmail.com, E-mail: juan.louis.francois@gmail.com [Universidad Nacional Autonoma de Mexico, Departamento de Sistemas Energeticos, Facultad de Ingenieria, Jiutepec, Mor (Mexico)

    2014-07-01

    The Supercritical Water Cooled Reactor (SCWR) is one of the most promising and innovative designs selected by the Generation IV International Forum. One of the concepts being studied is the High Performance Light Water Reactor (HPLWR), which is the European version of the SCWR. In this paper we present the numerical analysis of the behavior of a HPLWR with temperature feedback effects. The neutronic process, the heat transfer in the fuel rod and the thermalhydraulics in the core of the HPLWR were considered in this study. The neutronic calculations were performed with HELIOS-2 and the obtained results were used to evaluate the reactivity due to fuel temperature and supercritical water density. (author)

  1. Temperature feedback effects in a supercritical water reactor concept with multiple heat-up steps

    International Nuclear Information System (INIS)

    Barragan-Martinez, A.M.; Espinosa-Paredes, G.; Vazquez-Rodriguez, A.; Martin-del-Campo, C.; Francois, J.L.

    2014-01-01

    The Supercritical Water Cooled Reactor (SCWR) is one of the most promising and innovative designs selected by the Generation IV International Forum. One of the concepts being studied is the High Performance Light Water Reactor (HPLWR), which is the European version of the SCWR. In this paper we present the numerical analysis of the behavior of a HPLWR with temperature feedback effects. The neutronic process, the heat transfer in the fuel rod and the thermalhydraulics in the core of the HPLWR were considered in this study. The neutronic calculations were performed with HELIOS-2 and the obtained results were used to evaluate the reactivity due to fuel temperature and supercritical water density. (author)

  2. Investigation of R-134a as a modeling fluid for supercritical water

    International Nuclear Information System (INIS)

    Jouvin, J.C.; Pioro, I.

    2014-01-01

    The objective of this paper is to investigate the feasibility of using Refrigerant-134a (R-134a) as a potential modeling fluid by comparing the thermophysical properties with those of water. Operating conditions of SuperCritical Water-cooled Reactors (SCWRs) are scaled into those of R-134a, in order to provide proper SCWR-equivalent conditions. The thermophysical properties for R-134a are obtained from NIST REFPROP software. The results indicate that the thermophysical properties of R-134a undergo significant changes within the critical and pseudocritical regions similar to that of supercritical water. An investigation into the pseudocritical region of R-134a was also conducted. (author)

  3. Lamp system with conditioned water coolant and diffuse reflector of polytetrafluorethylene(PTFE)

    Science.gov (United States)

    Zapata, Luis E.; Hackel, Lloyd

    1999-01-01

    A lamp system with a very soft high-intensity output is provided over a large area by water cooling a long-arc lamp inside a diffuse reflector of polytetrafluorethylene (PTFE) and titanium dioxide (TiO.sub.2) white pigment. The water is kept clean and pure by a one micron particulate filter and an activated charcoal/ultraviolet irradiation system that circulates and de-ionizes and biologically sterilizes the coolant water at all times, even when the long-arc lamp is off.

  4. MIC damage in a water coolant header for remote process equipment

    International Nuclear Information System (INIS)

    Jenkins, C.F.

    1994-01-01

    Stainless steel water piping used to supply coolant for remote chemical separations equipment developed leaks during low flow conditions resulting from an extended interruption of operations. All the leaks occurred at welds in the bottom zone of the pipe, which was blanketed with silt deposits from the unfiltered well water used for cooling. Ultrasonic, radiographic, and metallographic examinations of leak sites revealed worm hole pitting adjacent to the welds. Seepage at the penetrations was strongly acidic and resulted in corrosion on the external pipe surfaces beneath brown crusty deposits which had developed. Analyses of the water and deposits suggest a strong propensity toward microbiologically influenced corrosion (MIC) and fouling

  5. MIC damage in a water coolant header for remote process equipment

    International Nuclear Information System (INIS)

    Jenkins, C.F.

    1996-01-01

    Stainless steel water piping, used to supply coolant for remote chemical separations equipment, developed several leaks during low flow conditions, the result of an extended interruption of operations. All the leaks occurred at welds in the bottom of the pipe, which was blanketed with silt deposits from unfiltered well water used for cooling. Ultrasonic, radiographic, and metallographic examinations of the leak sites revealed worm-hole pitting adjacent to the welds. Seepage at the penetrations was strongly acidic and corroded the external pipe surfaces. Analyses of the water and deposits suggested microbiologically influenced corrosion and fouling

  6. Transformation of heavy metals in lignite during supercritical water gasification

    International Nuclear Information System (INIS)

    Chen, Guifang; Yang, Xinfei; Chen, Shouyan; Dong, Yong; Cui, Lin; Zhang, Yong; Wang, Peng; Zhao, Xiqiang; Ma, Chunyuan

    2017-01-01

    Highlights: • The transformations of heavy metals during lignite SCWG were investigated. • The risks of heavy metals in lignite and residues after SCWG were evaluated. • The effects of experimental conditions on corrosion during SCWG were studied. - Abstract: Transformation characteristics of heavy metals during lignite supercritical water gasification (SCWG) were studied. A sequential extraction procedure (modified Tessier method) was used to selectively extract different fractions of Pb, Cd, Cr, Mn, Cu, Ni, and Zn. Heavy metals transformed into more stable fractions after SCWG. For Pb, Cd, Mn, Cu, and Zn, SCWG reduced the bioavailability and the risks posed by heavy metals in lignite. Under the experimental conditions, the conversion rates for Pb and Cd were 16.0%–25.2% and 16.3%–23.4%, respectively, whereas those for Mn, Cu, and Zn were much lower. Solid products enriched with Pb, Cd, Mn, Cu, and Zn were obtained after SCWG; the contents of these metals varied slightly in the liquid products under different experimental conditions. Excess Cr and Ni that did not originate from lignite were found in the residues, owing to reactor corrosion during lignite SCWG. Higher temperatures alleviated corrosion, whereas higher pressures and equivalence ratios (ER) had the opposite effect. None of the heavy metals were detected in the gas phase under the experimental conditions used in the present study. The correlation between the distributions of heavy metals and the experimental conditions were also studied. The transformation pathways of Pb, Cd, Mn, Cu, and Zn during SCWG were deduced according to the experimental results.

  7. Preliminary assessment of water-based nano-fluids for use as coolants in PWRs

    International Nuclear Information System (INIS)

    Jacopo Buongiorno

    2005-01-01

    Full text of publication follows: The impact of using water-based fluids with small additions (<2% vol.) of nano-sized (10-100 nm) particle populations as coolants for current and advanced PWRs is evaluated. Such 'engineered' fluids (known as nano-fluids) are attractive because the presence of the nano-particles enhances energy transport considerably. As a result, nano-fluids are known to have (i) higher thermal conductivity than water (up to 20% depending on nano-particle material, size and volumetric fraction), (ii) higher heat transfer coefficients (up to 40%), (iii) higher CHF (up to 300% in pool boiling), and (iv) comparable pressure drop. Furthermore, nano-fluids appear to be very stable suspensions with little or no sedimentation, because of the small size of the dispersed particles and their typically low volumetric fractions. The ultimate objective of this work is to assess whether existing PWRs could be retro-fitted with a water-based nano-fluid coolant, to increase safety margins, reduce stored energy, and/or allow for power up-rates. Also, advanced PWRs could be designed with nano-fluids. The linear heat generation rate in PWRs is limited by a) fuel centerline melting, b) cladding overheating (CHF), and c) stored energy release following a large-break LOCA. Mechanisms b) and c) are usually the most limiting. For given geometry and linear power, it is obvious that the core with the nano-fluid coolant will have higher margins to CHF and LOCA limits. Conversely, for given margins, a higher linear power can be accommodated by the nano-fluid-cooled core. Standard thermal-hydraulic models for the PWR hot fuel pin (including a RELAP model for the LOCA) have been used to quantify the benefit of using nano-fluid coolants on the performance of a PWR. (author)

  8. Experimental study of supercritical water flow and heat transfer in vertical tube

    International Nuclear Information System (INIS)

    Li Hongbo; Yang Jue; Lu Donghua; Gu Hanyang; Zhao Meng

    2012-01-01

    The experiment of flow and heat transfer of supercritical water has been performed on the supercritical water multipurpose test loop co-constructed by China Guangdong Nuclear Power Group and Shanghai Jiao Tong University with a 7.6 mm vertical tube. Heat transfer experimental data is obtained. The results of experimental research of thermal-hydraulic parameters on flow and heat transfer of supercritical water show that: (1) Heat transfer enhancement occurs when the bulk temperature reaches pseudo-critical point with low mass flow velocity; (2) The heat transfer co- efficient and Nusselt number are decreased with the increasing of heat flux; (3) The wall temperature is decreased, but the heat transfer coefficient and Nusselt number are increased with the increasing of mass flow velocity; (4) The wall temperature is increased, but the heat transfer coefficient and Nusselt number are decreased with the increasing of sys- tem pressure. (authors)

  9. Error analysis of supercritical water correlations using ATHLET system code under DHT conditions

    Energy Technology Data Exchange (ETDEWEB)

    Samuel, J., E-mail: jeffrey.samuel@uoit.ca [Univ. of Ontario Inst. of Tech., Oshawa, ON (Canada)

    2014-07-01

    The thermal-hydraulic computer code ATHLET (Analysis of THermal-hydraulics of LEaks and Transients) is used for analysis of anticipated and abnormal plant transients, including safety analysis of Light Water Reactors (LWRs) and Russian Graphite-Moderated High Power Channel-type Reactors (RBMKs). The range of applicability of ATHLET has been extended to supercritical water by updating the fluid-and transport-properties packages, thus enabling the code to the used in analysis of SuperCritical Water-cooled Reactors (SCWRs). Several well-known heat-transfer correlations for supercritical fluids were added to the ATHLET code and a numerical model was created to represent an experimental test section. In this work, the error in the Heat Transfer Coefficient (HTC) calculation by the ATHLET model is studied along with the ability of the various correlations to predict different heat transfer regimes. (author)

  10. Oxidation behavior of austenitic iron-base ODS alloy in supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Behnamian, Y.; Dong, Z.; Zahiri, R.; Kohandehghan, A.; Mitlin, D., E-mail: behnamia@ualberta.ca, E-mail: zdong@ualberta.ca, E-mail: kohandeh@ualberta.ca, E-mail: rzahiris@ualberta.ca, E-mail: dave.mitlin@ualberta.ca [Univ. of Alberta, Edmondon, AB (Canada); Zhou, Z., E-mail: zhouzhj@mater.ustb.edu.cn [Univ. of Science and Tech. Beijing, Beijing (China); Chen, W.; Luo, J., E-mail: weixing.chen@ualberta.ca, E-mail: Jingli.luo@ualberta.ca [Univ. of Alberta, Edmonton, AB (Canada); Zheng, W., E-mail: wenyue@nrcan.gc.ca [Natural Resources Canada, Canmet MATERIALS, Hamilton, ON (Canada); Guzonas, D. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2014-07-01

    In this study, the effect of exposure time on the corrosion of the 304 stainless steel based oxide dispersion strengthened alloy, SS304ODS, in supercritical water was investigated at 650 {sup o}C with constant dissolved oxygen concentration. The results show that the oxidation of SS304ODS in supercritical water followed a parabolic law at 650 {sup o}C. Discontinuous oxide scale with two distinct layers has formed after 550 hours. The inner layer was chromium-rich while the outer layer was iron-rich (Magnetite). The oxide islands grow with increasing the exposure time. With increasing exposure time, the quantity of oxide islands increased in which major preferential growth along oxide-substrate interface was observed. The possible mechanism of SS304ODS oxidation in supercritical water was also discussed. (author)

  11. The solubilities of phosphate and sulfate salts in supercritical water

    NARCIS (Netherlands)

    Leusbrock, Ingo; Metz, Sybrand J.; Rexwinkel, Glenn; Versteeg, Geert F.

    Inorganic compounds are regularly present in aqueous streams. To understand their influence and behavior on these streams at supercritical conditions, little to no property data is available, which can be used as starting point for further research or application design. Since inorganic compounds

  12. A potential of boiling water power reactors with a natural circulation of a coolant

    International Nuclear Information System (INIS)

    Osmachkin, V.S.; Sokolov, I.N.

    1998-01-01

    The use of the natural circulation of coolant in the boiling water reactors simplifies a reactor control and facilities the service of the equipment components. The moderated core power loads allows the long fuel burnup, good control ability and large water stock set up the enhancement of safety level. That is considered to be very important for isolated regions or small countries. In the paper a high safety level and effectiveness of BWRs with natural circulation are reviewed. The limitations of flow stability and protection measures are being discussed. Some recent efforts in designing of such reactors are described.(author)

  13. On the gasification of wet biomass in supercritical water : over de vergassing van natte biomassa in superkritiek water

    NARCIS (Netherlands)

    Withag, J.A.M.

    2013-01-01

    Supercritical water gasification (SCWG) is a challenging thermo-chemical conversion route for wet biomass and waste streams into hydrogen and/or methane. At temperatures and pressures above the critical point the physical properties of water differ strongly from liquid water or steam. Because of the

  14. Review and proposal for heat transfer predictions at supercritical water conditions using existing correlations and experiments

    Energy Technology Data Exchange (ETDEWEB)

    Jaeger, Wadim, E-mail: wadim.jaeger@kit.edu [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, DE-76344 Eggenstein-Leopoldshafen (Germany); Sanchez Espinoza, Victor Hugo [Karlsruhe Institute of Technology, Institute for Neutron Physics and Reactor Technology, DE-76344 Eggenstein-Leopoldshafen (Germany); Hurtado, Antonio [Technical University of Dresden, Institute of Power Engineering, DE-01062 Dresden (Germany)

    2011-06-15

    Highlights: > Implementation of heat transfer correlations for supercritical water into TRACE. > Simulation of several heat transfer experiments with modified TRACE version. > Most correlations are not able to reproduce the experimental results. > Bishop, Sandberg and Tong correlation is most suitable for TRACE applications. - Abstract: This paper summarizes the activities of the TRACE code validation at the Institute for Neutron Physics and Reactor Technology related to supercritical water conditions. In particular, the providing of the thermo physical properties and its appropriate use in the wall-to-fluid heat transfer models in the frame of the TRACE code is the object of this investigation. In a first step, the thermo physical properties of the original TRACE code were modified in order to account for supercritical conditions. In a second step, existing Nusselt correlations were reviewed and implemented into TRACE and available experiments were simulated to identify the most suitable Nusselt correlation(s).

  15. Review and proposal for heat transfer predictions at supercritical water conditions using existing correlations and experiments

    International Nuclear Information System (INIS)

    Jaeger, Wadim; Sanchez Espinoza, Victor Hugo; Hurtado, Antonio

    2011-01-01

    Highlights: → Implementation of heat transfer correlations for supercritical water into TRACE. → Simulation of several heat transfer experiments with modified TRACE version. → Most correlations are not able to reproduce the experimental results. → Bishop, Sandberg and Tong correlation is most suitable for TRACE applications. - Abstract: This paper summarizes the activities of the TRACE code validation at the Institute for Neutron Physics and Reactor Technology related to supercritical water conditions. In particular, the providing of the thermo physical properties and its appropriate use in the wall-to-fluid heat transfer models in the frame of the TRACE code is the object of this investigation. In a first step, the thermo physical properties of the original TRACE code were modified in order to account for supercritical conditions. In a second step, existing Nusselt correlations were reviewed and implemented into TRACE and available experiments were simulated to identify the most suitable Nusselt correlation(s).

  16. Experiments in a natural circulation loop with supercritical water at low powers

    International Nuclear Information System (INIS)

    Pilkhwal, D.S.; Sharma, Manish; Jana, S.S.; Vijayan, P.K.

    2013-05-01

    Earlier, 1/2 ″ uniform diameter Supercritical Pressure Natural Circulation Loop (SPNL) was set-up in hall-7, BARC for carrying out experiments related to supercritical fluids. The loop is a rectangular loop having two heaters and two coolers. Experiments were carried out with CO 2 under supercritical conditions for various pressures and different combinations of heater and cooler orientations. Since, the design conditions are more severe for supercritical water (SCW) experiments, the loop was modified for SCW by installing new test sections, pressurizer and power supply for operation with supercritical water. Experimental data were generated on steady state, heat transfer and stability under natural circulation conditions for the horizontal heater and horizontal cooler (HHHC) orientation with SCW up to a heater power of 8.5 kW. The flow rate data and instability data were compared with the predictions of in-house developed 1-D code NOLSTA, which showed reasonable agreement. The heat transfer coefficient data were also compared with the predictions of various correlations exhibit peak at bulk temperature lower than that obtained in the experiments. Most of these correlations predicted experimental data well in the pseudo-critical region. However, all correlations are matching well with experimental data beyond the pseudo-critical region. The details of the experimental facility, Experiments carried out and the results presented in this report. (author)

  17. Factors governing particulate corrosion product adhesion to surfaces in water reactor coolant circuits

    International Nuclear Information System (INIS)

    1979-03-01

    Gravity, van der Waals, magnetic, electrical double layer and hydrodynamic forces are considered as potential contributors to the adhesion of particulate corrosion products to surfaces in water reactor coolant circuits. These forces are renewed and evaluated, and the following are amongst the conclusions drawn; adequate theories are available to estimate the forces governing corrosion product particle adhesion to surfaces in single phase flow in water reactor coolant circuits. Some uncertainty is introduced by the geometry of real particle-surface systems. The major uncertainties are due to inadequate data on the Hamaker constant and the zeta potential for the relevant materials, water chemistry and radiation chemistry at 300 0 C; van der Waals force is dominant over the effect of gravity for particles smaller than about 100 m; quite modest zeta potentials, approximately 50mV, are capable of inhibiting particle deposition throughout the size range relevant to water reactors; for surfaces exposed to typical water reactor flow conditions, particles smaller than approximately 1 m will be stable against resuspension in the absence of electrical double layer repulsion; and the magnitude of the electrical double layer repulsion for a given potential depends on whether the interaction is assumed to occur at constant potential or constant change. (author)

  18. Numerical investigation of flow instability in parallel channels with supercritical water

    International Nuclear Information System (INIS)

    Shitsi, Edward; Debrah, Seth Kofi; Agbodemegbe, Vincent Yao; Ampomah-Amoako, Emmanuel

    2017-01-01

    Highlights: •Supercritical flow instability in parallel channels is investigated. •Flow dynamics and heat transfer characteristics are analyzed. •Mass flow rate, pressure, heating power, and axial power shape have significant effects on flow instability. •Numerical results are validated with experimental results. -- Abstract: SCWR is one of the selected Gen IV reactors purposely for electricity generation in the near future. It is a promising technology with higher efficiency compared to current LWRs but without the challenges of heat transfer and its associated flow instability. Supercritical flow instability is mainly caused by sharp change in the coolant properties around the pseudo-critical point of the working fluid and research into this phenomenon is needed to address concerns of flow instability at supercritical pressures. Flow instability in parallel channels at supercritical pressures is investigated in this paper using a three dimensional (3D) numerical tool (STAR-CCM+). The dynamics characteristics such as amplitude and period of out-of-phase inlet mass flow oscillation at the heated channel inlet, and heat transfer characteristic such as maximum outlet temperature of the heated channel outlet temperature oscillation are discussed. Influences of system parameters such as axial power shape, pressure, mass flow rate, and gravity are discussed based on the obtained mass flow and temperature oscillations. The results show that the system parameters have significant effect on the amplitude of the mass flow oscillation and maximum temperature of the heated outlet temperature oscillation but have little effect on the period of the mass flow oscillation. The amplitude of mass flow oscillation and maximum temperature of the heated channel outlet temperature oscillation increase with heating power. The numerical results when compared to experiment data show that the 3D numerical tool (STAR-CCM+) could capture dynamics and heat transfer characteristics of

  19. Heat Transfer to Supercritical Water in Gaseous State or Affected by Mixed Convection in Vertical Tubes

    International Nuclear Information System (INIS)

    Pis'menny, E.N.; Razumovskiy, V.G.; Maevskiy, E.M.; Koloskov, A.E.; Pioro, I.L.

    2006-01-01

    The results on heat transfer to supercritical water heated above the pseudo-critical temperature or affected by mixed convection flowing upward and downward in vertical tubes of 6.28-mm and 9.50-mm inside diameter are presented. Supercritical water heat-transfer data were obtained at a pressure of 23.5 MPa, mass flux within the range from 250 to 2200 kg/(m 2 s), inlet temperature from 100 to 415 deg. C and heat flux up to 3.2 MW/m 2 . Temperature regimes of the tubes cooled with supercritical water in a gaseous state (i.e., supercritical water at temperatures beyond the pseudo-critical temperature) were stable and easily reproducible within a wide range of mass and heat fluxes. An analysis of the heat-transfer data for upward and downward flows enabled to determine a range of Gr/Re 2 values corresponding to the maximum effect of free convection on the heat transfer. It was shown that: 1) the heat transfer coefficient at the downward flow of water can be higher by about 50% compared to that of the upward flow; and 2) the deteriorated heat-transfer regime is affected with the flow direction, i.e., at the same operating conditions, the deteriorated heat transfer may be delayed at the downward flow compared to that at the upward flow. These heat-transfer data are applicable as the reference dataset for future comparison with bundle data. (authors)

  20. Delocalized organic pollutant destruction through a self-sustaining supercritical water oxidation process

    International Nuclear Information System (INIS)

    Lavric, E.D.; Weyten, H.; Ruyck, J. de; Plesu, V.; Lavric, V.

    2005-01-01

    Supercritical water oxidation (SCWO) is a recent development aiming at the destruction of organic pollutants present with low concentrations in waste waters. The present paper focuses on the process simulation of SCWO with emphasis on the proper modelling of supercritical thermodynamic conditions and on the possibility to make the SCWO process self-sufficient from the energetic viewpoint. Self-sufficiency may be of interest to encourage more delocalization of waste water treatment. The process of SCWO for dilute waste water (no more than 5 wt.%) is modelled through the ASPEN Plus copyright process simulator. Studies were made to search for energetic self-sufficiency conditions using various technologies for power production from the heat of reaction, like supercritical water expansion in a turbine, use of a closed Brayton cycle (CBC) and use of an organic Rankine cycle (ORC). The results obtained showed that the process is energetically self-sufficient using either a small supercritical turbine, or an ORC. In less restrictive conditions regarding the component efficiencies, the CBC, in theory, also leads to self-sufficiency, but from the analysis, it appears that this solution is less realistic

  1. Effect of coolant velocity on the fragmentation of single melt drops in water

    International Nuclear Information System (INIS)

    Cunningham, M.H.; Frost, D.L.

    1997-01-01

    Flash X-ray radiography and high-speed photography are used to investigate the effect of the coolant velocity on the fine fragmentation of molten tin drops in water. A water cannot is used to accelerate the water to a constant speed of up to 30 m/s. The water is accelerated with a double piston arrangement including a foam shock absorber to eliminate the formation of a shock wave. In this way, the effect of coolant velocity on drop breakup is investigated in the absence of the strong shock wave that is present in most earlier studies. The results show that there is a transition from thermal to hydrodynamic fragmentation through an intermediate stage in which the drops initially undergo hydrodynamic fragmentation followed by the formation of a vapour bubble. For low velocities (9 m/s) this bubble collapses, fragmenting the remainder of the drop while at greater velocities (15 m/s) the drop breaks up within the bubble before it condenses. At 22 and 28 m/s there is no vapour formation and the drop fragments due to hydrodynamic effects. Quantitative analysis of the radiographs is used to determine the mass distribution of the melt during the drop fragmentation. Comparison with earlier work in which the ambient flow is preceded by a strong shock wave indicates that the transition from thermal to hydrodynamic breakup is strongly dependent on the characteristics of the pressure field experienced by the drop. (author)

  2. Thermodynamic analysis of the use a chemical heat pump to link a supercritical water-cooled nuclear reactor and a thermochemical water-splitting cycle for hydrogen production

    International Nuclear Information System (INIS)

    Granovskii, Mikhail; Dincer, Ibrahim; Rosen, Marc A.; Pioro, Igor

    2008-01-01

    Increases in the power generation efficiency of nuclear power plants (NPPs) are mainly limited by the permissible temperatures in nuclear reactors and the corresponding temperatures and pressures of the coolants in reactors. Coolant parameters are limited by the corrosion rates of materials and nuclear-reactor safety constraints. The advanced construction materials for the next generation of CANDU reactors, which employ supercritical water (SCW) as a coolant and heat carrier, permit improved 'steam' parameters (outlet temperatures up to 625degC and pressures of about 25 MPa). An increase in the temperature of steam allows it to be utilized in thermochemical water splitting cycles to produce hydrogen. These methods are considered by many to be among the most efficient ways to produce hydrogen from water and to have advantages over traditional low-temperature water electrolysis. However, even lower temperature water splitting cycles (Cu-Cl, UT-3, etc.) require an intensive heat supply at temperatures higher than 550-600degC. A sufficient increase in the heat transfer from the nuclear reactor to a thermochemical water splitting cycle, without jeopardizing nuclear reactor safety, might be effectively achieved by application of a heat pump, which increases the temperature of the heat supplied by virtue of a cyclic process driven by mechanical or electrical work. Here, a high-temperature chemical heat pump, which employs the reversible catalytic methane conversion reaction, is proposed. The reaction shift from exothermic to endothermic and back is achieved by a change of the steam concentration in the reaction mixture. This heat pump, coupled with the second steam cycle of a SCW nuclear power generation plant on one side and a thermochemical water splitting cycle on the other, increases the temperature of the 'nuclear' heat and, consequently, the intensity of heat transfer into the water splitting cycle. A comparative preliminary thermodynamic analysis is conducted of

  3. Safety analysis of high temperature reactor cooled and moderated by supercritical light water

    International Nuclear Information System (INIS)

    Ishiwatari, Yuki; Oka, Yoshiaki; Koshizuka, Seiichi

    2003-01-01

    This paper describes 'Safety' of a high temperature supercritical light water cooled and moderated reactor (SCRLWR-H) with descending flow water rods. The safety system of the SCLWR-H is similar to that of a BWR. It consists of reactor scram, high pressure auxiliary feedwater system (AFS), low pressure core injection system (LPCI), safety relief valves (SRV), automatic depressurization system (ADS), and main steam isolation valves (MSIV). Ten types of transients and five types of accidents are analyzed using a plant transient analysis code SPRAT-DOWN. The sequences are determined referring to LWRs. At the 'Loss of load without turbine bypass' transient, the coolant density and the core power are increased by the over-pressurization, and at the same time the core flow rate is decreased by the closure of the turbine control valves. The peak cladding temperature increases to 727degC. The high temperature at this type of transient is one of the characteristics of the SCLWR-H. Conversely at 'feedwater-loss' events, the core power decrease to some extend by density feedback before the reactor scram. The peak cladding temperatures at the 'Partial loss of feedwater' transient and the 'Total loss of feedwater' accident are only 702degC and 833degC, respectively. The cladding temperature does not increase so much at the transients 'Loss of feedwater heating' and 'CR withdrawal' because of the operation of the plant control system. All the transients and accidents satisfy the satisfy criteria with good margins. The highest cladding temperatures of the transients and the accidents are 727degC and 833degC at the 'Loss of load without turbine bypass' and 'Total loss of feedwater', respectively. The duration of the high cladding temperature is very short at the transients. According to the parametric survey, the peak cladding temperature are sensitive to the parameters such as the pump coast-down time, delay of pump trip, AFS capacity, AFS delay, CR worth, and SRV setpoint

  4. Analysis of an ultrasonic level device for in-core Pressurized Water Reactor coolant detection

    International Nuclear Information System (INIS)

    Johnson, K.R.

    1981-01-01

    A rigorous semi-empirical approach was undertaken to model the response of an ultrasonic level device (ULD) for application to in-core coolant detection in Pressurized Water Reactors (PWRs). An equation is derived for the torsional wave velocity v/sub t phi/ in the ULD. Existing data reduction techniques were analyzed and compared to results from use of the derived equation. Both methods yield liquid level measurements with errors of approx. 5%. A sensitivity study on probe performance at reactor conditions predicts reduced level responsivity from data at lower temperatures

  5. Analysis of a small break loss-of-coolant accident of pressurized water reactor by APROS

    Energy Technology Data Exchange (ETDEWEB)

    Al-Falahi, A. [Helsinki Univ. of Technology, Espoo (Finland); Haennine, M. [VTT Energy, Espoo (Finland); Porkholm, K. [IVO International, Ltd., Vantaa (Finland)

    1995-09-01

    The purpose of this paper is to study the capability of APROS (Advanced PROcess Simulator) code to simulate the real plant thermal-hydraulic transient of a Small Break Loss-Of-Coolant Accident (SBLOCA) of Loss-Of-Fluid Test (LOFT) facility. The LOFT is a scaled model of a Pressurized Water Reactor (PWR). This work is a part of a larger validation of the APROS thermal-hydraulic models. The results of SBLOCA transient calculated by APROS showed a reasonable agreement with the measured data.

  6. Recommended reactor coolant water chemistry requirements for WWER-1000 units with 235U higher enriched fuel

    International Nuclear Information System (INIS)

    Dobrevski, I.; Zaharieva, N.

    2011-01-01

    The last decade worldwide experience of PWRs and WWERs confirms the trends for the improvement of the nuclear power industry electricity production through the implementation of high burn-up or high fuel duty, which are usually accompanied with the usage of UO 2 fuel with higher content of 235 U - 4.0% - 4.5% (5.0%). It was concluded that the onset of sub-cooled nucleate boiling (SNB) on the fuel cladding surfaces and the initial excess reactivity of the core are the primary and basic factors accompanying the implementation of uranium fuel with higher 235 U content, aiming extended fuel cycles and higher burn-up of the fuel in Pressurized Water Reactors. As main consequences of the presence of these factors the modifications of chemical / electrochemical environments of nuclear fuel cladding- and reactor coolant system- surfaces are evaluated. These conclusions are the reason for: 1) The determination of the choices of the type of fuel cladding materials in respect with their enough corrosion resistance to the specific fuel cladding environment, created by the presence of SNB; 2) The development and implementation of primary circuit water chemistry guidelines ensuring the necessary low corrosion rates of primary circuit materials and limitation of cladding deposition and out-of-core radioactivity buildup; 3) Implementation of additional neutron absorbers which allow enough decrease of the initial concentration of H 3 BO 3 in coolant, so that its neutralization will be possible with the permitted alkalising agent concentrations. In this paper the specific features of WWER-1000 units in Bulgarian Nuclear Power Plant; use of 235 U higher enriched fuel in the WWER-1000 reactors in the Kozloduy NPP; coolant water chemistry and radiochemistry plant data during the power operation period of the Kozloduy NPP Unit 5, 15 th fuel cycle; evaluation of the approaches and results by the conversion of the WWER-1000 Units at the Kozloduy NPP to the uranium fuel with 4.3% 235 U as

  7. Isoelectric focusing in continuously tapered fused silica capillary prepared by etching with supercritical water

    Czech Academy of Sciences Publication Activity Database

    Šlais, Karel; Horká, Marie; Karásek, Pavel; Planeta, Josef; Roth, Michal

    2013-01-01

    Roč. 85, č. 9 (2013), s. 4296-4300 ISSN 0003-2700 R&D Projects: GA ČR(CZ) GAP106/12/0522; GA MV VG20102015023 Institutional support: RVO:68081715 Keywords : capillary isoelectric focusing * resolution of ampholytes * supercritical water Subject RIV: CB - Analytical Chemistry, Separation Impact factor: 5.825, year: 2013

  8. Solubility of fused silica in sub- and supercritical water: Estimation from a thermodynamic model

    Czech Academy of Sciences Publication Activity Database

    Karásek, Pavel; Šťavíková, Lenka; Planeta, Josef; Hohnová, Barbora; Roth, Michal

    2013-01-01

    Roč. 83, NOV (2013), s. 72-77 ISSN 0896-8446 R&D Projects: GA ČR(CZ) GAP106/12/0522 Institutional support: RVO:68081715 Keywords : amorphous silica * fused silica * supercritical water * aqueous solubility Subject RIV: CB - Analytical Chemistry, Separation Impact factor: 2.571, year: 2013

  9. Sensitivity analysis of CFD code FLUENT-12 for supercritical water in vertical bare tubes

    Energy Technology Data Exchange (ETDEWEB)

    Farah, A.; Haines, P.; Harvel, G.; Pioro, I., E-mail: amjad.farah@yahoo.com, E-mail: patrickjhaines@gmail.com, E-mail: glenn.harvel@uoit.ca, E-mail: igor.pioro@uoit.ca [Univ. of Ontario Inst. of Technology, Faculty of Energy Systems and Nuclear Science,Oshawa, Ontario (Canada)

    2012-07-01

    The ability to use FLUENT 12 or other CFD software to accurately model supercritical water flow through various geometries in diabatic conditions is integral to research involving coal-fired power plants as well as Supercritical Water-cooled Reactors (SCWR). The cost and risk associated with constructing supercritical water test loops are far too great to use in a university setting. Previous work has shown that FLUENT 12, specifically realizable k-ε model, can reasonably predict the bulk and wall temperature distributions of externally heated vertical bare tubes for cases with relatively low heat and mass fluxes. However, sizeable errors were observed for other cases, often those which involved large heat fluxes that produce deteriorated heat transfer (DHT) regimes. The goal of this research is to gain a more complete understanding of how FLUENT 12 models supercritical water cases and where errors can be expected to occur. One control case is selected where expected changes in bulk and wall temperatures occur and they match empirical correlations' predictions, and the operating parameters are varied individually to gauge their effect on FLUENT's solution. The model used is the realizable k-ε, and the parameters altered are inlet pressure, mass flux, heat flux, and inlet temperature. (author)

  10. Heat transfer in a seven-rod test bundle with supercritical pressure water (1). Experiments

    International Nuclear Information System (INIS)

    Ezato, Koichiro; Seki, Yohji; Dairaku, Masayuki; Suzuki, Satoshi; Enoeda, Mikio; Akiba, Masato; Mori, H.; Oka, Y.

    2009-01-01

    Heat transfer experiments in a seven-rod test bundle with supercritical pressure water has been carried out. The pressure drop and heat transfer coefficients (HTCs) in the test section are evaluated. In the present limited conditions, difference between HTCs at the surface facing the sub-channel center and those at the surface in the narrowest region between rods is not observed. (author)

  11. Near- and supercritical water as a diameter manipulation and surface roughening agent in fused silica capillaries

    Czech Academy of Sciences Publication Activity Database

    Karásek, Pavel; Planeta, Josef; Roth, Michal

    2013-01-01

    Roč. 85, č. 1 (2013), s. 327-333 ISSN 0003-2700 R&D Projects: GA ČR(CZ) GAP106/12/0522; GA ČR(CZ) GAP206/11/0138 Institutional support: RVO:68081715 Keywords : supercritical water * fused silica capillary * surface treatment Subject RIV: CB - Analytical Chemistry, Separation Impact factor: 5.825, year: 2013

  12. Catalytic reforming of glycerol in supercritical water over bimetallic Pt-Ni catalyst

    NARCIS (Netherlands)

    Chakinala, A.G.; van Swaaij, Willibrordus Petrus Maria; Kersten, Sascha R.A.; de Vlieger, Dennis; Seshan, Kulathuiyer; Brilman, Derk Willem Frederik

    2013-01-01

    Catalytic reforming of pure glycerol for the production of hydrogen at low temperature and short residence times in supercritical water was investigated using a bimetallic Pt–Ni catalyst supported on alumina. Initial tests were carried out to study the reforming activity of bimetallic Pt–Ni

  13. An Energy Analysis on Gasification of Sewage Sludge by a Direct Injection in Supercritical Water

    NARCIS (Netherlands)

    Yukananto, Riza; Louwes, Alexander Charnchai; Bramer, Eduard A.; Brem, Gerrit

    2017-01-01

    Supercritical Water Gasification is an efficient technology in converting wet biomass into H2 and CH4 in comparison to other conventional thermochemical processes. Coke deposition, however, remains as a major challenge in this technology. Coke formation is the result of polymerization reactions that

  14. Supercritical water gasification of sewage sludge: gas production and phosphorus recovery

    NARCIS (Netherlands)

    Acelas Soto, N.Y.; Lopez, D.P.; Brilman, Derk Willem Frederik; Kersten, Sascha R.A.; Kootstra, A.M.J.

    2014-01-01

    In this study, the feasibility of the gasification of dewatered sewage sludge in supercritical water (SCW) for energy recovery combined with P-recovery from the solid residue generated in this process was investigated. SCWG temperature (400 °C, 500 °C, 600 °C) and residence time (15 min, 30 min, 60

  15. Investigation of the precipitation of Na2SO4 in supercritical water

    DEFF Research Database (Denmark)

    Voisin, T.; Erriguible, A.; Philippot, G.

    2017-01-01

    solubility in sub-and supercritical water is determined on a wide temperature range using a continuous set-up. Crystallite sizes formed after precipitation are measured with in situ synchrotron wide angle X-ray scattering (WAXS). Combining these experimental results, a numerical modeling of the precipitation......SuperCritical Water Oxidation process (SCWO) is a promising technology for treating toxic and/or complex chemical wastes with very good efficiency. Above its critical point (374 degrees C, 22.1 MPa), water exhibits particular properties and organic compounds can be easily dissolved and degraded...... with the addition of oxidizing agents. But these interesting properties imply a main drawback regarding inorganic compounds. Highly soluble at ambient temperature in water, these inorganics (such as salts) are no longer soluble in supercritical water and precipitate into solids, creating plugs in SCWO processes...

  16. HANARO secondary coolant management

    International Nuclear Information System (INIS)

    Kim, Seon Duk.

    1998-02-01

    In this report, the basic theory for management of water quality, environmental factors influencing to the coolant, chemicals and its usage for quality control of coolant are mentioned, and water balance including the loss rate by evaporation (34.3 m 3 /hr), discharge rate (12.665 m 3 /hr), concentration ratio and feed rate (54.1 m 3 /hr) are calculated at 20 MW operation. Also, the analysis data of HANSU Limited for HANARO secondary coolant (feed water and circulating coolant) - turbidity, pH, conductivity, M-alkalinity, Ca-hardness, chloride ion, total iron ion, phosphoric ion and conversion rate are reviewed. It is confirmed that the feed water has good quality and the circulating coolant has been maintained within the control specification in general, but some items exceeded the control specification occasionally. Therefore it is judged that more regular discharge of coolant is needed. (author). 6 refs., 17 tabs., 18 figs

  17. CFD validation of a supercritical water flow for SCWR design heat and mass fluxes

    International Nuclear Information System (INIS)

    Roelofs, F.; Lycklama a Nijeholt, J.A.; Komen, E.M.J.; Lowenberg, M.; Starflinger, J.

    2007-01-01

    The applicability of Computational Fluid Dynamics (CFD) for water under supercritical conditions in supercritical water reactors (SCWR) has still to be verified. In the recent past, CFD validation analyses were performed by various institutes for supercritical water in vertical tubes based on the well known experimental data from Yamagata. However, validation using data from experiments with working conditions closer to the actual operational conditions of such reactors is needed. From a literature survey the experiments performed by Herkenrath are selected to perform validation analyses at higher heat fluxes and a higher mass flux. The accuracy of CFD using RANS (Reynolds Average Navier-Stokes) turbulence modelling for supercritical fluids under conditions close to the operational conditions of a supercritical water reactor is determined. It is concluded that the wall temperature can be predicted by RANS CFD, using the RNG k-ε turbulence model, with accuracy in the range of 5% for heat fluxes up to 1100 kW/m 2 and for a bulk enthalpy up to 2200 kJ/kg. For a bulk enthalpy exceeding 2200 kJ/kg, a significant lower accuracy of the CFD predictions (about 3%) is found for the simulations of the experiments of Yamagata in comparison with the simulations of the experiments of Herkenrath. For these experiments, the accuracy is about 18 per cent. This might be a result of the fact that the CFD analyses do not simulate the flattening of the temperature profile at about 2200 kJ/kg which is found in the experiments of Herkenrath. However, the obtained accuracies ranging from 3% to 18% are still deemed to be acceptable for many design purposes. (authors)

  18. Return momentum effect on reactor coolant water level distribution during mid-loop conditions

    International Nuclear Information System (INIS)

    Seo, Jae Kwang; Yang, Jae Young; Park, Goon Cherl

    2001-01-01

    An accurate prediction of the Reactor Coolant System( RCS) water level is of importance in the determination of the allowable operating range to ensure safety during mid-loop operations. However, complex hydrualic phenomena induced by the Shutdown Cooling System (SCS) return momentum causes different water levels from those in the loop where the water level indicators are located. This was apparently observed at the pre-core cold hydro test of the Younggwang Nuclear Unit 3 (YGN 3) in Korea. In this study, in order to analytically understand the effect of the SCS return momentum on the RCS water level distribution, a model using a one-dimensional momentum and energy conservation for cylindrical channel, hydraulic jump in operating cold leg, water level build-up at the Reactor Vessel (RV) inlet nozzle, Bernoulli constant in downcomer region, and total water volume conservation has been developed. The model predicts the RCS water levels at various RCS locations during the mid-loop conditions and the calculation results were compared with the test data. The analysis shows that the hydraulic jump in the operating cold legs, in conjuction with the pressure drop throughout the RCS, is the main cause creating the water level differences at various RCS locations. The prediction results provide good explanations for the test data and show the significant effect of the SCS return momentum on the RCS water levels

  19. Nuclear data sensitivity and uncertainty for the Canadian supercritical water-cooled reactor II: Full core analysis

    International Nuclear Information System (INIS)

    Langton, S.E.; Buijs, A.; Pencer, J.

    2015-01-01

    Highlights: • H-2, Pu-239, and Th-232 make large contributions to SCWR modelling sensitivity. • H-2, Pu-239, and Th-232 make large contributions to SCWR modelling uncertainty. • Isotopes of Zr make large contributions to SCWR modelling uncertainty. - Abstract: Uncertainties in nuclear data are a fundamental source of uncertainty in reactor physics calculations. To determine their contribution to uncertainties in calculated reactor physics parameters, a nuclear data sensitivity and uncertainty study is performed on the Canadian supercritical water reactor (SCWR) concept. The nuclear data uncertainty contributions to the neutron multiplication factor k eff are 6.31 mk for the SCWR at the beginning of cycle (BOC) and 6.99 mk at the end of cycle (EOC). Both of these uncertainties have a statistical uncertainty of 0.02 mk. The nuclear data uncertainty contributions to Coolant Void Reactivity (CVR) are 1.0 mk and 0.9 mk for BOC and EOC, respectively, both with statistical uncertainties of 0.1 mk. The nuclear data uncertainty contributions to other reactivity parameters range from as low as 3% of to as high as ten times the values of the reactivity coefficients. The largest contributors to the uncertainties in the reactor physics parameters are Pu-239, Th-232, H-2, and isotopes of zirconium

  20. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    Reference is made to coolant channels for pressurised water and boiling water reactors and the arrangement described aims to improve heat transfer between the fuel rods and the coolant. Baffle means extending axially within the channel are provided and disposed relative to the fuel rods so as to restrict flow oscillations occurring within the coolant from being propagated transversely to the axis of the channel. (UK)

  1. Design and operational parameters of transportable supercritical water oxidation waste destruction unit

    International Nuclear Information System (INIS)

    McFarland, R.D.; Brewer, G.R.; Rofer, C.K.

    1991-12-01

    Supercritical water oxidation (SCWO) is the destruction of hazardous waste by oxidation in the presence of water at temperatures and pressures above its critical point. A 1 gal/h SCWO waste destruction unit (WDU) has been designed, built, and operated at Los Alamos National Laboratory. This unit is transportable and is intended to demonstrate the SCWO technology on wastes at Department of Energy sites. This report describes the design of the WDU and the preliminary testing phase leading to demonstration

  2. Loss of coolant analysis for CIRENE-LATINA heavy water reactor

    International Nuclear Information System (INIS)

    Chiantore, B.; Dubbini, M.; Proto, G.

    1978-01-01

    CIRENE is a heavy-water moderated, boiling water cooled pressure tube reactor. Fuel is natural uranium. A variety of breaks in the primary coolant system have been postulated for the analysis of the CIRENE Latina Plant (now under construction) such as double-end break of inlet header, downcomer, steam line and inlet feeders. The basic tool for analysis is the TILT-N Code which has been purposely developed for simulating the nuclear, thermal and hydrodynamic behaviour of the CIRENE core and associated heat transport system. An extensive full-scale test programme has been carried out by CNEN and CISE which fully confirms the adequacy of the model. The main results of the analysis show that maximum temperatures are far from those leading to significant fuel damage and that adequate core cooling is provided over the whole transient. (author)

  3. Fact and fiction in ECP measurement and control in boiling water reactor primary coolant circuits

    International Nuclear Information System (INIS)

    Macdonald, D.D.

    2005-01-01

    A review is presented of various electrochemical potentials, including the electrochemical corrosion potential (ECP), that are used in the mitigation of stress corrosion cracking in the primary coolant circuits of boiling water reactors (BWRs). Attention is paid to carefully defining each potential in terms of fundamental electrochemical concepts, so as to counter the confusion that has arisen due to the misuse of previously accepted terminology. A brief discussion is also included of reference electrodes and it is shown on the basis of experimental data that the use of a platinum redox sensor as a reference electrode in the monitoring of ECP in BWR primary coolant circuits is inappropriate and should be discouraged. If platinum is used as a reference electrode, because of extenuating circumstances (e.g., potential measurements in high dose regions in a reactor core), the onus must be placed on the user to demonstrate quantitatively that the electrode behaves as an equilibrium electrode under the specified conditions and/or that its potential is invariant with changes in the independent variables of the system. Preferably, a means should also be demonstrated of transferring the measured potential to the standard hydrogen electrode (SHE) scale. (orig.)

  4. Applications of subcritical and supercritical water conditions for extraction, hydrolysis, gasification, and carbonization of biomass: a critical review

    Directory of Open Access Journals (Sweden)

    D. Lachos-Perez

    2017-06-01

    Full Text Available This review summarizes the recent essential aspects of subcritical and supercritical water technology applied tothe extraction, hydrolysis, carbonization, and gasification processes. These are clean and fast technologies which do not need pretreatment, require less reaction time, generate less corrosion and residues, do not usetoxic solvents, and reduce the synthesis of degradation byproducts. The equipment design, process parameters, and types of biomass used for subcritical and supercritical water process are presented. The benefits of catalysis to improve process efficiency are addressed. Bioactive compounds, reducing sugars, hydrogen, biodiesel, and hydrothermal char are the final products of subcritical and supercritical water processes. The present review also revisits advances of the research trends in the development of subcriticaland supercritical water process technologies.

  5. Preliminary Study on the High Efficiency Supercritical Pressure Water-Cooled Reactor for Electricity Generation

    Energy Technology Data Exchange (ETDEWEB)

    Bae, Yoon Yeong; Park, Jong Kyun; Cho, Bong Hyun and others

    2006-01-15

    This research has been performed to introduce a concept of supercritical pressure water cooled reactor(SCWR) in Korea The area of research includes core conceptual design, evaluation of candidate fuel, fluid systems conceptual design with mechanical consideration, preparation of safety analysis code, and construction of supercritical pressure heat transfer test facility, SPHINX, and preliminary test. As a result of the research, a set of tools for the reactor core design has been developed and the conceptual core design with solid moderator was proposed. The direct thermodynamic cycle has been studied to find a optimum design. The safety analysis code has also been adapted to supercritical pressure condition. A supercritical pressure CO2 heat transfer test facility has been constructed and preliminary test proved the facility works as expected. The result of this project will be good basis for the participation in the international collaboration under GIF GEN-IV program and next 5-year mid and long term nuclear research program of MOST. The heat transfer test loop, SPHINX, completed as a result of this project may be used for the power cycle study as well as further heat transfer study for the various geometries.

  6. Surface chemistry and corrosion behavior of Inconel 625 and 718 in subcritical, supercritical, and ultrasupercritical water

    International Nuclear Information System (INIS)

    Rodriguez, David; Merwin, Augustus; Karmiol, Zachary; Chidambaram, Dev

    2017-01-01

    Highlights: • Mixtures of oxides containing Ni, Fe, Cr and Nb formed on the surface. • Short term exposure tests observed breakdown of native film. • Formation of a Fe rich oxide layer on Inconel 718 prevents mass loss. - Abstract: Corrosion behavior of Inconel 625 and 718 in subcritical, supercritical and ultrasupercritical water was studied as a function of temperature and time. The change in the chemistry of the as-received surface film on Inconel 625 and 718 after exposure to subcritical water at 325 °C and supercritical water at 425 °C and 527.5 °C for 2 h was studied. After exposure to 325 °C subcritical water, the CrO_4"2"− based film formed; however minor quantities of NiFe_xCr_2_-_xO_4 spinel compounds were observed. The oxide film formed on both alloys when exposed to supercritical water at 425 °C consisted of NiFe_xCr_2_-_xO_4 spinel. The surface films on both alloys were identified as NiFe_2O_4 when exposed to supercritical water at 527.5 °C. To characterize the fully developed oxide layer, studies were conducted at test solution temperatures of 527.5 and 600 °C. Samples were exposed to these temperatures for 24, 96, and 200 h. Surface chemistry was analyzed using X-ray diffraction, as well as Raman and X-ray photoelectron spectroscopies. Inconel 718 exhibited greater mass gain than Inconel 625 for all temperatures and exposure times. The differences in corrosion behavior of the two alloys are attributed to the lower content of chromium and increased iron content of Inconel 718 as compared to Inconel 625.

  7. Surface chemistry and corrosion behavior of Inconel 625 and 718 in subcritical, supercritical, and ultrasupercritical water

    Science.gov (United States)

    Rodriguez, David; Merwin, Augustus; Karmiol, Zachary; Chidambaram, Dev

    2017-05-01

    Corrosion behavior of Inconel 625 and 718 in subcritical, supercritical and ultrasupercritical water was studied as a function of temperature and time. The change in the chemistry of the as-received surface film on Inconel 625 and 718 after exposure to subcritical water at 325 °C and supercritical water at 425 °C and 527.5 °C for 2 h was studied. After exposure to 325 °C subcritical water, the CrO42- based film formed; however minor quantities of NiFexCr2-xO4 spinel compounds were observed. The oxide film formed on both alloys when exposed to supercritical water at 425 °C consisted of NiFexCr2-xO4 spinel. The surface films on both alloys were identified as NiFe2O4 when exposed to supercritical water at 527.5 °C. To characterize the fully developed oxide layer, studies were conducted at test solution temperatures of 527.5 and 600 °C. Samples were exposed to these temperatures for 24, 96, and 200 h. Surface chemistry was analyzed using X-ray diffraction, as well as Raman and X-ray photoelectron spectroscopies. Inconel 718 exhibited greater mass gain than Inconel 625 for all temperatures and exposure times. The differences in corrosion behavior of the two alloys are attributed to the lower content of chromium and increased iron content of Inconel 718 as compared to Inconel 625.

  8. Surface chemistry and corrosion behavior of Inconel 625 and 718 in subcritical, supercritical, and ultrasupercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Rodriguez, David; Merwin, Augustus; Karmiol, Zachary; Chidambaram, Dev, E-mail: dcc@unr.edu

    2017-05-15

    Highlights: • Mixtures of oxides containing Ni, Fe, Cr and Nb formed on the surface. • Short term exposure tests observed breakdown of native film. • Formation of a Fe rich oxide layer on Inconel 718 prevents mass loss. - Abstract: Corrosion behavior of Inconel 625 and 718 in subcritical, supercritical and ultrasupercritical water was studied as a function of temperature and time. The change in the chemistry of the as-received surface film on Inconel 625 and 718 after exposure to subcritical water at 325 °C and supercritical water at 425 °C and 527.5 °C for 2 h was studied. After exposure to 325 °C subcritical water, the CrO{sub 4}{sup 2−} based film formed; however minor quantities of NiFe{sub x}Cr{sub 2-x}O{sub 4} spinel compounds were observed. The oxide film formed on both alloys when exposed to supercritical water at 425 °C consisted of NiFe{sub x}Cr{sub 2-x}O{sub 4} spinel. The surface films on both alloys were identified as NiFe{sub 2}O{sub 4} when exposed to supercritical water at 527.5 °C. To characterize the fully developed oxide layer, studies were conducted at test solution temperatures of 527.5 and 600 °C. Samples were exposed to these temperatures for 24, 96, and 200 h. Surface chemistry was analyzed using X-ray diffraction, as well as Raman and X-ray photoelectron spectroscopies. Inconel 718 exhibited greater mass gain than Inconel 625 for all temperatures and exposure times. The differences in corrosion behavior of the two alloys are attributed to the lower content of chromium and increased iron content of Inconel 718 as compared to Inconel 625.

  9. Effect of ionite decomposition products on the reactor coolant pH in a boiling-water reactor

    International Nuclear Information System (INIS)

    Bredikhin, V.Ya.; Moskvin, L.N.

    1982-01-01

    The effect of products resulting from thermal radiolysis of ionites on water-chemical regime of NPP with RBMK is considered basing on investigations conducted in a boiling type experimental reactor. Data are presented on dynamics of changes in the specific electric conductivity and pH of the coolant following destruction of ion exchange groups and ionite matrix under the effect of reactor radiation. The authors draw a conclusion that radiation destruction of ionito fine disperse suspension or high-molecular soluble compounds in the reactor are, probably, one of the main reasons for variations in pH values of the coolant at NPP in non-correction water chemical regime

  10. Supercritical Water Mixture (SCWM) Experiment in the High Temperature Insert-Reflight (HTI-R)

    Science.gov (United States)

    Hicks, Michael C.; Hegde, Uday G.; Garrabos, Yves; Lecoutre, Carole; Zappoli, Bernard

    2013-01-01

    Current research on supercritical water processes on board the International Space Station (ISS) focuses on salt precipitation and transport in a test cell designed for supercritical water. This study, known as the Supercritical Water Mixture Experiment (SCWM) serves as a precursor experiment for developing a better understanding of inorganic salt precipitation and transport during supercritical water oxidation (SCWO) processes for the eventual application of this technology for waste management and resource reclamation in microgravity conditions. During typical SCWO reactions any inorganic salts present in the reactant stream will precipitate and begin to coat reactor surfaces and control mechanisms (e.g., valves) often severely impacting the systems performance. The SCWM experiment employs a Sample Cell Unit (SCU) filled with an aqueous solution of Na2SO4 0.5-w at the critical density and uses a refurbished High Temperature Insert, which was used in an earlier ISS experiment designed to study pure water at near-critical conditions. The insert, designated as the HTI-Reflight (HTI-R) will be deployed in the DECLIC (Device for the Study of Critical Liquids and Crystallization) Facility on the International Space Station (ISS). Objectives of the study include measurement of the shift in critical temperature due to the presence of the inorganic salt, assessment of the predominant mode of precipitation (i.e., heterogeneously on SCU surfaces or homogeneously in the bulk fluid), determination of the salt morphology including size and shapes of particulate clusters, and the determination of the dominant mode of transport of salt particles in the presence of an imposed temperature gradient. Initial results from the ISS experiments will be presented and compared to findings from laboratory experiments on the ground.

  11. Stress corrosion cracking behavior of annealed and cold worked 316L stainless steel in supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Sáez-Maderuelo, A., E-mail: alberto.saez@ciemat.es; Gómez-Briceño, D.

    2016-10-15

    Highlights: • The alloy 316L is susceptible to stress corrosion cracking in supercritical water. • The susceptibility of alloy 316L increases with temperature and plastic deformation. • Dynamic strain ageing processes may be active in the material. - Abstract: The supercritical water reactor (SCWR) is one of the more promising designs considered by the Generation IV International Forum due to its high thermal efficiency and improving security. To build this reactor, standardized structural materials used in light water reactors (LWR), like austenitic stainless steels, have been proposed. These kind of materials have shown an optimum behavior to stress corrosion cracking (SCC) under LWR conditions except when they are cold worked. It is known that physicochemical properties of water change sharply with pressure and temperature inside of the supercritical region. Owing to this situation, there are several doubts about the behavior of candidate materials like austenitic stainless steel 316L to SCC in the SCWR conditions. In this work, alloy 316L was studied in deaerated SCW at two different temperatures (400 °C and 500 °C) and at 25 MPa in order to determine how changes in this variable influence the resistance of this material to SCC. The influence of plastic deformation in the behavior of alloy 316L to SCC in SCW was also studied at both temperatures. Results obtained from these tests have shown that alloy 316L is susceptible to SCC in supercritical water reactor conditions where the susceptibility of this alloy increases with temperature. Moreover, prior plastic deformation of 316L SS increased its susceptibility to environmental cracking in SCW.

  12. Direct potentiometric control of chloride-ion content in water coolant of nuclear reactors

    International Nuclear Information System (INIS)

    Moskvin, L.N.; Vilkov, N.Ya.; Krasnoperov, V.M.; Epimakhova, L.V.

    1979-01-01

    The work of automatic chloride measuring device designed for continuous determination of chloride-ion concentration in water coolants of nuclear power plants is investigated. A series of experiments have been performed to investigate a device with sensitive element in the form of potentiometric cell with two flowing porous metal silver electrodes (PSE), placed in series. A calibration circuit of chloride measuring devices and PSE is described. A comparison is made between the results obtained by means of automatic chloride measuring device and results of manual control of samples. A conclusion is drawn that automatic chloride measuring devices meet the requirements of nuclear power plants for methods and instruments of control of chloride-ions microconcentration. The development and implantation of automatic chloride-ion analizers will make the analytical control on nuclear power plants easier and make it possible to obtain more reliable information

  13. Study on severe accident induced by large break loss of coolant accident for pressureized water reactor

    International Nuclear Information System (INIS)

    Zhang Longfei; Zhang Dafa; Wang Shaoming

    2007-01-01

    Using the best estimate computer code SCDAP/RELAP5/MOD3.2 and taking US Westinghouse corporation Surry nuclear power plant as the reference object, a typical three-loop pressurized water reactor severe accident calculation model was established and 25 cm large break loss of coolant accident (LBLOCA) in cold and hot leg of primary loop induced core melt accident was analyzed. The calculated results show that core melt progression is fast and most of the core material melt and relocated to the lower plenum. The lower head of reactor pressure vessel failed at an early time and the cold leg break is more severe than the hot leg break in primary loop during LBLOCA. (authors)

  14. Modeling of biomass to hydrogen via the supercritical water pyrolysis process

    Energy Technology Data Exchange (ETDEWEB)

    Divilio, R.J. [Combustion Systems Inc., Silver Spring, MD (United States)

    1998-08-01

    A heat transfer model has been developed to predict the temperature profile inside the University of Hawaii`s Supercritical Water Reactor. A series of heat transfer tests were conducted on the University of Hawaii`s apparatus to calibrate the model. Results of the model simulations are shown for several of the heat transfer tests. Tests with corn starch and wood pastes indicated that there are substantial differences between the thermal properties of the paste compared to pure water, particularly near the pseudo critical temperature. The assumption of constant thermal diffusivity in the temperature range of 250 to 450 C gave a reasonable prediction of the reactor temperatures when paste is being fed. A literature review is presented for pyrolysis of biomass in water at elevated temperatures up to the supercritical range. Based on this review, a global reaction mechanism is proposed. Equilibrium calculations were performed on the test results from the University of Hawaii`s Supercritical Water Reactor when corn starch and corn starch and wood pastes were being fed. The calculations indicate that the data from the reactor falls both below and above the equilibrium hydrogen concentrations depending on test conditions. The data also indicates that faster heating rates may be beneficial to the hydrogen yield. Equilibrium calculations were also performed to examine the impact of wood concentration on the gas mixtures produced. This calculation showed that increasing wood concentrations favors the formation of methane at the expense of hydrogen.

  15. Ion exchange resins destruction in a stirred supercritical water oxidation reactor

    International Nuclear Information System (INIS)

    Leybros, A.; Roubaud, A.; Guichardon, P.; Boutin, O.

    2010-01-01

    Spent ion exchange resins (IERs) are radioactive process wastes for which there is no satisfactory industrial treatment. Supercritical water oxidation offers a viable treatment alternative to destroy the organic structure of resins, used to remove radioactivity. Up to now, studies carried out in supercritical water for IER destruction showed that degradation rates higher than 99% are difficult to obtain even using a catalyst or a large oxidant excess. In this study, a co-fuel, isopropanol, has been used in order to improve degradation rates by initiating the oxidation reaction and increasing temperature of the reaction medium. Concentrations up to 20 wt% were tested for anionic and cationic resins. Total organic carbon reduction rates higher than 99% were obtained from this process, without the use of a catalyst. The influence of operating parameters such as IERs feed concentration, nature and counterions of exchanged IERs were also studied. (authors)

  16. Direct Conversion of Cellulose into Ethyl Lactate in Supercritical Ethanol-Water Solutions.

    Science.gov (United States)

    Yang, Lisha; Yang, Xiaokun; Tian, Elli; Lin, Hongfei

    2016-01-08

    Biomass-derived ethyl lactate is a green solvent with a growing market as the replacement for petroleum-derived toxic organic solvents. Here we report, for the first time, the production of ethyl lactate directly from cellulose with the mesoporous Zr-SBA-15 silicate catalyst in a supercritical mixture of ethanol and water. The relatively strong Lewis and weak Brønsted acid sites on the catalyst, as well as the surface hydrophobicity, were beneficial to the reaction and led to synergy during consecutive reactions, such as depolymerization, retro-aldol condensation, and esterification. Under the optimum reaction conditions, ∼33 % yield of ethyl lactate was produced from cellulose with the Zr-SBA-15 catalyst at 260 °C in supercritical 95:5 (w/w) ethanol/water. © 2016 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  17. Optimization of the fuel assembly for the Canadian Supercritical Water-cooled Reactor (SCWR)

    Energy Technology Data Exchange (ETDEWEB)

    French, C.; Bonin, H.; Chan, P., E-mail: Corey.French@rmc.ca [Royal Military College of Canada, Dept. of Chemistry and Chemical Engineering, Kingston, Ontario (Canada)

    2013-07-01

    A parametric optimization of the Canadian Supercritical Water-cooled Reactor (SCWR) lattice geometry and fresh fuel content is performed in this work. With the potential to improve core physics and performance, significant gains to operating and safety margins could be achieved through slight progressions. The fuel performance codes WIMS-AECL and SERPENT are used to calculate performance factors, and use them as inputs to an optimization algorithm. (author)

  18. Supercritical water-cooled reactor fuel management and economic comparison and analysis

    International Nuclear Information System (INIS)

    Cai Guangming; Ruan Liangcheng; Liu Xuechun

    2014-01-01

    The supercritical water-cooled reactor (SCWR) is expected to have an excellent fuel economical efficiency because of its high thermal efficiency. This article compares CSR1OOO with the current mainstream PWR and ABWR on the aspect of the economical efficiency of fuel management, and finally makes an unexpected conclusion that the SCWR has worse fuel economy than others. And it remains to be deliberated whether the SCWR will be the fourth generation of nuclear system. (authors)

  19. Oxidization and stress corrosion cracking initiation of austenitic alloys in supercritical water

    International Nuclear Information System (INIS)

    Behnamian, Y.; Li, M.; Luo, J.L.; Chen, W.X.; Zheng, W.; Guzonas, D.A.

    2012-01-01

    This study determined the stress corrosion cracking behaviour of austenitic alloys in pure supercritical water. Austenitic stainless steels 310S, 316L, and Inconel 625 were tested as static capsule samples at 500 o C for up to 5000 h. After that period, crack initiations were readily observed in all samples, signifying susceptibility to stress corrosion cracking. The microcracks in 316L stainless steel and Inconel 625 were almost intergranular, whereas transgranular microcrack initiation was observed in 310S stainless steel. (author)

  20. Supercritical water-treated fused silica capillaries in analytical separations: Status review

    Czech Academy of Sciences Publication Activity Database

    Karásek, Pavel; Horká, Marie; Šlais, Karel; Planeta, Josef; Roth, Michal

    2018-01-01

    Roč. 1539, MAR (2018), s. 1-11 ISSN 0021-9673 R&D Projects: GA MV VI20172020069; GA ČR(CZ) GA16-03749S; GA MZd(CZ) NV16-29916A Institutional support: RVO:68081715 Keywords : supercritical water * fused silica capillary * surface treatment Subject RIV: CB - Analytical Chemistry, Separation OBOR OECD: Analytical chemistry Impact factor: 3.981, year: 2016

  1. Corrosion fatigue studies on F82H mod. martensitic steel in reducing water coolant environments

    Energy Technology Data Exchange (ETDEWEB)

    Maday, M F; Masci, A [ENEA, Casaccia (Italy). Centro Ricerche Energia

    1998-03-01

    Load-controlled low cycle fatigue tests have been carried out on F82H martensitic steel in 240degC oxygen-free water with and without dissolved hydrogen, in order to simulate realistic coolant boundary conditions to be approached in DEMO. It was found that water independently of its hydrogen content, determined the same fatigue life reduction compared to the base-line air results. Water cracks exhibited in their first propagation stages similar fracture morphologies which were completely missing on the air cracks, and were attributed to the action of an environment related component. Lowering frequency gave rise to an increase in F82H fatigue lifetimes without any change in cracking mode in air, and to fatigue life reduction by microvoid coalescence alone in water. The data were discussed in terms of (i) frequency dependent concurrent processes for crack initiation and (ii) frequency-dependent competitive mechanisms for crack propagation induced by cathodic hydrogen from F82H corrosion. (author)

  2. Study of thermal hydraulic behavior of supercritical water flowing through fuel rod bundles

    International Nuclear Information System (INIS)

    Thakre, Sachin; Lakshmanan, S.P.; Kulkarni, Vinayak; Pandey, Manmohan

    2009-01-01

    Investigations on thermal-hydraulic behavior in Supercritical Water Reactor (SCWR) fuel assembly have obtained a significant attention in the international SCWR community because of its potential to obtain high thermal efficiency and compact design. Present work deals with CFD analysis to study the flow and heat transfer behavior of supercritical water in 4 metre long 7-pin fuel bundle using commercial CFD package ANSYS CFX for single phase steady state conditions. Considering the symmetric conditions, 1/12th part of the fuel rod bundle is taken as a domain of analysis. RNG K-epsilon model with scalable wall functions is used for modeling the turbulence behavior. Constant heat flux boundary condition is applied at the fuel rod surface. IAPWS equations of state are used to compute thermo-physical properties of supercritical water. Sharp variations in its thermo-physical properties (specific heat, density) are observed near the pseudo-critical temperature causing sharp change in heat transfer coefficient. The pseudo-critical point initially appears in the gaps among heated fuel rods, and then spreads radially outward reaching the adiabatic wall as the flow goes downstream. The enthalpy gain in the centre of the channel is much higher than that in the wall region. Non-uniformity in the circumferential distribution of surface temperature and heat transfer coefficient is observed which is in agreement with published literature. Heat transfer coefficient is high on the rod surface near the tight region and decreases as the distance between rod surfaces increases. (author)

  3. Challenges of selecting materials for the process of biomass gasification in supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Boukis, N.; Habicht, W.; Hauer, E.; Dinjus, E. [Karlsruher Institut fuer Technologie (KIT), Karlsruhe (Germany). Inst. fuer Technische Chemie

    2010-07-01

    A new process for the gasification of wet biomass is the reaction in supercritical water. The product is a combustible gas, rich in hydrogen with a high calorific value. The reaction is performed under high temperatures - up to 700 C - and pressures up to 30 MPa. The combination of these physical conditions and the corrosive environment is very demanding for the construction materials of the reactor. Only few alloys exhibit the required mechanical properties, especially the mechanical strength at temperatures higher than 600 C. Ni-Base alloys like alloy 625 can be applied up to a temperature of 700 C and are common materials for application under supercritical water conditions. During gasification experiments with corn silage and other biomasses, corrosion of the reactor material alloy 625 appears. The gasification of an aqueous methanol solution in supercritical water at temperatures up to 600 C and 25 - 30 MPa pressure results in an product gas rich in hydrogen, carbon dioxide and some methane. Alloy 625 shows very low corrosion rates in this environment. It is obvious that the heteroatoms and salts present in biomass cause corrosion of the reactor material. (orig.)

  4. Formation mechanism and luminescence appearance of Mn-doped zinc silicate particles synthesized in supercritical water

    International Nuclear Information System (INIS)

    Takesue, Masafumi; Suino, Atsuko; Hakuta, Yukiya; Hayashi, Hiromichi; Smith, Richard Lee

    2008-01-01

    Luminescence appearance of Mn-doped zinc silicate (Zn 2 SiO 4 :Mn 2+ , ZSM) formed in supercritical water at 400 deg. C and 29 MPa at reaction times from 1 to 4320 min was studied in the relation to its phase formation mechanism. Appearance of luminescent ZSM from green emission by α-ZSM and yellow emission by β-ZSM occurred over the same time period during the onset of phase formation at a reaction time of 2 min. Luminescence appeared at a much lower temperature and at shorter reaction times than the conventional solid-state reaction. Needle-like-shaped α-ZSM was the most stable particle shape and phase in the supercritical water reaction environment and particles formed via two routes: a homogenous nucleation route and a heterogenous route that involves solid-state diffusion and recrystallization. - Graphical abstract: Luminescence appearance of Mn-doped zinc silicate (Zn 2 SiO 4 :Mn 2+ , ZSM) formed in supercritical water at 400 deg. C and 29 MPa were studied in the relation to its phase formation mechanism. Green emission by α-ZSM and yellow emission by β-ZSM occurred over the same time period during the onset of phase formation

  5. Control-rod, pressure and flow-induced accident and transient analysis of a direct-cycle, supercritical-pressure, light-water-cooled fast breeder reactor

    International Nuclear Information System (INIS)

    Kitoh, Kazuaki; Koshizuka, Seiichi; Oka, Yoshiaki

    1996-01-01

    The features of the direct-cycle, supercritical-pressure, light-water-cooled fast breeder reactor (SCFBR) are high thermal efficiency and simple reactor system. The safety principle is basically the same as that of an LWR since it is a water-cooled reactor. Maintaining the core flow is the basic safety requirement of the reactor, since its coolant system is the one through type. The transient behaviors at control rod, pressure and flow-induced abnormalities are analyzed and presented in this paper. The results of flow-induced transients of SCFBR were reported at ICONE-3, though pressure change was neglected. The change of fuel temperature distribution is also considered for the analysis of the rapid reactivity-induced transients such as control rod withdrawal. Total loss of flow and pump seizure are analyzed as the accidents. Loss of load, control rod withdrawal from the normal operation, loss of feedwater heating, inadvertent start of an auxiliary feedwater pump, partial loss of coolant flow and loss of external power are analyzed as the transients. The behavior of the flow-induced transients is not so much different from the analyses assuming constant pressure. Fly wheels should be equipped with the feedwater pumps to prolong the coast-down time more than 10s and to cope with the total loss of flow accident. The coolant density coefficient of the SCFBR is less than one tenth of a BWR in which the recirculation flow is used for the power control. The over pressurization transients at the loss of load is not so severe as that of a BWR. The power reaches 120%. The minimum deterioration heat flux ratio (MDHFR) and the maximum pressure are sufficiently lower than the criteria; MDHFR above 1.0 and pressure ratio below 1.10 of 27.5 MPa, maximum pressure for operation. Among the reactivity abnormalities, the control rod withdrawal transient from the normal operation is analyzed

  6. A study of the tritium behavior in coolant and moderator system of heavy water reactors

    Energy Technology Data Exchange (ETDEWEB)

    Kim, S. P.; Song, S. S.; Chae, K. S. and others [Chosun Univ., Gwangju (Korea, Republic of)

    1993-12-15

    The objectives of this report is to present a regulatory policy on the environmental impact and personnel exposure by understanding the generation, accumulation, environmental release and management of tritium in heavy water reactors. By estimating the tritium concentration at Wolsong nuclear plant site by estimating and forecasting the generation and accumulation of tritium in coolant and moderator systems at Wolsong unit 1, we will study the management and release of tritium at Wolsong units 3 and 4 which are ready for construction. The major activities of this study are as follows : tritium generation and accumulation in heavy water reactor, a quantitative assessment of the accumulation and release of tritium at Wolsong nuclear plant site, heavy water management at Wolsong nuclear plants. The tritium concentration and accumulation trends in the systems at Wolsong unit 1 was estimated. A quantitative assessment of the tritium accumulation and release for Wolsong 2, 3 and 4 based on data from Wolsong 1 was performed. The tritium removal schemes and its long-term management plan were made.

  7. Development of in-situ laser cutting technique for removal of single selected coolant channel from pressurized heavy water reactor

    International Nuclear Information System (INIS)

    Vishwakarma, S.C.; Upadhyaya, B.N.

    2016-01-01

    We report on the development of a pulsed Nd:YAG laser based cutting technique for removal of single coolant channel from pressurized heavy water reactor (PHWR). It includes development of special tools/manipulators and optimization of laser cutting process parameters for cutting of liner tube, end fitting, bellow lip weld joint, and pressure tube stubs. For each cutting operation, a special tool with precision motion control is utilized. These manipulators/tools hold and move the laser cutting nozzle in the required manner and are fixed on the same coolant channel, which has to be removed. This laser cutting technique has been successfully deployed for removal of selected coolant channels Q-16, Q-15 and N-6 of KAPS-2 reactor with minimum radiation dose consumption and in short time. (author)

  8. The research of materials and water chemistry for supercritical water-cooled reactors in Research Centre Rez

    International Nuclear Information System (INIS)

    Zychova, Marketa; Fukac, Rostislav; Vsolak, Rudolf; Vojacek, Ales; Ruzickova, Mariana; Vonkova, Katerina

    2012-09-01

    Research Centre Rez (CVR) is R and D company based in the Czech Republic. It was established as the subsidiary of the Nuclear Research Institute Rez plc. One of the main activities of CVR is the research of materials and chemistry for the generation IV reactor systems - especially the supercritical water-cooled one. For these experiments is CVR equipped by a supercritical water loop (SCWL) and a supercritical water autoclave (SCWA) serving for research of material and Supercritical Water-cooled Reactor (SCWR) environment compatibility experiments. SCWL is a research facility designed to material, water chemistry, radiolysis and other testing in SCWR environment, SCWA serves for complementary and supporting experiments. SCWL consists of auxiliary circuits (ensuring the required parameters as temperature, pressure and chemical conditions in the irradiation channel, purification and measurements) and irradiation channel (where specimens are exposed to the SCWR environment). The design of the loop is based on many years of experience with loop design for various types of corrosion/water chemistry experiments. Designed conditions in the test area of SCWL are 600 deg. C and 25 MPa. SCWL was designed in 2008 within the High Performance Light Water Reactor Phase 2 project and built during 2008 and 2009. The trial operations were performed in 2010 and 2011 and were divided into three phases - the first phase to verify the functionality of auxiliary circuits of the loop, the second phase to verify the complete facility (auxiliary circuits and functional irradiation channel internals) and the third phase to verify the feasibility of corrosion tests with the complete equipment and specimens. All three trial operations were very successful - designed conditions and parameters were reached. (authors)

  9. Improving Safety, Economic, Substantiality, and Security of Nuclear Energy with Canadian Super-Critical Water-cooled Reactor Concept

    International Nuclear Information System (INIS)

    Hamilton, Holly; Pencer, Jeremy; Yetisir, Metin; Leung, Laurence

    2012-01-01

    Super-Critical Water-cooled Reactor is one of the six design concepts being developed under the Generation IV International Forum. It is the only concept evolving from the water-cooled reactors and taking advantages of the balance-of-plant design and operation experience of the fossil-power plants. Canada is developing the SCR concept from the well-established pressure-tube reactor technology. The Canadian SCWR maintains modular design approach using relative small fuel channels with the separation of coolant and moderator. It is equipped with an advanced fuel channel design that is capable to transfer decay heat from the fuel to the moderator under the long-term cooling stage. Coupled with the advanced passive-moderator cooling system, cooling of fuel and fuel channel is continuous even without external power or operator intervention. The Canadian SCWR is operating at a pressure of 25 MPa with a core outlet temperature of 625 deg. C. This has led to a drastic increase in thermal efficiency to 48% from 34% of the current fleet of reactors (a 40% rise in relative efficiency). With the high core outlet temperature, a direct thermal cycle has been adopted and has led to simplification in plant design attributing to the cost reduction compared to the current reactor designs. The Canadian SCWR adopts the advanced Thorium fuel cycle to enhance the substantiality, economic, and security. than uranium in the world (estimated to be three times more). This provides the long-term fuel supply. Thorium's price is stable compared to uranium and is consistently lower than uranium. This would maintain the predictability and economic of fuel supply. Thorium itself is a non-fissile material and once irradiated requires special handling. This improves proliferative resistance. The objective of this paper is to highlight these improvements in generating nuclear energy with the Canadian SCWR

  10. FY1995 generic supercritical water technology; 1995 nendo generic technology to shite no chorinkai riyo gijutsu

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-03-01

    For the establishment of the basis of supercritical fluid technology, we perform elucidation of the specific feature of the supercritical fluid as a reaction media and development of some new process. In this study, we first studied the fluid structure of SCF through in-situ spectroscopy and MD simulation. As a result, significant hydrogen bonding amongst water molecules and a solvation structure around the solute were observed in the supercritical state. This fluid structure has new features different from that of high temperature steam or liquid water. We found that this is closely related to the difference of bulk properties of SCF and local one around the solute. On the basis of these fundamental findings and with the better understanding of the specific features of SCF as a reaction media, development of some new process had been conducted more efficiently and successfully. The processes being developed in this study include 1) waste biomass and plastic conversion to recover chemicals, 2) hydrogenation of heavy oil for desulphurization through partial oxidation 1 and 3) hydrothermal synthesis of metal oxide fine particles. (NEDO)

  11. Identification of significant process variables for a flow-through supercritical water oxidation reactor

    International Nuclear Information System (INIS)

    Rossi, R.E.

    1992-05-01

    The effects of four process variables on the destruction efficiency of a flow-through supercritical water oxidation reactor were investigated. These process variables included: (1) reactor throughput (GPH), (2) concentration of the surrogate waste (% acetone), (3) maximum reactor tube-wall temperature (OC), and (4) applied stoichiometric oxygen. The analysis was conducted utilizing two-level factorial experiments, steepest ascent methods, and central composite designs. This experimental protocol assures efficient experimentation and allows for an empirical response surface model of the system to be developed. This experimentation identified a significant positive effect for stoichiometric oxygen applied and temperature variations between 400 to 500 degrees C. The increase in destruction efficiency due to stoichiometric 0 2 provides strong evidence that supercritical water oxidations are catalyzed by excess oxygen, and the strong temperature effect is a result of large increases in the kinetic rates for this temperature range. However, increasing temperature between 550 to 650 degrees C does not provide substantial increases in destruction efficiency. In addition, destruction efficiency is significantly unproved by increasing the Reynolds number and residence time. The destruction efficiency of the reactor is also dependent upon the initial concentration of surrogate waste. This concentration dependence may indicate first-order supercritical CO kinetics is inadequate for describing all waste types and reactor configurations. Alternatively, it may indicate reactant mixing, caused by local turbulence at the oxidation fronts of these higher concentration waste streams, results in higher destruction efficiencies

  12. Determination of fat- and water-soluble vitamins by supercritical fluid chromatography: A review.

    Science.gov (United States)

    Tyśkiewicz, Katarzyna; Dębczak, Agnieszka; Gieysztor, Roman; Szymczak, Tomasz; Rój, Edward

    2018-01-01

    Vitamins are compounds that take part in all basic functions of an organism but also are subject of number of studies performed by different researchers. Two groups of vitamins are distinguished taking into consideration their solubility. Chromatography with supercritical CO 2 has found application in the determination, separation, and quantitative analyses of both fat- and water-soluble vitamins. The methods of vitamins separation have developed and improved throughout the years. Both groups of compounds were separated using supercritical fluid chromatography with different detection on different stationary phases. The main aim of this review is to provide an overview of the studies of vitamins separation that have been determined so far. © 2017 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.

  13. Model for cobalt 60/58 deposition on primary coolant piping in a boiling water reactor

    International Nuclear Information System (INIS)

    Dehollander, W.R.

    1979-01-01

    A first principles model for deposition of radioactive metals into the corrosion films of primary coolant piping is proposed. It is shown that the predominant mechanism is the inclusion of the radioactive species such as Cobalt 60 into the spinel structure of the corrosion film during the act of active corrosion. This deposition can occupy only a defined fraction of the available plus 2 valence sites of the spinel. For cobalt ions, this ratio is roughly 4.6 x 10 -3 of the total iron sites. Since no distinction is made between Cobalt 60, Cobalt 58, and Cobalt 59 in this process, the radioactivity associated with this inclusion is a function of the ratio of the radioactive species to the nonradioactive species in the water causing the corrosion of the pipe metal. The other controlling parameter is the corrosion rate of the pipe material. This can be a function of time, for example, and it shown that freshly descaled metal when exposed to the cobalt containing water can incorporate as much as 10 x 10 -3 cobalt ions per iron atom in the initial corrosion period. This has implications for the problem of decontaminating nuclear reactor piping. Equations and selected observations are presented without reference to any specifically identified reactor or utility, so as to protect any proprietary interest

  14. Symposium on operational and environmental issues concerning use of water as a coolant in power plants and industries: proceedings

    International Nuclear Information System (INIS)

    2008-12-01

    The symposium is organised to bring together researchers, plant operators and regulatory agencies working in the area of operational and environmental problems associated with use of water as a coolant in power plants and other allied industries. The symposium targets chemists, biologists, environmental scientists, power plant operating engineers and plant designers working in various academic, governmental and non-governmental organisations. The major themes of the symposium are: water chemistry of coolant systems in power plants and other industries, chemistry of primary and moderator systems in nuclear power plants and research reactors, corrosion issues including Flow-Accelerated Corrosion (FAC) and its control in water coolant systems, chemistry of steam and water at elevated temperature in nuclear power plants, once through steam generator chemistry, industrial fire water systems, ion-exchange purification, innovative water treatment in power and industrial units, chemical cleaning and chemical decontamination, biofouling and biocorrosion, cooling water treatment chemicals and their environmental fate and environmental impact of thermal effluents. Papers relevant to INIS are indexed separately

  15. Multi-Phase Equilibrium and Solubilities of Aromatic Compounds and Inorganic Compounds in Sub- and Supercritical Water: A Review.

    Science.gov (United States)

    Liu, Qinli; Ding, Xin; Du, Bowen; Fang, Tao

    2017-11-02

    Supercritical water oxidation (SCWO), as a novel and efficient technology, has been applied to wastewater treatment processes. The use of phase equilibrium data to optimize process parameters can offer a theoretical guidance for designing SCWO processes and reducing the equipment and operating costs. In this work, high-pressure phase equilibrium data for aromatic compounds+water systems and inorganic compounds+water systems are given. Moreover, thermodynamic models, equations of state (EOS) and empirical and semi-empirical approaches are summarized and evaluated. This paper also lists the existing problems of multi-phase equilibria and solubility studies on aromatic compounds and inorganic compounds in sub- and supercritical water.

  16. Gasification of fruit wastes and agro-food residues in supercritical water

    International Nuclear Information System (INIS)

    Nanda, Sonil; Isen, Jamie; Dalai, Ajay K.; Kozinski, Janusz A.

    2016-01-01

    Highlights: • Supercritical water gasification of various fruit wastes and agro-food residues. • Coconut shell had superior carbon content and calorific value due to high lignin. • Maximum H_2 yields at 600 °C with 1:10 biomass-to-water ratio, 45 min and 23–25 MPa. • High H_2 yields from coconut shell, bagasse and aloe vera rind with 2 wt% K_2CO_3. • High CH_4 yields from coconut shell with 2 wt% NaOH due to methanation reaction. - Abstract: Considerable amounts of fruit wastes and agro-food residues are generated worldwide as a result of food processing. Converting the bioactive components (e.g., carbohydrates, lipids, fats, cellulose, hemicellulose and lignin) in food wastes to biofuels is a potential remediation approach. This study highlights the characterization and hydrothermal conversion of several fruit wastes and agro-food residues such as aloe vera rind, banana peel, coconut shell, lemon peel, orange peel, pineapple peel and sugarcane bagasse to hydrogen-rich syngas through supercritical water gasification. The agro-food wastes were gasified in supercritical water to study the impacts of temperature (400–600 °C), biomass-to-water ratio (1:5 and 1:10) and reaction time (15–45 min) at a pressure range of 23–25 MPa. The catalytic effects of NaOH and K_2CO_3 were also investigated to maximize the hydrogen yields and selectivity. The elevated temperature (600 °C), longer reaction time (45 min) and lower feed concentration (1:10 biomass-to-water ratio) were optimal for higher hydrogen yield (0.91 mmol/g) and total gas yield (5.5 mmol/g) from orange peel. However, coconut shell with 2 wt% K_2CO_3 at 600 °C and 1:10 biomass-to-water ratio for 45 min revealed superior hydrogen yield (4.8 mmol/g), hydrogen selectivity (45.8%) and total gas yield (15 mmol/g) with enhanced lower heating value of the gas product (1595 kJ/Nm"3). The overall findings suggest that supercritical water gasification of fruit wastes and agro-food residues could serve as

  17. RELAP5-3D Code for Supercritical-Pressure Light-Water-Cooled Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Riemke, Richard Allan; Davis, Cliff Bybee; Schultz, Richard Raphael

    2003-04-01

    The RELAP5-3D computer program has been improved for analysis of supercritical-pressure, light-water-cooled reactors. Several code modifications were implemented to correct code execution failures. Changes were made to the steam table generation, steam table interpolation, metastable states, interfacial heat transfer coefficients, and transport properties (viscosity and thermal conductivity). The code modifications now allow the code to run slow transients above the critical pressure as well as blowdown transients (modified Edwards pipe and modified existing pressurized water reactor model) that pass near the critical point.

  18. Numerical comparison of thermal hydraulic aspects of supercritical carbon dioxide and subcritical water-based natural circulation loop

    Energy Technology Data Exchange (ETDEWEB)

    Sarkar, Milan Krishna Singhar; Basu, Dipankar Narayan [Dept. of Mechanical Engineering, Indian Institute of Technology Guwahati, Guwahati (India)

    2017-02-15

    Application of the supercritical condition in reactor core cooling needs to be properly justified based on the extreme level of parameters involved. Therefore, a numerical study is presented to compare the thermalhydraulic performance of supercritical and single-phase natural circulation loops under low-to-intermediate power levels. Carbon dioxide and water are selected as respective working fluids, operating under an identical set of conditions. Accordingly, a three-dimensional computational model was developed, and solved with an appropriate turbulence model and equations of state. Large asymmetry in velocity and temperature profiles was observed in a single cross section due to local buoyancy effect, which is more prominent for supercritical fluids. Mass flow rate in a supercritical loop increases with power until a maximum is reached, which subsequently corresponds to a rapid deterioration in heat transfer coefficient. That can be identified as the limit of operation for such loops to avoid a high temperature, and therefore, the use of a supercritical loop is suggested only until the appearance of such maxima. Flow-induced heat transfer deterioration can be delayed by increasing system pressure or lowering sink temperature. Bulk temperature level throughout the loop with water as working fluid is higher than supercritical carbon dioxide. This is until the heat transfer deterioration, and hence the use of a single-phase loop is prescribed beyond that limit.

  19. Experimental study on the minimum drag coefficient of supercritical pressure water in horizontal tubes

    International Nuclear Information System (INIS)

    Lei, Xianliang; Li, Huixiong; Guo, YuMeng; Zhang, Qing; Zhang, Weiqiang; Zhang, Qian

    2016-01-01

    Highlights: • The minimum drag coefficient phenomenon (MDC) has been observed and further investigated. • Effects of heat flux, mass flux and pressure to MDC have been discussed. • A series of comparisons between existing correlations and data have been conducted. • Two correlations of drag coefficient are proposed for isothermal and nonisothermal flow. - Abstract: Hydraulic resistance and its components are of great importance for understanding the turbulence nature of supercritical fluid and establishing prediction methods. Under supercritical pressures, the hydraulic resistance of the fluid exhibits a “pit” in the regions near its pseudo-critical point, which is hereafter called the minimum drag coefficient phenomenon. However, this special phenomenon was paid a little attention before. Hence systematical experiments have been carried out to investigate the hydraulic resistance of supercritical pressure water in both adiabatic and heated horizontal tubes. Parametric effects of heat flux, pressure and mass fluxes to drag coefficient are further compared. It is found that almost all of the existing correlations don’t agree well with the experimental data due to the insufficient consideration of thermal-properties near the pseudocritical point. Two correlations of the drag coefficients are finally proposed by introducing the new variable of the derivative of density with respect to temperature or Prandtl number, which can better predict the drag coefficient of isothermal and nonisothermal flow respectively.

  20. Effect of sub- and supercritical water treatments on the physicochemical properties of crab shell chitin and its enzymatic degradation.

    Science.gov (United States)

    Osada, Mitsumasa; Miura, Chika; Nakagawa, Yuko S; Kaihara, Mikio; Nikaido, Mitsuru; Totani, Kazuhide

    2015-12-10

    This study examined the effects of sub- and supercritical water pretreatments on the physicochemical properties of crab shell α-chitin and its enzymatic degradation to obtain N,N'-diacetylchitobiose (GlcNAc)2. Following sub- and supercritical water pretreatments, the protein in the crab shell was removed and the residue of crab shell contained α-chitin and CaCO3. Prolonged pretreatment led to α-chitin decomposition. The reaction of pure α-chitin in sub- and supercritical water pretreatments was investigated separately; we observed lower mean molecular weight and weaker hydrogen bonds compared with untreated α-chitin. (GlcNAc)2 yields from enzymatic degradation of subcritical (350 °C, 7 min) and supercritical water (400 °C, 2.5 min) pretreated crab shell were 8% and 6%, compared with 0% without any pretreatment. This study shows that sub- and supercritical water pretreatments of crab shell provide to an alternative method to the use of acid and base for decalcification and deproteinization of crab shell required for (GlcNAc)2 production. Copyright © 2015 Elsevier Ltd. All rights reserved.

  1. An assessment of ex-vessel fuel-coolant interaction energetics for advanced light water reactors

    International Nuclear Information System (INIS)

    Murphy, J.G.; Corradini, M.L.

    1997-01-01

    The occurrence of an energetic fuel/coolant interaction (FCI) below the reactor pressure vessel in the cavity of advanced light water reactors (ALWRs) are analyzed to determine the possible hazard to structural walls as a result of dynamic liquid phase pressures. Such analyses are important to demonstrate that these cavity walls will maintain their integrity so that ex-vessel core debris coolability is possible. Past studies that have examined this or related issues are reviewed, and a methodology is proposed to analyze the occurrence of this physical event using the IFCI and TEXAS models for the FCI as well as dynamic shock wave propagation estimates using hand calculations as well as the CTH hydro model. Scenarios for the ALWRs are reviewed, and one severe accident scenario is used as an example to demonstrate the methodology. Such methodologies are recommended for consideration in future safety studies. These methodologies should be verified with direct comparison to energetic FCI data such as that being produced in KROTOS at the Joint Research Centre, Ispra

  2. Interfacing systems LOCAs [Loss of Coolant Accidents] at boiling water reactors

    International Nuclear Information System (INIS)

    Chu, Tsong-Lun; Fitzpatrick, R.; Stoyanov, S.

    1987-01-01

    The work presented in this paper was performed by Brookhaven National Laboratory (BNL) in support of Nuclear Regulatory Commission's (NRC) effort towards the resolution of Generic Issue 105 ''Interfacing System Loss of Coolant Accidents (LOCAs) at Boiling Water Reactors (BWRs).'' For BWRs, intersystem LOCA have typically either not been considered in probabilistic risk analyses, or if considered, were judged to contribute little to the risk estimates because of their perceived low frequency of occurrence. However, recent operating experience indicates that the pressure isolation valves (PIVs) in BWRs may not adequately protect against overpressurization of low pressure systems. The objective of this paper is to present the results of a study which analyzed interfacing system LOCA at several BWRs. The BWRs were selected to best represent a spectrum of BWRs in service using industry operating event experience and plant-specific information/configurations. The results presented here include some possible changes in test requirements/practices as well as an evaluation of their reduction potential in terms of core damage frequency

  3. Startup of a high-temperature reactor cooled and moderated by supercritical-pressure light water

    International Nuclear Information System (INIS)

    Yi, Tin Tin; Ishiwatari, Yuki; Koshizuka, Seiichi; Oka, Yoshiaki

    2003-01-01

    The startup schemes of high-temperature reactors cooled and moderated by supercritical pressure light water (SCLWR-H) with square lattice and descending flow type water rods are studied by thermal-hydraulic analysis. In this study, two kinds of startup systems are investigated. In the constant pressure startup system, the reactor starts at a supercritical pressure. A flash tank and pressure reducing valves are necessary. The flash tank is designed so that the moisture content in the steam is less than 0.1%. In sliding pressure startup system, the reactor starts at a subcritical pressure. A steam-water separator and a drain tank are required for two-phase flow at startup. The separator is designed by referring to the water separator used in supercritical fossil-fired power plants. The maximum cladding surface temperature during the power-raising phase of startup is restricted not to exceed the rated value of 620degC. The minimum feedwater flow rate is 25% for constant pressure startup and 35% for sliding pressure startup system. It is found that both constant pressure startup system and sliding pressure startup system are feasible in SCLWR-H from the thermal hydraulic point of view. The core outlet temperature as high as 500degC can be achieved in the present design of SCLWR-H. Since the feedwater flow rate of SCLWR-H (1190 kg/s) is lower than that of the previous SCR designs the weight of the component required for startup is reduced. The sliding pressure startup system is better than constant pressure startup system in order to reduce the required component weight (and hence material expenditure) and to simplify the startup plant system. (author)

  4. Modelling the power conversion unit of a generic nuclear fusion plant, with a dual coolant blanket and a supercritical CO2 power cycle, by means of RELAP5-3D

    Energy Technology Data Exchange (ETDEWEB)

    Batet, L.

    2015-07-01

    In the framework of the Spanish fusion program TECNO-FUS, a dual coolant blanket design was proposed for DEMO. A generic power conversion system (supercritical recompression CO2 cycle) based on this proposal has been simulated using RELAP5-3D, a multipurpose system thermal-hydraulic code developed by the Idaho National Laboratory (USA). The code allows the dynamic simulation of thermal-hydraulic systems, including the control features. A model has been set up by assembling the available RELAP5-3D components: pipe, branch, pump, compressor, turbine, etc. Thermal fluxes between fluids in heat exchangers are simulated by means of heat structures, which are used as well to simulate the heating from plasma. A number of control features have been designed for the simulated plant, and their parameters have been adjusted. The code is then able to simulate robustly the dynamics of the system with a few boundary conditions. This paper exemplifies the usefulness of the code and model to understand the behavior of the plant and to perform sensitivity analyses of the control parameters or other design features. (Author)

  5. Development of a test facility for analyzing transients in supercritical water-cooled reactors by fractional scaling analysis

    Energy Technology Data Exchange (ETDEWEB)

    Roberto, Thiago D., E-mail: thiagodbtr@gmail.com [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil); Silva, Mário A. B. da, E-mail: mabs500@gmail.com [Departamento de Energia Nuclear (CTG/UFPE), Av. Professor Luiz Freire, 1000, Recife 50740-540, PE (Brazil); Lapa, Celso M.F., E-mail: lapa@ien.gov.br [Instituto de Engenharia Nuclear (IEN/CNEN—RJ), Rua Hélio de Almeida, 75 21941-972, Rio de Janeiro Caixa-Postal: 68550, RJ (Brazil)

    2016-01-15

    The feasibility of performing experiments using water under supercritical conditions is limited by technical and financial difficulties. These difficulties can be overcome by using model fluids that are characterized by feasible supercritical conditions, that is, lower critical pressure and critical temperature. Experimental investigations are normally used to determine the conditions under which model fluids reliably represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine the model fluids that represent supercritical fluids in a transient state. Recently, a similar technique known as fractional scaling analysis was developed to establish the conditions under which experiments can be performed using models that represent transients in prototypes. This paper presents a fractional scaling analysis application to determine parameters for a test facility in which transient conditions in supercritical water-cooled reactors are simulated by using carbon dioxide as a model fluid, whose critical point conditions are more feasible than those of water. Similarity is obtained between water (prototype) and carbon dioxide (model) by depressurization in a simple vessel. The main parameters required for the construction of a future test facility are obtained using the proposed method.

  6. Development of a test facility for analyzing transients in supercritical water-cooled reactors by fractional scaling analysis

    International Nuclear Information System (INIS)

    Roberto, Thiago D.; Silva, Mário A. B. da; Lapa, Celso M.F.

    2016-01-01

    The feasibility of performing experiments using water under supercritical conditions is limited by technical and financial difficulties. These difficulties can be overcome by using model fluids that are characterized by feasible supercritical conditions, that is, lower critical pressure and critical temperature. Experimental investigations are normally used to determine the conditions under which model fluids reliably represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine the model fluids that represent supercritical fluids in a transient state. Recently, a similar technique known as fractional scaling analysis was developed to establish the conditions under which experiments can be performed using models that represent transients in prototypes. This paper presents a fractional scaling analysis application to determine parameters for a test facility in which transient conditions in supercritical water-cooled reactors are simulated by using carbon dioxide as a model fluid, whose critical point conditions are more feasible than those of water. Similarity is obtained between water (prototype) and carbon dioxide (model) by depressurization in a simple vessel. The main parameters required for the construction of a future test facility are obtained using the proposed method.

  7. Methodologies and technologies for life assessment and management of coolant channels of Indian pressurised heavy water reactors

    International Nuclear Information System (INIS)

    Rupani, B.B.; Sinha, S.K.; Sinha, R.K.

    2002-01-01

    Zirconium alloy coolant channels are central to the design of Indian Pressurised Heavy Water Reactors (PHWRs) and form the individual pressure boundaries. These coolant channels consist of horizontal pressure tubes made of zirconium alloys, which are separated from cold calandria tubes using garter spring spacers. High temperature heavy water coolant flows through the pressure tube which supports the fuel bundles. A typical coolant channel in a PHWR is shown. These pressure tubes are subjected to several life limiting degradation mechanisms like creep and growth, hydrogen pick-up, reduction in fracture toughness and delayed hydride cracking phenomena because of their operation under high temperature, high stress and high fast neutron flux environment. Considering the early onset of these degradation mechanisms in Zircaloy-2 pressure tubes used in the early generation of Indian PHWRs, the life management of these coolant channels becomes a challenging task, involving multidisciplinary R and D efforts in areas like analytical modelling of degradation mechanisms, evolution of methodologies for assessment of fitness for service and, tools and techniques for remote on line monitoring of integrity, maintenance and replacement. The degradation mechanisms have been modelled and incorporated into specially developed computer codes, such as SCAPCA for irradiation induced creep and growth deformation modelling, HYCON for hydrogen pick-up modelling, BLIST for hydrogen diffusion, blister nucleation and growth modelling and CEAL for assessment of leak before break behaviour. These codes have been validated with respect to the results of in-service inspection and post irradiation examination. Development of analytical models actually paved the way for the evolution of more refined methodologies for assessing the safe residual life of coolant channel. Information gathered from various experiments simulating the degradation mechanisms, results of post-irradiation examination of the

  8. An experimental investigation of flow instability between two heated parallel channels with supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Xi, Xi; Xiao, Zejun, E-mail: fabulous_2012@sina.com; Yan, Xiao; Li, Yongliang; Huang, Yanping

    2014-10-15

    Highlights: • Flow instability experiment between two heated channels with supercritical water is carried out. • Two kinds of out of phase flow instability are found and instability boundaries under different working conditions are obtained. • Dynamics characteristics of flow instability are analyzed. - Abstract: Super critical water reactor (SCWR) is the generation IV nuclear reactor in the world. Under normal operation, water enters SCWR from cold leg with a temperature of 280 °C and then leaves the core with a temperature of 500 °C. Due to the sharp change of temperature, there is a huge density change in the core, which could result in potential flow instability and the safety of reactor would be threatened consequently. So it is necessary to carry out relevant investigation in this field. An experimental investigation which concerns with out of phase flow instability between two heated parallel channels with supercritical water has been carried out in this paper. Due to two INCONEL 625 pipes with a thickness of 6.5 mm are adopted, more experimental results are attained. To find out the influence of axial power shape on the onset of flow instability, each heated channel is divided into two sections and the heating power of each section can be controlled separately. Finally the instability boundaries are obtained under different inlet temperatures, axial power shapes, total inlet mass flow rates and system pressures. The dynamics characteristics of out of phase oscillation are also analyzed.

  9. Hydrogen production from high-moisture content biomass in supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Antal, M.J. Jr.; Adschiri, T.; Ekbom, T. [Univ. of Hawaii, Honolulu, HI (United States)] [and others

    1996-10-01

    Most hydrogen is produced by steam reforming methane at elevated pressures. The goal of this research is to develop commercial processes for the catalytic steam reforming of biomass and other organic wastes at high pressures. This approach avoids the high cost of gas compression and takes advantage of the unique properties of water at high pressures. Prior to this year the authors reported the ability of carbon to catalyze the decomposition of biomass and related model compounds in supercritical water. The product gas consists of hydrogen, carbon dioxide, carbon monoxide, methane, and traces of higher hydrocarbons. During the past year the authors have: (a) developed a method to extend the catalyst life, (b) begun studies of the role of the shift reaction, (c) completed studies of carbon dioxide absorption from the product effluent by high pressure water, (d) measured the rate of carbon catalyst gasification in supercritical water, (e) discovered the pumpability of oil-biomass slurries, and (f) completed the design and begun fabrication of a flow reactor that will steam reform whole biomass feedstocks (i.e. sewage sludge) and produce a hydrogen rich synthesis gas at very high pressure (>22 MPa).

  10. Numerical investigation of heat transfer in parallel channels with water at supercritical pressure

    Directory of Open Access Journals (Sweden)

    Edward Shitsi

    2017-11-01

    Full Text Available Thermal phenomena such as heat transfer enhancement, heat transfer deterioration, and flow instability observed at supercritical pressures as a result of fluid property variations have the potential to affect the safety of design and operation of Supercritical Water-cooled Reactor SCWR, and also challenge the capabilities of both heat transfer correlations and Computational Fluid Dynamics CFD physical models. These phenomena observed at supercritical pressures need to be thoroughly investigated.An experimental study was carried out by Xi to investigate flow instability in parallel channels at supercritical pressures under different mass flow rates, pressures, and axial power shapes. Experimental data on flow instability at inlet of the heated channels were obtained but no heat transfer data along the axial length was obtained. This numerical study used 3D numerical tool STAR-CCM+ to investigate heat transfer at supercritical pressures along the axial lengths of the parallel channels with water ahead of experimental data. Homogeneous axial power shape HAPS was adopted and the heating powers adopted in this work were below the experimental threshold heating powers obtained for HAPS by Xi. The results show that the Fluid Centre-line Temperature FCLT increased linearly below and above the PCT region, but flattened at the PCT region for all the system parameters considered. The inlet temperature, heating power, pressure, gravity and mass flow rate have effects on WT (wall temperature values in the NHT (normal heat transfer, EHT (enhanced heat transfer, DHT (deteriorated heat transfer and recovery from DHT regions. While variation of all other system parameters in the EHT and PCT regions showed no significant difference in the WT and FCLT values respectively, the WT and FCLT values respectively increased with pressure in these regions. For most of the system parameters considered, the FCLT and WT values obtained in the two channels were nearly the

  11. Numerical investigation of heat transfer in parallel channels with water at supercritical pressure.

    Science.gov (United States)

    Shitsi, Edward; Kofi Debrah, Seth; Yao Agbodemegbe, Vincent; Ampomah-Amoako, Emmanuel

    2017-11-01

    Thermal phenomena such as heat transfer enhancement, heat transfer deterioration, and flow instability observed at supercritical pressures as a result of fluid property variations have the potential to affect the safety of design and operation of Supercritical Water-cooled Reactor SCWR, and also challenge the capabilities of both heat transfer correlations and Computational Fluid Dynamics CFD physical models. These phenomena observed at supercritical pressures need to be thoroughly investigated. An experimental study was carried out by Xi to investigate flow instability in parallel channels at supercritical pressures under different mass flow rates, pressures, and axial power shapes. Experimental data on flow instability at inlet of the heated channels were obtained but no heat transfer data along the axial length was obtained. This numerical study used 3D numerical tool STAR-CCM+ to investigate heat transfer at supercritical pressures along the axial lengths of the parallel channels with water ahead of experimental data. Homogeneous axial power shape HAPS was adopted and the heating powers adopted in this work were below the experimental threshold heating powers obtained for HAPS by Xi. The results show that the Fluid Centre-line Temperature FCLT increased linearly below and above the PCT region, but flattened at the PCT region for all the system parameters considered. The inlet temperature, heating power, pressure, gravity and mass flow rate have effects on WT (wall temperature) values in the NHT (normal heat transfer), EHT (enhanced heat transfer), DHT (deteriorated heat transfer) and recovery from DHT regions. While variation of all other system parameters in the EHT and PCT regions showed no significant difference in the WT and FCLT values respectively, the WT and FCLT values respectively increased with pressure in these regions. For most of the system parameters considered, the FCLT and WT values obtained in the two channels were nearly the same. The

  12. Experimental and numerical investigation of heat transfer from a narrow annulus to supercritical pressure water

    International Nuclear Information System (INIS)

    Wang, Han; Bi, Qincheng; Yang, Zhendong; Wang, Linchuan

    2015-01-01

    Highlights: • Heat transfer of supercritical water in a narrow annulus is investigated. • Effects of system parameters and flow direction on heat transfer are studied. • Deteriorated heat transfer is analyzed both experimentally and numerically. - Abstract: Heat transfer characteristics of supercritical pressure water in a narrow annulus with vertically upward and downward flows were investigated experimentally and numerically. The outer diameter of the inner heated rod is 8 mm with an effective heated length of 620 mm. Experimental parameters covered the pressure of 23–28 MPa, mass flux of 400–1000 kg/m 2 s and heat flux on the outer surface of the heated rod from 200 to 1000 kW/m 2 . The general heat transfer behaviors were discussed with respect to various mass fluxes and pressures. According to the experimental data, it was found that the effect of flow direction on heat transfer depends on the heat-flux to mass-flux ratio (q/G). Heat transfer is much improved in the downward flow compared to that of upward flow at high q/G ratios. At the pressure of 25 MPa, low-mass-flux deteriorated heat transfer occurred in the upward flow but not in the downward flow. At the same test parameters, however, heat transfer deterioration was observed at both of the two flow directions when the pressure was lowered to 23 MPa. The experimental results indicate that buoyancy plays an important role for this type of deterioration, but is not the only mechanism that leads to the heat transfer deterioration. Three turbulence models were assessed against the annulus test data, it was found that the SST k-ω model gives a satisfying prediction of heat transfer deterioration especially for the case of downward flow. The mechanisms for the low-mass-flow heat transfer deterioration were investigated from the viewpoints of buoyancy and property variations of the supercritical water

  13. Reactivities of polystyrenic polymers with supercritical water under nitrogen or air. Identification and formation of degradation compounds

    International Nuclear Information System (INIS)

    Dubois, M.A.; Dozol, J.F.; Massiani, C.; Ambrosio, M.

    1996-01-01

    Supercritical water oxidation (SCWO) could offer a viable treatment alternative to destroy the organic structure of ion-exchange resins (IER) that are radioactive process wastes and which contain radioactivity. The GC/MS technique was used successfully to identify the low-concentration degradation compounds that are present in the cold liquid effluent after SCWO of polystyrenic IER at 380 C (25.5 MPa). The study of the behavior of these IER in supercritical water enhances the role of temperature and the role of supercritical water in the degradation process. With the exception of acetic acid, the identified compounds are aromatic. The functional groups are released during the heating time, and they do not interfere in the degradation process. The oxidation involves a complex set of reaction pathways. A mechanism including parallel and competitive reactions is proposed

  14. Nuclear reactor coolant channels

    International Nuclear Information System (INIS)

    Macbeth, R.V.

    1978-01-01

    A nuclear reactor coolant channel is described that is suitable for sub-cooled reactors as in pressurised water reactors as well as for bulk boiling, as in boiling water reactors and steam generating nuclear reactors. The arrangement aims to improve heat transfer between the fuel elements and the coolant. Full constructional details are given. See also other similar patents by the author. (U.K.)

  15. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    International Nuclear Information System (INIS)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-01-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean/US/laboratory/university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program

  16. CFD analysis of supercritical water flow and heat transfer in single channel with mixing vane

    International Nuclear Information System (INIS)

    Zuo Guoping; Xie Hongyan; Yu Tao

    2012-01-01

    Three-dimensional rectangular channel with the mixing wane in supercritical water reactor is investigated with CFX. The mixing vane elevation influenced on temperature distribution and flow field are simulated in the model. The results showed the mixing vane cause fluid circumferential flow, making flow hot and cold fluids mixed and fluid temperature uniform distribution, effectively improve the fuel rod surface temperature distribution and reduced hot temperature. Among the mixing wing elevation of 15, 30, 45, 50, 60 and 70 angle, the 30 angle is the best case in improving temperature distribution. (authors)

  17. Oxidization and stress corrosion cracking initiation of austenitic alloys in supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Behnamian, Y.; Li, M.; Luo, J.L.; Chen, W.X. [Univ. of Alberta, Dept. of Chemical and Materials Engineering, Edmonton, Alberta (Canada); Zheng, W. [Materials Technology Laboratory, NRCan, Ottawa, Ontario (Canada); Guzonas, D.A. [Atomic Energy of Canada Limited, Chalk River, Ontario (Canada)

    2012-07-01

    This study determined the stress corrosion cracking behaviour of austenitic alloys in pure supercritical water. Austenitic stainless steels 310S, 316L, and Inconel 625 were tested as static capsule samples at 500{sup o}C for up to 5000 h. After that period, crack initiations were readily observed in all samples, signifying susceptibility to stress corrosion cracking. The microcracks in 316L stainless steel and Inconel 625 were almost intergranular, whereas transgranular microcrack initiation was observed in 310S stainless steel. (author)

  18. Effect of Heating Method on Hydrogen Production by Biomass Gasification in Supercritical Water

    Directory of Open Access Journals (Sweden)

    Qiuhui Yan

    2014-01-01

    Full Text Available The glucose as a test sample of biomass is gasified in supercritical water with different heating methods driven by renewable solar energy. The performance comparisons of hydrogen production of glucose gasification are investigated. The relations between temperature raising speed of reactant fluid, variation of volume fraction, combustion enthalpy, and chemical exergy of H2 of the product gases with reactant solution concentration are presented, respectively. The results show that the energy quality of product gases with preheating process is higher than that with no preheating unit for hydrogen production. Hydrogen production quantity and gasification rate of glucose decrease obviously with the increase of concentration of material in no preheating system.

  19. Final report on the oxidation of energetic materials in supercritical water. Final Air Force report

    Energy Technology Data Exchange (ETDEWEB)

    Buelow, S.J.; Allen, D.; Anderson, G.K. [and others

    1995-04-03

    The objective of this project was to determine the suitability of oxidation in supercritical fluids (SCO), particularly water (SCWO), for disposal of propellants, explosives, and pyrotechnics (PEPs). The SCO studies of PEPs addressed the following issues: The efficiency of destruction of the substrate. The products of destruction contained in the effluents. Whether the process can be conducted safely on a large scale. Whether energy recovery from the process is economically practicable. The information essential for process development and equipment design was also investigated, including issues such as practical throughput of explosives through a SCWO reactor, reactor materials and corrosion, and models for process design and optimization.

  20. Advanced Computational Thermal Fluid Physics (CTFP) and Its Assessment for Light Water Reactors and Supercritical Reactors

    Energy Technology Data Exchange (ETDEWEB)

    D.M. McEligot; K. G. Condie; G. E. McCreery; H. M. McIlroy; R. J. Pink; L.E. Hochreiter; J.D. Jackson; R.H. Pletcher; B.L. Smith; P. Vukoslavcevic; J.M. Wallace; J.Y. Yoo; J.S. Lee; S.T. Ro; S.O. Park

    2005-10-01

    Background: The ultimate goal of the study is the improvement of predictive methods for safety analyses and design of Generation IV reactor systems such as supercritical water reactors (SCWR) for higher efficiency, improved performance and operation, design simplification, enhanced safety and reduced waste and cost. The objective of this Korean / US / laboratory / university collaboration of coupled fundamental computational and experimental studies is to develop the supporting knowledge needed for improved predictive techniques for use in the technology development of Generation IV reactor concepts and their passive safety systems. The present study emphasizes SCWR concepts in the Generation IV program.

  1. Optimization of a fuel bundle within a CANDU supercritical water reactor

    International Nuclear Information System (INIS)

    Schofield, M.E.

    2009-01-01

    The supercritical water reactor is one of six nuclear reactor concepts being studied under the Generation IV International Forum. Generation IV nuclear reactors will improve the metrics of economics, sustainability, safety and reliability, and physical protection and proliferation resistance over current nuclear reactor designs. The supercritical water reactor has specific benefits in the areas of economics, safety and reliability, and physical protection. This work optimizes the fuel composition and bundle geometry to maximize the fuel burnup, and minimize the surface heat flux and the form factor. In optimizing these factors, improvements can be achieved in the areas of economics, safety and reliability of the supercritical water reactor. The WIMS-AECL software was used to model a fuel bundle within a CANDU supercritical water reactor. The Gauss' steepest descent method was used to optimize the above mentioned factors. Initially the fresh fuel composition was optimized within a 43-rod CANFLEX bundle and a 61-rod bundle. In both the 43-rod and 61-rod bundle scenarios an online refuelling scheme and non-refuelling scheme were studied. The geometry of the fuel bundles was then optimized. Finally, a homogeneous mixture of thorium and uranium fuel was studied in a 60-rod bundle. Each optimization process showed definitive improvements in the factors being studied, with the most significant improvement being an increase in the fuel burnup. The 43-rod CANFLEX bundle was the most successful at being optimized. There was little difference in the final fresh fuel content when comparing an online refuelling scheme and non-refuelling scheme. Through each optimization scenario the ratio of the fresh fuel content between the annuli was a significant determining cause in the improvements in the factors being optimized. The geometry optimization showed that improvement in the design of a fuel bundle is indeed possible, although it would be more advantageous to pursue it

  2. Validation of Computational Fluid Dynamics Calculation Using Rossendorf Coolant Mixing Model Flow Measurements in Primary Loop of Coolant in a Pressurized Water Reactor Model

    Directory of Open Access Journals (Sweden)

    Istvan Farkas

    2016-08-01

    Full Text Available The aim of this work is to simulate the thermohydraulic consequences of a main steam line break and to compare the obtained results with Rossendorf Coolant Mixing Model (ROCOM 1.1 experimental results. The objective is to utilize data from steady-state mixing experiments and computational fluid dynamics (CFD calculations to determine the flow distribution and the effect of thermal mixing phenomena in the primary loops for the improvement of normal operation conditions and structural integrity assessment of pressurized water reactors. The numerical model of ROCOM was developed using the FLUENT code. The positions of the inlet and outlet boundary conditions and the distribution of detailed velocity/turbulence parameters were determined by preliminary calculations. The temperature fields of transient calculation were averaged in time and compared with time-averaged experimental data. The perforated barrel under the core inlet homogenizes the flow, and therefore, a uniform temperature distribution is formed in the pressure vessel bottom. The calculated and measured values of lowest temperature were equal. The inlet temperature is an essential parameter for safety assessment. The calculation predicts precisely the experimental results at the core inlet central region. CFD results showed a good agreement (both qualitatively and quantitatively with experimental results.

  3. Development of supercritical water reactors in Russia and abroad

    International Nuclear Information System (INIS)

    Glebov, A.P.; Klushin, A.V.

    2014-01-01

    The results of Russian and foreign studies on the water-cooled high critical parameters reactors are analyzed. Developments on this subject are conducted in more than 15 countries. The advantages of WWER- SCP and characteristics of experimental reactor of WWER-SCP-30 are discussed. It is noted that priority task is to develop a reactor with thermal neutron spectrum with a subsequent transition to the reactor with a fast neutron spectrum [ru

  4. Failure Analysis of 600 MW Supercritical Boiler Water Wall

    OpenAIRE

    Fu Huilin; Cai Zhengchun; Yan Xiaozhong; He Jinqiao; Zhou Yucai

    2013-01-01

    Boiler tube often causes abnormal boiler outage, bringing greater economic losses. This thesis mainly comes from the dynamics of boiler water, boiler furnace accident location of wall temperature distribution to explore the cause of the accident boiler. Calculation results show that the deformation will seriously reduce the boiler allowable maximum temperature difference between the screens. And the boiler is not over-temperature, low temperature difference between the screens, which have bur...

  5. Partial oxidation of n-hexadecane through decomposition of hydrogen peroxide in supercritical water

    KAUST Repository

    Alshammari, Y.M.

    2015-01-01

    © 2014 The Institution of Chemical Engineers. This work reports the experimental analysis of partial oxidation of n-hexadecane under supercritical water conditions. A novel reactor flow system was developed which allows for total decomposition of hydrogen peroxide in a separate reactor followed partial oxidation of n-hexadecane in a gasification reactor instead of having both reactions in one reactor. The kinetics of hydrothermal decomposition of hydrogen peroxide was studied in order to confirm its full conversion into water and oxygen under the desired partial oxidation conditions, and the kinetic data were found in a good agreement with previously reported literature. The gas yield and gasification efficiency were investigated under different operating parameters. Furthermore, the profile of C-C/C=C ratio was studied which showed the favourable conditions for maximising yields of n-alkanes via hydrogenation of their corresponding 1-alkenes. Enhanced hydrogenation of 1-alkenes was observed at higher O/C ratios and higher residence times, shown by the increase in the C-C/C=C ratio to more than unity, while increasing the temperature has shown much less effect on the C-C/C=C ratio at the current experimental conditions. In addition, GC-MS analysis of liquid samples revealed the formation of heavy oxygenated compounds which may suggest a new addition reaction to account for their formation under the current experimental conditions. Results show new promising routes for hydrogen production with in situ hydrogenation of heavy hydrocarbons in a supercritical water reactor.

  6. Enhancing the solubility and bioavailability of poorly water-soluble drugs using supercritical antisolvent (SAS) process.

    Science.gov (United States)

    Abuzar, Sharif Md; Hyun, Sang-Min; Kim, Jun-Hee; Park, Hee Jun; Kim, Min-Soo; Park, Jeong-Sook; Hwang, Sung-Joo

    2018-03-01

    Poor water solubility and poor bioavailability are problems with many pharmaceuticals. Increasing surface area by micronization is an effective strategy to overcome these problems, but conventional techniques often utilize solvents and harsh processing, which restricts their use. Newer, green technologies, such as supercritical fluid (SCF)-assisted particle formation, can produce solvent-free products under relatively mild conditions, offering many advantages over conventional methods. The antisolvent properties of the SCFs used for microparticle and nanoparticle formation have generated great interest in recent years, because the kinetics of the precipitation process and morphologies of the particles can be accurately controlled. The characteristics of the supercritical antisolvent (SAS) technique make it an ideal tool for enhancing the solubility and bioavailability of poorly water-soluble drugs. This review article focuses on SCFs and their properties, as well as the fundamentals of overcoming poorly water-soluble drug properties by micronization, crystal morphology control, and formation of composite solid dispersion nanoparticles with polymers and/or surfactants. This article also presents an overview of the main aspects of the SAS-assisted particle precipitation process, its mechanism, and parameters, as well as our own experiences, recent advances, and trends in development. Copyright © 2017 Elsevier B.V. All rights reserved.

  7. Corrosion behavior of ceramic-coated ZIRLO™ exposed to supercritical water

    Science.gov (United States)

    Mandapaka, Kiran K.; Cahyadi, Rico S.; Yalisove, Steven; Kuang, Wenjun; Sickafus, K.; Patel, Maulik K.; Was, Gary S.

    2018-01-01

    The corrosion behavior of ceramic coated ZIRLO™ tubing was evaluated in a supercritical water (SCW) environment to determine its behavior in high temperature water. Two coating architectures were analyzed; a 4 bi-layer TiAlN/TiN coating with Ti bond coat, and a TiN monolithic coating with Ti bond layer on ZIRLO™ tubes using cathodic arc physical vapor deposition (CA-PVD) technique. Femtosecond laser ablation was used to introduce reproducible defects in some of the coated tubes. On exposure to deaerated supercritical water at 542 °C for 48 h, coated tubes exhibited significantly higher weight gain compared to uncoated ZIRLO™. Examination revealed formation of a uniform ZrO2 layer beneath the coating of a thickness similar to that on the uncoated tube inner surface. The defects generated during the coating process acted as preferential paths for diffusion of oxygen resulting in the oxidation of substrate ZIRLO™. However, there was no delamination of the coating. There were insignificant differences in the oxidation weight gain between laser ablated and non-ablated tubes and the laser induced defects did not spread beyond their original size.

  8. Supercritical Water Oxidation: A Solution for the Elimination of Back-End Organic Reprocessing Wastes

    International Nuclear Information System (INIS)

    Leybros, A.; Roubaud, A.; Turc, H.A.; Fournel, B.

    2008-01-01

    Supercritical water oxidation (SCWO) is a very efficient technique for total elimination of organic wastes from reprocessing activities on the way of 'zero wastes' facilities. This technology uses the properties of supercritical water (P > 221 bars and T > 647 K) to obtain a good mixing between oxygen (the oxidant) and the organic waste. Thereby, the oxidation reaction is fast and complete. Using the SCWO process, contamination contained in organic materials like spent solvents can be confined in a closed space, like a reactor in a glovebox. A new application is tested for the treatment of solid organic wastes like ion exchange resins (IER). Experiments are made with suspensions of IER in water and isopropyl-alcohol. A nuclear version of the process with the double shell reactor has been constructed and is being tested. The aim of this work is to obtain a treatment capacity of 1 kg/h for the nuclear version with the same global set-up, concept of process and security as well as contamination management as for a 200 g/h pilot. (authors)

  9. Supercritical Water Oxidation: A Solution for the Elimination of Back-End Organic Reprocessing Wastes

    Energy Technology Data Exchange (ETDEWEB)

    Leybros, A.; Roubaud, A.; Turc, H.A.; Fournel, B. [Supercritical fluids and membranes Laboratory, CEA Valrho, BP 17171, 30207 Bagnols/Ceze Cedex (France)

    2008-07-01

    Supercritical water oxidation (SCWO) is a very efficient technique for total elimination of organic wastes from reprocessing activities on the way of 'zero wastes' facilities. This technology uses the properties of supercritical water (P > 221 bars and T > 647 K) to obtain a good mixing between oxygen (the oxidant) and the organic waste. Thereby, the oxidation reaction is fast and complete. Using the SCWO process, contamination contained in organic materials like spent solvents can be confined in a closed space, like a reactor in a glovebox. A new application is tested for the treatment of solid organic wastes like ion exchange resins (IER). Experiments are made with suspensions of IER in water and isopropyl-alcohol. A nuclear version of the process with the double shell reactor has been constructed and is being tested. The aim of this work is to obtain a treatment capacity of 1 kg/h for the nuclear version with the same global set-up, concept of process and security as well as contamination management as for a 200 g/h pilot. (authors)

  10. COAL CONVERSION WASTEWATER TREATMENT BY CATALYTIC OXIDATION IN SUPERCRITICAL WATER; FINAL

    International Nuclear Information System (INIS)

    Phillip E. Savage

    1999-01-01

    Wastewaters from coal-conversion processes contain phenolic compounds in appreciable concentrations. These compounds need to be removed so that the water can be discharged or re-used. Catalytic oxidation in supercritical water is one potential means of treating coal-conversion wastewaters, and this project examined the reactions of phenol over different heterogeneous oxidation catalysts in supercritical water. More specifically, we examined the oxidation of phenol over a commercial catalyst and over bulk MnO(sub 2), bulk TiO(sub 2), and CuO supported on Al(sub 2) O(sub 3). We used phenol as the model pollutant because it is ubiquitous in coal-conversion wastewaters and there is a large database for non-catalytic supercritical water oxidation (SCWO) with which we can contrast results from catalytic SCWO. The overall objective of this research project is to obtain the reaction engineering information required to evaluate the utility of catalytic supercritical water oxidation for treating wastes arising from coal conversion processes. All four materials were active for catalytic supercritical water oxidation. Indeed, all four materials produced phenol conversions and CO(sub 2) yields in excess of those obtained from purely homogeneous, uncatalyzed oxidation reactions. The commercial catalyst was so active that we could not reliably measure reaction rates that were not limited by pore diffusion. Therefore, we performed experiments with bulk transition metal oxides. The bulk MnO(sub 2) and TiO(sub 2) catalysts enhance both the phenol disappearance and CO(sub 2) formation rates during SCWO. MnO(sub 2) does not affect the selectivity to CO(sub 2), or to the phenol dimers at a given phenol conversion. However, the selectivities to CO(sub 2) are increased and the selectivities to phenol dimers are decreased in the presence of TiO(sub 2) , which are desirable trends for a catalytic SCWO process. The role of the catalyst appears to be accelerating the rate of formation of

  11. Some aspects of influence of coolant water chemistry on reliability of WWER and RBMK type fuels

    International Nuclear Information System (INIS)

    Solyany, V.I.; Bibilashvili, Yu.K.; Sukhanov, G.I.; Pimenov, Yu.V.; Gosudarstvennyj Komitet po Ispol'zovaniyu Atomnoj Ehnergii SSSR, Moscow)

    1983-01-01

    In WWER and RBMK reactors now in operation a good quality of primary coolant is achieved and the required corrosion resistance of structural materials and normal irradiation conditions are ensured. Data on commercial fuel operation and clad material (Zr 1% Nb alloy) condition are briefly generalized. Some results of reactor investigations of corrosion behaviour of commercial Zr 1% Nb alloy under the condition of WWER and RBMK coolant are discussed and compared. It is established that the chemical effect of coolant on fuel cladding does not in itself limit its serviceability at design burn-ups but due to the possible processes of crud formation, corrosion (total and local), fretting-corrosion and hydriding it can influence the fuel reliability. This influence is qualitatively assessed through a rise in the clad temperature, a reduction of material plasticity and clad thickness. (author)

  12. Some aspects of influence of coolant water chemistry on reliability of WWER and RBMK type fuels

    Energy Technology Data Exchange (ETDEWEB)

    Solyany, V I; Bibilashvili, Yu K; Sukhanov, G I; Pimenov, Yu V [Vsesoyuznyj Nauchno-Issledovatel' skij Inst. Neorganicheskikh Materialov, Moscow (USSR); Gosudarstvennyj Komitet po Ispol' zovaniyu Atomnoj Ehnergii SSSR, Moscow)

    1983-12-01

    In WWER and RBMK reactors now in operation a good quality of primary coolant is achieved and the required corrosion resistance of structural materials and normal irradiation conditions are ensured. Data on commercial fuel operation and clad material (Zr 1% Nb alloy) condition are briefly generalized. Some results of reactor investigations of corrosion behaviour of commercial Zr 1% Nb alloy under the condition of WWER and RBMK coolant are discussed and compared. It is established that the chemical effect of coolant on fuel cladding does not in itself limit its serviceability at design burn-ups but due to the possible processes of crud formation, corrosion (total and local), fretting-corrosion and hydriding it can influence the fuel reliability. This influence is qualitatively assessed through a rise in the clad temperature, a reduction of material plasticity and clad thickness.

  13. Material and water chemistry for a ferritic reactor coolant system in pressure water reactors

    International Nuclear Information System (INIS)

    Stieding, L.

    1979-04-01

    The use of unplated, low-alloy steels in a boric acid controlled PWR is not considered possible without changing the water conditions during the start-up and shut-down periods of the reactor. The significant pH reduction of the water due to boric acid during these periods most probably leads to damage of the magnetite protective layers followed by selective corrosion. As this highly important process has not been sufficiently evaluated with respect to our specific application problem, more detailed information will be necessary. KWU test facilities provide a means of performing such tests. In order to avoid corrosion attack during the above operating conditions, an inhibition of the water with 7 Li-borate is recommended which, however, will amount to approx. DM 60.000,-- per period of use. (orig.) [de

  14. Reducing the fuel temperature for pressure-tube supercritical-water-cooled reactors and the effect of fuel burnup

    Energy Technology Data Exchange (ETDEWEB)

    Nichita, E., E-mail: eleodor.nichita@uoit.ca; Kovaltchouk, V., E-mail: vitali.kovaltchouk@uoit.ca

    2015-12-15

    Highlights: • Typical PT-SCWR fuel uses single-region pins consisting of a homogeneous mixture of ThO{sub 2} and PuO{sub 2}. • Using two regions (central for the ThO{sub 2} and peripheral for the PuO{sub 2}) reduces the fuel temperature. • Single-region-pin melting-to-average power ratio is 2.5 at 0.0 MW d/kg and 2.3 at 40 MW d/kg. • Two-region-pin melting-to-average power ratio is 36 at 0.0 MW d/kg and 10.5 at 40 MW d/kg. • Two-region-pin performance drops with burnup due to fissile-element buildup in the ThO{sub 2} region. - Abstract: The Pressure-Tube Supercritical-Water-Cooled Reactor (PT-SCWR) is one of the concepts under investigation by the Generation IV International Forum for its promise to deliver higher thermal efficiency than nuclear reactors currently in operation. The high coolant temperature (>625 K) and high linear power density employed by the PT-SCWR cause the fuel temperature to be fairly high, leading to a reduced margin to fuel melting, thus increasing the risk of actual melting during accident scenarios. It is therefore desirable to come up with a fuel design that lowers the fuel temperature while preserving the high linear power ratio and high coolant temperature. One possible solution is to separate the fertile (ThO{sub 2}) and fissile (PuO{sub 2}) fuel materials into different radial regions in each fuel pin. Previously-reported work found that by locating the fertile material at the centre and the fissile material at the periphery of the fuel pin, the fuel centreline temperature can be reduced by ∼650 K for fresh fuel compared to the case of a homogeneous (Th–Pu)O{sub 2} mixture for the same coolant temperature and linear power density. This work provides a justification for the observed reduction in fuel centreline temperature and suggests a systematic approach to lower the fuel temperature. It also extends the analysis to the dependence of the radial temperature profile on fuel burnup. The radial temperature profile is

  15. Fundamental Understanding of Crack Growth in Structural Components of Generation IV Supercritical Light Water Reactors

    Energy Technology Data Exchange (ETDEWEB)

    Iouri I. Balachov; Takao Kobayashi; Francis Tanzella; Indira Jayaweera; Palitha Jayaweera; Petri Kinnunen; Martin Bojinov; Timo Saario

    2004-11-17

    This work contributes to the design of safe and economical Generation-IV Super-Critical Water Reactors (SCWRs) by providing a basis for selecting structural materials to ensure the functionality of in-vessel components during the entire service life. During the second year of the project, we completed electrochemical characterization of the oxide film properties and investigation of crack initiation and propagation for candidate structural materials steels under supercritical conditions. We ranked candidate alloys against their susceptibility to environmentally assisted degradation based on the in situ data measure with an SRI-designed controlled distance electrochemistry (CDE) arrangement. A correlation between measurable oxide film properties and susceptibility of austenitic steels to environmentally assisted degradation was observed experimentally. One of the major practical results of the present work is the experimentally proven ability of the economical CDE technique to supply in situ data for ranking candidate structural materials for Generation-IV SCRs. A potential use of the CDE arrangement developed ar SRI for building in situ sensors monitoring water chemistry in the heat transport circuit of Generation-IV SCWRs was evaluated and proved to be feasible.

  16. Numerical investigation of supercritical water-cooled nuclear reactor in horizontal rod bundles

    Energy Technology Data Exchange (ETDEWEB)

    Shang Zhi, E-mail: shangzhi@tsinghua.org.c [Faculty of Engineering, Kingston University, London SW15 3DW (United Kingdom); Science and Technology Facilities Council, Daresbury Laboratory, Warrington WA4 4AD (United Kingdom); Lo, Simon, E-mail: simon.lo@uk.cd-adapco.co [CD-adapco, Trident House, Basil Hill Road, Didcot OX11 7HJ (United Kingdom)

    2010-04-15

    The commercial CFD code STAR-CD v4.02 is used as a numerical simulation tool for flows in the supercritical water-cooled nuclear reactor (SCWR). The basic heat transfer element in the reactor core can be considered as round rods and rod bundles. Reactors with vertical or horizontal flow in the core can be found. In vertically oriented core, symmetric characters of flow and heat transfer can be found and two-dimensional analyses are often performed. However, in horizontally oriented core the flow and heat transfer are fully three-dimensional due to the buoyancy effect. In this paper, horizontal rods and rod bundles at SCWR conditions are studied. Special STAR-CD subroutines were developed by the authors to correctly represent the dramatic change in physical properties of the supercritical water with temperature. In the rod bundle simulations, it is found that the geometry and orientation of the rod bundle have strong effects on the wall temperature distributions and heat transfers. In one orientation the square bundle has a higher wall temperature difference than other bundles. However, when the bundles are rotated by 90 deg. the highest wall temperature difference is found in the hexagon bundle. Similar analysis could be useful in design and safety studies to obtain optimum fuel rod arrangement in a SCWR.

  17. Hazard classification for the supercritical water oxidation test bed. Revision 1

    International Nuclear Information System (INIS)

    Ramos, A.G.

    1994-10-01

    A hazard classification of ''routinely accepted by the public'' has been determined for the operation of the supercritical water oxidation test bed at the Idaho National Engineering Laboratory. This determination is based on the fact that the design and proposed operation meet or exceed appropriate national standards so that the risks are equivalent to those present in similar activities conducted in private industry. Each of the 17 criteria for hazards ''routinely accepted by the public,'' identified in the EG and G Idaho, Inc., Safety Manual, were analyzed. The supercritical water oxidation (SCWO) test bed will treat simulated mixed waste without the radioactive component. It will be designed to operate with eight test wastes. These test wastes have been chosen to represent a broad cross-section of candidate mixed wastes anticipated for storage or generation by DOE. In particular, the test bed will generate data to evaluate the ability of the technology to treat chlorinated waste and other wastes that have in the past caused severe corrosion and deposition in SCWO reactors

  18. Chemical recycling of carbon fibers reinforced epoxy resin composites in oxygen in supercritical water

    International Nuclear Information System (INIS)

    Bai, Yongping; Wang, Zhi; Feng, Liqun

    2010-01-01

    The carbon fibers in carbon fibers reinforced epoxy resin composites were recovered in oxygen in supercritical water at 30 ± 1 MPa and 440 ± 10 o C. The microstructure of the recovered carbon fibers was observed using scanning electron microscopy (SEM) and atom force microscopy (AFM). The results revealed that the clean carbon fibers were recovered and had higher tensile strength relative to the virgin carbon fibers when the decomposition rate was above 85 wt.%, although the recovered carbon fibers have clean surface, the epoxy resin on the surface of the recovered carbon fibers was readily observed. As the decomposition rate increased to above 96 wt.%, no epoxy resin was observed on the surface of the carbon fibers and the oxidation of the recovered carbon fibers was readily measured by X-ray photoelectron spectroscopy (XPS) analysis. The carbon fibers were ideally recovered and have original strength when the decomposition rates were between 94 and 97 wt.%. This study clearly showed the oxygen in supercritical water is a promising way for recycling the carbon fibers in carbon fibers reinforced resin composites.

  19. Corrosion behavior of oxide dispersion strengthened ferritic steels in supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Gao, Wenhua [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China); Guo, Xianglong, E-mail: guoxianglong@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China); Shen, Zhao [Department of Materials Science, University of Oxford, Parks Road, Oxford OX1 3PH (United Kingdom); Zhang, Lefu, E-mail: lfzhang@sjtu.edu.cn [School of Nuclear Science and Engineering, Shanghai Jiao Tong University, No. 800 Dongchuan Road, Shanghai 200240 (China)

    2017-04-01

    The corrosion resistance of three different Cr content oxide dispersion strengthened (ODS) ferritic steels in supercritical water (SCW) and their passive films formed on the surface have been investigated. The results show that the dissolved oxygen (DO) and chemical composition have significant influence on the corrosion behavior of the ODS ferritic steels. In 2000 ppb DO SCW at 650 °C, the 14Cr-4Al ODS steel forms a tri-layer oxide film and the surface morphologies have experienced four structures. For the tri-layer oxide film, the middle layer is mainly Fe-Cr spinel and the Al is gradually enriched in the inner layer. - Highlights: • We evaluated the corrosion resistance of three different Cr content ODS steels at 650 °C in supercritical water. • Corrosion behavior of ODS steels is rarely reported and ODS steel may be promising material for generation IV reactors. • We found total opposite phenomenon compared to Lee's work before. Our result may be more reasonable.

  20. Partial oxidation of landfill leachate in supercritical water: Optimization by response surface methodology

    International Nuclear Information System (INIS)

    Gong, Yanmeng; Wang, Shuzhong; Xu, Haidong; Guo, Yang; Tang, Xingying

    2015-01-01

    Highlights: • Partial oxidation of landfill leachate in supercritical water was investigated. • The process was optimized by Box–Behnken design and response surface methodology. • GY H2 , TRE and CR could exhibit up to 14.32 mmol·gTOC −1 , 82.54% and 94.56%. • Small amounts of oxidant can decrease the generation of tar and char. - Abstract: To achieve the maximum H 2 yield (GY H2 ), TOC removal rate (TRE) and carbon recovery rate (CR), response surface methodology was applied to optimize the process parameters for supercritical water partial oxidation (SWPO) of landfill leachate in a batch reactor. Quadratic polynomial models for GY H2 , CR and TRE were established with Box–Behnken design. GY H2 , CR and TRE reached up to 14.32 mmol·gTOC −1 , 82.54% and 94.56% under optimum conditions, respectively. TRE was invariably above 91.87%. In contrast, TC removal rate (TR) only changed from 8.76% to 32.98%. Furthermore, carbonate and bicarbonate were the most abundant carbonaceous substances in product, whereas CO 2 and H 2 were the most abundant gaseous products. As a product of nitrogen-containing organics, NH 3 has an important effect on gas composition. The carbon balance cannot be reached duo to the formation of tar and char. CR increased with the increase of temperature and oxidation coefficient

  1. CFD investigation of vertical rod bundles of supercritical water-cooled nuclear reactor

    International Nuclear Information System (INIS)

    Shang Zhi

    2009-01-01

    The commercial CFD code STAR-CD v4.02 is used as the numerical simulation tool for the supercritical water-cooled nuclear reactor (SCWR). The numerical simulation is based on the real full 3D rod bundles' geometry of the nuclear reactors. For satisfying the near-wall resolution of y + ≤ 1, the structure mesh with the stretched fine mesh near wall is employed. The validation of the numerical simulation for mesh generation strategy and the turbulence model for the heat transfer of supercritical water is carried out to compare with 3D tube experiments. After the validation, the same mesh generation strategy and the turbulence model are employed to study three types of the geometry frame of the real rod bundles. Through the numerical investigations, it is found that the different arrangement of the rod bundles will induce the different temperature distribution at the rods' walls. The wall temperature distributions are non-uniform along the wall and the values depend on the geometry frame. At the same flow conditions, downward flow gets higher wall temperature than upward flow. The hexagon geometry frame has the smallest wall temperature difference comparing with the others. The heat transfer is controlled by P/D ratio of the bundles.

  2. Supercritical water corrosion of high Cr steels and Ni-base alloys

    International Nuclear Information System (INIS)

    Jang, Jin Sung; Han, Chang Hee; Hwang, Seong Sik

    2004-01-01

    High Cr steels (9 to 12% Cr) have been widely used for high temperature high pressure components in fossil power plants. Recently the concept of SCWR (supercritical water-cooled reactor) has aroused a keen interest as one of the next generation (Generation IV) reactors. Consequently Ni-base (or high Ni) alloys as well as high Cr steels that have already many experiences in the field are among the potential candidate alloys for the cladding or reactor internals. Tentative inlet and outlet temperatures of the anticipated SCWR are 280 and 510 .deg. C respectively. Among many candidate alloys there are austenitic stainless steels, Ni base alloys, ODS alloys as well as high Cr steels. In this study the corrosion behavior of the high Cr steels and Ni base (or high Ni) alloys in the supercritical water were investigated. The corrosion behavior of the unirradiated base metals could be used in the near future as a guideline for the out-of-pile or in-pile corrosion evaluation tests

  3. One-phase and two-phase homologous curves for coolant pumps of the pressurized light water nuclear reactors

    International Nuclear Information System (INIS)

    Santos, G.A. dos.

    1990-01-01

    The two-phase coolant pump model of pressurized light water nuclear reactors is an important point for the loss of primary coolant accident analysis. The single-phase pump characteristics are an essential feature for operational transients studies, for example, the shut-down and start-up of pump. These parameters, in terms of the homologous curves, set up the complete performance of the pump and are input for transients and accidents analysis thermal-hydraulic codes. This work propose a mathematical model able to predict the single-phase and two-phase homologous curves where it was incorporated geometric and operational pump condition. The results were compared with the experimental tests data from literature and it has showed a good agreement. (author)

  4. A real-time tritium-in-water monitor for measurement of heavy water leak to the secondary coolant

    International Nuclear Information System (INIS)

    Rathnakaran, M.; Ravetkar, R.M.; Samant, R.K.; Abani, M.C.

    2000-01-01

    The paper describes the development and evaluation of on-line, real-time tritium in water monitor for detection and measurement of heavy water leak to the secondary coolant in a Pressurised Heavy Water Reactor. The detector used for this is a plastic scintillator film, made in the form of sponge and housed in a flow cell which is used for measurement of tritium activity present in heavy water. Two photomultiplier tubes are optically coupled on either face of the flow cell detector and measurement is done in coincidence mode. The sample water is continuously passed through the flow cell detector and a continuous measurement of tritium activity is carried out. It is observed that the impurities in the process water sample are gradually trapped in the flow cell, which affects the transparency of the detector with use. This reduces the sensitivity of the system. In addition, chlorine, which is added in the sample water, to arrest the fungus formation, creates chemiluminescence which interfere the measurement. To improve the sample quality as well as to eliminate the chemiluminescence created by chlorine, sample conditioner consisting of polypropylene candle, activated charcoal and glass fibre filter paper is developed. Polypropylene candle traps particulates above 5 μm pore size, activated charcoal absorbs organic compounds, free chlorine, fungus and turbidity and glass fibre filter paper stops submicron size particles. The measurement is also affected by the interference of dissolved argon-41 in the sample water. A bubbler system developed at BARC is used to strip the dissolved Ar-41 present in the sample which enables the system to measure tritium in presence of this interfering radioactive gas. The microprocessor based electronic system, used in the monitor provides the facility for selection of counting time and thereby improving the counting statistics. Alarm circuit is provided to give timely alarm when the tritium activity concentration exceeds the preset level

  5. Experimental study on heat transfer to supercritical water flowing in 1- and 4-m-long vertical tubes

    International Nuclear Information System (INIS)

    Kirillov, Pavel; Pomet'ko, Richard; Smirnov, Aleksandr; Grabezhnaia, Vera; Pioro, Igor; Duffey, Romney; Khartabil, Hussam

    2005-01-01

    This paper presents selected on heat transfer to supercritical water flowing upward in 1- and 4-m-long vertical tubes. Supercritical water heat-transfer data were obtained at pressures of 24-25 MPa, mass fluxes of 200 - 1500 kg/m 2 s, heat fluxes up to 1050 kW/m 2 and inlet temperature from 300 to 380degC for several combinations of wall and bulk fluid temperatures that were below, at or above the pseudocritical temperature. In general, the experiments confirmed that there are three heat transfer modes for water at supercritical pressures: (1) normal heat transfer characterized in general with heat transfer coefficients (HTCs) similar to those of subcritical convective heat transfer far from critical or pseudocritical regions, which are calculated according to the Dittus-Boelter type correlations, (2) deteriorated heat transfer with lower values of the HTC and hence higher values of wall temperature within some part of a test section compared to those of normal heat transfer and (3) improved heat transfer with higher values of the HTC and hence lower values of wall temperature within some part of a test section compared to those of normal heat transfer. These new heat-transfer data are applicable as a reference dataset for future comparison with supercritical water bundle data and for the verification of scaling parameters between water and modelling fluids. (author)

  6. Compartmentalized safety coolant injection system

    International Nuclear Information System (INIS)

    Johnson, F.T.

    1983-01-01

    A safety coolant injection system for nuclear reactors wherein a core reflood tank is provided to afford more reliable reflooding of the reactor core in the event of a break in one of the reactor coolant supply loops. Each reactor coolant supply loop is arranged in a separate compartment in the containment structure to contain and control the flow of spilled coolant so as to permit its use during emergency core cooling procedures. A spillway allows spilled coolant in the compartment to pass into the emergency water storage tank from where it can be pumped back to the reactor vessel. (author)

  7. Reduction of corrosion products in water coolant - basic way of increase in efficiency and improvement of ecological safety of NPU

    International Nuclear Information System (INIS)

    Prozorov, V.V.

    2004-01-01

    Corrosion of oxidated steel in water with additives of inhibitors or oxygen was considered. It is shown that preliminary oxidation of steel makes possible declining concentration of inhibitors or oxygen. Experiments demonstrate possibilities of the neutral-oxygen water regime for supply of the effective protection. Corrosion resistance of steel may be increased in many times through correct aqua-chemical regimes. Also concentration of corrosion products may be decreased in many times in coolant and their activation in neutron flux of nuclear reactor, amount of radioisotopes [ru

  8. Zinc corrosion after loss-of-coolant accidents in pressurized water reactors – Thermo- and fluid-dynamic effects

    Energy Technology Data Exchange (ETDEWEB)

    Seeliger, André, E-mail: a.seeliger@hszg.de [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Alt, Sören; Kästner, Wolfgang; Renger, Stefan [Hochschule Zittau/Görlitz, Institute of Process Technology, Process Automation and Measuring Technology, Theodor-Körner-Allee 16, D-02763 Zittau (Germany); Kryk, Holger; Harm, Ulrich [Helmholtz-Zentrum Dresden-Rossendorf, Institute of Fluid Dynamics, P.O. Box 510119, D-01314 Dresden (Germany)

    2016-08-15

    zinc compounds (mainly borates) were observed at the heatable zircaloy surfaces and characterized in detail during the heating-up to several coolant temperatures. As a strict consequence of their proven influence on heat removal and coolant flow behavior in the PWR core, preventive water-chemical methods were defined and tested.

  9. A comparison of the mechanisms of photooxidative degradation of organic molecules on irradiated semiconductor powders and in aerated supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Fox, Marye Anne [Department of Chemistry and Biochemistry, University of Texas at Austin, Austin, TX (United States)

    1995-08-01

    It is the purpose of this paper to survey evidence that suggests that control of the local environment is important in both heterogeneous TiO{sub 2} photocatalysis and in thermal oxidation reactions taking place in supercritical fluids, i.e. that the expected influences of these very different methods for microcompartmentalization do indeed influence the observed reaction kinetics in an easily observable way. Variations in reaction kinetics and the photophysical properties are described for (1) small semiconductor clusters, including their altered photocatalytic activity in and on inert supports; and (2) molecular probes dispersed within the self-aggregating clusters formed within supercritical water

  10. Formation of ZnO at zinc oxidation by near- and supercritical water under the constant electric field

    Science.gov (United States)

    Shishkin, A. V.; Sokol, M. Ya.; Shatrova, A. V.; Fedyaeva, O. N.; Vostrikov, A. A.

    2014-12-01

    The work has detected an influence of a constant electric field (up to E = 300 kV/m) on the structure of a nanocrystalline layer of zinc oxide, formed on the surface of a planar zinc anode in water under supercritical (673 K and 23 MPa) and near-critical (673 K and 17. 5 MPa) conditions. The effect of an increase of zinc oxidation rate with an increase in E is observed under supercritical conditions and is absent at near-critical ones. Increase in the field strength leads to the formation of a looser structure in the inner part of the zinc oxide layer.

  11. Study on corrosion behavior of candidate materials in 650℃ supercritical water

    International Nuclear Information System (INIS)

    Ma Shuli; Luo Ying; Zhang Qiang; Wang Hao; Qiu Shaoyu

    2014-01-01

    The general corrosion behavior of three candidate materials (347, HR3C and In-718) was investigated in 650 ℃/25 MPa deionized water. Morphology and composition of the surface oxide film with different exposure time were observed through FEG-SEM and EDS. The phase constitute was analyzed by GIXRD. For all the test materials, the weight loss follows typical parabolic law and the weight loss of 347 shows more than 40 times higher than that of HR3C and In-718. The oxide film of three alloys mainly consists of Ni(Cr, Fe) 2 O 4 . In-718 shows severe pitting and the oxide film of 347 appears significant spalling, while HR3C has compact oxide film. In the high temperature supercritical water, the high Cr content may enhance the general corrosion property of the alloys, while addition of Nb may be detrimental to the pitting resistance of alloys. (authors)

  12. Computational fluid dynamic model for glycerol gasification in supercritical water in a tee junction shaped cylindrical reactor

    NARCIS (Netherlands)

    Yukananto, Riza; Pozarlik, Artur K.; Brem, Gerrit

    2018-01-01

    Gasification in supercritical water is a very promising technology to process wet biomass into a valuable gas. Providing insight of the process behavior is therefore very important. In this research a computational fluid dynamic model is developed to investigate glycerol gasification in

  13. Heat-Transfer characteristics of Supercritical Water flowing upward in bare-tubes

    Energy Technology Data Exchange (ETDEWEB)

    Sidawi, K., E-mail: khalil.sidawi@uoit.ca [University of Ontario Institute of Technology, Faculty of Energy Systems and Nuclear Sciences, Oshawa, ON (Canada)

    2015-07-01

    There has been many correlations developed for Supercritical Water (SCW) flowing in bare-tubes. These correlations, generally, have limits based on the experimental trials. However, this does not indicate the true range to which these correlations can be applied. Furthermore, increases in heat flux and decreases in mass flux have been known to lead to Deteriorated Heat-Transfer (DHT). One way to classify fluids in the supercritical region is to use the Eckert Number to differentiate between two different sub-states{sup 1} ; when T < T{sub pc}, SCW is considered to be liquid-like, whereas at T > T{sub pc}, SCW is considered to be gas-like. There is a significant decrease in RMS error for calculated HTC in trials where there is a single sub-state across the cross-section. Trials where there is a combination of sub-states have drastically higher RMS error for HTC. Furthermore, some trials indicate a decrease in HTC at the interphase between the two sub-states. (author)

  14. Supercritical Water Gasification of Biomass in a Ceramic Reactor: Long-Time Batch Experiments

    Directory of Open Access Journals (Sweden)

    Daniele Castello

    2017-10-01

    Full Text Available Supercritical water gasification (SCWG is an emerging technology for the valorization of (wet biomass into a valuable fuel gas composed of hydrogen and/or methane. The harsh temperature and pressure conditions involved in SCWG (T > 375 °C, p > 22 MPa are definitely a challenge for the manufacturing of the reactors. Metal surfaces are indeed subject to corrosion under hydrothermal conditions, and expensive special alloys are needed to overcome such drawbacks. A ceramic reactor could be a potential solution to this issue. Finding a suitable material is, however, complex because the catalytic effect of the material can influence the gas yield and composition. In this work, a research reactor featuring an internal alumina inlay was utilized to conduct long-time (16 h batch tests with real biomasses and model compounds. The same experiments were also conducted in batch reactors made of stainless steel and Inconel 625. The results show that the three devices have similar performance patterns in terms of gas production, although in the ceramic reactor higher yields of C2+ hydrocarbons were obtained. The SEM observation of the reacted alumina surface revealed a good resistance of such material to supercritical conditions, even though some intergranular corrosion was observed.

  15. Research and development of supercritical water-cooled reactor (SCWR) in Japan

    International Nuclear Information System (INIS)

    Yamada, Katsumi; Oka, Yoshiaki

    2005-01-01

    The SCWR is an innovative LWR operating at supercritical pressure with a once-through direct cycle. It has the potential advantage of low capital cost due to its high thermal efficiency and substantial plant system simplifications. This paper outlines the completed and on-going R and D in Japan, and describes plans of the next phase projects for SCWR development. The concept was born at the University of Tokyo fifteen years ago. After a feasibility study by an industry team, a project for key technology development and plant conceptual design was launched in fiscal year (FY) 2000 funded by METI, followed by another project for fundamental study on supercritical water chemistry under radiation field and an I-NERI project for material development, and was completed in FY 2004 presenting an SCWR plant concept. To advance and optimize the plant concept, a new project is proposed in Japan. In addition, another project for developing the SCWR with fast spectrum core is proposed. The SCWR concept has acquired worldwide interest and was selected as one of the six Generation IV nuclear energy systems under GIF Program in FY 2002, and international collaboration for the SCWR RD and D is being established with an aggressive target of constructing a prototype reactor in the next fifteen years. The projects in Japan are expected to promote the development of the SCWR and to contribute the GIF activities. (author)

  16. Improvement of lifetime availability through design, inspection, repair and replacement of coolant channels of Indian Pressurized Heavy Water Reactors

    International Nuclear Information System (INIS)

    Rupani, B.B.; Sinha, R.K.

    1998-01-01

    This paper covers an overview of the work carried out for the life management of the coolant channels of Indian Pressurised Heavy Water Reactors. In order to improve maintainability of the coolant channels and reduce down time needed for periodical creep adjustment, improved design of channel hardware were developed. The modular insulation panel, designed as a substitute for the jig saw panels, reduces the time needed for accessing the space around the end-fitting significantly. A compact mechanical snubber has been developed to totally eliminate the need for periodic creep adjustment. In addition, the paper also describes the technologies developed for performing some special inspection, repair and replacement tasks for the coolant channels. These include systems for garter spring repositioning by Mechanical Flexing Technique for fresh reactors and Integrated Garter Spring Repositioning System for operating reactors. A tooling system, developed for in-situ retrieval of sliver scrape samples from pressure tubes, is also described. These samples can be analysed in laboratories to yield valuable information on hydrogen concentration in pressure tube material. The current and planned activities towards development of technologies for improvement of the life time availability of the power plants are addressed. (author)

  17. Contribution to the optimization of the chemical and radiochemical purification of pressurized water nuclear power plants primary coolant

    International Nuclear Information System (INIS)

    Elain, L.

    2004-12-01

    The primary coolant of pressurised water reactors is permanently purified thanks to a device, composed of filters and the demineralizers furnished with ion exchange resins (IER), located in the chemical and volume control system (CVCS). The study of the retention mechanisms of the radio-contaminants by the IER implies, initially, to know the speciation of the primary coolant percolant through the demineralizers. Calculations of theoretical speciation of the primary coolant were carried out on the basis of known composition of the primary coolant and thanks to the use of an adapted chemical speciation code. A complementary study, dedicated to silver behaviour, considered badly extracted, suggests metallic aggregates existence generated by the radiolytic reduction of the Ag + ions. An analysis of the purification curves of the elements Ni, Fe, Co, Cr, Mn, Sb and their principal radionuclides, relating to the cold shutdown of Fessenheim 1-cycle 20 and Tricastin 2-cycle 21, was carried out, in the light of a model based on the concept of a coupling well term - source term. Then, a thermodynamic modelling of ion exchange phenomena in column was established. The formation of the permutation front and the enrichment zones planned was validated by frontal analysis experiments of synthetic fluids (mixtures of Ni(B(OH) 4 ) 2 , LiB(OH) 4 and AgB(OH) 4 in medium B(OH) 3 )), and of real fluid during the putting into service of the device mini-CVCS at the time of Tricastin 2 cold shutdown. New tools are thus proposed, opening the way with an optimised management of demineralizers and a more complete interpretation of the available experience feedback. (author)

  18. Q-factor of coolant flow in the primary circuit of NPP with pressurised water reactors

    International Nuclear Information System (INIS)

    Proskuryakov, K.N.; Belikov, S.O.; Novikov, K.S.

    2011-01-01

    Systems of preoperational vibration dynamic monitoring in of WWER are presented. The results of measurements during commission of NPP with WWER are presented. The paper provides the result of the research, that estimation of coolant fluctuations caused by pulse perturbation of pressure in the primary circuit NPP. It is shown that results could be received at known value of a Q - factor of acoustical oscillatory system only. The research demonstrates the results of dependence of the sound speed from the mass steam content in the coolant flow thru reactor core. The worked out results can be used for identification of the reasons of abnormal growth of level of vibrations of fuel assembly, fuel rod, equipment and internals, and for forecasting the operation conditions which provide of vibration - acoustical resonances in the primary loop equipment. (author)

  19. Research on loss of coolant accident of pressurized-water reactor based on PSO algorithm

    International Nuclear Information System (INIS)

    Ma Jie; Guo Lifeng; Peng Qiao

    2012-01-01

    In order to improve the diagnosis performance of Loss of Coolant Accident (LOCA), based on Back Propagation (BP) algorithm study, a fault diagnosis network is established based on Particle Swarm Optimization (PSO) algorithm in this paper. The PSO algorithm is used to train the weights and the thresholds of neural network, which can conquer part convergence problem of BP algorithm. The test results show that the diagnosis network has higher accuracy of LOCA. (authors)

  20. Behaviour of a pressurized-water reactor nuclear power plant during loss-of-coolant accident

    International Nuclear Information System (INIS)

    Adam, E.; Carl, H.; Kubis, K.

    1979-01-01

    Starting from the foundation of the design basis accident in a PWR-type nuclear power plant - Loss of Coolant Accident -the actual status of the processes to be expected in the reactor are described. Operating behaviour of the heat removal system and efficiency of the safety systems are evaluated. Final considerations are concerned with the overall behaviour of the plant under such conditions. Probable failures, shut down times and possibilities of repair are estimated. (author)

  1. Life cycle assessment of hydrogen and power production by supercritical water reforming of glycerol

    International Nuclear Information System (INIS)

    Galera, S.; Gutiérrez Ortiz, F.J.

    2015-01-01

    Highlights: • The environmental performance of the supercritical water reforming (SCWR) of glycerol was assessed. • Biogenic CO 2 emissions allowed quantifying a realistic GHG inventory of 3.8 kg CO 2 -eq/kg H 2 . • The environmental profile of SCWR process was compared to those of other technologies. • A good environmental performance of H 2 and power production by SCWR of glycerol was obtained. - Abstract: The environmental performance of hydrogen and electricity production by supercritical water reforming (SCWR) of glycerol was evaluated following a Life Cycle Assessment (LCA) approach. The heat-integrated process was designed to be energy self-sufficient. Mass and energy balances needed for the study were performed using Aspen Plus 8.4, and the environmental assessment was carried out through SimaPro 8.0. CML 2000 was selected as the life cycle impact assessment method, considering as impact categories the global warming, ozone layer depletion, abiotic depletion, photochemical oxidant formation, eutrophication, acidification, and cumulative energy demand. A distinction between biogenic and fossil CO 2 emissions was done to quantify a more realistic GHG inventory of 3.77 kg CO 2 -eq per kg H 2 produced. Additionally, the environmental profile of SCWR process was compared to other H 2 production technologies such as steam methane reforming, carbon gasification, water electrolysis and dark fermentation among others. This way, it is shown that SCWR of glycerol allows reducing greenhouse gas emissions and obtaining a favorable positive life cycle energy balance, achieving a good environmental performance of H 2 and power production by SCWR of glycerol

  2. Transient simulation of coolant peak temperature due to prolonged fan and/or water pump operation after the vehicle is keyed-off

    Science.gov (United States)

    Pang, Suh Chyn; Masjuki, Haji Hassan; Kalam, Md. Abul; Hazrat, Md. Ali

    2014-01-01

    Automotive designers should design a robust engine cooling system which works well in both normal and severe driving conditions. When vehicles are keyed-off suddenly after some distance of hill-climbing driving, the coolant temperature tends to increase drastically. This is because heat soak in the engine could not be transferred away in a timely manner, as both the water pump and cooling fan stop working after the vehicle is keyed-off. In this research, we aimed to visualize the coolant temperature trend over time before and after the vehicles were keyed-off. In order to prevent coolant temperature from exceeding its boiling point and jeopardizing engine life, a numerical model was further tested with prolonged fan and/or water pump operation after keying-off. One dimensional thermal-fluid simulation was exploited to model the vehicle's cooling system. The behaviour of engine heat, air flow, and coolant flow over time were varied to observe the corresponding transient coolant temperatures. The robustness of this model was proven by validation with industry field test data. The numerical results provided sensible insights into the proposed solution. In short, prolonging fan operation for 500 s and prolonging both fan and water pump operation for 300 s could reduce coolant peak temperature efficiently. The physical implementation plan and benefits yielded from implementation of the electrical fan and electrical water pump are discussed.

  3. Safety analysis of a high temperature supercritical pressure light water cooled and moderated reactor

    International Nuclear Information System (INIS)

    Ishiwatari, Y.; Oka, Y.; Koshizuka, S.

    2002-01-01

    A safety analysis code for a high temperature supercritical pressure light water cooled reactor (SCLWR-H) with water rods cooled by descending flow, SPRAT-DOWN, is developed. The hottest channel, a water rod, down comer, upper and lower plenums, feed pumps, etc. are modeled as junction of nodes. Partial of the feed water flows downward from the upper dome of the reactor pressure vessel to the water rods. The accidents analyzed here are total loss of feed water flow, feed water pump seizure, and control rods ejection. All the accidents satisfy the criteria. The accident event at which the maximum cladding temperature is the highest is total loss of feedwater flow. The transients analyzed here are loss of feed water heating, inadvertent start-up of an auxiliary water supply system, partial loss of feed water flow, loss of offsite power, loss of load, and abnormal withdrawal of control rods. All the transients satisfied the criteria. The transient event for which the maximum cladding temperature is the highest is control rod withdrawal at normal operation. The behavior of loss of load transient is different from that of BWR. The power does not increase because loss of flow occurs and the density change is small. The sensitivities of the system behavior to various parameters during transients and accidents are analyzed. The parameters having strong influence are the capacity of the auxiliary water supply system, the coast down time of the main feed water pumps, and the time delay of the main feed water pumps trip. The control rod reactivity also has strong influence. (authors)

  4. In reactor performance of defected zircaloy-clad U3Si fuel elements in pressurized and boiling water coolants

    International Nuclear Information System (INIS)

    Feraday, M.A.; Allison, G.M.; Ambler, J.F.R.; Chalder, G.H.; Lipsett, J.J.

    1968-05-01

    The results of two in-reactor defect tests of Zircaloy-clad U 3 Si are reported. In the first test, a previously irradiated element (∼5300 MWd/ tonne U) was defected then exposed to first pressurized water then boiling water at ∼270 o C. In the second test, an unirradiated element containing a central void was defected, waterlogged, then exposed to pressurized water for 50 minutes. Both tests were terminated because of high activity in the loop coolant detected by both gamma and delayed neutron monitors. Post-irradiation examination showed that both elements had suffered major sheath failures which were attributed to the volume increase accompanying the formation of large quantities of corrosion product formed by the reaction of water with the hot central part of the fuel. It was concluded that the corrosion resistance of U 3 Si at 300 o C is not seriously affected by irradiation, but the corrosion rate increases rapidly with temperature. (author)

  5. Preliminary phenomena identification and ranking tables for simplified boiling water reactor Loss-of-Coolant Accident scenarios

    International Nuclear Information System (INIS)

    Kroeger, P.G.; Rohatgi, U.S.; Jo, J.H.; Slovik, G.C.

    1998-04-01

    For three potential Loss-of-Coolant Accident (LOCA) scenarios in the General Electric Simplified Boiling Water Reactors (SBWR) a set of Phenomena Identification and Ranking Tables (PIRT) is presented. The selected LOCA scenarios are typical for the class of small and large breaks generally considered in Safety Analysis Reports. The method used to develop the PIRTs is described. Following is a discussion of the transient scenarios, the PIRTs are presented and discussed in detailed and in summarized form. A procedure for future validation of the PIRTs, to enhance their value, is outlined. 26 refs., 25 figs., 44 tabs

  6. Linear titration plot for the determination of boron in the primary coolant of a pressurized water reactor

    International Nuclear Information System (INIS)

    Midgley, D.; Gatford, C.

    1992-01-01

    A linear titration plot method has been devised for the determination of boron as boric acid in partly neutralized solution, such as occurs in the primary coolant of pressurized water reactors. The total boron and the alkali in the sample are determined simultaneously. Although it is not essential to add mannitol in this method, it is more accurate when the solution is saturated with mannitol. Comparisons are made with other modes of titration: Gran plots, first and second differential potentiometric titrations and indicator titrations. None of these gives the total boron directly in partly neutralized solutions. (author)

  7. Nonlinear dynamic response analysis in piping system for a loss of coolant accident in primary loop of pressurized water reactor

    International Nuclear Information System (INIS)

    Zhang Xiwen; He Feng; Hao Pengfei; Wang Xuefang

    2000-01-01

    Based on the elaborate force and moment analysis with characteristics method and control-volume integrating method for the piping system of primary loop under pressurized water reactor' loss of coolant accident (LOCA) conditions, the nonlinear dynamic response of this system is calculated by the updated Lagrangian formulation (ADINA code). The piping system and virtual underpinning are specially processed, the move displacement of the broken pipe with time is accurately acquired, which is very important and useful for the design of piping system and virtual underpinning

  8. Unique rod lens/video system designed to observe flow conditions in emergency core coolant loops of pressurized water reactors

    International Nuclear Information System (INIS)

    Carter, G.W.

    1979-01-01

    Techniques and equipment are described which are used for video recordings of the single- and two-phase fluid flow tests conducted with the PKL Spool Piece Measurement System designed by Lawrence Livermore Laboratory and EG and G Inc. The instrumented spool piece provides valuable information on what would happen in pressurized water reactor emergency coolant loops should an accident or rupture result in loss of fluid. The complete closed-circuit television video system, including rod lens, light supply, and associated spool mounting fixtures, is discussed in detail. Photographic examples of test flows taken during actual spool piece system operation are shown

  9. Numerical investigation of heat transfer in upward flows of supercritical water in circular tubes and tight fuel rod bundles

    International Nuclear Information System (INIS)

    Yang Jue; Oka, Yoshiaki; Ishiwatari, Yuki; Liu Jie; Yoo, Jaewoon

    2007-01-01

    Heat transfer in upward flows of supercritical water in circular tubes and in tight fuel rod bundles is numerically investigated by using the commercial CFD code STAR-CD 3.24. The objective is to have more understandings about the phenomena happening in supercritical water and for designs of supercritical water cooled reactors. Some turbulence models are selected to carry out numerical simulations and the results are compared with experimental data and other correlations to find suitable models to predict heat transfer in supercritical water. The comparisons are not only in the low bulk temperature region, but also in the high bulk temperature region. The two-layer model (Hassid and Poreh) gives a better prediction to the heat transfer than other models, and the standard k-ε high Re model with the standard wall function also shows an acceptable predicting capability. Three-dimensional simulations are carried out in sub-channels of tight square lattice and triangular lattice fuel rod bundles at supercritical pressure. Results show that there is a strong non-uniformity of the circumferential distribution of the cladding surface temperature, in the square lattice bundle with a small pitch-to-diameter ratio (P/D). However, it does not occur in the triangular lattice bundle with a small P/D. It is found that this phenomenon is caused by the large non-uniformity of the flow area in the cross-section of sub-channels. Some improved designs are numerically studied and proved to be effective to avoid the large circumferential temperature gradient at the cladding surface

  10. Superior corrosion resistance properties of TiN-based coatings on Zircaloy tubes in supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Khatkhatay, Fauzia [Department of Electrical and Computer Engineering, Texas A and M University, College Station, TX 77843-3128 (United States); Jiao, Liang [Materials Science and Engineering Program, Texas A and M University, College Station, TX 77843-3003 (United States); Jian, Jie [Department of Electrical and Computer Engineering, Texas A and M University, College Station, TX 77843-3128 (United States); Zhang, Wenrui [Materials Science and Engineering Program, Texas A and M University, College Station, TX 77843-3003 (United States); Jiao, Zhijie [Nuclear Engineering and Radiological Sciences, University of Michigan, Ann Arbor, MI 48109-2104 (United States); Gan, Jian; Zhang, Hongbin [Idaho National Laboratory, Idaho Falls, ID 83415-6188 (United States); Zhang, Xinghang [Materials Science and Engineering Program, Texas A and M University, College Station, TX 77843-3003 (United States); Department of Mechanical Engineering, Texas A and M University, College Station, TX 77843-3123 (United States); Wang, Haiyan, E-mail: wangh@ece.tamu.edu [Department of Electrical and Computer Engineering, Texas A and M University, College Station, TX 77843-3128 (United States); Materials Science and Engineering Program, Texas A and M University, College Station, TX 77843-3003 (United States)

    2014-08-01

    Thin films of TiN and Ti{sub 0.35}Al{sub 0.65}N nanocomposite were deposited on polished Zircaloy-4 tubes. After exposure to supercritical water for 48 h, the coated tubes are remarkably intact, while the bare uncoated tube shows severe oxidation and breakaway corrosion. X-ray diffraction patterns, secondary electron images, backscattered electron images, and energy dispersive X-ray spectroscopy data from the tube surfaces and cross-sections show that a protective oxide, formed on the film surface, effectively prevents further oxidation and corrosion to the Zircaloy-4 tubes. This result demonstrates the effectiveness of thin film ceramics as protective coatings under extreme environments.

  11. Analysing supercritical water reactor's (SCWR's) special safety systems using probabilistic tools

    International Nuclear Information System (INIS)

    Ituen, I.; Novog, D.R.

    2011-01-01

    The next generation of reactors, termed Generation IV, has very attractive features -- its superior safety characteristics, high thermal efficiency, and fuel cycle sustainability. A key element of the Generation IV designs is the improvement in safety, which in turn requires improvements in safety system performance and reliability, as well as a reduction in initiating event frequencies. This study compares the response of the systems important to safety in the CANDU-Supercritical Water Reactor to those of the generic CANDU under a main steamline break accident and loss of forced circulation events -- to quantify the improvements in safety for the pre-conceptual CANDU SCWR design. Probabilistic safety analysis is the tool used in this study to test the behavior of the pre- conceptual design during these events. (author)

  12. Optimization of the fuel assembly for the Canadian SuperCritical Water-cooled Reactor (SCWR)

    Energy Technology Data Exchange (ETDEWEB)

    French, C., E-mail: Corey.French@cnsc-ccsn.gc.ca [Canadian Nuclear Safety Commission, Ottawa, Ontario (Canada); Bonin, H.; Chan, P.K. [Royal Military College of Ontario, Kingston, Ontario (Canada)

    2013-07-01

    An approach to develop a parametric optimization tool to support the Canadian Supercritical Water-cooled Reactor (SCWR) fuel design is presented in this work. The 2D benchmark lattices for 78-pin and 64-pin fuel assemblies are used as the initial models from which fuel performance and subsequent optimization stem from. A tandem optimization procedure is integrated which employs the steepest descent method. The physics codes WIMS-AECL, MCNP6 and SERPENT are used to calculate and verify select performance factors. The results are used as inputs to an optimization algorithm that yield optimal fresh fuel isotopic composition and lattice geometry. Preliminary results on verifications of infinite lattice reactivity are demonstrated in this paper. (author)

  13. A Review of Laboratory-Scale Research on Upgrading Heavy Oil in Supercritical Water

    Directory of Open Access Journals (Sweden)

    Ning Li

    2015-08-01

    Full Text Available With the growing demand for energy and the depletion of conventional crude oil, heavy oil in huge reserve has attracted extensive attention. However, heavy oil cannot be directly refined by existing processes unless they are upgraded due to its complex composition and high concentration of heteroatoms (N, S, Ni, V, etc.. Of the variety of techniques for heavy oil upgrading, supercritical water (SCW is gaining popularity because of its excellent ability to convert heavy oil into valued, clean light oil by the suppression of coke formation and the removal of heteroatoms. Based on the current status of this research around the world, heavy oil upgrading in SCW is summarized from three aspects: Transformation of hydrocarbons, suppression of coke, and removal of heteroatoms. In this work, the challenge and future development of the orientation of upgrading heavy oil in SCW are pointed out.

  14. Hydrogen production by supercritical water gasification of wastewater from food waste treatment processes

    Energy Technology Data Exchange (ETDEWEB)

    Lee, In-Gu [Korea Institute of Energy Research (Korea, Republic of)

    2010-07-01

    Korean food wastes have high moisture content (more than 85 wt%) and their major treatment processes such as drying or biological fermentations generate concentrated organic wastewater (CODs of about 100,000 mgO{sub 2}/L). For obtaining both wastewater treatment and hydrogen production from renewable resources, supercritical water gasification (SCWG) of the organic wastewater was carried out in this work. The effect of catalyst, reaction temperature, and reactor residence time on COD destruction and composition of gas products was examined. As a result, a SCWG of the wastewater over Ni- Y/activated charcoal at 700 C, 28 MPa yielded 99 % COD destruction and hydrogen-rich gas production (45 vol% H{sub 2}). A liquid-phase thermal pretreatment to destroy solid particles from the wastewater was proposed for more effective operation of the SCWG system. (orig.)

  15. Oxidation kinetics of model compounds of metabolic waste in supercritical water

    Science.gov (United States)

    Webley, Paul A.; Holgate, Henry R.; Stevenson, David M.; Tester, Jefferson W.

    1990-01-01

    In this NASA-funded study, the oxidation kinetics of methanol and ammonia in supercritical water have been experimentally determined in an isothermal plug flow reactor. Theoretical studies have also been carried out to characterize key reaction pathways. Methanol oxidation rates were found to be proportional to the first power of methanol concentration and independent of oxygen concentration and were highly activated with an activation energy of approximately 98 kcal/mole over the temperature range 480 to 540 C at 246 bar. The oxidation of ammonia was found to be catalytic with an activation energy of 38 kcal/mole over temperatures ranging from 640 to 700 C. An elementary reaction model for methanol oxidation was applied after correction for the effect of high pressure on the rate constants. The conversion of methanol predicted by the model was in good agreement with experimental data.

  16. Design and study of water supply system for supercritical unit boiler in thermal power station

    Science.gov (United States)

    Du, Zenghui

    2018-04-01

    In order to design and optimize the boiler feed water system of supercritical unit, the establishment of a highly accurate controlled object model and its dynamic characteristics are prerequisites for developing a perfect thermal control system. In this paper, the method of mechanism modeling often leads to large systematic errors. Aiming at the information contained in the historical operation data of the boiler typical thermal system, the modern intelligent identification method to establish a high-precision quantitative model is used. This method avoids the difficulties caused by the disturbance experiment modeling for the actual system in the field, and provides a strong reference for the design and optimization of the thermal automation control system in the thermal power plant.

  17. Corrosion mechanism of a Ni-based alloy in supercritical water: Impact of surface plastic deformation

    International Nuclear Information System (INIS)

    Payet, Mickaël; Marchetti, Loïc; Tabarant, Michel; Chevalier, Jean-Pierre

    2015-01-01

    Highlights: • The dissolution of Ni and Fe cations occurs during corrosion of Ni-based alloys in SCW. • The nature of the oxide layer depends locally on the alloy microstructure. • The corrosion mechanism changes when cold-work increases leading to internal oxidation. - Abstract: Ni–Fe–Cr alloys are expected to be a candidate material for the generation IV nuclear reactors that use supercritical water at temperatures up to 600 °C and pressures of 25 MPa. The corrosion resistance of Alloy 690 in these extreme conditions was studied considering the surface finish of the alloy. The oxide scale could suffer from dissolution or from internal oxidation. The presence of a work-hardened zone reveals the competition between the selective oxidation of chromium with respect to the oxidation of nickel and iron. Finally, corrosion mechanisms for Ni based alloys are proposed considering the effects of plastically deformed surfaces and the dissolution.

  18. CFD in supercritical water-cooled nuclear reactor (SCWR) with horizontal tube bundles

    Energy Technology Data Exchange (ETDEWEB)

    Zhi Shang, E-mail: zhi.shang@stfc.ac.uk [Science and Technology Facilities Council, Daresbury Laboratory, Warrington WA4 4AD (United Kingdom); Lo, Simon, E-mail: simon.lo@uk.cd-adapco.com [CD-adapco, Trident House, Basil Hill Road, Didcot OX11 7HJ (United Kingdom)

    2011-11-15

    The commercial CFD code STAR-CD 4.02 is used as a numerical simulation tool for flows in the supercritical water-cooled nuclear reactor (SCWR). The basic heat transfer element in the reactor core can be considered as round tubes and tube bundles. Reactors with vertical or horizontal flow in the core can be found. In a vertically oriented core, symmetric characters of flow and heat transfer can be found and two-dimensional analyses are often performed. However, in a horizontally oriented core the flow and heat transfer are fully three-dimensional due to the buoyancy effect. In this paper, horizontal tubes and tube bundles at SCWR conditions are studied. Special STAR-CD subroutines were developed by the authors to correctly represent the dramatic change in physical properties of the supercritical water with temperature. From the study of single round tubes, the Speziale quadratic non-linear high-Re k-{epsilon} turbulence model with the two-layer model for near wall treatment is found to produce the best results in comparison with experimental data. In tube bundle simulations, it is found that the temperature is higher in the top half of the bundle and the highest tube wall temperature is located at the outside tubes where the flow rate is the lowest. The secondary flows across the bundle are highly complex. Their main effect is to even out the temperature over the area within each individual recirculation region. Similar analysis could be useful in design and safety studies to obtain optimum fuel rod arrangement in a SCWR.

  19. CFD in supercritical water-cooled nuclear reactor (SCWR) with horizontal tube bundles

    International Nuclear Information System (INIS)

    Shang, Zhi; Lo, Simon

    2009-01-01

    The commercial CFD code STAR-CD 4.02 is used as a numerical simulation tool for flows in the supercritical water-cooled nuclear reactor (SCWR). The basic heat transfer element in the reactor core can be considered as round tubes and tube bundles. Reactors with vertical or horizontal flow in the core can be found. In vertically oriented core, symmetric characters of flow and heat transfer can be found and two-dimensional analyses are often performed. However, in horizontally oriented core the flow and heat transfer are fully three-dimensional due to the buoyancy effect. In this paper, horizontal tubes and tube bundles at SCWR conditions are studied. Special STAR-CD subroutines were developed by the authors to correctly represent the dramatic change in physical properties of the supercritical water with temperature. From the study of single round tubes, the Speziale quadratic non-linear high-Re k-ε turbulence model with the two-layer model for near wall treatment is found to produce the best results in comparison with experimental data. In tube bundle simulations, it is found that the temperature is higher in the top half of the bundle and the highest tube wall temperature is located at the outside tubes where the flow rate is the lowest. The secondary flows across the bundle are highly complex. Their main effect is to even out the temperature over the area within each individual recirculating region. Similar analysis could be useful in design and safety studies to obtain optimum fuel rod arrangement in a SCWR. (author)

  20. Partial oxidation of landfill leachate in supercritical water: Optimization by response surface methodology

    Energy Technology Data Exchange (ETDEWEB)

    Gong, Yanmeng; Wang, Shuzhong; Xu, Haidong; Guo, Yang; Tang, Xingying

    2015-09-15

    Highlights: • Partial oxidation of landfill leachate in supercritical water was investigated. • The process was optimized by Box–Behnken design and response surface methodology. • GY{sub H2}, TRE and CR could exhibit up to 14.32 mmol·gTOC{sup −1}, 82.54% and 94.56%. • Small amounts of oxidant can decrease the generation of tar and char. - Abstract: To achieve the maximum H{sub 2} yield (GY{sub H2}), TOC removal rate (TRE) and carbon recovery rate (CR), response surface methodology was applied to optimize the process parameters for supercritical water partial oxidation (SWPO) of landfill leachate in a batch reactor. Quadratic polynomial models for GY{sub H2}, CR and TRE were established with Box–Behnken design. GY{sub H2}, CR and TRE reached up to 14.32 mmol·gTOC{sup −1}, 82.54% and 94.56% under optimum conditions, respectively. TRE was invariably above 91.87%. In contrast, TC removal rate (TR) only changed from 8.76% to 32.98%. Furthermore, carbonate and bicarbonate were the most abundant carbonaceous substances in product, whereas CO{sub 2} and H{sub 2} were the most abundant gaseous products. As a product of nitrogen-containing organics, NH{sub 3} has an important effect on gas composition. The carbon balance cannot be reached duo to the formation of tar and char. CR increased with the increase of temperature and oxidation coefficient.

  1. Estimation of Oxidation Kinetics and Oxide Scale Void Position of Ferritic-Martensitic Steels in Supercritical Water

    Directory of Open Access Journals (Sweden)

    Li Sun

    2017-01-01

    Full Text Available Exfoliation of oxide scales from high-temperature heating surfaces of power boilers threatened the safety of supercritical power generating units. According to available space model, the oxidation kinetics of two ferritic-martensitic steels are developed to predict in supercritical water at 400°C, 500°C, and 600°C. The iron diffusion coefficients in magnetite and Fe-Cr spinel are extrapolated from studies of Backhaus and Töpfer. According to Fe-Cr-O ternary phase diagram, oxygen partial pressure at the steel/Fe-Cr spinel oxide interface is determined. The oxygen partial pressure at the magnetite/supercritical water interface meets the equivalent oxygen partial pressure when system equilibrium has been attained. The relative error between calculated values and experimental values is analyzed and the reasons of error are suggested. The research results show that the results of simulation at 600°C are approximately close to experimental results. The iron diffusion coefficient is discontinuous in the duplex scale of two ferritic-martensitic steels. The simulation results of thicknesses of the oxide scale on tubes (T91 of final superheater of a 600 MW supercritical boiler are compared with field measurement data and calculation results by Adrian’s method. The calculated void positions of oxide scales are in good agreement with a cross-sectional SEM image of the oxide layers.

  2. Corrosion of low alloy steel containing 0.5% chromium in supercritical CO2-saturated brine and water-saturated supercritical CO2 environments

    Science.gov (United States)

    Wei, Liang; Gao, Kewei; Li, Qian

    2018-05-01

    The corrosion behavior of P110 low-Cr alloy steel in supercritical CO2-saturated brine (aqueous phase) and water-saturated supercritical CO2 (SC CO2 phase) was investigated. The results show that P110 steel primarily suffered general corrosion in the aqueous phase, while severe localized corrosion occurred in the SC CO2 phase. The formation of corrosion product scale on P110 steel in the aqueous phase divided into three stages: formation of the initial corrosion layer containing amorphous Cr(OH)3, FeCO3 and a small amount of Fe3C; transformation of initial corrosion layer to mixed layer, which consisted of FeCO3 and a small amount of Cr(OH)3 and Fe3C; growth and dissolution of the mixed layer. Finally, only a single mixed layer covered on the steel in the aqueous phase. However, the scale formed in SC CO2 phase consisted of two layers: the inner mixed layer and the dense outer FeCO3 crystalline layer.

  3. Loss of Coolant Accident (LOCA) / Emergency Core Coolant System (ECCS Evaluation of Risk-Informed Margins Management Strategies for a Representative Pressurized Water Reactor (PWR)

    Energy Technology Data Exchange (ETDEWEB)

    Szilard, Ronaldo Henriques [Idaho National Lab. (INL), Idaho Falls, ID (United States)

    2016-09-01

    A Risk Informed Safety Margin Characterization (RISMC) toolkit and methodology are proposed for investigating nuclear power plant core, fuels design and safety analysis, including postulated Loss-of-Coolant Accident (LOCA) analysis. This toolkit, under an integrated evaluation model framework, is name LOCA toolkit for the US (LOTUS). This demonstration includes coupled analysis of core design, fuel design, thermal hydraulics and systems analysis, using advanced risk analysis tools and methods to investigate a wide range of results.

  4. Coolant Mixing in a Pressurized Water Reactor: Deboration Transients, Steam-Line Breaks, and Emergency Core Cooling Injection

    International Nuclear Information System (INIS)

    Prasser, Horst-Michael; Grunwald, Gerhard; Hoehne, Thomas; Kliem, Soeren; Rohde, Ulrich; Weiss, Frank-Peter

    2003-01-01

    The reactor transient caused by a perturbation of boron concentration or coolant temperature at the inlet of a pressurized water reactor (PWR) depends on the mixing inside the reactor pressure vessel (RPV). Initial steep gradients are partially lessened by turbulent mixing with coolant from the unaffected loops and with the water inventory of the RPV. Nevertheless the assumption of an ideal mixing in the downcomer and the lower plenum of the reactor leads to unrealistically small reactivity inserts. The uncertainties between ideal mixing and total absence of mixing are too large to be acceptable for safety analyses. In reality, a partial mixing takes place. For realistic predictions it is necessary to study the mixing within the three-dimensional flow field in the complicated geometry of a PWR. For this purpose a 1:5 scaled model [the Rossendorf Coolant Mixing Model (ROCOM) facility] of the German PWR KONVOI was built. Compared to other experiments, the emphasis was put on extensive measuring instrumentation and a maximum of flexibility of the facility to cover as much as possible different test scenarios. The use of special electrode-mesh sensors together with a salt tracer technique provided distributions of the disturbance within downcomer and core entrance with a high resolution in space and time. Especially, the instrumentation of the downcomer gained valuable information about the mixing phenomena in detail. The obtained data were used to support code development and validation. Scenarios investigated are the following: (a) steady-state flow in multiple coolant loops with a temperature or boron concentration perturbation in one of the running loops, (b) transient flow situations with flow rates changing with time in one or more loops, such as pump startup scenarios with deborated slugs in one of the loops or onset of natural circulation after boiling-condenser-mode operation, and (c) gravity-driven flow caused by large density gradients, e.g., mixing of cold

  5. SUPERCRITICAL WATER PARTIAL OXIDATION PHASE I - PILOT-SCALE TESTING / FEASIBILITY STUDIES FINAL REPORT

    Energy Technology Data Exchange (ETDEWEB)

    SPRITZER,M; HONG,G

    2005-01-01

    Under Cooperative Agreement No. DE-FC36-00GO10529 for the Department of Energy, General Atomics (GA) is developing Supercritical Water Partial Oxidation (SWPO) as a means of producing hydrogen from low-grade biomass and other waste feeds. The Phase I Pilot-scale Testing/Feasibility Studies have been successfully completed and the results of that effort are described in this report. The Key potential advantages of the SWPO process is the use of partial oxidation in-situ to rapidly heat the gasification medium, resulting in less char formation and improved hydrogen yield. Another major advantage is that the high-pressure, high-density aqueous environment is ideal for reaching and gasifying organics of all types. The high water content of the medium encourages formation of hydrogen and hydrogen-rich products and is especially compatible with high water content feeds such as biomass materials. The high water content of the medium is also effective for gasification of hydrogen-poor materials such as coal. A versatile pilot plant for exploring gasification in supercritical water has been established at GA's facilities in San Diego. The Phase I testing of the SWPO process with wood and ethanol mixtures demonstrated gasification efficiencies of about 90%, comparable to those found in prior laboratory-scale SCW gasification work carreid out at the University of Hawaii at Manoa (UHM) as well as other biomass gasification experience with conventional gasifiers. As in the prior work at UHM, a significant amount of the hydrogen found in the gas phase products is derived from the water/steam matrix. The studies at UHM utilized an indirectly heated gasifier with an acitvated carbon catalyst. In contrast, the GA studies utilized a directly heated gasifier without catalyst, plus a surrogate waste fuel. Attainment of comparable gasification efficiencies without catalysis is an important advancement for the GA process, and opens the way for efficient hydrogen production from low

  6. Reactor coolant cleanup device

    International Nuclear Information System (INIS)

    Igarashi, Noboru.

    1986-01-01

    Purpose: To enable to introduce reactor water at high temperature and high pressure as it is, as well as effectively adsorb to eliminate cobalt in reactor water. Constitution: The coolant cleanup device comprises a vessel main body inserted to coolant pipeway circuits in a water cooled reactor power plant and filters contained within the vessel main body. The filters are prepared by coating and baking powder of metal oxides such as manganese ferrite having a function capable of adsorbing cobalt in the coolants onto the surface of supports made of metals or ceramics resistant to strong acids and alkalies in the form of three-dimensional network structure, for example, zircaloy-2, SUS 303 and the zirconia (baking) to form a basic filter elements. The basic filter elements are charged in plurality to the vessel main body. (Kawaiami, Y.)

  7. A test facility for heat transfer, pressure drop and stability studies under supercritical conditions

    International Nuclear Information System (INIS)

    Sharma, Manish; Pilkhwal, D.S.; Jana, S.S.; Vijayan, P.K.

    2013-02-01

    Supercritical water (SCW) exhibits excellent heat transfer characteristics and high volumetric expansion coefficient (hence high mass flow rates in natural circulation systems) near pseudo-critical temperature. SCW is being considered as a coolant in some advanced nuclear reactor designs on account of its potential to offer high thermal efficiency, compact size, elimination of steam generator, separator and dryer, making it economically competitive. The elimination of phase change results in elimination of the Critical Heat Flux (CHF) phenomenon. Cooling a reactor at full power with natural instead of forced circulation is generally considered as enhancement of passive safety. In view of this, it is essential to study natural circulation, heat transfer and pressure drop characteristics of supercritical fluids. Carbon-dioxide can be considered to be a good simulant of water for natural circulation at supercritical conditions since the density and viscosity variation of carbon-dioxide follows a parallel curve as that of water at supercritical conditions. Hence, a supercritical pressure natural circulation loop (SPNCL) has been set up in Hall-7, BARC to investigate the heat transfer, pressure drop and stability characteristics of supercritical carbon-dioxide under natural circulation conditions. The details of the experimental facility are presented in this report. (author)

  8. Wollastonite Carbonation in Water-Bearing Supercritical CO2: Effects of Particle Size.

    Science.gov (United States)

    Min, Yujia; Li, Qingyun; Voltolini, Marco; Kneafsey, Timothy; Jun, Young-Shin

    2017-11-07

    The performance of geologic CO 2 sequestration (GCS) can be affected by CO 2 mineralization and changes in the permeability of geologic formations resulting from interactions between water-bearing supercritical CO 2 (scCO 2 ) and silicates in reservoir rocks. However, without an understanding of the size effects, the findings in previous studies using nanometer- or micrometer-size particles cannot be applied to the bulk rock in field sites. In this study, we report the effects of particle sizes on the carbonation of wollastonite (CaSiO 3 ) at 60 °C and 100 bar in water-bearing scCO 2 . After normalization by the surface area, the thickness of the reacted wollastonite layer on the surfaces was independent of particle sizes. After 20 h, the reaction was not controlled by the kinetics of surface reactions but by the diffusion of water-bearing scCO 2 across the product layer on wollastonite surfaces. Among the products of reaction, amorphous silica, rather than calcite, covered the wollastonite surface and acted as a diffusion barrier to water-bearing scCO 2 . The product layer was not highly porous, with a specific surface area 10 times smaller than that of the altered amorphous silica formed at the wollastonite surface in aqueous solution. These findings can help us evaluate the impacts of mineral carbonation in water-bearing scCO 2 .

  9. Molten fuel-coolant interactions resulting from power transients in aluminium plate/water moderated reactors

    International Nuclear Information System (INIS)

    Storr, G.J.

    1989-08-01

    The behaviour of two reactors SL1 and SPERT D12, which underwent fast nuclear power transients prior to core destruction by a molten fuel-coolant interaction (MFCI) has been analysed and the results compared with measured data. The calculated spatial melt distribution and the mechanical work done during the events leads to high (∼ 250 kJ/kg) conversion efficiencies for this type of interaction when compared with molten drop experiments. A simple model for the steam explosion, using static thermodynamic properties of high temperature and pressure steam is used to calculate the dynamics of the reactors following the MFCI. 26 refs., 5 figs., 5 tabs

  10. A universal salt model based on under-ground precipitation of solid salts due to supercritical water `out-salting'

    Science.gov (United States)

    Rueslåtten, H.; Hovland, M. T.

    2010-12-01

    One of the common characteristics of planets Earth and Mars is that both host water (H2O) and large accumulations of salt. Whereas Earth’s surface-environment can be regarded as ‘water-friendly’ and ‘salt hostile’, the reverse can be said for the surface of Mars. This is because liquid water is stable on Earth, and the atmosphere transports humidity around the globe, whereas on planet Mars, liquid water is unstable, rendering the atmosphere dry and, therefore, ‘salt-friendly’. The riddle as to how the salt accumulated in various locations on those two planets, is one of long-lasting and great debate. The salt accumulations on Earth are traditionally termed ‘evaporites’, meaning that they formed as a consequence of the evaporation of large masses of seawater. How the accumulations on Mars formed is much harder to explain, as an ocean only existed briefly. Although water molecules and OH-groups may exist in abundance in bound form (crystal water, adsorbed water, etc.), the only place where free water is expected to be stable on Mars is within underground faults, fractures, and crevices. Here it likely occurs as brine or in the form of ice. Based on these conditions, a key to understanding the accumulation of large deposits of salt on both planets is linked to how brines behave in the subsurface when pressurized and heated beyond their supercritical point. At depths greater than about 3 km (P>300 bars) water will no longer boil in a steam phase. Rather, it becomes supercritical and will attain the phase of supercritical water vapor (SCRIW) with a specific gravity of typically 0.3 g/cm3. An important characteristic of SCRIW is its inability to dissolve the common sea salts. The salt dissolved in the brines will therefore precipitate as solid particles when brines (seawater on the Earth) move into the supercritical P&T-domain (T>400°C, P>300 bars). Numerical modeling of a hydrothermal system in the Atlantis II Deep of the Red Sea indicates that a

  11. Stress corrosion cracking and oxidation of austenitic stainless steel 316 L and model alloy in supercritical water reactor

    International Nuclear Information System (INIS)

    Saez-Maderuelo, A.; Gomez-Briceno, D.; Diego, G.

    2015-01-01

    In this work, an austenitic stainless steel type 316 L was tested in deaerated supercritical water at 400 deg. C and 500 deg. C and 25 MPa to determine how variations in water conditions influence its stress corrosion cracking behaviour and to make progress in the understanding of mechanisms involved in SCC processes in this environment. Moreover, the influence of plastic deformation in the resistance of the material to SCC was also studied at both temperatures. In addition to this, previous oxidation experiments at 400 deg. C and 500 deg. C and at 25 MPa were taken into account to gain some insight in this kind of processes. Furthermore, a cold worked model alloy based on the stainless steel 316 L with some variations in the chemical composition in order to simulate the composition of the grain boundary after irradiation was tested at 400 deg. C and 25 MPa in deaerated supercritical water. (authors)

  12. Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

    Energy Technology Data Exchange (ETDEWEB)

    Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R. [Atomic Energy of Canada Limited, Chalk River, ON (Canada)

    2011-07-01

    This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

  13. Conceptual mechanical design for a pressure-tube type supercritical water-cooled reactor

    International Nuclear Information System (INIS)

    Yetisir, M.; Diamond, W.; Leung, L.K.H.; Martin, D.; Duffey, R.

    2011-01-01

    This paper presents a conceptual mechanical design for a heavy-water-moderated pressure-tube supercritical water (SCW) reactor, which has evolved from the well-established CANDU nuclear reactor. As in the current designs, the pressure-tube SCW reactor uses a calandria vessel and, as a result, many of today's technologies (such as the shutdown safety systems) can readily be adopted with small changes. Because the proposed concept uses a low-pressure moderator, it does not require a pressure vessel that is subject to the full SCW pressure and temperature conditions. The proposed design uses batch refueling and hence, the reactor core is orientated vertically. Significant simplifications result in the design with the elimination of on line fuelling systems, fuel channel end fittings and fuel channel closure seals and thus utilize the best features of Light Water Reactor (LWR) and Heavy Water Reactor (HWR) technologies. The safety goal is based on achieving a passive 'no core melt' configuration for the channels and core, so the mechanical features and systems directly reflect this desired attribute. (author)

  14. Stress corrosion cracking susceptibility of austenitic stainless steels in supercritical water conditions

    International Nuclear Information System (INIS)

    Novotny, R.; Haehner, P.; Ripplinger, S.; Siegl, J.; Penttilae, Sami; Toivonen, Aki

    2009-01-01

    Within the 6th Framework Program HPLWR-2 project (High Performance Light Water Reactor - Phase 2), stress corrosion cracking (SCC) susceptibilities of selected austenitic stainless steels, 316L and 316NG, were studied in supercritical water (SCW) with the aim to identify and describe the specific failure mechanisms prevailing during slow strain-rate tensile (SSRT) tests in ultra-pure demineralised SCW water solution. The SSRT tests were performed using a step-motor controlled loading device in an autoclave at 350 deg. C, 500 deg. C and 550 deg. C. Besides water temperature, the pressure, the oxygen content and the strain rate (resp. crosshead speed) were varied in the series of tests. The specimens SSRT tested to failure were subjected to fractographic analysis, in order to characterise the failure mechanisms. The fractography confirmed that failure was due to a combination of transgranular SCC and transgranular ductile fracture. The share of SCC and ductile fracture in the failure process of individual specimens was affected by the parameters of the SSRT tests, so that the environmental influence on SCC susceptibility could be assessed, in particular, the SCC sensitising effects of increasing oxygen content, decreasing strain rate and increasing test temperature. (author)

  15. Component evaluation for intersystem loss-of-coolant accidents in advanced light water reactors

    International Nuclear Information System (INIS)

    Ware, A.G.

    1994-07-01

    Using the methodology outlined in NUREG/CR-5603 this report evaluates (on a probabilistic basis) design rules for components in ALWRs that could be subjected to intersystem loss-of-coolant accidents (ISLOCAs). The methodology is intended for piping elements, flange connections, on-line pumps and valves, and heat exchangers. The NRC has directed that the design rules be evaluated for BWR pressures of 7.04 MPa (1025 psig), PWR pressures of 15.4 MPa (2235 psig), and 177 degrees C (350 degrees F), and has established a goal of 90% probability that system rupture will not occur during an ISLOCA event. The results of the calculations in this report show that components designed for a pressure of 0.4 of the reactor coolant system operating pressure will satisfy the NRC survival goal in most cases. Specific recommendations for component strengths for BWR and PWR applications are made in the report. A peer review panel of nationally recognized experts was selected to review and critique the initial results of this program

  16. Development of out-of-core concepts for a supercritical-water, pressure-tube reactor

    International Nuclear Information System (INIS)

    Diamond, W.T.

    2010-01-01

    One of the Generation IV programs at Chalk River Laboratories has as its prime focus the development of out-of-core concepts for the SuperCritical Water (SCW) pressure tube reactor under development in Canada. A number of technical issues associated with the interface of out-of-core components and the pressure tubes of a SCW pressure tube reactor are being investigated. This article focuses on several aspects of out-of-core components and layouts, building upon concepts that have been developed during the past few years. The efforts are strongly focused on concepts for a fuel channel that can be fabricated with the tight lattice pitch (typically 230 to 250 mm) that may be required for some applications such as utilization of a thorium fuel cycle. It is not practical to adapt concepts with a tight lattice pitch while using the thicker materials required for the higher temperatures and pressures required for supercritical operation. A change in lattice pitch or configuration is required to accommodate the component size increases. This presentation will cover a number of new concepts developed to produce feeders and end fittings for the harsh conditions of a SCW pressure tube reactor. These components are then developed into conceptual models of a Gen IV pressure tube reactor mounted in both horizontal and vertical orientations. Full 3-D solid models of both concepts will be demonstrated as well as a 1/10th-scale model of one face of a horizontal concept that has been built from components made with a 3-D printer. (author)

  17. Destruction of DOE/DP surrogate wastes with supercritical water oxidation technology

    International Nuclear Information System (INIS)

    Bramlette, T.T.; Mills, B.E.; Hencken, K.R.; Brynildson, M.E.; Johnston, S.C.; Hruby, J.M.; Freemster, H.C.; Odegard, B.C.; Modell, M.

    1990-11-01

    Surrogate wastes of specific interest to DOE/DP production facilities (Hanford and Rocky Flats), and the electronics industry have been successfully processed in a laboratory-scale, supercritical water oxidation flow reactor. In all cases, the observed destruction/reduction efficiencies for the organic components were in excess of 99.9%, limited by instrumentation detection capability. Separation of the inorganic components of the Hanford process stream was more difficult to accomplish than destruction of the organic component. Large fractions of all metals contained in this stream were found both in the solids separator effluent and in deposits removed from the reactor. Mass closure was not achieved. Of the process stream's non-metallic, inorganic components, the sulfates and phosphates precipitated, while the nitrates tended to stay in solution. The inorganic material that did precipitate from the simulated Hanford mixed waste accumulated in zones that may be associated with changes in the chemical and physical properties of the supercritical fluid. Corrosion is expected to be a significant problem. Witness wires of Inconel 625, Hastalloy C-276, and titanium placed in the preheater, reactor and cooldown exchanger indicated selective dissolution of chromium, nickel, and molybdenum for some conditions, and non-selective dissolution for others. While these results are very promising, further research is required to evaluate the scalability, reliability, and economics of SCWO reactor components and systems, particularly for mixed wastes. Future research must explore a parameter space (temperature, pressure, pH, residence time, etc.) focused on selecting conditions and materials for specific process streams

  18. Drying of supercritical carbon dioxide with membrane processes

    NARCIS (Netherlands)

    Lohaus, Theresa; Scholz, Marco; Koziara, Beata; Benes, Nieck Edwin; Wessling, Matthias

    2015-01-01

    In supercritical extraction processes regenerating the supercritical fluid represents the main cost constraint. Membrane technology has potential for cost efficient regeneration of water-loaded supercritical carbon dioxide. In this study we have designed membrane-based processes to dehydrate

  19. Evaluation of tubular reactor designs for supercritical water oxidation of U.S. Department of Energy mixed waste

    International Nuclear Information System (INIS)

    Barnes, C.M.

    1994-12-01

    Supercritical water oxidation (SCWO) is an emerging technology for industrial waste treatment and is being developed for treatment of the US Department of Energy (DOE) mixed hazardous and radioactive wastes. In the SCWO process, wastes containing organic material are oxidized in the presence of water at conditions of temperature and pressure above the critical point of water, 374 C and 22.1 MPa. DOE mixed wastes consist of a broad spectrum of liquids, sludges, and solids containing a wide variety of organic components plus inorganic components including radionuclides. This report is a review and evaluation of tubular reactor designs for supercritical water oxidation of US Department of Energy mixed waste. Tubular reactors are evaluated against requirements for treatment of US Department of Energy mixed waste. Requirements that play major roles in the evaluation include achieving acceptable corrosion, deposition, and heat removal rates. A general evaluation is made of tubular reactors and specific reactors are discussed. Based on the evaluations, recommendations are made regarding continued development of supercritical water oxidation reactors for US Department of Energy mixed waste

  20. Investigation of forced convection heat transfer of supercritical pressure water in a vertically upward internally ribbed tube

    International Nuclear Information System (INIS)

    Wang Jianguo; Li Huixiong; Guo Bin; Yu Shuiqing; Zhang Yuqian; Chen Tingkuan

    2009-01-01

    In the present paper, the forced convection heat transfer characteristics of water in a vertically upward internally ribbed tube at supercritical pressures were investigated experimentally. The six-head internally ribbed tube is made of SA-213T12 steel with an outer diameter of 31.8 mm and a wall thickness of 6 mm and the mean inside diameter of the tube is measured to be 17.6 mm. The experimental parameters were as follows. The pressure at the inlet of the test section varied from 25.0 to 29.0 MPa, and the mass flux was from 800 to 1200 kg/(m 2 s), and the inside wall heat flux ranged from 260 to 660 kW/m 2 . According to experimental data, the effects of heat flux and pressure on heat transfer of supercritical pressure water in the vertically upward internally ribbed tube were analyzed, and the characteristics and mechanisms of heat transfer enhancement, and also that of heat transfer deterioration, were also discussed in the so-called large specific heat region. The drastic changes in thermophysical properties near the pseudocritical points, especially the sudden rise in the specific heat of water at supercritical pressures, may result in the occurrence of the heat transfer enhancement, while the covering of the heat transfer surface by fluids lighter and hotter than the bulk fluid makes the heat transfer deteriorated eventually and explains how this lighter fluid layer forms. It was found that the heat transfer characteristics of water at supercritical pressures were greatly different from the single-phase convection heat transfer at subcritical pressures. There are three heat transfer modes of water at supercritical pressures: (1) normal heat transfer, (2) deteriorated heat transfer with low HTC but high wall temperatures in comparison to the normal heat transfer, and (3) enhanced heat transfer with high HTC and low wall temperatures in comparison to the normal heat transfer. It was also found that the heat transfer deterioration at supercritical pressures was

  1. Heat Transfer Characteristics of the Supercritical CO{sub 2} Flowing in a Vertical Annular Channel

    Energy Technology Data Exchange (ETDEWEB)

    Yoo, Tae Ho; Bae, Yoon Yeong; Kim, Hwan Yeol [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

    2010-05-15

    Heat transfer test facility, SPHINX(Supercritical Pressure Heat transfer Investigation for NeXt generation), has been operated at KAERI for an investigation of the thermal-hydraulic characteristics of supercritical CO{sub 2} at several test sections with a different geometry. The loop uses CO{sub 2} because it has much lower critical pressure and temperature than those of water. Experimental study of heat transfer to supercritical CO{sub 2} in a vertical annular channel with and hydraulic diameter of 4.5 mm has been performed. CO{sub 2} flows downward through the annular channel simulating the downward-flowing coolant in a multi-pass reactor or water rod moderator in a single pass reactor. The heat transfer characteristics in a downward flow were analyzed and compared with the upward flow test results performed previously with the same test section at KAERI

  2. Heat Transfer Characteristics of the Supercritical CO2 Flowing in a Vertical Annular Channel

    International Nuclear Information System (INIS)

    Yoo, Tae Ho; Bae, Yoon Yeong; Kim, Hwan Yeol

    2010-01-01

    Heat transfer test facility, SPHINX(Supercritical Pressure Heat transfer Investigation for NeXt generation), has been operated at KAERI for an investigation of the thermal-hydraulic characteristics of supercritical CO 2 at several test sections with a different geometry. The loop uses CO 2 because it has much lower critical pressure and temperature than those of water. Experimental study of heat transfer to supercritical CO 2 in a vertical annular channel with and hydraulic diameter of 4.5 mm has been performed. CO 2 flows downward through the annular channel simulating the downward-flowing coolant in a multi-pass reactor or water rod moderator in a single pass reactor. The heat transfer characteristics in a downward flow were analyzed and compared with the upward flow test results performed previously with the same test section at KAERI

  3. Continuous control of pH value and chloride concentration in a water coolant of nuclear reactors

    International Nuclear Information System (INIS)

    Moskvin, L.N.; Krasnoperov, V.M.; Fokina, K.G.; Vilkov, N.Ya.

    1975-01-01

    Potentiometry method with the use of flowing cells with two identical electrodes is the simplest and most safe for continuous pH value and chloride control in nuclear reactor circulating circuits. The constant potential on the comparison electrode may be provided by supplying the analyzed solution to it through the ion resin filter of mixed operation. The pos--sibility of a continuous pH value monitoring in a flowing cell with two glass electrodes in parallel is considered. To monitor clorides a cell with two porous chlorine-silver electrodes positioned in series is used. The cells of the design described are shown to be workable in water simulating coolants for water-cooled reactors

  4. Correlations between the electrochemical behaviour and surface film composition of TZM alloy exposed to divertor water coolant environments

    International Nuclear Information System (INIS)

    Maday, M.-F.; Giorgi, R.; Dikonimos-Makris, T.

    1997-01-01

    X-ray photoelectron spectroscopy (XPS) has been carried out on TZM alloy surfaces after short and long immersion tests in high temperature (250 C) aqueous environments simulating possible fusion reactor coolant conditions during operation. Phase identification by XPS was used in connection with the open circuit potential trends to suggest plausible hypotheses about TZM corrosion behaviour in the various chemical environments considered in this study. It was proposed that exposure of TZM to oxidizing water conditions produced poorly protective layers, which consist essentially of low (IV) and intermediate (V) valency Mo oxides/hydroxides. Conversely the results obtained in deaerated and reducing water conditions suggested that barrier films could develop in these environments: the phases exhibit a bilayered structure and consisted of an inner tetravalent Mo oxide/hydroxide and an outer hexavalent Mo oxide. The protective properties of such layers were attributed to the hexavalent Mo species. (orig.)

  5. Study of a fuel assembly for the nuclear reactor of IV generation cooled with supercritical water

    International Nuclear Information System (INIS)

    Barragan M, A.; Martin del Campo M, C.; Francois L, J. L.; Espinosa P, G.

    2011-11-01

    In this work a neutron study is presented about a square assembly design of double line of fuel rods, with moderator box to the center of the arrangement, for a nuclear reactor cooled with supercritical water (SCWR). The SCWR reactor was chosen by the characteristics of its design, mainly because is based in light water reactors (PWR and BWR), and the operational experience that has of them allow to use models and similar programs to simulate the fuel and the nucleus of this type of reactors. To develop the necessary models and to carry out the design and analysis of the SCWR reactor, the neutron codes MCNPX and Helios were used. The reason of using both codes, is because the code MCNPX used thoroughly in the neutron simulation of these reactors, it has been our reference code to analyze the results obtained with the Helios code which results are more efficient because its calculation times are minors. In the nucleus design the same parameters for both codes were considered. The results show that the design with Helios is a viable option to simulate these reactors since their values of the neutrons multiplication factor are very similar to those obtained with MCNPX. On the other hand, it could be corroborated that the CASMO-4 code is inadequate to simulate the fuel to the temperature conditions and water pressure in the SCWR. (Author)

  6. Destruction of an industrial wastewater by supercritical water oxidation in a transpiring wall reactor

    International Nuclear Information System (INIS)

    Bermejo, M.D.; Cocero, M.J.

    2006-01-01

    The supercritical water oxidation (SCWO) is a technology that takes advantage of the special properties of water in the surroundings of critical point of water to completely oxidize wastes in residence times lower than 1 min. The problems caused by the harsh operational conditions of the SCWO process are being solved by new reactor designs, such as the transpiring wall reactor (TWR). In this work, the operational parameters of a TWR have been studied for the treatment of an industrial wastewater. As a result, the process has been optimized for a feed flow of 16 kg/h with feed inlet temperatures higher than 300 deg. C and transpiring flow relation (R) between 0.2 and 0.6 working with an 8% (w/w) isopropanol (IPA) as a fuel. The experimental data and a mathematical model have been applied for the destruction of an industrial waste containing acetic acid and crotonaldehyde as main compounds. As the model predicted, removal efficiencies higher than 99.9% were obtained, resulting in effluents with 2 ppm total organic carbon (TOC) at feed flow of 16 kg/h, 320 deg. C of feed temperature and R = 0.32. An effluent TOC of 35 ppm under conditions feed flow of 18 kg/h, feed inlet temperatures of 290 deg. C, reaction temperatures of 570 deg. C and R = 0.6

  7. Reactive turbulent flow CFD study in supercritical water oxidation process: application to a stirred double shell reactor

    International Nuclear Information System (INIS)

    Moussiere, S.

    2006-12-01

    Supercritical water oxidation is an innovative process to treat organic liquid waste which uses supercritical water properties to mix efficiency the oxidant and the organic compounds. The reactor is a stirred double shell reactor. In the step of adaptation to nuclear constraints, the computational fluid dynamic modeling is a good tool to know required temperature field in the reactor for safety analysis. Firstly, the CFD modeling of tubular reactor confirms the hypothesis of an incompressible fluid and the use of k-w turbulence model to represent the hydrodynamic. Moreover, the EDC model is as efficiency as the kinetic to compute the reaction rate in this reactor. Secondly, the study of turbulent flow in the double shell reactor confirms the use of 2D axisymmetric geometry instead of 3D geometry to compute heat transfer. Moreover, this study reports that water-air mixing is not in single phase. The reactive turbulent flow is well represented by EDC model after adaptation of initial conditions. The reaction rate in supercritical water oxidation reactor is mainly controlled by the mixing. (author)

  8. Turbulent heat transfer in a coolant channel of a pressurized water reactor (PWR) core

    International Nuclear Information System (INIS)

    Kumar, Sanjeev; Saha, Arun K.; Munshi, Prabhat

    2016-01-01

    Exact predictions in nuclear reactors are more crucial, because of the safety aspects. It necessitates the appropriate modeling of heat transfer phenomena in the reactors core. A two-dimensional thermal-hydraulics model is used to study the detailed analysis of the coolant region of a fuel pin. Governing equations are solved using Marker and Cell (MAC) method. Standard wall functions k-ε turbulence model is incorporated to consider the turbulent behaviour of the flow field. Validation of the code and a few results for a typical PWR running at normal operating conditions reported earlier. There were some discrepancies in the old calculations. These discrepancies have been resolved and updated results are presented in this work. 2D thermal-hydraulics model results have been compared with the 1D thermal-hydraulics model results and conclusions have been drawn. (author)

  9. Frontier between medium and large break loss of coolant accidents of pressurized water reactor

    Science.gov (United States)

    Kim, Taewan

    2017-10-01

    In order to provide the probabilistic safety assessment with more realistic condition to calculate the frequency of the initiating event, a study on the frontier between medium-break and large-break loss-of-coolant-accidents has been performed by using best-estimate thermal hydraulic code, TRACE. A methodology based on the combination of the essential safety features and system parameter has been applied to the Zion nuclear power plant to evaluate the validity of the frontier utilized for the probabilistic safety assessment. The peak cladding temperature has been chosen as a relevant system parameter that represents the system behavior during the transient. The results showed that the frontier should be extended from 6 in. to 10 in. based on the required safety functions and system response.

  10. Loss-of-coolant accident for large pipe breaks in light water reactor plants

    International Nuclear Information System (INIS)

    Keusenhoff, J.

    1980-01-01

    The importance of loss-of-coolant accidents (LOCA) and their control for nuclear reactor safety is explained. Showing the cooling circuits and emergency core cooling systems (ECCS) of both, PWR and BWR, the possible break spectrum and the general sequence of events is discussed. The governing physical phenomena for the different LOCA phases are pointed out in more detail. Special emphasis is taken on rules, regulations and failure criteria for licensing purposes. Analysis methods and codes for both, evaluation and best-estimate model are compared under deterministic and probabilistic approach, respectively. Some insight in present integral and separate effect tests demonstrates the interdependency of analysis and experiment. Results of LOCA analysis and experiments show the present state of the art. (orig.)

  11. Application of the extended Kalman filtering for the estimation of core coolant flow rate in pressurized water reactors

    International Nuclear Information System (INIS)

    Shieh, D.J.; Upadhyaya, B.R.

    1986-01-01

    In-core neutron detector and core-exit temperature signals in a pressurized water reactor (PWR) satisfy the condition of observability of the core dynamic system, and can be used to estimate nonmeasurable state variables and model parameters. The extension of the Kalman filtering technique is very useful for direct parameter estimation. This approach is applied to the determination of core coolant mass flow rate in PWRs and is evaluated using in-core measurements at the Loss-of-Fluid Test (LOFT) reactor. The influence of model uncertainties on the estimation accuracy was studied using the ambiguity function analysis. A sequential discretization method was developed to achieve faster convergence to the true value, avoiding model discretization at each sample point. The performance of the extended Kalman filter and the computational innovations were evaluated using a reduced order core dynamic model of the LOFT reactor and random data simulation. The technique was then applied to the determination of LOFT core coolant flow rate from operational data at 100% and 65% flow conditions

  12. Feasibility analysis of modified AL-6XN steel for structure component application in supercritical water-cooled reactor

    Institute of Scientific and Technical Information of China (English)

    Xinggang LI; Qingzhi YAN; Rong MA; Haoqiang WANG; Changchun GE

    2009-01-01

    Modified AL-6XN austenite steel was patterned after AL-6XN superaustenitic stainless steel by introducing microalloy elements such as zirconium and titanium in order to adapt to recrystallizing thermo-mechanical treatment and further improve crevice corrosion resistance. Modified AL-6XN exhibited comparable tensile strength, and superior plasticity and impact toughness to commercial AL-6XN steel. The effects of aging behavior on corrosion resistance and impact toughness were measured to evaluate the qualification of modified AL-6XN steel as an in-core component and cladding material in a supercritical water-cooled reactor. Attention should be paid to degradation in corrosion resistance and impact toughness after aging for 50 hours when modified AL-6XN steel is considered as one of the candidate materials for in-core components and cladding tubes in supercritical water-cooled reactors.

  13. A flow reactor for the flow supercritical water oxidation of wastes to mitigate the reactor corrosion problem

    International Nuclear Information System (INIS)

    Chitanvis, S.M.

    1994-01-01

    We have designed a flow tube reactor for supercritical water oxidation of wastes that confines the oxidation reaction to the vicinity of the axis of the tube. This prevents high temperatures and reactants as well as reaction products from coming in intimate contact with reactor walls. This implies a lessening of corrosion of the walls of the reactor. We display numerical simulations for a vertical reactor with conservative design parameters that illustrate our concept. We performed our calculations for the destruction of sodium nitrate by ammonium hydroxide In the presence of supercritical water, where the production of sodium hydroxide causes corrosion. We have compared these results with that for a horizontal set-up where the sodium hydroxide created during the reaction ends up on the floor of the tube, implying a higher probability of corrosion

  14. Effect of surface modification on the corrosion resistivity in supercritical water

    International Nuclear Information System (INIS)

    Penttila, S.; Horvath, A.; Toivonen, A.; Zolnai, Z.

    2011-01-01

    This paper summarizes the results of high temperature corrosion studies of the candidate austenitic alloys at relevant operating conditions for SCWR. The high temperature and pressure above the thermodynamic critical point of water result in higher oxidation rate which might be critical for thin-wall components like fuel cladding. The goal of this work was to study the effect of surface preparation on the oxidation rate on Ti-stabilized austenitic alloy 1.4970. Surfaces were prepared with ion implantation using He"+- and N"+-ions. Samples were immersed in supercritical water at 650"oC/25 MPa, for up to 2000 hours. Added to this, conventional surface treatments were conducted for austenitic alloy 316L tube samples in order to study the effect of cold work in sample surface on corrosion resistance. The corrosion rate was evaluated by measuring the weight change of the samples. The compositions of the oxide layers were analyzed using scanning electron microscope (SEM) in conjunction with Energy Dispersive Spectroscopy (EDS). (author)

  15. The effect of alkali on the product distribution from black liquor conversion under supercritical water.

    Science.gov (United States)

    Hawangchu, Y; Atong, D; Sricharoenchaikul, V

    2017-07-01

    Lignin in chemical pulping waste, or black liquor (BL), can be converted into various products via supercritical water gasification (SCWG). However, the inherited alkaline contents from the pulping chemicals may affect the product yields and properties. In this research, the influence of the residual alkali on the product distribution via SCWG of soda BL and kraft BL was evaluated. The SCWG was performed in a batch quartz reactor for 10 min at various temperatures (673, 773 and 873 K) and pressures (250, 300 and 400 bar). The highest hydrogen (H 2 ) production occurred at 873 K for the soda BL. The water-gas shift reaction with sodium ions played an important part in the H 2 production, while only small amounts of methane and carbon monoxide were detected. Hydrocarbons, carboxylic acids and esters were the dominant substrates in the liquid products, which denoted the potential of this method for bond cleaving of the lignin macromolecule. As a result, BL, which typically contains alkali salt, was an appropriate feedstock for the SCWG reaction to produce renewable fuel. This method not only has a positive influence on the generation of value added products from highly corrosive waste but also helps avoid some technical problems commonly encountered with direct firing in a recovery boiler.

  16. Reacting flow simulations of supercritical water oxidation of PCB-contaminated transformer oil in a pilot plant reactor

    Directory of Open Access Journals (Sweden)

    V. Marulanda

    2011-06-01

    Full Text Available The scale-up of a supercritical water oxidation process, based on recent advancements in kinetic aspects, reactor configuration and optimal operational conditions, depends on the research and development of simulation tools, which allow the designer not only to understand the complex multiphysics phenomena that describe the system, but also to optimize the operational parameters to attain the best profit for the process and guarantee its safe operation. Accordingly, this paper reports a multiphysics simulation with the CFD software Comsol Multiphysics 3.3 of a pilot plant reactor for the supercritical water oxidation of a heavily PCB-contaminated mineral transformer oil. The proposed model was based on available information for the kinetic aspects of the complex mixture and the optimal operational conditions obtained in a lab-scale continuous supercritical water oxidation unit. The pilot plant simulation results indicate that it is not feasible to scale-up directly the optimal operational conditions obtained in the isothermal lab-scale experiments, due to the excess heat released by the exothermic oxidation reactions that result in outlet temperatures higher than 600°C, even at reactor inlet temperatures as low as 400°C. Consequently, different alternatives such as decreasing organic flowrates or a new reactor set-up with multiple oxidant injections should be considered to guarantee a safe operation.

  17. Assessment of hydrogen bonding effect on ionization of water from ambient to supercritical region–MD simulation approach

    International Nuclear Information System (INIS)

    Swiatla-Wojcik, D.; Mozumder, A.

    2014-01-01

    We present a novel, molecular dynamics (MD) simulation based, strategy to analyze how the degree of hydrogen bonding may influence the ionization and dissociation of water upon heating from ambient to supercritical temperatures. Calculations show a negligible change in the ionization energy up to 200 °C. At higher temperatures the ionization energy increases due to the decreasing degree of hydrogen bonding. The influence of density (pressure) is pronounced in the supercritical region. The ionization is more energy consuming in the less dense fluid. We also show that high temperature and low density may promote dissociation of the electronically excited water molecules. Implications on the initial radiation chemical yields of the hydrated electron, hydrogen atom and hydroxyl radical are discussed. - Highlights: • Up to 200 °C changes in the vertical and adiabatic ionization potentials are negligible. • At higher temperatures ionization is more energy consuming. • Ionization potential increases with decreasing density of supercritical water. • High temperature and low density promote dissociation of the excited molecules

  18. Disposition of nonflammable low-level radioactive wastes using supercritical water with ruthenium(IV) oxide catalyst

    International Nuclear Information System (INIS)

    Sugiyama, Wataru

    2013-01-01

    This paper presents the distribution behavior of iron, cobalt, cesium, iodine and strontium attached to nonflammable organic materials, in solid, liquid and gas phases during the decomposition of these materials using supercritical water with ruthenium(IV) oxide (RuO 2 ) catalyst. The distributions of these elements under various conditions (initial amounts, with/without precipitation reagent) were determined by using their radioisotopes as simulated low-level radioactive wastes (LLW) in order to ease the detection of trace amounts of elements even in solid and gas phases. Iron and cobalt were found only in the solid phase when iron hydroxide was added as a precipitation reagent before the supercritical water reaction. Cesium, iodine and strontium were found in the liquid phase after the reaction. Therefore, by adding precipitation reagents such as sodium tetraphenylborate, and sodium carbonate (Na 2 CO 3 ) (or sodium hydrogen carbonate (NaHCO 3 )) and silver nitrate (AgNO 3 ) aqueous solutions to each resultant liquid phase containing cesium, strontium and iodine, respectively, these elements can be successfully recovered only in the solid phase. The gases produced during the decomposition of the organic material contain no radioactivity under all conditions in this study. These results indicate that all of the elements investigated in this study (iron, cobalt, cesium, iodine and strontium) can be recovered successfully by this supercritical water process using RuO 2 Consequently, this process is suggested as a predominant candidate for the treatment of nonflammable organic materials in LLW. (author)

  19. Experimental study on thermal-hydraulic behaviors of a pressure balanced coolant injection system for a passive safety light water reactor JPSR

    Energy Technology Data Exchange (ETDEWEB)

    Satoh, Takashi; Watanabe, Hironori; Araya, Fumimasa; Nakajima, Katsutoshi [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment; Iwamura, Takamichi; Murao, Yoshio

    1998-02-01

    A conceptual design study of a passive safety light water reactor JPSR has been performed at Japan Atomic Energy Research Institute JAERI. A pressure balanced coolant injection experiment has been carried out, with an objective to understand thermal-hydraulic characteristics of a passive coolant injection system which has been considered to be adopted to JPSR. This report summarizes experimental results and data recorded in experiment run performed in FY. 1993 and 1994. Preliminary experiments previously performed are also briefly described. As the results of the experiment, it was found that an initiation of coolant injection was delayed with increase in a subcooling in the pressure balance line. By inserting a separation device which divides the inside of core make-up tank (CMT) into several small compartments, a diffusion of a high temperature region formed just under the water surface was restrained and then a steam condensation was suppressed. A time interval from an uncovery of the pressure balance line to the initiation of the coolant injection was not related by a linear function with a discharge flow rate simulating a loss-of-coolant accident (LOCA) condition. The coolant was injected intermittently by actuation of a trial fabricated passive valve actuated by pressure difference for the present experiment. It was also found that the trial passive valve had difficulties in setting an actuation set point and vibrations noises and some fraction of the coolant was remained in CMT without effective use. A modification was proposed for resolving these problems by introducing an anti-closing mechanism. (author)

  20. Specificities of reactor coolant pumps units with lead and lead-bismuth coolant

    International Nuclear Information System (INIS)

    Beznosov, A.V.; Anotonenkov, M.A.; Bokov, P.A.; Baranova, V.S.; Kustov, M.S.

    2009-01-01

    The analysis results of impact of lead and lead-bismuth coolants specific properties on the coolants flow features in flow channels of the main and auxiliary circulating pumps are presented. Impossibility of cavitation initiation in flow channels of vane pumps pumping lead and lead-bismuth coolants was demonstrated. The experimental research results of discontinuity of heavy liquid metal coolant column were presented and conditions of gas cavitation initiation in coolant flow were discussed. Invalidity of traditional calculation methods of water and sodium coolants circulation pumps calculations for lead and lead-bismuth coolants circulation pumps was substantiated [ru

  1. Development of Hplc Techniques for the Analysis of Trace Metal Species in the Primary Coolant of a Pressurised Water Reactor.

    Science.gov (United States)

    Barron, Keiron Robert Philip

    Available from UMI in association with The British Library. The need to monitor corrosion products in the primary circuit of a pressurised water reactor (PWR), at a concentration of 10pg ml^{-1} is discussed. A review of trace and ultra-trace metal analysis, relevant to the specific requirements imposed by primary coolant chemistry, indicated that high performance liquid chromatography (HPLC), coupled with preconcentration of sample was an ideal technique. A HPLC system was developed to determine trace metal species in simulated PWR primary coolant. In order to achieve the desired detection limit an on-line preconcentration system had to be developed. Separations were performed on Aminex A9 and Benson BC-X10 analytical columns. Detection was by post column reaction with Eriochrome Black T and Calmagite Linear calibrations of 2.5-100ng of cobalt (the main species of interest), were achieved using up to 200ml samples. The detection limit for a 200ml sample was 10pg ml^{-1}. In order to achieve the desired aim of on-line collection of species at 300^circ C, the use of inorganic ion-exchangers is essential. A novel application, utilising the attractive features of the inorganic ion-exchangers titanium dioxide, zirconium dioxide, zirconium arsenophosphate and pore controlled glass beads, was developed for the preconcentration of trace metal species at temperature and pressure. The performance of these exchangers, at ambient and 300^ circC was assessed by their inclusion in the developed analytical system and by the use of radioisotopes. The particular emphasis during the development has been upon accuracy, reproducibility of recovery, stability of reagents and system contamination, studied by the use of radioisotopes and response to post column reagents. This study in conjunction with work carried out at Winfrith, resulted in a monitoring system that could follow changes in coolant chemistry, on deposition and release of metal species in simulated PWR water loops. On

  2. Assumptions used for evaluating the potential radiological consequences of a less of coolant accident for pressurized water reactors - June 1974

    International Nuclear Information System (INIS)

    Anon.

    1974-01-01

    Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility. The design basis loss of coolant accident is one of the postulated accidents used to evaluate the adequacy of these structures, systems, and components with respect to the public health and safety. This guide gives acceptable assumptions that may be used in evaluating the radiological consequences of this accident for a pressurized water reactor. In some cases, unusual site characteristics, plant design features, or other factors may require different assumptions which will be considered on an individual case basis. The Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concurred in the regulatory position

  3. Assumptions used for evaluating the potential radiological consequences of a loss of coolant accident for boiling water reactors - June 1974

    International Nuclear Information System (INIS)

    Anon.

    1974-01-01

    Section 50.34 of 10 CFR Part 50 requires that each applicant for a construction permit or operating license provide an analysis and evaluation of the design and performance of structures, systems, and components of the facility with the objective of assessing the risk to public health and safety resulting from operation of the facility. The design basis loss of coolant accident is one of the postulated accidents used to evaluate the adequacy of these structures, systems, and components with respect to the public health and safety. This guide gives acceptable assumptions that may be used in evaluating the radiological consequences of this accident for a pressurized water reactor. In some cases, unusual site characteristics, plant design features, or other factors may require different assumptions which will be considered on an individual case basis. The Advisory Committee on Reactor Safeguards has been consulted concerning this guide and has concurred in the regulatory position

  4. Determination of boron as boric acid by automatic potentiometric titration using Gran plots [in pressurized water reactor coolant

    International Nuclear Information System (INIS)

    Midgley, D.; Gatford, C.

    1989-11-01

    Boron in PWR primary coolant and related waters may be determined as boric acid by titration with sodium hydroxide, using a glass electrode as a pH indicator. Earlier work has shown that this analysis can conveniently be carried out automatically with adequate precision and accuracy for routine use, although bias became apparent at the lowest concentrations tested. The latest titrators enable the titration data to be transformed mathematically to give two linear segments, before and after the end-point (Gran plots). The results are as precise as those from other titration methods (in which the end-point is found from the point of inflexion of a plot of pH against volume of titrant), but the bias at low concentrations is much reduced. This is achieved without extra time or involvement of the operator. (author)

  5. Axial distribution of deformation in the cladding of pressurized water reactor fuel rods in a loss-of-coolant accident

    International Nuclear Information System (INIS)

    Rose, K.M.; Mann, C.A.; Hindle, E.D.

    1979-01-01

    In the event of a loss-of-coolant accident in a pressurized water reactor, the cladding of the fuel rods would undergo a temperature excursion while being subject to tensile hoop stress. The deformation behavior of 470-mm lengths of Zircaloy-4 fuel cladding has been studied experimentally; under a range of stress levels in the high-alpha range of zirconium (600 to 850 0 C), diametral strains of up to 70% were observed over the greater part of their length. A negative-feedback mechanism is suggested, based on the reduction of secondary creep rate following cooling by enhanced heat loss at swelling areas. An approximate analysis based on this mechanism was found to be in reasonable agreement with the experimental results. A computer modeling code is being developed to predict cladding deformation under realistic conditions

  6. Axial distribution of deformation in the cladding of pressurized water reactor fuel rods in a loss-of-coolant accident

    Energy Technology Data Exchange (ETDEWEB)

    Rose, K.M.; Mann, C.A.; Hindle, E.D.

    1979-12-01

    In the event of a loss-of-coolant accident in a pressurized water reactor, the cladding of the fuel rods would undergo a temperature excursion while being subject to tensile hoop stress. The deformation behavior of 470-mm lengths of Zircaloy-4 fuel cladding has been studied experimentally; under a range of stress levels in the high-alpha range of zirconium (600 to 850/sup 0/C), diametral strains of up to 70% were observed over the greater part of their length. A negative-feedback mechanism is suggested, based on the reduction of secondary creep rate following cooling by enhanced heat loss at swelling areas. An approximate analysis based on this mechanism was found to be in reasonable agreement with the experimental results. A computer modeling code is being developed to predict cladding deformation under realistic conditions.

  7. Residual-stresses in austenitic stainless-steel primary coolant pipes and welds of pressurized-water reactors

    International Nuclear Information System (INIS)

    Faure, F.; Leggatt, R.H.

    1996-01-01

    Surface and through thickness residual stress measurements were performed on an aged cast austenitic-ferritic stainless steel pipe and on an orbital TIG weld representative of those of primary coolant pipes in pressurized water reactors. An abrasive-jet hole drilling method and a block removal and layering method were used. Surface stresses and through thickness stress profiles are strongly dependent upon heat treatments, machining and welding operations. In the aged cast stainless steel pipe, stresses ranged between -250 and +175 MPa. On and near the orbital TIG weld, the outside surface of the weld was in tension both in the axial and hoop directions, with maximum values reaching 420 MPa in the weld. On the inside surface, the hoop stresses were compressive, reaching -300 MPa. However, the stresses in the axial direction at the root of the weld were tensile within 4 mm depth from the inside surface, locally reaching 280 MPa. (author)

  8. An experimental investigation of the post-CHF enhancement factor for a prototypical ITER divertor plate with water coolant

    Energy Technology Data Exchange (ETDEWEB)

    Marshall, T.D. [Rensselaer Polytechnic Institute, Troy, NY (United States); Watson, R.D.; McDonald, J.M. [Sandia National Lab., Albuquerque, NM (United States)] [and others

    1995-09-01

    In an off-normal event, water-cooled copper divertor plates in the International Thermonuclear Experimental Reactor (ITER) may either experience heat loads beyond their design basis, or the normal heat loads may be accompanied by low coolant pressure and velocity. The purpose of this experiment was to illustrate that during one-sided heating, as in ITER, a copper divertor plate with the proper side wall thickness, at low system pressure and velocity can absorb without failing an incident heat flux, q{sub i}, that significantly exceed the value, q{sub i}{sup CHF}, which is associated with local CHF at the wall of the coolant channel. The experiment was performed using a 30 kW electron beam test system for heating of a square cross-section divertor heat sink with a smooth circular channel of 7.63 mm diameter. The heated width, length, and wall thickness were 16, 40, and 3 mm, respectively. Stable surface temperatures were observed at incident heat fluxes greater than the local CHF point, presumably due to circumferential conduction around the thick tube walls when q{sub i}{sup CHF} was exceeded. The Post-CHF enhancement factor, {eta}, is defined as the ratio of the incident burnout heat flux, q{sub i}{sup BO}, to q{sub i}{sup CHF}. For this experiment with water at inlet conditions of 70{degrees}C, 1 m/s, and 1 MPa, q{sub i}{sup CHF} and q{sub i}{sup BO} were 600 and 1100 W/cm{sup 2}, respectively, which gave an {eta} of 1.8.

  9. Visualization study for forced convection heat transfer of supercritical carbon dioxide near pseudo-boiling point

    International Nuclear Information System (INIS)

    Sakurai, K.; Ko, H.S.; Okamoto, K.; Madarame, H.

    2001-01-01

    For development of new reactor, supercritical water is expected to be used as coolant to improve thermal efficiency. However, the thermal characteristics of supercritical fluid is not revealed completely because its difficulty for experiment. Specific phenomena tend to occur near the pseudo-boiling point which is characterised by temperature corresponding to the saturation point in ordinary fluid. Around this point, the physic properties such as density, specific heat and thermal conductivity are drastically varying. Although there is no difference between gas and liquid phases in supercritical fluids, phenomena similar to boiling (with heat transfer deterioration) can be observed round the pseudo-boiling point. Experiments of heat transfer have been done for supercritical fluid in forced convective condition. However, these experiments were mainly realised inside stainless steel cylinder pipes, for which flow visualisation is difficult. Consequently, this work has been devoted to the development of method allowing the visualisation of supercritical flows. The experiment setup is composed of main loop and test section for the visualisation. Carbon dioxide is used as test fluid. Supercritical carbon dioxide flows upward in rectangular channel and heated by one-side wall to generate forced convection heat transfer. Through window at mid-height of the test section, shadowgraphy was applied to visualize density gradient distribution. The behavior of the density wave in the channel is visualized and examined through the variation of the heat transfer coefficient. (author)

  10. Processing of high level waste: Spectroscopic characterization of redox reactions in supercritical water. 1998 annual progress report

    International Nuclear Information System (INIS)

    Arrington, C.A. Jr.

    1998-01-01

    'The author is engaged in a collaborative research effort with Los Alamos staff scientists Steven Buelow, Jeanne Robinson, and Bernie Foy all staff members in group CST-6. The work proposed by these LANL staff scientists is directed towards the destruction of complexants and oxidation of chromium and technetium by hydrothermal processing in near critical or supercritical aqueous solutions. The work addresses two areas of investigation related to ongoing efforts at LANL: (1) kinetic studies of oxidation-reduction reactions in supercritical water; (2) measurement of physical properties of ionic solutes in supercritical water. All of the work during this first year was carried out at Los Alamos National Lab. During the Summer program at LANL all equipment and supplies were provided through Dr. Buelow''s program at LANL. The author has now set up a Raman spectroscopy lab at Furman. Using departmental funds he purchased an optical bench, a laser, and a CCD detector, and a grant from the Dreyfus Foundation assisted in the purchase of a Raman spectrometer. He is now able to carry out experiments using the Raman system at Furman. The plan is to continue the Summer collaboration at LANL and carry out experiments at Furman during the academic year.'

  11. Numerical analysis of flow instability in the water wall of a supercritical CFB boiler with annular furnace

    Science.gov (United States)

    Xie, Beibei; Yang, Dong; Xie, Haiyan; Nie, Xin; Liu, Wanyu

    2016-08-01

    In order to expand the study on flow instability of supercritical circulating fluidized bed (CFB) boiler, a new numerical computational model considering the heat storage of the tube wall metal was presented in this paper. The lumped parameter method was proposed for wall temperature calculation and the single channel model was adopted for the analysis of flow instability. Based on the time-domain method, a new numerical computational program suitable for the analysis of flow instability in the water wall of supercritical CFB boiler with annular furnace was established. To verify the code, calculation results were respectively compared with data of commercial software. According to the comparisons, the new code was proved to be reasonable and accurate for practical engineering application in analysis of flow instability. Based on the new program, the flow instability of supercritical CFB boiler with annular furnace was simulated by time-domain method. When 1.2 times heat load disturbance was applied on the loop, results showed that the inlet flow rate, outlet flow rate and wall temperature fluctuated with time eventually remained at constant values, suggesting that the hydrodynamic flow was stable. The results also showed that in the case of considering the heat storage, the flow in the water wall is easier to return to stable state than without considering heat storage.

  12. Development of a porous wall reactor for Oxidation in Supercritical Water. Hydrodynamic Modelling and application to salty wastes

    International Nuclear Information System (INIS)

    Fauvel, E.

    2002-01-01

    This report deals with a transpiring wall reactor for supercritical water oxidation of organic effluents. The singularity of the reactor lies on the inner porous tube made of alumina to minimise both limiting problems, corrosion and salt precipitation. The presence of the inner tube implies a rather complex hydrodynamics. Thus, an hydrodynamic study was performed, in an original way, in a supercritical fluid using the method of the residence time distribution. It enabled to determine the hydrodynamic model of the reactor. Moreover, an inspecting device of the resistance of the inner tube to thermal gradients was developed. Lastly, the performances of the transpiring wall reactor were tested on model compounds such as sodium sulphate and the mixture of dodecane/tributylphosphate. (author) [fr

  13. A Simplified Supercritical Fast Reactor with Thorium Fuel

    OpenAIRE

    Peng Zhang; Kan Wang; Ganglin Yu

    2014-01-01

    Super-Critical water-cooled Fast Reactor (SCFR) is a feasible option for the Gen-IV SCWR designs, in which much less moderator and thus coolant are needed for transferring the fission heat from the core compared with the traditional LWRs. The fast spectrum of SCFR is useful for fuel breeding and thorium utilization, which is then beneficial for enhancing the sustainability of the nuclear fuel cycle. A SCFR core is constructed in this work, with the aim of simplifying the mechanical structure ...

  14. Survey of Water Chemistry and Corrosion of NPP

    Energy Technology Data Exchange (ETDEWEB)

    Jung, Ki Sok; Hong, Bong Geon

    2008-06-15

    Status of water chemistry of nuclear power plant and materials corrosion has been surveyed. For PWR, system chemistry of primary coolant and secondary coolant as well as the related corrosion of materials was surveyed. For BWR, system chemistry as whole has been surveyed with its accompanying corrosion problems. Radiolysis of coolant water and activation of corrosion products also was surveyed. Future NPP such as supercritical water cooled reactor and fusion reactor has also been surveyed for their water chemistry and corrosion problems. As a result, proposal for some research items has been suggested. Some related corrosion research techniques and electrochemical fundamentals are also presented.

  15. Survey of Water Chemistry and Corrosion of NPP

    International Nuclear Information System (INIS)

    Jung, Ki Sok; Hong, Bong Geon

    2008-06-01

    Status of water chemistry of nuclear power plant and materials corrosion has been surveyed. For PWR, system chemistry of primary coolant and secondary coolant as well as the related corrosion of materials was surveyed. For BWR, system chemistry as whole has been surveyed with its accompanying corrosion problems. Radiolysis of coolant water and activation of corrosion products also was surveyed. Future NPP such as supercritical water cooled reactor and fusion reactor has also been surveyed for their water chemistry and corrosion problems. As a result, proposal for some research items has been suggested. Some related corrosion research techniques and electrochemical fundamentals are also presented

  16. Corrosion and deuterium uptake of Zr-based alloys in supercritical water

    International Nuclear Information System (INIS)

    Khatamian, D.

    2010-01-01

    To increase the thermodynamic efficiency above 40% in nuclear power plants, the use of supercritical water as the heat transport fluid has been suggested. Zircaloy-2, -4, Zr-Cr-Fe, Zr-1Nb and Zr-2.5Nb were tested as prospective fuel cladding materials in 30 MPa D 2 O at 500 o C. Zircaloy-2 showed the highest rates of corrosion and hydriding. Although Zr-Cr-Fe initially showed a very low corrosion rate, it displayed breakaway corrosion kinetics after 50 h exposure. The best-behaved material both from a corrosion and hydrogen uptake point of view was Zr-2.5Nb. However, the Zr-2.5Nb oxide growth rate was still excessive and beyond the current CANDU design allowance. Similar coupons, coated with Cr, were also tested. The coated layer effectively prevented oxidation of the coupons except on the edges, where the coating was thinner and had some flaws. In addition, the Cr-coated Zr-2.5Nb coupons had the lowest deuterium pickup of all the alloys tested and showed no signs of accelerated or nonuniform corrosion. (author)

  17. Control of Canadian once-through direct cycle supercritical water-cooled reactors

    International Nuclear Information System (INIS)

    Sun, Peiwei; Wang, Baosheng; Zhang, Jianmin; Su, Guanghui

    2015-01-01

    Highlights: • Dynamic characteristics of Canadian SCWR are analyzed. • Hybrid feedforward and feedback control is adopted to deal with cross-coupling. • Gain scheduling control with smooth weight is applied to deal with nonlinearity. • It demonstrates through simulation that the control requirements are satisfied. - Abstract: Canadian supercritical water-cooled reactor (SCWR) can be modelled as a Multiple-input Multiple-output (MIMO) system. It has a high power-to-flow ratio, strong cross-coupling and high degree of nonlinearity in its dynamic characteristics. Among the outputs, the steam temperature is strongly affected by the reactor power and the most challenging to control. It is difficult to adopt a traditional control system design methodology to obtain a control system with satisfactory performance. In this paper, feedforward control is applied to reduce the effect on steam temperature from the reactor power. Single-input Single-output (SISO) feedback controllers are synthesized in the frequency domain. Using the feedforward controller, the steam temperature variation due to disturbances at the reactor power has been significantly suppressed. The control system can effectively maintain the overall system stability and regulate the plant around a specified operating condition. To deal with the nonlinearities, gain scheduling control strategy is adopted. Different sets of controllers combined by smooth weight functions are used for the plant at different load conditions. The proposed control strategies have been evaluated under various operating scenarios. Simulation results show that satisfactory performance can successfully achieved by the designed control system

  18. Effect of yttria addition on the stability of porous chromium oxide ceramics in supercritical water

    International Nuclear Information System (INIS)

    Dong Ziqiang; Chen Weixing; Zheng Wenyue; Guzonas, Dave

    2013-01-01

    Porous chromium oxide (Cr 2 O 3 ) ceramics were prepared by oxidizing highly porous chromium carbides that were obtained by a reactive sintering method, and were evaluated at temperatures ranging from 375 °C to 625 °C in supercritical water (SCW) environments with a fixed pressure of 25–30 MPa. Reactive element yttrium was introduced to the porous oxide ceramic by adding various amounts of yttria of 5, 10 and 20 wt.%, respectively, prior to reactive sintering. The exposure in SCW shows that the porous chromium oxide is quite stable in SCW at 375 °C. However, the stability decreased with increasing temperature. It is well known that chromium oxide can be oxidized to soluble chromium (VI) species in SCW when oxygen is present. Adding yttria increases the stability of chromium oxide in SCW environments. However, adding yttria higher than 5 wt.% increased the weight loss of porous chromium oxide samples because of the direct dissociation of Y 2 O 3 in SCW.

  19. Solubility behavior of quartz and corundum in supercritical water: A quantitative thermodynamic interpretation

    International Nuclear Information System (INIS)

    Ziemniak, S.E.

    1995-05-01

    Dissolution reaction equilibria for α-quartz (SiO 2 ) and corundum (α-Al 2 0 3 ) in pure, supercritical water are quantified using a density-dependent thermodynamic model. The database of existing solubility literature for α-quartz (0.2-10 kb, 200--575 degrees C) is shown to be consistent with the presence of two hydrolyzed SI(IV) ion forms: Si(OH) 4 (aq) and Si 2 O(OH) 6 (aq); the corundum database (1-20 kb, 400--700 degrees C) is consistent with Al(OH) 3 (aq) and Al(OH) 4 - . A third Si(IV) ion hydroxocomplex, Si 2 O 2 (OH) 5 - , is indicated at lower pressures (0.03-0.10 kb). The characteristic sigmoidal nature of the solubility isobars is explained by dimerization of Si(OH) 4 (aq) (at high densities) or the formation of anionic hydrolysis products, Si 2 0 2 (OH) 5 - and Al(OH) 4 - , in the low density region (p 2 O(OH) 6 (aq) and Si 2 O 2 (OH) 5 - are made available for the first time

  20. Development of high temperature reference electrodes for potentiometric analyses in supercritical water environments

    International Nuclear Information System (INIS)

    Tung Yuming; Yeh Tsungkuang; Wang Meiya

    2014-01-01

    A specifically designed reference electrode was developed for analyzing the electrochemical behaviors of alloy materials in supercritical water (SCW) environments and identifying the associated electrochemical parameters. In this study, Ag/AgCl reference electrodes and Zr/ZrO 2 reference electrodes suitable for high-temperature applications were manufactured and adopted to measure the electrochemical corrosion potentials (ECPs) of 304L stainless steel (SS) and nickel-based alloy 625 in SCW environments with various amounts of dissolved oxygen (DO). The Ag/AgCl reference electrode made in this laboratory was used as a calibration base for the laboratory-made Zr/ZrO 2 reference electrode at high temperatures up to 400degC. The two reference electrodes were then used for ECP measurements of 304L SS and alloy 625 specimens in 400degC SCW with various DO levels of 300 ppb, 1 ppm, 8.3 ppm, and 32 ppm and under deaerated conditions. The outcome indicated that concentration increases in DO in the designated SCW environment would yield increases in ECP of the two alloys and they exhibited different ECP responses to DO levels. In addition, the laboratory-made Zr/ZrO 2 reference electrode was able to continuously operate for several months and delivered consistent and steady ECP data of the specimens in SCW environments. (author)

  1. Supercritical water gasification of biomass for H2 production: process design.

    Science.gov (United States)

    Fiori, Luca; Valbusa, Michele; Castello, Daniele

    2012-10-01

    The supercritical water gasification (SCWG) of biomass for H(2) production is analyzed in terms of process development and energetic self-sustainability. The conceptual design of a plant is proposed and the SCWG process involving several substrates (glycerol, microalgae, sewage sludge, grape marc, phenol) is simulated by means of AspenPlus™. The influence of various parameters - biomass concentration and typology, reaction pressure and temperature - is analyzed. The process accounts for the possibility of exploiting the mechanical energy of compressed syngas (later burned to sustain the SCWG reaction) through expansion in turbines, while purified H(2) is fed to fuel cells. Results show that the SCWG reaction can be energetically self-sustained if minimum feed biomass concentrations of 15-25% are adopted. Interestingly, the H(2) yields are found to be maximal at similar feed concentrations. Finally, an energy balance is performed showing that the whole process could provide a net power of about 150 kW(e)/(1000 kg(feed)/h). Copyright © 2012 Elsevier Ltd. All rights reserved.

  2. Partial oxidation of municipal sludge with activited carbon catalyst in supercritical water

    International Nuclear Information System (INIS)

    Guo Yang; Wang Shuzhong; Gong Yanmeng; Xu Donghai; Tang Xingying; Ma Honghe

    2010-01-01

    The partial oxidation (POX) characteristics of municipal sludge in supercritical water (SCW) were investigated by using batch reactor. Effects of reaction parameters such as oxidant equivalent ratio (OER), reaction time and temperature were investigated. Activated carbon (AC) could effectively improve the mole fraction of H 2 in gas product at low OER. However, high OER (greater than 0.3) not only led to the combustion reaction of CO and H 2 , but also caused corrosion of reactor inner wall. Hydrogenation and polymerization of the intermediate products are possible reasons for the relative low COD removal rate in our tests. Metal oxide leached from the reactor inner wall and the main components of the granular sludge were deposited in the AC catalyst. Reaction time had more significant effect on BET surface area of AC than OER had. Long reaction time led to the methanation reaction following hydrolysis and oxidation reaction of AC in SCW in the presence of oxygen. Correspondingly, the possible reaction mechanisms were proposed.

  3. Supercritical water oxidation of colored smoke, dye, and pyrotechnic compositions. Final report: Pilot plant conceptual design

    Energy Technology Data Exchange (ETDEWEB)

    LaJeunesse, C.A.; Chan, Jennifer P.; Raber, T.N.; Macmillan, D.C.; Rice, S.F.; Tschritter, K.L.

    1993-11-01

    The existing demilitarization stockpile contains large quantities of colored smoke, spotting dye, and pyrotechnic munitions. For many years, these munitions have been stored in magazines at locations within the continental United States awaiting completion of the life-cycle. The open air burning of these munitions has been shown to produce toxic gases that are detrimental to human health and harmful to the environment. Prior efforts to incinerate these compositions have also produced toxic emissions and have been unsuccessful. Supercritical water oxidation (SCWO) is a rapidly developing hazardous waste treatment method that can be an alternative to incineration for many types of wastes. The primary advantage SCWO affords for the treatment of this selected set of obsolete munitions is that toxic gas and particulate emissions will not occur as part of the effluent stream. Sandia is currently designing a SCWO reactor for the US Army Armament Research, Development & Engineering Center (ARDEC) to destroy colored smoke, spotting dye, and pyrotechnic munitions. This report summarizes the design status of the ARDEC reactor. Process and equipment operation parameters, process flow equations or mass balances, and utility requirements for six wastes of interest are developed in this report. Two conceptual designs are also developed with all process and instrumentation detailed.

  4. Heat transfer to water at supercritical pressures in a circular and square annular flow geometry

    International Nuclear Information System (INIS)

    Licht, Jeremy; Anderson, Mark; Corradini, Michael

    2008-01-01

    A supercritical water heat transfer facility has been built at the University of Wisconsin to study heat transfer in a circular and square annular flow channel. Operating conditions included mass velocities of 350-1425 kg/m 2 s, heat fluxes up to 1.0 MW/m 2 , and bulk inlet temperatures up to 400 o C; all at a pressure of 25 MPa. The accuracy and validity of selected heat transfer correlations and buoyancy criterion were compared with heat transfer measurements. Jackson's Nusselt correlation was able to best predict the test data, capturing 86% of the data within 25%. Watts Nusselt correlation showed a similar trend but under predicted measurements by 10% relative to Jackson's. Comparison of experimental results with results of previous investigators has shown general agreement with high mass velocity data. Low mass velocity data have provided some insight into the difficulty in applying these Nusselt correlations to a region of deteriorated heat transfer. Geometrical differences in heat transfer were seen when deterioration was present. Jackson's buoyancy criterion predicted the onset of deterioration while modifications were applied to Seo's Froude number based criterion

  5. Data acquisition testing in supercritical water oxidation using machine cutting oils and metals

    International Nuclear Information System (INIS)

    Garcia, K.M.

    1996-01-01

    The Department of Energy, the Navy, and SERDP provided funding for an extensive series of testing of a Supercritical Water Oxidation (SCWO) system. The goal of the testing was to create performance data on the process when dealing with highly chlorinated wastes containing heavy metals, and radionuclides. The testing was performed in a MODAR vessel oxidizer. Performance was measured by the ability of the process to achieve greater than 99.99% destruction of the organic content, to partition the metals and radionuclide surrogates for mass balance, and survive the highly corrosive species in the effluent. The test data has shown that these goals were accomplished. 30 gal/day of highly chlorinated machine cutting oil was treated for 130 hrs. There were no significant corrosion or solids handling problems. This machine cutting oil, TRIM reg-sign SOL was chosen by DOE for its complex nature and has proven to be one of the more refractory organic feeds encountered by MODAR. The Navy provided 8 waste streams collected from their shore facilities operation. These paints varied in solids content with wastes such as paint chips, and adhesives. The ninth test run was with all 8 series of wastes combined. The MODAR system successfully treated all of these waste streams providing performance data on the ability of SCWO to treat difficult sludges

  6. Destruction of chemical agent simulants in a supercritical water oxidation bench-scale reactor

    Energy Technology Data Exchange (ETDEWEB)

    Veriansyah, Bambang [Supercritical Fluid Research Laboratory, Clean Technology Research Center, Korea Institute of Science and Technology (KIST), 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of) and Department of Green Process and System Engineering, University of Science and Technology, 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of)]. E-mail: vaveri@kist.re.kr; Kim, Jae-Duck [Supercritical Fluid Research Laboratory, Clean Technology Research Center, Korea Institute of Science and Technology (KIST), 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of) and Department of Green Process and System Engineering, University of Science and Technology, 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of)]. E-mail: jdkim@kist.re.kr; Lee, Jong-Chol [Agency for Defense Development (ADD), P.O. Box 35-1, Yuseong-gu, Daejeon (Korea, Republic of)]. E-mail: jcleeadd@hanafos.com

    2007-08-17

    A new design of supercritical water oxidation (SCWO) bench-scale reactor has been developed to handle high-risk wastes resulting from munitions demilitarization. The reactor consists of a concentric vertical double wall in which SCWO reaction takes place inside an inner tube (titanium grade 2, non-porous) whereas pressure resistance is ensured by a Hastelloy C-276 external vessel. The performances of this reactor were investigated with two different kinds of chemical warfare agent simulants: OPA (a mixture of isopropyl amine and isopropyl alcohol) as the binary precursor for nerve agent of sarin and thiodiglycol [TDG (HOC{sub 2}H{sub 4}){sub 2}S] as the model organic sulfur heteroatom. High destruction rates based on total organic carbon (TOC) were achieved (>99.99%) without production of chars or undesired gases such as carbon monoxide and methane. The carbon-containing product was carbon dioxide whereas the nitrogen-containing products were nitrogen and nitrous oxide. Sulfur was totally recovered in the aqueous effluent as sulfuric acid. No corrosion was noticed in the reactor after a cumulative operation time of more than 250 h. The titanium tube shielded successfully the pressure vessel from corrosion.

  7. Local density inhomogeneities and dynamics in supercritical water: A molecular dynamics simulation approach.

    Science.gov (United States)

    Skarmoutsos, Ioannis; Samios, Jannis

    2006-11-02

    Molecular dynamics atomistic simulations in the canonical ensemble (NVT-MD) have been used to investigate the "Local Density Inhomogeneities and their Dynamics" in pure supercritical water. The simulations were carried out along a near-critical isotherm (Tr = T/Tc = 1.03) and for a wide range of densities below and above the critical one (0.2 rho(c) - 2.0 rho(c)). The results obtained reveal the existence of significant local density augmentation effects, which are found to be sufficiently larger in comparison to those reported for nonassociated fluids. The time evolution of the local density distribution around each molecule was studied in terms of the appropriate time correlation functions C(Delta)rhol(t). It is found that the shape of these functions changes significantly by increasing the density of the fluid. Finally, the local density reorganization times for the first and second coordination shell derived from these correlations exhibit a decreasing behavior by increasing the density of the system, signifying the density effect upon the dynamics of the local environment around each molecule.

  8. Effects of a hypothetical loss-of-coolant accident on a Mark I Boiling Water Reactor pressure-suppression system

    International Nuclear Information System (INIS)

    Pitts, J.H.; McCauley, E.W.

    1977-01-01

    A loss-of-coolant accident (LOCA) in a boiling-water-reactor (BWR) power plant has never occurred. However, because this type of accident could be particularly severe, it is used as a principal theoretical basis for design. A series of consistent, versatile, and accurate air-water tests that simulate LOCA conditions has been completed on a 1 / 5 -scale Mark I BWR pressure-suppression system. Results from these tests are used to quantify the vertical-loading function and to study the associated fluid dynamics phenomena. Detailed histories of vertical loads on the wetwell are shown. In particular, variation of hydrodynamic-generated vertical loads with changes in drywell-pressurization rate, downcomer submergence, and the vent-line loss coefficient are established. Initial drywell overpressure, which partially preclears the downcomers of water, substantially reduces the peak vertical loads. Scaling relationships, developed from dimensional analysis and verified by bench-top experiments, allow the 1 / 5 -scale results to be applied to a full-scale BWR power plant. This analysis leads to dimensionless groupings that are invariant. These groupings show that, if water is used as the working fluid, the magnitude of the forces in a scaled facility is reduced by the cube of the scale factor and occurs in a time reduced by the square root of the scale factor

  9. The effect of inertia force in water lubricated thrust bearings of canned reactor coolant pump

    International Nuclear Information System (INIS)

    Deng Liping

    1994-01-01

    The water lubricated thrust bearings are analyzed. According to characteristic of low viscosity of water the lubricated equation for design and calculation of water lubricated thrust bearings is established. The calculation and analyses show that the effect of inertia force in water lubricated thrust bearings should not be neglected except the conditions of low speed, high angle of inclination and low radius ratio of pad

  10. Measurement data of cesium 137 yields in primary coolant of an in-pile water loop in fission products release experiment

    International Nuclear Information System (INIS)

    Ishiwatari, Nasumi; Nagai, Hitoshi; Takeda, Tsuneo

    1979-03-01

    Series of fuel rods (UO 2 pellets sheathed with stainless steel) having an artificial pinhole were irradiated in the in-pile test section of water loop JMTR OWL-1. Presented are the results of measurements of cesium 137 yields in primary coolant of OWL-1 from 1975 to 1978. (author)

  11. Infrared thermal measurements of laser soft tissue ablation as a function of air/water coolant for Nd:YAG and diode lasers

    Science.gov (United States)

    Gekelman, Diana; Yamamoto, Andrew; Oto, Marvin G.; White, Joel M.

    2003-06-01

    The purpose of this investigation was to measure the maximum temperature at the Nd:YAG and Diode lasers fiberoptic tips as a function of air/water coolant, during soft tissue ablation in pig jaws. A pulsed Nd:YAG laser (1064nm) and a Diode laser (800-830 nm) were used varying parameters of power, conditioning or not of the fiber tip, under 4 settings of air/water coolant. The maximum temperature at the fiber tip was measured using an infra-red camera and the interaction of the fiber with the porcine soft tissue was evaluated. A two-factor ANOVA was used for statistical analysis (plaser interaction with soft tissues produced temperatures levels directly proportional to power increase, but the conditioning of the fiber tip did not influence the temperature rise. On the other hand, conditioning of the fiber tip did influence the temperature rise for Diode laser. The addition of air/water coolant, for both lasers, did not promote temperature rise consistent with cutting and coagulation of porcine soft tissue. Laser parameters affect the fiberoptic surface temperature, and the addition of air/water coolant significantly lowered surface temperature on the fiberoptic tip for all lasers and parameters tested.

  12. Thermal-hydraulic analysis of heat transfer in subchannels of the European high performance supercritical Water-Cooled Reactor for different CFD turbulence models

    Energy Technology Data Exchange (ETDEWEB)

    Castro, Landy Y.; Rojas, Leorlen Y.; Gamez, Abel; Rosales, Jesus; Gonzalez, Daniel; Garcia, Carlos, E-mail: lcastro@instec.cu, E-mail: leored1984@gmail.com, E-mail: agamezgmf@gmail.com, E-mail: jrosales@instec.cu, E-mail: danielgonro@gmail.com, E-mail: cgh@instec.cu [Instituto Superior de Tecnologias y Ciencias Aplicadas (InSTEC), La Habana (Cuba); Oliveira, Carlos Brayner de, E-mail: cabol@ufpe.br [Universidade Federal de Pernambuco (UFPE), Recife, PE (Brazil); Dominguez, Dany S., E-mail: dsdominguez@gmail.com [Universidade Estadual de Santa Cruz (UESC), Ilheus, BA (Brazil). Pos-Graduacao em Modelagem Computacional

    2015-07-01

    Chosen as one of six Generation‒IV nuclear-reactor concepts, Supercritical Water-cooled Reactors (SCWRs) are expected to have high thermal efficiencies within the range of 45 - 50% owing to the reactor's high pressures and outlet temperatures. In this reactor, the primary water enters the core under supercritical-pressure condition (25 MPa) at a temperature of 280 deg C and leaves it at a temperature of up to 510 deg C. Due to the significant changes in the physical properties of water at supercritical-pressure, the system is susceptible to local temperature, density and power oscillations. The behavior of supercritical water into the core of the SCWR, need to be sufficiently studied. Most of the methods available to predict the effects of the heat transfer phenomena within the pseudocritical region are based on empirical one-directional correlations, which do not capture the multidimensional effects and do not provide accurate results in regions such as the deteriorated heat transfer regime. In this paper, computational fluid dynamics (CFD) analysis was carried out to study the thermal-hydraulic behavior of supercritical water flows in sub-channels of a typical European High Performance Light Water Reactor (HPLWR) fuel assembly using commercial CFD code CFX-14. It was determined the steady-state equilibrium parameters and calculated the temperature and density distributions. A comparative study for different turbulence models were carried out and the obtained results are discussed. (author)

  13. Thermal-hydraulic analysis of heat transfer in subchannels of the European high performance supercritical Water-Cooled Reactor for different CFD turbulence models

    International Nuclear Information System (INIS)

    Castro, Landy Y.; Rojas, Leorlen Y.; Gamez, Abel; Rosales, Jesus; Gonzalez, Daniel; Garcia, Carlos; Oliveira, Carlos Brayner de; Dominguez, Dany S.

    2015-01-01

    Chosen as one of six Generation‒IV nuclear-reactor concepts, Supercritical Water-cooled Reactors (SCWRs) are expected to have high thermal efficiencies within the range of 45 - 50% owing to the reactor's high pressures and outlet temperatures. In this reactor, the primary water enters the core under supercritical-pressure condition (25 MPa) at a temperature of 280 deg C and leaves it at a temperature of up to 510 deg C. Due to the significant changes in the physical properties of water at supercritical-pressure, the system is susceptible to local temperature, density and power oscillations. The behavior of supercritical water into the core of the SCWR, need to be sufficiently studied. Most of the methods available to predict the effects of the heat transfer phenomena within the pseudocritical region are based on empirical one-directional correlations, which do not capture the multidimensional effects and do not provide accurate results in regions such as the deteriorated heat transfer regime. In this paper, computational fluid dynamics (CFD) analysis was carried out to study the thermal-hydraulic behavior of supercritical water flows in sub-channels of a typical European High Performance Light Water Reactor (HPLWR) fuel assembly using commercial CFD code CFX-14. It was determined the steady-state equilibrium parameters and calculated the temperature and density distributions. A comparative study for different turbulence models were carried out and the obtained results are discussed. (author)

  14. Analysis of water hammer-structure interaction in piping system for a loss of coolant accident in primary loop of pressurized water reactor

    International Nuclear Information System (INIS)

    Zhang Xiwen; Yang Jinglong; He Feng; Wang Xuefang

    2000-01-01

    The conventional analysis of water hammer and dynamics response of structure in piping system is divided into two parts, and the interaction between them is neglected. The mechanism of fluid-structure interaction under the double-end break pipe in piping system is analyzed. Using the characteristics method, the numerical simulation of water hammer-structure interaction in piping system is completed based on 14 parameters and 14 partial differential equations of fluid-piping cell. The calculated results for a loss of coolant accident (LOCA) in primary loop of pressurized water reactor show that the waveform and values of pressure and force with time in piping system are different from that of non-interaction between water hammer and structure in piping system, and the former is less than the later

  15. Development of computational methods for the safety assessment of gas-cooled high-temperature and supercritical light-water reactors. Final report; Rechenmethoden zur Bewertung der Sicherheit von gasgekuehlten Hochtemperaturreaktoren und superkritischen Leichtwasserreaktoren. Abschlussbericht

    Energy Technology Data Exchange (ETDEWEB)

    Buchholz, S.; Cron, D. von der; Hristov, H.; Lerchl, G.; Papukchiev, A.; Seubert, A.; Sureda, A.; Weis, J.; Weyermann, F.

    2012-12-15

    . Performed test and validation calculations for short and long term transients like withdrawal and ejection of all or single control rods, cold helium ingress or depressurized loss of forced cooling (DLOFC) demonstrate the applicability of TORT-TD/ATTICA3D to 3-D analyses of pebble bed HTR. Chapter 6 documents the extension made in ATHLET regarding application to supercritical water reactors. This includes the implementation of supercritical water as a working fluid and extensions of the model equations for the physics of heat transfer and pressure drop at supercritical water pressure as well as the extension of the material properties package to pressures above the critical point and the modeling of supercritical discharge. The extensions in ATHLET to simulate pebble-bed HTR are described in chapter 7. In ATHLET, the coolant helium has been implemented both as gas component and a working fluid. The material properties package has been properly extended. For the thermal hydraulic modeling of the reactor pressure vessel, a generic parallel channel model including cross connections has been developed for the PBMR-400 design. The HECU model in ATHLET has been extended to spherical geometries in order to simulate the heat transfer processes in HTR fuel pebbles with detailed representation of the TRISO particle fuel. In addition, ATHLET models of gas turbine and compressor have been developed and tested. Finally, chapter 8 documents the development and validation of ANSYS CFX for application to alternative reactor concepts. This includes extensions and applications of the CFX code regarding HPLWR requirements. Accuracy demonstrations of ANSYS CFX models for heat transfer and wall interfaces of gas cooled systems have been performed for several turbulence models by comparing with experimental data. Finally, the development and validation of the coupled code system ATHLET/ANSYS CFX for alternative reactor concepts is described and first coupled steam and helium simulations are

  16. Pourbaix diagrams for the iron–water system extended to high-subcritical and low-supercritical conditions

    International Nuclear Information System (INIS)

    Cook, William G.; Olive, Robert P.

    2012-01-01

    Highlights: ► Pourbaix diagrams for iron–water are extended to low-supercritical temperatures. ► Thermodynamic properties for use in R-HKF model re-evaluated. ► Above the critical point, magnetite solubility is between 10 −11 and 10 −10 mol/kg. - Abstract: The supercritical water-cooled reactor (SCWR) is a Generation IV reactor concept that will operate at temperatures and pressures above water’s thermodynamic critical point. Pourbaix diagrams for the iron–water system at temperatures slightly below and above the critical point at 25 MPa have been constructed to aid the evaluation and development of potential construction materials. High temperature data extrapolation was performed using a revised Helgeson–Kirkham–Flowers model and fit to data on magnetite and hematite solubility in high-temperature water. A low-concentration diagram at 350 °C reveals the importance of water chemistry control to avoid transitioning to an active corrosion region.

  17. The radionuclides of primary coolant in HANARO and the recent activities performed to reduce the radioactivity or reactor pool water

    Energy Technology Data Exchange (ETDEWEB)

    Kim, Minjin [HANARO Research Reactor Centre, Korea Atomic Energy Research Inst., Taejon (Korea, Republic of)

    1998-10-01

    In HANARO reactor, there have been activities to identify the principal radionuclides and to quantify them under the normal operation. The purposes of such activities were to establish the measure by which we can reduce the radioactivity of the reactor pool water and detect, in early stage, the abnormal symptoms due to the leakage of radioactive materials from the irradiation sample or the damage of the nuclear fuel, etc. The typical radionuclides produced by the activation of reactor coolant are N{sup 16} and Ar{sup 41}. The radionuclides produced by the activation of the core structural material consist of Na{sup 24}, Mn{sup 56}, and W{sup 187}. Of the various radionuclides, governing the radiation level at the pool surface are Na{sup 24}, Ar{sup 41}, Mn{sup 58}, and W{sup 187}. By establishing the hot water layer system on the pool surface, we expected that the radionuclides such as Ar{sup 41} and Mn{sup 56} whose half-life are relatively short could be removed to a certain extent. Since the content of radioactivity of Na{sup 24} occupies about 60% of the total radioactivity, we assumed that the total radiation level would be greatly reduced if we could decrease the radiation level of Na{sup 24}. However the actual radiation level has not been reduced as much as we expected. Therefore, some experiments have been carried out to find the actual causes afterwards. What we learned through the experiments are that any disturbance in reactor pool water layer causes increase of the pool surface radiation level and even if we maintain the hot water layer well, reactor shutdown will be very much likely to happen once the hot water layer is disturbed. (author)

  18. Coolant system decontamination

    International Nuclear Information System (INIS)

    Anstine, L.D.; James, D.B.; Melaika, E.A.; Peterson, J.P.

    1981-01-01

    An improved method for decontaminating the coolant system of water cooled nuclear power reactors and for regenerating the decontamination solution is described. A small amount of one or more weak-acid organic complexing agents is added to the reactor coolant, and the pH is adjusted to form a decontamination solution which is circulated throughout the coolant system to dissolve metal oxides from the interior surfaces and complex the resulting metal ions and radionuclide ions. The coolant containing the complexed metal ions and radionuclide ions is passed through a strong-base anion exchange resin bed which has been presaturated with a solution containing the complexing agents in the same ratio and having the same pH as the decontamination solution. As the decontamination solution passes through the resin bed, metal-complexed anions are exchanged for the metal-ion-free anions on the bed, while metal-ion-free anions in the solution pass through the bed, thus removing the metal ions and regenerating the decontamination solution. (author)

  19. Steam--water mixing in nuclear reactor safety loss-of-coolant experiments

    International Nuclear Information System (INIS)

    Naff, S.A.; Schwarz, W.F.

    1978-01-01

    Computer models used to predict the response of reactors to hypothesized accidents necessarily incorporate approximating assumptions. To verify the models by comparing predicted and measured responses in test facilities, these assumptions must be confirmed to be realistic. Recent experiments in facilities capable of repeatedly duplicating the transient behavior of a pressurized water reactor undergoing a pipe rupture show that the assumption of complete water-steam mixing during the transient results in the predicted decompression being faster than that observed. Water reactor safety studies currently in progress include programs aimed at the verification of computer models or ''codes'' used to predict reactor system responses to various hypothesized accidents. The approach is to compare code predictions of transients with the actual test transients in experimental facilities. The purpose of this paper is to explain an important instance in which predictions and data are not in complete agreement and to indicate the significance to water reactor safety studies

  20. Treatment of EDTA contained reactor coolant using water dielectric barrier discharge plasma

    International Nuclear Information System (INIS)

    Song, Sang Heon; Kwon, Daniel; Kim, Gon Ho

    2005-01-01

    EDTA (Ethylene Diamine Tetraacetic Acid) is used as a main absorbent for the metal ion in the secondary loop of the nuclear reactor. Dissolving the wasted EDTA with low cost, therefore, is important issue for the maintenance of the nuclear power reactor and the protection of environment. EDTA is not easily biodegradable, furthermore these methods could make remained another pollutant as complex chemical compounds. Compared to chemical method, the physical methods, using the energetic particles and UVs, are more favorable because they dissociate the bonds of organic compounds directly without the secondary chemical reactions during the treatment. Recently, high energy electron beam, the plasma torch, and the water breakdown by high voltage pulse are applied to treatment of the waste water contained chemicals. Here consideration is narrow down to improve the interaction between the plasma and the chemical bonds of EDTA because the energetic particles; activated radicals, and UVs, are abundant in plasmas. The new method adapted of the water DBD (dielectric barrier discharge) which plasma generates directly on the top of the water contained EDTA is proposed. The application of DBD plasmas has been extended for cleaning the organic compounds from the contaminated surface and also for removing volatile organic chemicals (VOC) such as NO x and SO x from the exhausted gases. Here, the water DBD reactor (SEMTECH, SD-DWG-04-1) is consisted that the one electrode is a ceramic insulator and another one is the water itself. Interestingly, the one electrode, the water, is not the solid dielectric electrode. In this study, therefore, the characteristics with driving frequency are considered and the feasibility of this new method for the DBD treatment of EDTA contained water is demonstrated

  1. Recycling high-performance carbon fiber reinforced polymer composites using sub-critical and supercritical water

    Science.gov (United States)

    Knight, Chase C.

    Carbon fiber reinforced plastics (CFRP) are composite materials that consist of carbon fibers embedded in a polymer matrix, a combination that yields materials with properties exceeding the individual properties of each component. CFRP have several advantages over metals: they offer superior strength to weight ratios and superior resistance to corrosion and chemical attack. These advantages, along with continuing improvement in manufacturing processes, have resulted in rapid growth in the number of CFRP products and applications especially in the aerospace/aviation, wind energy, automotive, and sporting goods industries. Due to theses well-documented benefits and advancements in manufacturing capabilities, CFRP will continue to replace traditional materials of construction throughout several industries. However, some of the same properties that make CFRP outstanding materials also pose a major problem once these materials reach the end of service life. They become difficult to recycle. With composite consumption in North America growing by almost 5 times the rate of the US GDP in 2012, this lack of recyclability is a growing concern. As consumption increases, more waste will inevitably be generated. Current composite recycling technologies include mechanical recycling, thermal processing, and chemical processing. The major challenge of CFRP recycling is the ability to recover materials of high-value and preserve their properties. To this end, the most suitable technology is chemical processing, where the polymer matrix can be broken down and removed from the fiber, with limited damage to the fibers. This can be achieved using high concentration acids, but such a process is undesirable due to the toxicity of such materials. A viable alternative to acid is water in the sub-critical and supercritical region. Under these conditions, the behavior of this abundant and most environmentally friendly solvent resembles that of an organic compound, facilitating the breakdown

  2. Method for removing cesium from aqueous liquid, method for purifying the reactor coolant in boiling water and pressurized water reactors and a mixed ion exchanged resin bed, useful in said purification

    International Nuclear Information System (INIS)

    Otte, J.N.A.; Liebmann, D.

    1989-01-01

    The invention relates to a method for removing cesium from an aqueous liquid, and to a resin bed containing a mixture of an anion exchange resin and cation exchange resin useful in said purification. In a preferred embodiment, the present invention is a method for purifying the reactor coolant of a presurized water or boiling water reactor. Said method, which is particularly advantageously employed in purifying the reactor coolant in the primary circuit of a pressurized reactor, comprises contacting at least a portion of the reactor coolant with a strong base anion exchange resin and the strong acid cation exchange resin derived from a highly cross-linked, macroporous copolymer of a monovinylidene aromatic and a cross-linking monomer copolymerizable therewith. Although the reactor coolant can sequentially be contacted with one resin type and thereafter with the second resin type, the contact is preferably conducted using a resin bed comprising a mixture of the cation and anion exchange resins. 1 fig., refs

  3. Supercritical Carbon Dioxide turbomachinery design for water-cooled Small Modular Reactor application

    International Nuclear Information System (INIS)

    Lee, Jekyoung; Lee, Jeong Ik; Yoon, Ho Joon; Cha, Jae Eun

    2014-01-01

    Highlights: • We described the concept of coupling the S-CO 2 Brayton cycle to the water-cooled SMRs. • We describe a turbomachinery design code called KAISD T MD that can use real gases too. • We suggest changes to the S-CO 2 cycle layout with multiple-independent shafts. • KAIST T MD was used to design the turbomachinery of suggested layout. - Abstract: The Supercritical Carbon Dioxide (S-CO 2 ) Brayton cycle has been gaining attention due to its compactness and high efficiency at moderate turbine inlet temperature. Previous S-CO 2 cycle research works in the field of nuclear engineering were focused on its application to the next generation reactor with higher turbine inlet temperature than the existing conventional water-cooled nuclear power plants. However, it was shown in authors’ previous paper that the advantages of the S-CO 2 Brayton cycle can be also further applied to the water-cooled Small Modular Reactor (SMR) with a success, since SMR requires minimal overall footprint while retaining high performance. One of the major issues in the S-CO 2 Brayton cycle is the selection and design of appropriate turbomachinery for the designed cycle. Because most of the nuclear industry uses incompressible working fluids or ideal gases in the turbomachinery, a more detailed examination of the design of the turbomachinery is required for a power system that uses S-CO 2 as working fluid. This is because the S-CO 2 Brayton cycle high efficiency is the result of the non-ideal variation of properties near the CO 2 critical point. Thus, the major focus of this paper is to suggest the design of the turbomachinery necessary for the S-CO 2 Brayton cycle coupled to water cooled SMRs. For this reason, a S-CO 2 Brayton cycle turbomachinery design methodology was suggested and the suggested design methodology was first tested with the existing experimental data to verify its capability. After then, it was applied to the proposed reference system to demonstrate its

  4. Supercritical Carbon Dioxide turbomachinery design for water-cooled Small Modular Reactor application

    Energy Technology Data Exchange (ETDEWEB)

    Lee, Jekyoung, E-mail: leejaeky85@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Lee, Jeong Ik, E-mail: jeongiklee@kaist.ac.kr [Department of Nuclear and Quantum Engineering, Korea Advanced Institute of Science and Technology, 373-1 Guseong-dong, Yuseong-gu, Daejeon 305-701 (Korea, Republic of); Yoon, Ho Joon, E-mail: hojoon.yoon@kustar.ac.ae [Khalifa University of Science, Technology and Research (KUSTAR), P.O. Box 127788, Abu Dhabi (United Arab Emirates); Cha, Jae Eun, E-mail: jecha@kaeri.re.kr [Korea Atomic Energy Research Institute, 1045 Daedeok-daero, Yuseong-gu, Daejeon 305-353 (Korea, Republic of)

    2014-04-01

    Highlights: • We described the concept of coupling the S-CO{sub 2} Brayton cycle to the water-cooled SMRs. • We describe a turbomachinery design code called KAISD{sub T}MD that can use real gases too. • We suggest changes to the S-CO{sub 2} cycle layout with multiple-independent shafts. • KAIST{sub T}MD was used to design the turbomachinery of suggested layout. - Abstract: The Supercritical Carbon Dioxide (S-CO{sub 2}) Brayton cycle has been gaining attention due to its compactness and high efficiency at moderate turbine inlet temperature. Previous S-CO{sub 2} cycle research works in the field of nuclear engineering were focused on its application to the next generation reactor with higher turbine inlet temperature than the existing conventional water-cooled nuclear power plants. However, it was shown in authors’ previous paper that the advantages of the S-CO{sub 2} Brayton cycle can be also further applied to the water-cooled Small Modular Reactor (SMR) with a success, since SMR requires minimal overall footprint while retaining high performance. One of the major issues in the S-CO{sub 2} Brayton cycle is the selection and design of appropriate turbomachinery for the designed cycle. Because most of the nuclear industry uses incompressible working fluids or ideal gases in the turbomachinery, a more detailed examination of the design of the turbomachinery is required for a power system that uses S-CO{sub 2} as working fluid. This is because the S-CO{sub 2} Brayton cycle high efficiency is the result of the non-ideal variation of properties near the CO{sub 2} critical point. Thus, the major focus of this paper is to suggest the design of the turbomachinery necessary for the S-CO{sub 2} Brayton cycle coupled to water cooled SMRs. For this reason, a S-CO{sub 2} Brayton cycle turbomachinery design methodology was suggested and the suggested design methodology was first tested with the existing experimental data to verify its capability. After then, it was

  5. Experimental investigation of heat transfer for supercritical pressure water flowing in vertical annular channels

    International Nuclear Information System (INIS)

    Gang Wu; Bi Qincheng; Yang Zhendong; Wang Han; Zhu Xiaojing; Hao Hou; Leung, L.K.H.

    2011-01-01

    Highlights: → Two annular test sections were constructed with annular gaps of 4 and 6 mm. → Two heat transfer regions have been observed: normal and deteriorated heat transfer. → The spacer enhances the heat transfer at downstream locations. → The Jackson correlation agrees quite closely with the experimental data. - Abstract: An experiment has recently been completed at Xi'an Jiaotong University (XJTU) to obtain wall-temperature measurements at supercritical pressures with upward flow of water inside vertical annuli. Two annular test sections were constructed with annular gaps of 4 and 6 mm, respectively, and an internal heater of 8 mm outer diameter. Experimental-parameter ranges covered pressures of 23-28 MPa, mass fluxes of 350-1000 kg/m 2 /s, heat fluxes of 200-1000 kW/m 2 , and bulk inlet temperatures up to 400 deg. C. Depending on the flow conditions and heat fluxes, two distinctive heat transfer regimes, referring to as the normal heat transfer and deteriorated heat transfer, have been observed. At similar flow conditions, the heat transfer coefficients for the 6 mm gap annular channel are larger than those for the 4 mm gap annular channel. A strong effect of spiral spacer on heat transfer has been observed with a drastic reduction in wall temperature at locations downstream of the device in the annuli. Two tube-data-based correlations have been assessed against the experimental heat transfer results. The Jackson correlation agrees with the experimental trends and overpredicts slightly the heat transfer coefficients. The Dittus-Boelter correlation is applicable only for the normal heat transfer region but not for the deteriorated heat transfer region.

  6. Supercritical water oxidation of 2-, 3- and 4-nitroaniline: A study on nitrogen transformation mechanism.

    Science.gov (United States)

    Yang, Bowen; Cheng, Zhiwen; Fan, Maohong; Jia, Jinping; Yuan, Tao; Shen, Zhemin

    2018-08-01

    Supercritical water oxidation (SCWO) of 2-, 3- and 4-nitroaniline (NA) was investigated under residence time of 1-6 min, pressure of 18-26 MPa, temperature of 350-500 °C, with initial concentration of 1 mM and 300% excess oxygen. Among these operating conditions, temperature and residence time played a more significant role in decomposing TOC and TN than pressure. Moreover, the products of N-containing species were mainly N 2 , ammonia and nitrate. When temperature, pressure and retention time enhanced, the yields of NO 3 - and org-N were reduced, the amount of N 2 was increasing, the proportion of NH 4 + , however, presented a general trend from rise to decline in general. The experiment of aniline/nitrobenzene indicated that TN removal behavior between amino and nitro groups would prefer to happen in the molecule rather than between the molecules, therefore, the smaller interval between the amino and nitro group was the more easily to interreact. This might explain the reason why TN removal efficiency was in an order that 2-NA > 3-NA > 4-NA. The NH 4 + /NO 3 - experiment result demonstrated that ammonia and nitrate did convert into N 2 during SCWO, however, the formation of N 2 was little without auxiliary fuel. Density functional theory (DFT) method was used to calculate the molecular structures of 2-, 3- and 4-NA to further explore reaction mechanism, which verified that amino group was more easily to be attacked than nitro group. Based on these results, the conceivable reaction pathways of 2-, 3- and 4-NA were proposed, which contained three parts, namely denitrification, ring-open and mineralization. Copyright © 2018. Published by Elsevier Ltd.

  7. Supercritical water treatment of heavy metal and arsenic metalloid-bioaccumulating-biomass.

    Science.gov (United States)

    Li, Jianxin; Chen, Jinbo; Chen, Shan

    2018-08-15

    Hyperaccumulator biomass, as a promising resource for renewable energy that can be converted into valuable fuel productions with high conversion efficiency, must be considered as hazardous materials and be carefully treated before further reuse due to the high contents of heavy metals. In this study, Pteris vittata L., an As-hyperaccumulator biomass was treated by an effective and environmental friendly method-supercritical water gasification (SCWG) using a bench-scale batch reactor. The contents of heavy metals (Cd, Pb and Zn) and arsenic metalloid in solid, liquid and gaseous products during SCWG process were thoroughly investigated. The speciation fractions including exchangeable, reducible, oxidizable and residual fractions of each heavy metal as the proportion of the total contents in solid residue were presented and the transformations trend of these heavy metals during the SCWG process was especially demonstrated. The significant operating parameters, including reaction temperature (395-445 °C), pressure (21-27 MPa) and residence time (0-40 min) were varied to explore their effects on the contents and forms. Moreover, the environmental risks of heavy metals in solid residues were evaluated based on risk assessment code, taking into consideration the speciation fractions and bioavailability. It was highlighted that although heavy metals particularly Pb and Zn tended to accumulate in solid residues with a maximum increment of about 50% in the total content, they were mostly converted to more stable oxidizable and residual fractions, and thus the ecotoxicity and bioavailability were greatly mitigated with no obvious increase in direct toxicity fractions. Each tested heavy metal presented no or low risk to the environments after SCWG treatments, meaning that the environmental pollution levels were markedly reduced with no or low risk to the environment. This study highlights the remarkable ability of SCWG for the heavy metal stabilization. Copyright

  8. Flow method for rapid production of Batio3 nanoparticles in supercritical water

    International Nuclear Information System (INIS)

    Atashfaraz, M.; Shariati-Niassar, M.; Ohara, Satoshi; Takami, S.; Umetsu, M.; Naka, T.; Adschiri, T.

    2006-01-01

    Fine BaTiO 3 nanoparticles were obtained by hydrothermal synthesis under supercritical conditions with batch and flow type experimental methods. Mixture of barium hydroxide and titanium oxide starting solution was treated in the supercritical wafer at 400 d eg C and 30 MPa. The size of nanoparticles synthesized in the flow type experiment was smaller than that in the batch type. Rapid heating in a flow, reactor is effective to synthesize smaller size and narrower particle size distribution for the BaTiO 3 , nanoparticles. The mechanism for this result was discussed based on the solubility of titanium oxide

  9. Severe accident in pressurized water reactors: molten fuel-coolant interaction

    International Nuclear Information System (INIS)

    Battail-Claret, Sylvie

    1993-01-01

    In order to study the phenomenon of interaction between corium and water, the author of this research thesis proposes a scenario to describe the behaviour of a drop of molten iron oxide suddenly plunged into a bath of liquid at room temperature. First, she addresses the modelling of the evolution of the vapour film which surrounds the hot drop and comprises a phase of establishment of a steady film and the phase of destabilisation of this film when an external pressure wave passes by. Besides, she modelled the process of fragmentation of a hot body induced by the destabilisation of a process due to the impact of liquid water micro-jets with water trapping in the hot body. Finally, a model of 'bubble dynamics' is proposed to describe the evolution of the vapour bubble fed by fragments. Theoretical results are compared with experimental results [fr

  10. Modeling the transport of hydrogen in the primary coolant of pressurized heavy water reactors

    International Nuclear Information System (INIS)

    Subramanian, H.; Velmurugan, S.; Narasimhan, S.V.; Jain, A.K.; Dash, S.C.

    2008-01-01

    Heavy water (D 2 O) is used in primary heat transport systems of PHWRs. To suppress the radiolysis of heavy water and to control oxygen, hydrogen is added at regular intervals to the primary heat transport system. The added hydrogen finds it way to the heavy water storage tank after passing through the bleed condenser. Owing to the different temperatures and two phase region present in these systems, hydrogen gets redistributed. It is important to know the concentration of dissolved hydrogen in these regions in order to ensure a steady state dissolved hydrogen concentration in the primary system. Different power stations report variations in the frequency and quantity of hydrogen added to achieve the prescribed steady state level. This paper makes an attempt to account for the inventory of hydrogen and model its transport in PHT system. (author)

  11. Reflooding phase after loss of coolant of an advanced pressurized water reactor with high conversion ratio

    International Nuclear Information System (INIS)

    Schumann, S.

    1984-01-01

    The emergency core cooling behaviour of an advanced pressurized water reactor (APWR) during the reflooding phase of the LOCA with double-ended break is analysed and compared to a common pressurized water reactor (PWR). The code FLUT-BS, its models and correlations are explained in detail and have been verified by numerous PWR-reflood experiments with large parameter range. The influence of core-design on ECC-behaviour as well as the influences of initial and boundary values are examined. The results show the essential differences of ECC-behaviour between PWR and APWR. (orig.) [de

  12. Mass transfer of SCWO processes: Molecular diffusion and mass transfer coefficients of inorganic nitrate species in sub- and supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Goemans, M.G.E.; Gloyna, E.F. [Univ. of Texas, Austin, TX (United States). Dept. of Civil Engineering; Buelow, S.J. [Los Alamos National Lab., NM (United States)

    1996-04-01

    Molecular diffusion coefficients of lithium-, sodium-, potassium-, cesium-, calcium-, and strontium nitrate in subcritical water were determined by analysis of Taylor dispersion profiles. Pressures ranged from 300 to 500 bar at temperatures ranging from 25{degrees}C to 300{degrees}C. The reported diffusion values were determined at infinite dilution. Molecular diffusion coefficients were 10 to 20 times faster in near-critical subcritical water than in water at ambient temperature and pressure (ATP). These findings implied that the diffusion rates were more liquid like than they were gas like, hence experimental results were correlated with diffusion models for liquids. The subcritical diffusion data presented in this work, and supercritical diffusion results published elsewhere were correlated with hydrodynamic diffusion equations. Both the Wilke-Chang correlation and the Stokes-Einstein equation yielded predictions within 10% of the experimental results if the structure of the diffusing species could be estimated. The effect of the increased diffusion rates on mass transfer rates in supercritical water oxidation applications was quantified, with emphasis on heterogeneous oxidation processes. This study and results published elsewhere showed that diffusion limited conditions are much more likely to be encountered in SCWO processes than commonly acknowledged.

  13. The effect of water chemistry on a change in the composition of gas phase in the steam-water path of a supercritical-pressure boiler

    Science.gov (United States)

    Belyakov, I. I.; Belokonova, A. F.

    2010-07-01

    We present the results from an experimental research work on studying the behavior of the gas phase in the path of a supercritical-pressure boiler during its operation with different water chemistries, including all-volatile (hydrazine-ammonia), complexone, neutral oxygenated, and combined oxygenated-ammonia chemistries. It is shown that the minimal content of hydrogen in steam is achieved if feedwater is treated with oxygen.

  14. SEAFP cooling system design. Task M8 - water coolant option (final report)

    International Nuclear Information System (INIS)

    Stubley, P.; Natalizio, A.

    1994-01-01

    This report contains the ex-vessel portions of the outline designs for first wall, blanket and divertor cooling using water as the heat transport fluid. Equipment layout, key components and main system parameters are also described. (author). 7 tabs., 14 figs

  15. Research on Coolant Radiochemistry

    International Nuclear Information System (INIS)

    Ha, Yeong Keong; Kim, W. H.; Yeon, J. W.; Jung, Y. J.; Choi, K. C.; Choi, K. S.; Park, Y. J.; Cho, Y. H.

    2007-06-01

    The final objective of this study is to develop a method for reducing radioactive materials formed in the reactor coolant circuit. This second stage research was categorized into the following three subgroups: the development of the estimation technique of microscopic chemical variation at high temperatures and pressures, the fundamental study on the thermodynamics at high temperatures and pressures, and the study on the deposition of metal oxides and the determination of the main factors responsible for the growth of CRUD. First, in the development of the estimation technique of microscopic chemical change at high temperatures and pressures, the technique for measuring coolant chemistry such as pH, conductivity and Eh was developed to be appropriate for the high temperature and pressure condition. The coolant chemistry measuring system including the self-devised high temperature pH sensor can be applied to the field of nuclear reactor and contribute on a large scale in the automation of the coolant chemistry control and the establishment of the real-time on-line measuring technique. Secondly, the dissociation constant of water and the solubility of metal oxides were measured in the fundamental study on the thermodynamics at high temperatures and pressures. Finally, in the study on the deposition of metal oxides and the determination of the main factors responsible for the growth of CRUD, the careful investigation of the deposition phenomena of micro particles on the cladding surface showed that subcooled boiling and the dissolved hydrogen are the main factors responsible for the growth of CRUD. In addition, the basis was provided for the construction of a new particle behavior model in the reactor coolant circuit

  16. The initial study on supercritical water flow and heat transfer in square rod bundle channel with mixing vane

    International Nuclear Information System (INIS)

    Zuo Guoping; Cao Can; Yu Tao

    2010-01-01

    Three-dimensional rectangular channel with the mixing wine in supercritical water reactor was studied in the paper using the FLUENT software. The mixing wing elevation influence on temperature distribution and flow field were studied in the model. The results showed the mixing wing caused fluid circumferential flow, making flow hot and cold fluids mixed and fluid temperature uniform distribution, effectively improved the fuel rod surface temperature distribution and reduced hot temperature. Among the four cases of mixing wing elevation of 15, 30, 45 and 50 angle, 30 angle is the best case in improving temperature distribution. (authors)

  17. Study on neutronics performance of flower shape advanced supercritical water cooled fast reactor with different solid moderators

    International Nuclear Information System (INIS)

    Yu Tao; Li Zhifeng; Xie Jinsen; Peng Honghua

    2015-01-01

    The supercritical water cooled fast reactors worked at such harsh condition with high temperature and high pressure, huge hydrogen balance pressure and thermal shock can result in a great loss of hydrogen. The released hydrogen would be out of control under accident situations. K_e_f_f, conversion ratio, moderator temperature effect, Doppler effect and void effect of different material such as ZrH_1_._7, Bp, BeO, C and SiC are discussed. BeO and SiC hold better integrated performance among these materials. Besides, moderators have less effect on the Doppler effect of fuel. (authors)

  18. A study on physical characteristics of supercritical light - water reactor loaded with (232U-238Th-238U) oxide fuel

    International Nuclear Information System (INIS)

    Kulikov, E. G.; Shmelev, A. N.; Apse, V. A.; Kulikov, G. G.

    2007-01-01

    The attractiveness of using (U-Th)-fuel in supercritical light water reactor is considered. The dilution of 2 33U in 2 38U is proposed with the purpose of increasing non-proliferation of this fissile isotope. Comparison of different fuel compositions is accomplished from the point of view of fissile isotope breeding and achieved burn-up; parasitic neutron absorption cross-sections are also compared. It is analyzed the impact for neutron balance of both cladding materials: zirconium alloy and stainless steel

  19. Development of a lab-scale contaminated organic effluents treatment process using evaporation and supercritical water oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Turc, H.A.; Joussot-Dubien, C

    2004-07-01

    The organic liquid waste produced in the ATALANTE facility have to be treated in order to reduce the fire and contamination risks. Therefore, the Mini-DELOS process has been developed, which combines a low pressure evaporator in a shielded enclosure and a continuous supercritical water oxidation (SCWO) reactor in a glovebox. Evaporation makes it possible to evacuate the main organic stream as decontaminated distillates to an industrial incinerator. The remaining residue, concentrating the radioactivity can be converted through SCWO into a contaminated aqueous effluent, fully compatible with the existing outlets of the facility. The preliminary results of the first year of active operation of the Mini- DELOS process are here presented. (authors)

  20. Analysis of product distribution and characteristics in hydrothermal liquefaction of barley straw in subcritical and supercritical water

    DEFF Research Database (Denmark)

    Zhu, Zhe; Toor, Saqib; Rosendahl, Lasse

    2014-01-01

    In this study, hydrothermal liquefaction of barley straw in subcritical and supercritical water with potassium carbonate catalyst was performed in the temperatures range of 280-400°C. The influence of final reaction temperature on products yield was investigated and some physicochemical properties...... yield (35.24 wt %) as well as the maximum energy recovery of 55.33% were obtained at 300°C. The products obtained were characterized in terms of CHNS elemental composition, higher heating values (HHVs), Fourier transform infrared spectroscopy (FTIR) and gas chromatography-mass spectrometer (GC...

  1. SUPERCRITICAL WATER PARTIAL OXIDATION PHASE I - PILOT-SCALE TESTING/FEASIBILTY SUDIES FINAL REPORT

    Energy Technology Data Exchange (ETDEWEB)

    SPRITZER.M; HONG,G

    2005-01-01

    General Atomics (GA) is developing Supercritical Water Partial Oxidation (SWPO) as a means of producing hydrogen from low-grade biomass and other waste feeds. The Phase I Pilot-scale Testing/Feasibility Studies have been successfully completed and the results of that effort are described in this report. The key potential advantage of the SWPO process is the use of partial oxidation in-situ to rapidly heat the gasification medium, resulting in less char formation and improved hydrogen yield. Another major advantage is that the high-pressure, high-density aqueous environment is ideal for reacting and gasifying organics of all types. The high water content of the medium encourages formation of hydrogen and hydrogen-rich products and is especially compatible with high water content feeds such as biomass materials. The high water content of the medium is also effective for gasification of hydrogen-poor materials such as coal. A versatile pilot plant for exploring gasification in supercritical water has been established at GA's facilities in San Diego. The Phase I testing of the SWPO process with wood and ethanol mixtures demonstrated gasification efficiencies of about 90%, comparable to those found in prior laboratory-scale SCW gasification work carried out at the University of Hawaii at Manoa (UHM), as well as other biomass gasification experience with conventional gasifiers. As in the prior work at UHM, a significant amount of the hydrogen found in the gas phase products is derived from the water/steam matrix. The studies at UHM utilized an indirectly heated gasifier with an activated carbon catalyst. In contrast, the GA studies utilized a directly heated gasifier without catalyst, plus a surrogate waste fuel. Attainment of comparable gasification efficiencies without catalysis is an important advancement for the GA process, and opens the way for efficient hydrogen production from low-value, dirty feed materials. The Phase I results indicate that a practical

  2. Interfacing systems LOCA (loss-of-coolant accidents): Pressurized water reactors

    International Nuclear Information System (INIS)

    Bozoki, G.; Kohut, P.; Fitzpatrick, R.

    1989-02-01

    This report summarizes a study performed by Brookhaven National Laboratory for the Office of Nuclear Regulatory Research, Reactor and Plant Safety Issues Branch, Division of Reactor and Plant Systems, US Nuclear Regulatory Commission. This study was requested by the NRC in order to provide a technical basis for the resolution of Generic Issue 105 ''Interfacing LOCA at LWRs.'' This report deals with pressurized water reactors (PWRs). A parallel report was also accomplished for boiling water reactors. This study focuses on three representative PWRs and extrapolates the plant-specific findings for their generic applicability. In addition, a generic analysis was performed to investigate the cost-benefit aspects of imposing a testing program that would require some minimum level of leak testing of the pressure isolation valves on plants that presently have no such requirements. 28 refs., 31 figs., 64 tabs

  3. Review of boiling water reactor small break loss of coolant accidents

    International Nuclear Information System (INIS)

    Gururaj, P.M.; Dua, S.S.; Rao, A.S.

    1981-01-01

    This paper presents a review of the analytical and the experimental work performed by the General Electric Company to determine the performance of boiling water reactors (BWR) following postulated small break accidents (SBA). This review paper addresses the following issues: (1) the response of the BWR following small loss of inventory events; (2) methods of analysis and their justification; (3) necessity, if any, of operator action and the length of time available in which such action can be performed; and (4) operator interface following the SBA event. The results from these SBA studies for different BWR product lines show that even with the multiple system failures assumed, the BWR can successfully withstand an SBA. For a typical BWR/6, it takes the failure of 13 water delivery pumps to cause any significant core heatup. The only operator actions determined to be necessary are simple ones and ample time is available to the operator to perform these actions, if needed

  4. Continuous radiochemical analysis of fission products in a nuclear reactor water coolant

    International Nuclear Information System (INIS)

    Moskvin, L.N.; Zakharov, L.K.; Leont'ev, G.G.; Mel'nikov, V.A.; Orlenkov, I.S.; Slutskij, G.K.

    1975-01-01

    Method for continuous radiochemical analysis of I, Cs, Ba, Sr and Ce isotopes in a reactor water heat-transfer agent was developed. A continuous two-dimensional chromatographic process of complex mixtures separation of substances proved to be feasible on several parallel sorbent layers, which moved at constant velocities and separated by stationary intermediate collectors. Tests on model solutions containing I, Ce, Cs and Ba isotopes and on heat-carrier samples showed quantitative separation of elements. The results were indicative of a basic possibility of using multisorbent chromatographs for continuous control of multicomponent mixtures, particularly for control of radioactive fission product compositions in water heat-transfer agents in nuclear power plants. A diagram is shown for a two-dimensional chromatographic separation of a multicomponent mixture. Also shown is a flow chart of an installation for continuous control of iodine and cesium isotope activities

  5. Nonthermal inactivation of Escherichia coli K12 in buffered peptone water using a pilot-plant scale supercritical carbon dioxide system with gas-liquid porous metal contractor

    Science.gov (United States)

    This study evaluated the effectiveness of a supercritical carbon dioxide (SCCO2) system, with a gas-liquid CO2 contactor, for reducing Escherichia coli K12 in diluted buffered peptone water. 0.1% (w/v) buffered peptone water inoculated with E. coli K12 was processed using the SCCO2 system at CO2 con...

  6. Application of the regulations on pressurized components or light water reactor primary coolant circuits

    International Nuclear Information System (INIS)

    Barthelemy, F.; Menjon, G.

    1977-01-01

    This paper describes the philosophy and the provisions of the Order of 26 February 1974 concerning application of the regulations on pressurized components for light water reactor steam supply systems. The aim is to show how these regulations which differ from other regulations on pressurized components and is more detailed on many points, is applied in practice in France in the various stages of the design, construction and operation of PWRs. (NEA) [fr

  7. Preparation of Nanocrystalline Titania Thin Films by Using Pure and Water-modified Supercritical Carbon Dioxide.

    Czech Academy of Sciences Publication Activity Database

    Sajfrtová, Marie; Cerhová, Marie; Dřínek, Vladislav; Daniš, S.; Matějová, L.

    2016-01-01

    Roč. 117, NOV 2016 (2016), s. 289-296 ISSN 0896-8446 R&D Projects: GA ČR GA14-23274S Institutional support: RVO:67985858 Keywords : titania thin films * supercritical carbon dioxide * crystallization Subject RIV: CA - Inorganic Chemistry Impact factor: 2.991, year: 2016

  8. Radiological consequence analyses of loss of coolant accidents of various break sizes of Pressurized Heavy Water Reactor

    International Nuclear Information System (INIS)

    Sanyasi Rao, V.V.S.; Hari Prasad, M.; Ghosh, A.K.

    2010-01-01

    For any advanced technology, it is essential to ensure that the consequences associated with the accident sequences arising, if any, from the operation of the plant are as low as possible and certainly below the guidelines/limits set by the regulatory bodies. Nuclear power is no exception to this. In this paper consequences of the events arising from Loss of Coolant Accident (LOCA) sequences in Pressurized Heavy Water Reactor (PHWR), are analysed. The sequences correspond to different break sizes of LOCA followed by the operation or otherwise of Emergency Core Cooling System (ECCS). Operation or otherwise of the containment safety systems has also been considered. It has been found that there are no releases to the environment when ECCS is available. The releases, when ECCS is not available, arise from the slack and the ground. The radionuclides considered include noble gases, iodine, and cesium. The hourly meteorological parameters (wind speed, wind direction, precipitation and stability category), considered for this study, correspond to those of Kakrapar site. The consequences evaluated are the thyroid dose and the bone marrow dose received by a person located at various distances from the release point. Isodose curves are generated. From these evaluations, it has been found that the doses are very low. The complementary cumulative frequency distributions (CCFD) for thyroid and bone marrow doses have also been presented for the cases analysed. (author)

  9. Study on characteristics for different moderation ratios of heavy water coolant with different reactor types in equilibrium states

    International Nuclear Information System (INIS)

    Permana, Sidik; Takaki, Naoyuki; Sekimoto, Hiroshi

    2005-01-01

    Several characteristics for different moderation ratios of heavy water coolant with different reactor types in equilibrium states have been investigated. Performances of PWR and CANDU reactors are compared. A calculation method for determining the required uranium enrichment for criticality of the systems has been developed by coupling the equilibrium fuel cycle burn-up calculation and cell calculation of PIJ module of SRAC2000 code. In the present study, we have compared the characteristics for different moderator to fuel ratio (MFR, 0.1 to 30), different burn-up for CANDU type and four fuels cycle schemes. Nuclide density of 235 U at MFR=1.9 decreases with increasing number of confined HM, while 235 U at higher MFR has the opposite trend. However, the nuclide density of fissile material at higher MFR is lower except 238 U. CANDU type requires lower uranium enrichment and obtains higher conversion ratio than PWR type. Lowest burn-up requires the lowest uranium enrichment and obtains the highest conversion ratio. The breeding condition can be obtained for plutonium recycle cases at MFR=2.1 of Case 4 and MFR=1.4 of Case 3. The natural uranium can be achieved at MFR=14 of plutonium recycle cases, and it can be used easier by increasing number of confined HM. (author)

  10. Requirements on cast steel for the primary coolant circuit of water cooled reactors

    International Nuclear Information System (INIS)

    The most important requirements placed on the structural components of water cooled nuclear reactors include corrosion resistance and mechanical materials properties. Intercrystalline corrosion resistance was tested using the Strauss Test in compliance with the DIN 50914 Standard. Following sensitization between 600 to 700 degC with a dwell time between 15 minutes and 100 hours, a specimen homogeneously annealed with the casting and rapidly water cooled showed no intercrystalline corrosion. Specimens cooled from 1050 degC at a rate of 100 degC per hour showed no unambiguous tendency for intercrystalline corrosion after sensitization; in some cases, however, an initial attack of intercrystalline corrosion was found. It was found that austenitic Cr-Ni cast steel containing 2.5% Mo and about 15% ferrite showed the sensitive intercrystalline corrosion range at higher temperatures and longer dwell times than rolled Cr-Ni steels. In plating the ferritic cast steel with a corrosion resistant plating material, annealing temperature after welding must not exceed 600 to 620 degC otherwise the resistance of the plated layer against intercrystalline corrosion would not be safeguarded, and following annealing for stress removal at a temperature of 600 to 620 degC all requirements must be satisfied by the weld metal and weld transition placed on the initial material. Martensite materials are used for the manufacture of components which are not used under pressure, such as alloys with 13% Cr and 1% to 6% Ni and alloys with 17% Cr and 4% Ni. Carbon content is maintained below 0.10% to guarantee good weldability and the highest corrosion resistance. Cast steels with 13% Cr and 4% Ni after a dwell of 2500 hours in fully desalinated water without oxygen and with 3600 ppm of boron at a test temperature of 95 to 300 degC showed a surface reduction of 0.005 mm annually. In identical conditions except for the water containing oxygen the reduction in surface was 0.05 mm per year. (J.B.)

  11. Study on the LLFPs transmutation in a super-critical water-cooled fast reactor

    International Nuclear Information System (INIS)

    Lu Haoliang; Ishiwatari, Yuki; Oka, Yoshiaki

    2011-01-01

    Research highlights: → Transmutation of LLFPs with a super-criticial water cooled fast reactor. → Transmutation of iodine and cesium without the isotopic separation. → The transmuted isotope was mixed with UO 2 to reduce the effect of self-shielding. → A weak neutron moderator Al 2 O 3 was used to suppress the creation of 135 Cs from 133 Cs. - Abstract: The performance of the super-critical water-cooled fast reactor (Super FR) for the transmutation treatment of long-lived fission products (LLFPs) was evaluated. Two regions with the soft neutron spectrum, which is of great benefit to the LLFPs transmutation, can be utilized in the Super FR. First region is in the blanket assembly due to the ZrH 1.7 layer which was utilized to slow down the fast neutrons to achieve a negative void reactivity. Second region is in the reflector region of core like other metal-cooled fast reactors. The LLFPs selected in the transmutation analysis include 99 Tc, 129 I and 135 Cs discharged from LWR or fast reactor. Their isotopes, such as 127 I, 133 Cs, 134 Cs and 137 Cs were also considered to avoid the separation. By loading the isotopes ( 99 Tc or 127 I and 129 I) in the blanket assembly and the reflector region simultaneously, the transmutation rates of 5.36%/GWe year and 2.79%/GWe year can be obtained for 99 Tc and 129 I, respectively. The transmuted amounts of 99 Tc and 129 I are equal to the yields from 11.8 and 6.2 1000 MWe-class PWRs. Because of the very low capture cross section of 135 Cs and the effect of other cesium isotopes, 135 Cs was loaded with three rings of assemblies in the reflector region to make the transmuted amount be larger than the yields of two 1000 MWe-class PWRs. Based on these results, 99 Tc and 129 I can be transmuted conveniently and higher transmutation performance can be obtained in the Super FR. However, the transmutation of 135 Cs is very difficult and the transmuted amount is less than that produced by the Super FR. It turns out that the

  12. Continuous Hydrothermal Flow Synthesis of LaCrO3 in Supercritical Water and Its Application in Dual-Phase Oxygen Transport Membranes

    DEFF Research Database (Denmark)

    Xu, Yu; Pirou, Stéven; Zielke, Philipp

    2018-01-01

    The continuous production of LaCrO3 particles (average edge size 639 nm, cube-shaped) by continuous hydrothermal flow synthesis using supercritical water is reported for the first time. By varying the reaction conditions, it was possible to suggest a reaction mechanism for the formation of this p......The continuous production of LaCrO3 particles (average edge size 639 nm, cube-shaped) by continuous hydrothermal flow synthesis using supercritical water is reported for the first time. By varying the reaction conditions, it was possible to suggest a reaction mechanism for the formation...

  13. Coolant clean-up system in the primary coolant circuit for nuclear reactor

    International Nuclear Information System (INIS)

    Saito, Michio.

    1981-01-01

    Purpose: To maintain the quality of coolants at a prescribed level by distillating coolants in the primary coolant circuit for a BWR type reactor to remove impurities therefrom, taking out the condensates from the top of the distillation column and extracting impurities in a concentrated state from the bottom. Constitution: Coolant water for cooling the core is recycled by a recycling pump by way of a recycling pipeway in a reactor. The coolants extracted from an extraction pipeway connected to the recycling pipeway are fed into a distillation column, where distillation is taken place. Impurities in the coolants, that is, in-core corrosion products, fission products generated in the reactor core, etc. are separated by the distillation, concentrated and solidified in the bottom of the distillation column. While on the other hand, condensates removed with the impurities, that is, coolants cleaned-up are recycled to the coolant water for cooling the reactor core. (Moriyama, K.)

  14. In-Situ Synchrotron Radiation Study of Formation and Growth of Crystalline CexZr1-xO2 Nanoparticles Synthesized in Supercritical Water

    DEFF Research Database (Denmark)

    Tyrsted, Christoffer; Becker-Christensen, Jacob; Hald, Peter

    2010-01-01

    -zirconia system, the growth of ceria and zirconia nanoparticles is fundamentally different under supercritical water conditions. For comparison, ex situ synthesis has also been performed using an in-house supercritical flow reactor. The resulting samples were analyzed using PXRD, small-angle X-ray scattering......In situ synchrotron powder X-ray diffraction (PXRD) measurements have been conducted to follow the nucleation and growth of crystalline CexZr1-xO2 nanoparticles synthesized in supercritical water with a full substitution variation (x = 0, 0.2, 0.5, 0.8, and 1.0). Direction-dependent growth curves...... are determined and described using reaction kinetic models. A distinct change in growth kinetics is observed with increasing cerium content. For x = 0.8 and 1.0 (high cerium content), the growth is initially limited by the surface reaction kinetics; however, at a size of ∼6 nm, the growth changes and becomes...

  15. Effect of thermal treatment on the corrosion resistance of Type 316L stainless steel exposed in supercritical water

    Energy Technology Data Exchange (ETDEWEB)

    Jiao, Y. [Department of Materials Science & Engineering, McMaster University, Hamilton, ON (Canada); Zheng, W. [CanmetMATERIALS, Natural Resources Canada, Hamilton, ON (Canada); Guzonas, D.A. [Canadian Nuclear Laboratories Chalk River Laboratories, ON (Canada); Cook, W.G. [Department of Chemical Engineering, University of New Brunswick, Fredericton, NB (Canada); Kish, J.R., E-mail: kishjr@mcmaster.ca [Department of Materials Science & Engineering, McMaster University, Hamilton, ON (Canada)

    2015-09-15

    There are still unknown aspects about the growth mechanism of oxide scales formed on candidate stainless steel fuel cladding materials during exposure in supercritical water (SCW) under the conditions relevant to the Canadian supercritical water-cooled reactor (SCWR). The tendency for intermetallic precipitates to form within the grains and on grain boundaries during prolonged exposure at high temperatures represents an unknown factor to corrosion resistance, since they tend to bind alloyed Cr. The objective of this study was to better understand the extent to which intermetallic precipitates affects the mode and extent of corrosion in SCW. Type 316L stainless steel, used as a model Fe–Cr–Ni–Mo alloy, was exposed to 25 MPa SCW at 550 °C for 500 h in a static autoclave for this purpose. Mechanically-abraded samples were tested in the mill-annealed (MA) and a thermally-treated (TT) condition. The thermal treatment was conducted at 815 °C for 1000 h to precipitate the carbide (M{sub 23}C{sub 6}), chi (χ), laves (η) and sigma (σ) phases. It was found that although relatively large intermetallic precipitates formed at the scale/alloy interface locally affected the oxide scale formation, their discontinuous formation did not affect the short-term overall apparent corrosion resistance.

  16. Effect of thermal treatment on the corrosion resistance of Type 316L stainless steel exposed in supercritical water

    Science.gov (United States)

    Jiao, Y.; Zheng, W.; Guzonas, D. A.; Cook, W. G.; Kish, J. R.

    2015-09-01

    There are still unknown aspects about the growth mechanism of oxide scales formed on candidate stainless steel fuel cladding materials during exposure in supercritical water (SCW) under the conditions relevant to the Canadian supercritical water-cooled reactor (SCWR). The tendency for intermetallic precipitates to form within the grains and on grain boundaries during prolonged exposure at high temperatures represents an unknown factor to corrosion resistance, since they tend to bind alloyed Cr. The objective of this study was to better understand the extent to which intermetallic precipitates affects the mode and extent of corrosion in SCW. Type 316L stainless steel, used as a model Fe-Cr-Ni-Mo alloy, was exposed to 25 MPa SCW at 550 °C for 500 h in a static autoclave for this purpose. Mechanically-abraded samples were tested in the mill-annealed (MA) and a thermally-treated (TT) condition. The thermal treatment was conducted at 815 °C for 1000 h to precipitate the carbide (M23C6), chi (χ), laves (η) and sigma (σ) phases. It was found that although relatively large intermetallic precipitates formed at the scale/alloy interface locally affected the oxide scale formation, their discontinuous formation did not affect the short-term overall apparent corrosion resistance.

  17. Recovery of copper and lead from waste printed circuit boards by supercritical water oxidation combined with electrokinetic process

    International Nuclear Information System (INIS)

    Xiu Furong; Zhang Fushen

    2009-01-01

    An effective and benign process for copper and lead recovery from waste printed circuit boards (PCBs) was developed. In the process, the PCBs was pre-treated in supercritical water, then subjected to electrokinetic (EK) process. Experimental results showed that supercritical water oxidation (SCWO) process was strong enough to decompose the organic compounds of PCBs, and XRD spectra indicated that copper and lead were oxidized into CuO, Cu 2 O and β-PbO 2 in the process. The optimum SCWO treatment conditions were 60 min, 713 K, 30 MPa, and EK treatment time, constant current density were 11 h, 20 mA cm -2 , respectively. The recovery percentages of copper and lead under optimum SCWO + EK treatment conditions were around 84.2% and 89.4%, respectively. In the optimized EK treatment, 74% of Cu was recovered as a deposit on the cathode with a purity of 97.6%, while Pb was recovered as concentrated solutions in either anode (23.1%) or cathode (66.3%) compartments but little was deposited on the electrodes. It is believed that the process is effective and practical for Cu and Pb recovery from waste electric and electronic equipments.

  18. Supercritical water oxidation of dioxins and furans in waste incinerator fly ash, sewage sludge and industrial soil.

    Science.gov (United States)

    Zainal, Safari; Onwudili, Jude A; Williams, Paul T

    2014-08-01

    Three environmental samples containing dioxins and furans have been oxidized in the presence of hydrogen peroxide under supercritical water oxidation conditions. The samples consisted of a waste incinerator fly ash, sewage sludge and contaminated industrial soil. The reactor system was a batch, autoclave reactor operated at temperatures between 350 degrees C and 450degrees C, corresponding to pressures of approximately 20-33.5 MPa and with hydrogen peroxide concentrations from 0.0 to 11.25 vol%. Hydrogen peroxide concentration and temperature/pressure had a strong positive effect on the oxidation of dioxins and furans. At the highest temperatures and pressure of supercritical water oxidation of 4500C and 33.5 MPa and with 11.25 vol% of hydrogen peroxide, the destruction efficiencies of the individual polychlorinated dibenzo-p-dioxins/polychlorinated dibenzofurans (PCDD/PCDF) isomers were between 90% and 99%. There did not appear to be any significant differences in the PCDD/PCDF destruction efficiencies in relation to the different sample matrices of the waste incinerator fly ash, sewage sludge and contaminated industrial soil.

  19. Hydrogen production by supercritical water gasification of biomass. Phase 1 -- Technical and business feasibility study, technical progress report

    Energy Technology Data Exchange (ETDEWEB)

    NONE

    1997-12-01

    The nine-month Phase 1 feasibility study was directed toward the application of supercritical water gasification (SCWG) for the economical production and end use of hydrogen from renewable energy sources such as sewage sludge, pulp waste, agricultural wastes, and ultimately the combustible portion of municipal solid waste. Unique in comparison to other gasifier systems, the properties of supercritical water (SCW) are ideal for processing biowastes with high moisture content or contain toxic or hazardous contaminants. During Phase I, an end-to-end SCWG system was evaluated. A range of process options was initially considered for each of the key subsystems. This was followed by tests of sewage sludge feed preparation, pumping and gasification in the SCW pilot plant facility. Based on the initial process review and successful pilot-scale testing, engineering evaluations were performed that defined a baseline system for the production, storage and end use of hydrogen. The results compare favorably with alternative biomass gasifiers currently being developed. The results were then discussed with regional wastewater treatment facility operators to gain their perspective on the proposed commercial SCWG systems and to help define the potential market. Finally, the technical and business plans were developed based on perceived market needs and the projected capital and operating costs of SCWG units. The result is a three-year plan for further development, culminating in a follow-on demonstration test of a 5 MT/day system at a local wastewater treatment plant.

  20. Recycling acetic acid from polarizing film of waste liquid crystal display panels by sub/supercritical water treatments.

    Science.gov (United States)

    Wang, Ruixue; Chen, Ya; Xu, Zhenming

    2015-05-19

    Waste liquid crystal display (LCD) panels mainly contain inorganic materials (glass substrate) and organic materials (polarizing film and liquid crystal). The organic materials should be removed first since containing polarizing film and liquid crystal is to the disadvantage of the indium recycling process. In the present study, an efficient and environmentally friendly process to obtain acetic acid from waste LCD panels by sub/supercritical water treatments is investigated. Furthermore, a well-founded reaction mechanism is proposed. Several highlights of this study are summarized as follows: (i) 99.77% of organic matters are removed, which means the present technology is quite efficient to recycle the organic matters; (ii) a yield of 78.23% acetic acid, a quite important fossil energy based chemical product is obtained, which can reduce the consumption of fossil energy for producing acetic acid; (iii) supercritical water acts as an ideal solvent, a requisite reactant as well as an efficient acid-base catalyst, and this is quite significant in accordance with the "Principles of Green Chemistry". In a word, the organic matters of waste LCD panels are recycled without environmental pollution. Meanwhile, this study provides new opportunities for alternating fossil-based chemical products for sustainable development, converting "waste" into "fossil-based chemicals".

  1. Development of sub-channel/system coupled code and its application to a supercritical water-cooled test loop

    International Nuclear Information System (INIS)

    Liu, X.J.; Yang, T.; Cheng, X.

    2014-01-01

    To analyze the local thermal-hydraulic parameters in the supercritical water reactor-fuel qualification test (SCWR-FQT) fuel bundle with a flow blockage, a coupled sub-channel and system code system is developed in this paper. Both of the sub-channel code and system code are adapted to transient analysis of SCWR. Two codes are coupled by data transfer and data adaptation at the interface. In the coupled code, the whole system behavior including safety system characteristic is analyzed by system code ATHLET-SC, whereas the local thermal-hydraulic parameters are predicted by the sub-channel code COBRA-SC. Sensitivity analysis are carried out respectively in ATHLET-SC and COBRA-SC code, to identify the appropriate models for description of the flow blockage phenomenon in the test loop. Some measures to mitigate the accident consequence are also trialed to demonstrate their effectiveness. The results indicate that the new developed code has good feasibility to transient analysis of supercritical water-cooled test. And the peak cladding temperature caused by blockage in the fuel assembly can be reduced effectively by the safety measures of SCWR-FQT. (author)

  2. Coolant mixing in pressurized water reactors. Pt. 1. Feasibility of closed analytical solutions and simulation of the mixing with CFX-4. Final report

    International Nuclear Information System (INIS)

    Grunwald, G.; Hoehne, T.; Prasser, H.M.; Rohde, U.

    2001-10-01

    The project was aimed at the analytical and numerical simulation of coolant mixing in the downcomer and the lower plenum of PWRs. Generally, the coolant mixing is of relevance for two classes of accident scenarios - boron dilution and cold water transients. For the investigation of the relevant mixing phenomena, the Rossendorf test facility ROCOM has been designed. ROCOM is a 1:5 scaled Plexiglas trademark model of the PWR Konvoi allowing velocity measurements by the LDA technique. Design and construction of the ROCOM facility including the measurement equipment were performed in a second part of the project. For the design of the facility, CFD calculations were performed to analyze the scaling of the model. It was found, that the scaling of 1:5 to the prototype meets both: physical and economical demands. A theoretical 2D-model of the downcomer flow was developed based on the potential theory. The coolant inlet is represented by mass sources. Potential vortices were superposed to describe large scale recirculations. However, the method requires an a-priory knowledge of the location and intensity of the vorticity sources. Therefore, the main goal of the project was the numerical simulation of the coolant mixing of different PWRs. The temperature and boron concentration fields established by the coolant mixing during nominal and transient flow conditions in the pressure vessel of the PWR Konvoi and the Russian type WWER-440 were investigated. The calculations were carried out with the CFD-code CFX 4. The results of the CFD calculation are found in the final report. The report is based on the Ph.D. work of T. Hoehne. (orig.) [de

  3. EXPERIMENTAL STUDY OF LOCAL HYDRODYNAMICS AND MASS EXCHANGE PROCESSES OF COOLANT IN FUEL ASSEMBLIES OF PRESSURIZED WATER REACTORS

    Directory of Open Access Journals (Sweden)

    S. M. Dmitriev

    2016-01-01

    Full Text Available The results of experimental studies of local hydrodynamics and mass exchange of coolant flow behind spacer and mixing grids of different structural versions that were developed for fuel assemblies of domestic and foreign nuclear reactors are presented in the article. In order to carry out the study the models of the following fuel assemblies have been fabricated: FA for VVER and VBER, FA-KVADRAT for PWR-reactor and FA for KLT-40C reactor. All the models have been fabricated with a full geometrical similarity with full-scale fuel assemblies. The study was carried out by simulating the flow of coolant in a core by air on an aerodynamic test rig. In order to measure local hydrodynamic characteristics of coolant flow five-channel Pitot probes were used that enable to measure the velocity vector in a point by its three components. The tracerpropane method was used for studying mass transfer processes. Flow hydrodynamics was studied by measuring cross-section velocities of coolant flow and coolant rates according to the model cells. The investigation of mass exchange processes consisted of a study of concentration distribution for tracer in experimental model, in determination of attenuation lengths of mass transfer processes behind mixing grids, in calculating of inter-cellar mass exchange coefficient. The database on coolant flow in fuel assemblies for different types of reactors had been accumulated that formed the basis of the engineering substantiation of reactor cores designs. The recommendations on choice of optimal versions of mixing grids have been taken into consideration by implementers of the JSC “OKBM Afrikantov” when creating commissioned fuel assemblies. The results of the study are used for verification of CFD-codes and CFD programs of detailed cell-by-cell calculation of reactor cores in order to decrease conservatism for substantiation of thermal-mechanical reliability.

  4. Development of liquefaction process of coal and biomass in supercritical water; Chorinkaisui wo mochiita sekitan biomass doji ekika process no kaihatsu

    Energy Technology Data Exchange (ETDEWEB)

    Nonaka, H.; Matsumura, Y.; Tsutsumi, A.; Yoshida, K. [The University of Tokyo, Tokyo (Japan). Faculty of Engineering; Masuno, Y.; Inaba, A. [National Institute for Resources and Environment, Tsukuba (Japan)

    1996-10-28

    Liquefaction of coal and biomass in supercritical water has been investigated, in which strong solubilization force of supercritical water against hydrocarbons is utilized. Free radicals are formed through the cleavage of covalent bonds in coal under the heating condition at around 400{degree}C during coal liquefaction. It is important to stabilize these unstable intermediate products by hydrogen transfer. On the other hand, hydrogen is not required for the liquefaction of biomass having higher H/C atomic ratio and oxygen content than those of coal. Co-liquefaction of coal and biomass was conducted using supercritical water, in which excess hydrogen from the liquefaction of biomass would be transferred to coal, resulting in the effective liquefaction of coal. Mixture of coal and cellulose was liquefied in supercritical water at 390{degree}C under the pressure of 25 MPa using a semi-continuous reactor, and the results were compared with those from the separate liquefaction of them. The co-liquefaction of coal and cellulose did not show any difference in the residue yield from the separate liquefaction of these, but led to the increased production of compounds with lower molecular weight. The liquefaction was completed in 15 minutes. 5 refs., 3 figs., 3 tabs.

  5. A numerical thermal-hydraulic model to simulate the fast transients in a supercritical water channel subjected to sharp pressure variations

    NARCIS (Netherlands)

    Dutta, G.; Jiang, J.; Maitri, R.; Zhang, C.

    2016-01-01

    The present work demonstrates the extension of a thermal-hydraulic model, THRUST, with an objective to simulate the fast transient flow dynamics in a supercritical water channel of circular cross section. THRUST is a 1-D model which solves the nonlinearly coupled mass, axial momentum and energy

  6. Heat transfer to sub- and supercritical water flowing upward in a vertical tube at low mass fluxes: numerical analysis and experimental validation

    NARCIS (Netherlands)

    Odu, Samuel Obarinu; Koster, P.; van der Ham, Aloysius G.J.; van der Hoef, Martin Anton; Kersten, Sascha R.A.

    2016-01-01

    Heat transfer to supercritical water (SCW) flowing upward in a vertical heated tube at low mass fluxes (G ≤ 20 kg/m2 s) has been numerically investigated in COMSOL Multiphysics and validated with experimental data. The turbulence models, essential to describing local turbulence, in COMSOL have been

  7. 2D and 3D CFD modelling of a reactive turbulent flow in a double shell supercritical water oxidation reactor

    International Nuclear Information System (INIS)

    Moussiere, S.; Roubaud, A.; Fournel, B.; Joussot-Dubien, C.; Boutin, O.; Guichardon, P.

    2012-01-01

    In order to design and define appropriate dimensions for a supercritical oxidation reactor, a comparative 2D and 3D simulation of the fluid dynamics and heat transfer during an oxidation process has been performed. The solver used is a commercial code, Fluent 6.2 (R). The turbulent flow field in the reactor, created by the stirrer, is taken into account with a k-omega model and a swirl imposed to the fluid. In the 3D case the rotation of the stirrer can be modelled using the sliding mesh model and the moving reference frame model. This work allows comparing 2D and 3D velocity and heat transfer calculations. The predicted values (mainly species concentrations and temperature profiles) are of the same order in both cases. The reactivity of the system is taken into account with a classical Eddy Dissipation Concept combustion model. Comparisons with experimental temperature measurements validate the ability of the CFD modelling to simulate the supercritical water oxidation reactive medium. Results indicate that the flow can be considered as plug flow-like and that heat transfer is strongly enhanced by the stirring. (authors)

  8. Experimental investigation of heat transfer potential of Al2O3/Water-Mono Ethylene Glycol nanofluids as a car radiator coolant

    Directory of Open Access Journals (Sweden)

    Dattatraya G. Subhedar

    2018-03-01

    Full Text Available In this research, the heat transfer potential of Al2O3/Water-Mono Ethylene Glycol nanofluids is investigated experimentally as a coolant for car radiators. The base fluid was the mixture of water and mono ethylene glycol with 50:50 proportions by volume. The stable nanofluids obtained by ultra-sonication are used in all experiments. In this study nanoparticle volume fraction, coolant flow rate, inlet temperature used in the ranges of 0.2–0.8%, 4–9 l per minute and 65–85 °C. The results show that the heat transfer performance of radiator is enhanced by using nanofluids compared to conventional coolant. Nanofluid with lowest 0.2% volume fraction 30% rise in heat transfer is observed. Also the estimation of reduction in frontal area of radiator if base fluid is replaced by Nanofluid is done which will make lighter cooling system, produce less drag and save the fuel cost.

  9. Determination of blade-to-coolant heat-transfer coefficients on a forced-convection, water-cooled, single-stage turbine

    Science.gov (United States)

    Freche, John C; Schum, Eugene F

    1951-01-01

    Blade-to-coolant convective heat-transfer coefficients were obtained on a forced-convection water-cooled single-stage turbine over a large laminar flow range and over a portion of the transition range between laminar and turbulent flow. The convective coefficients were correlated by the general relation for forced-convection heat transfer with laminar flow. Natural-convection heat transfer was negligible for this turbine over the Grashof number range investigated. Comparison of turbine data with stationary tube data for the laminar flow of heated liquids showed good agreement. Calculated average midspan blade temperatures using theoretical gas-to-blade coefficients and blade-to-coolant coefficients from stationary-tube data resulted in close agreement with experimental data.

  10. Blade-to-coolant heat-transfer results and operating data from a natural-convection water-cooled single-stage turbine

    Science.gov (United States)

    Diaguila, Anthony J; Freche, John C

    1951-01-01

    Blade-to-coolant heat-transfer data and operating data were obtained with a natural-convection water-cooled turbine over range of turbine speeds and inlet-gas temperatures. The convective coefficients were correlated by the general relation for natural-convection heat transfer. The turbine data were displaced from a theoretical equation for natural convection heat transfer in the turbulent region and from natural-convection data obtained with vertical cylinders and plates; possible disruption of natural convection circulation within the blade coolant passages was thus indicated. Comparison of non dimensional temperature-ratio parameters for the blade leading edge, midchord, and trailing edge indicated that the blade cooling effectiveness is greatest at the midchord and least at the trailing edge.

  11. Subchannel analysis of Al2O3 nanofluid as a coolant in VMHWR

    International Nuclear Information System (INIS)

    Zarifi, Ehsan; Tashakor, Saman

    2015-01-01

    The main objective of this study is to predict the thermal hydraulic behavior of nanofluids as the coolant in the fuel assembly of variable moderation high performance light water reactor (VMHWR). VMHWR is the new version of high performance light water reactor (HPLWR) conceptual design. Light water reactors at supercritical pressure (VMHWR, HPLWR), being currently under design, are the new generation of nuclear reactors. Water-based nanofluids containing various volume fractions of Al 2 O 3 nanoparticles are analyzed. The conservation equations and conduction heat transfer equation for fuel and clad have been derived and discretized by the finite volume method. The transfer of mass, momentum and energy between adjacent subchannels are split into diversion crossflow and turbulent mixing components. The governed non linear algebraic equations are solved by using analytical iteration methods. Finally the nanofluid analysis results are compared with the pure water results.

  12. Vent clearing during a simulated loss-of-coolant accident in Mark I boiling-water-reactor pressure-suppression system

    International Nuclear Information System (INIS)

    Pitts, J.H.; McCauley, E.W.

    1978-01-01

    The response of the pressure-suspension containment system of Mark I boiling-water reactors to a loss-of-coolant accident (LOCA) is being studied. This response is a design basis for light-water nuclear reactors. Part of the study is being carried out on a 1 / 5 -scale experimental facility that models the pressure-suppression containment system of the Peach Bottom 2 nuclear power plant. The test series reported here focused on the initial or air-clearing phase of a hypothetical LOCA. Measured forces, measured pressures, and the hydrodynamic phenomena (observed with high-speed cameras) show a logical interrelationship

  13. Conducting water chemistry of the secondary coolant circuit of VVER-based nuclear power plant units constructed without using copper containing alloys

    Science.gov (United States)

    Tyapkov, V. F.

    2014-07-01

    The secondary coolant circuit water chemistry with metering amines began to be put in use in Russia in 2005, and all nuclear power plant units equipped with VVER-1000 reactors have been shifted to operate with this water chemistry for the past seven years. Owing to the use of water chemistry with metering amines, the amount of products from corrosion of structural materials entering into the volume of steam generators has been reduced, and the flow-accelerated corrosion rate of pipelines and equipment has been slowed down. The article presents data on conducting water chemistry in nuclear power plant units with VVER-1000 reactors for the secondary coolant system equipment made without using copper-containing alloys. Statistical data are presented on conducting ammonia-morpholine and ammonia-ethanolamine water chemistries in new-generation operating power units with VVER-1000 reactors with an increased level of pH. The values of cooling water leaks in turbine condensers the tube system of which is made of stainless steel or titanium alloy are given.

  14. Effects of Specific Fuel Consumption and Exhaust Emissions of Four Stroke Diesel Engine with CuO/Water Nanofluid as Coolant

    Directory of Open Access Journals (Sweden)

    Senthilraja S.

    2017-03-01

    Full Text Available This article reports the effects of CuO/water based coolant on specific fuel consumption and exhaust emissions of four stroke single cylinder diesel engine. The CuO nanoparticles of 27 nm were used to prepare the nanofluid-based engine coolant. Three different volume concentrations (i.e 0.05%, 0.1%, and 0.2% of CuO/water nanofluids were prepared by using two-step method. The purpose of this study is to investigate the exhaust emissions (NOx, exhaust gas temperature and specific fuel consumption under different load conditions with CuO/water nanofluid. After a series of experiments, it was observed that the CuO/water nanofluids, even at low volume concentrations, have a significant influence on exhaust emissions. The experimental results revealed that, at full load condition, the specific fuel consumption was reduced by 8.6%, 15.1% and 21.1% for the addition of 0.05%, 0.1% and 0.2% CuO nanoparticles with water, respectively. Also, the emission tests were concluded that 881 ppm, 853 ppm and 833 ppm of NOx emissions were observed at high load with 0.05%, 0.1% and 0.2% volume concentrations of CuO/water nanofluids, respectively.

  15. Experimental studies on heat transfer to supercritical water in 2 × 2 rod bundle with two channels

    International Nuclear Information System (INIS)

    Gu, H.Y.; Hu, Z.X.; Liu, D.; Xiao, Y.; Cheng, X.

    2015-01-01

    Highlights: • Heat transfer to supercritical water in a 2 × 2 rod bundle is investigated. • Effects of system parameters on heat transfer in bundle are analyzed. • The test data were compared with twenty heat transfer correlations. - Abstract: The experiment of heat transfer to supercritical water in 2 × 2 rod bundle is performed at Shanghai Jiao Tong University. The test section consists of two channels separated by a square steel assembly box with rounded corners. Water flows downward in the first channel and then turns upward in the second channel to cool the 2 × 2 rod bundle installed inside the assembly box. The bundle consists of four heated rods of 10 mm in O.D. and 1.18 in pitch-to-diameter ratio. The fluid enthalpy in the first channel increases due to the heat transfer through the assembly box when flowing downward. The minimum fluid enthalpy increase in the first channel appears at the pseudo-critical region due to the small temperature difference between the two channels. Effects of various parameters on heat transfer behavior inside the 2 × 2 rod bundle are similar to those observed in tube or annuli. No special phenomenon in heat transfer is observed during the mass flux and power transient. The steady-state heat transfer correlation is applicable to predict the heat transfer in the mass or power transient sequence. In addition, the importance of several dimensionless numbers and the accuracy of 20 heat transfer correlations are assessed. It is concluded that the buoyancy parameter proposed by Cheng et al. (2009) shows unique effect on heat transfer coefficient. Among the 20 selected heat transfer correlations, the correlations of Jackson and Fewster (1975) and Bishop et al. (1964) give the best predictions when compared with the experimental data

  16. Assessment of a general methodology for the analysis of natural circulation stability with water at supercritical pressure

    International Nuclear Information System (INIS)

    Debrah, K. S.

    2014-07-01

    To advance nuclear energy to meet future energy needs, the concept of Super Critical Water-Cooled Reactor (SCWR) as part or Generation IV (Gen IV) reactors was introduced with plans to deploy by 2030. Supercritical water-cooled reactors pose new challenges in stability and natural circulation phenomena at supercritical pressures because of the strong variability of thermodynamic and thermo-physical properties. ln this research, included in the frame work of the International Atomic Energy Agency (lAEA) fellowship and Coordinated Research Project (CRP) on H eat transfer Behavior and Thermo hydraulics Codes Testing for SCWRs , the natural circulation H 2 O experimental data at supercritical pressures of 25 MPa obtained at the China Institute of Atomic Energy (CIAE) of China, was used to evaluate the predictions of different system codes: RELAP5/MOD3.3, STAR-CCM+ as well as three (3) different and independent developed in-house codes (Ishii-sup loop, NCLoop T ran and NCLoop L ine). Stability analyses of an idealized loop (loop equivalent to CIAE natural circulation loop) of uniform diameter equivalent to the CIAE natural circulation loop at 25 MPa was performed using RELAP5 and an in-house code (Ishii-sup Loop). It was found for both RELAP and Ishii-sup Loop that, when heat structures are accounted for in models equipped with heat transfer and friction correlations for 'normal' fluids, the comparison with experimental data is not completely satisfactory because the observed experimental oscillations were delayed in simulation. It has also been found that the stability margin was slightly earlier than the peak of the flow rate-power curve at a given inlet enthalpy. Results from STAR-CCM+ was also compared with results obtained with RELAP5 and the in-house code of NCLoop. Even though STAR-CCM+ predicted a lower flow rate than the in-house codes, all codes exhibited the ability to predict the instability and results from all codes compared favorably. Stability

  17. CFD study of convective heat transfer to carbon dioxide and water at supercritical pressures in vertical circular pipes

    Energy Technology Data Exchange (ETDEWEB)

    Zhou, F.; Novog, D.R. [McMaster Univ., Hamilton, ON (Canada)

    2014-07-01

    Computational simulations of convective heat transfer of both carbon dioxide and water at supercritical pressures have been carried out using the commercial Computational Fluid Dynamics code STAR-CCM+. Detailed comparisons between four turbulence models, including two low-Reynolds k-ε models, SST k-ω model and the Reynolds Stress Transport (RST) model, are made under different flow conditions against two independent experiments on upward flow in vertical circular pipes. The heat-flux effect and mass-flux effect on the occurrence of heat transfer deterioration (HTD) are discussed, along with sensitivity studies of the boundary conditions and turbulent Prandtl number. The thresholds and mechanisms of HTD are also investigated using selected turbulence models. (author)

  18. Effect of property variations on the mixing of turbulent supercritical water streams in a T-junction

    Energy Technology Data Exchange (ETDEWEB)

    Bu, L.; Zhao, J. [Centre for E-City, School of Electrical and Electronics Engineering, Nanyang Technological Univ., Singapore, 639798 (Singapore)

    2012-07-01

    The supercritical water mixing phenomenon is investigated with a wide range of conditions, i.e. the inlet temperature of the streams ranges from 323.15 K to 723.15 K and the pressure ranges from 25 MPa to 45 MPa. A sensitivity study is carried out for the jet and main flow velocity ratio (VR) which is varying from 1 to 40. In addition, the effect of the inject angles of branch flow to main flow on the mixing is conducted by varying the inject angle from 80 deg. to 100 deg.. The results show that the maximum temperature gradient appears on the wall of the upstream side in all the cases, and the inclined angles can be optimized to mitigate the thermal stress. (authors)

  19. About calculation results of heat transfer in the fuel assembly clusters cooled by water with supercritical parameters

    International Nuclear Information System (INIS)

    Grabezhnaya, V.A.

    2008-01-01

    Paper reviews the numerical investigation into the heat transfer in the supercritical water cooled fuel assemblies on the basis of the various commercial codes. The turbulence available models specified in the codes describe adequately the experimental data in tubes within the range of flow temperatures away from the pseudocritical point, as well as under high mass velocities. There are k-ε type turbulence models that show qualitatively the local acceleration (slowdown) of the heat transfer in tubes, but they fail to describe the mentioned phenomena quantitatively. To determine the effect of grid spacers on the suppression of the heat transfer local slowdown and on the heat transfer acceleration in fuel assemblies and to ensure more accurate calculation of the fuel element cladding maximum temperature one should perform a number of the experiments making use of the fuel assembly models [ru

  20. Supercritical heat transfer in an annular channel with two-sided heaing

    International Nuclear Information System (INIS)

    Sergeev, V.V.; Remizov, O.V.; Gal'chenko, Eh.F.

    1986-01-01

    The paper deals with experimental inestigation into worsening of heat transfer at forced up flow in steam-water mixture in a vertical annular channel with two-sided heating and development of technique for calculation of supercritical heat exchange in this channel. Bench-scale experiments are carried out at high-pressure at mass rates of the coolant equal to 300-865 kg/(m 2 x s), pressure of 9.8-17.8 MPa and heat flux on the internal surface - 20-400 kW/m 2 , on the external surface - 35-450 kW/m 2 . Technique for calculation of supercritical heat exchange in channels with one- and two-sided heating is suggested. Analysis of the obtained experimental data permits to determine conditions for arising departure nucleate boiling on the internal and external surfaces and on both surfaces simultaneously. It is concluded that the suggested technique of calculation adequately reflects the effect of regime parameters of coolant flow on temperature regime of heat transferring surfaces in the supercritical area

  1. Algorithms and programs for solution of static and dynamic characteristics of counterflow heat exchangers with dissociating coolant

    International Nuclear Information System (INIS)

    Nitej, N.V.; Sharovarov, G.A.

    1982-01-01

    The method of estimation of counterflow heat exchanger characteristics is presented. Mathematical description of the processes is presented by the mass, energy and pulse conservation equations for both coolants and energy conservation equation for the wall which devides them. In the presence of chemical reactions the system is supplemented by equations, characterizing the kinetics of their progress. The methods of numerical solution of static and dynamic problems have been chosen, and the computer programs on the Fortran language have been developed. The schemes of solution of both problems are so constructed, that the conservation equations are placed in the main program, and such characteristics of the coolants as properties, heat transfer and friction coefficients, the mechanism of chemical reaction are concentrated in the subprogram unit. This allows to create the single method of solution with the flow of single-phase and two-phase coolants of abovecritical and supercritical paramters. The evaluation results of three heat exchangers are given: with heating of N 2 O 4 gas phase by heat of flue gas; with cooling of N 2 O 4 supercritical parameters by water; regenerator on N 2 O 4

  2. Synergetic effect of copper-plating wastewater as a catalyst for the destruction of acrylonitrile wastewater in supercritical water oxidation

    Energy Technology Data Exchange (ETDEWEB)

    Shin, Young Ho; Lee, Hong-shik; Lee, Young-Ho [School of Chemical and Biological Engineering and Institute of Chemical Processes, Seoul National University, 599 Gwanangno, Gwanak-gu, Seoul 151-744 (Korea, Republic of); Kim, Jaehoon; Kim, Jae-Duck [Supercritical Fluid Research Laboratory, Energy and Environment Research Division, Korea Institute of Science and Technology (KIST), 39-1 Hawolgok-dong, Seongbuk-gu, Seoul 136-791 (Korea, Republic of); Lee, Youn-Woo, E-mail: ywlee@snu.ac.kr [School of Chemical and Biological Engineering and Institute of Chemical Processes, Seoul National University, 599 Gwanangno, Gwanak-gu, Seoul 151-744 (Korea, Republic of)

    2009-08-15

    A new supercritical water oxidation process for the simultaneous treatment of mixed wastewater containing wastewater from acrylonitrile manufacturing processes and copper-plating processes was investigated using a continuous tubular reactor system. Experiments were carried out at temperatures ranging from 400 to 600 deg. C and a pressure of 25 MPa. The residence time was fixed at 2 s by changing the flow rates of feeds, depending on reaction temperature. The initial total organic carbon (TOC) concentration of the wastewaters and the O{sub 2} concentration at the reactor inlet were kept constant at 0.49 and 0.74 mol/L. It was confirmed that the copper-plating wastewater accelerated the TOC conversion of acrylonitrile wastewater from 17.6% to 67.3% at a temperature of 450 deg. C. Moreover, copper and copper oxide nanoparticles were generated in the process of supercritical water oxidation (SCWO) of mixed wastewater. 99.8% of copper in mixed wastewater was recovered as solid copper and copper oxides at a temperature of 600 deg. C, with their average sizes ranging from 150 to 160 nm. Our study showed that SCWO provides a synergetic effect for simultaneous treatment of acrylonitrile and copper-plating wastewater. During the reaction, the oxidation rate of acrylonitrile wastewater was enhanced due to the in situ formation of nano-catalysts of copper and/or copper oxides, while the exothermic decomposition of acrylonitrile wastewater supplied enough heat for the recovery of solid copper and copper oxides from copper-plating wastewater. The synergetic effect of wastewater treatment by the newly proposed SCWO process leads to full TOC conversion, color removal, detoxification, and odor elimination, as well as full recovery of copper.

  3. Literature survey of heat transfer and hydraulic resistance of water, carbon dioxide, helium and other fluids at supercritical and near-critical pressures

    Energy Technology Data Exchange (ETDEWEB)

    Pioro, I.L.; Duffey, R.B

    2003-04-01

    This survey consists of 430 references, including 269 Russian publications and 161 Western publications devoted to the problems of heat transfer and hydraulic resistance of a fluid at near-critical and supercritical pressures. The objective of the literature survey is to compile and summarize findings in the area of heat transfer and hydraulic resistance at supercritical pressures for various fluids for the last fifty years published in the open Russian and Western literature. The analysis of the publications showed that the majority of the papers were devoted to the heat transfer of fluids at near-critical and supercritical pressures flowing inside a circular tube. Three major working fluids are involved: water, carbon dioxide, and helium. The main objective of these studies was the development and design of supercritical steam generators for power stations (utilizing water as a working fluid) in the 1950s, 1960s, and 1970s. Carbon dioxide was usually used as the modeling fluid due to lower values of the critical parameters. Helium, and sometimes carbon dioxide, were considered as possible working fluids in some special designs of nuclear reactors. (author)

  4. Cleaning of aluminum after machining with coolants

    International Nuclear Information System (INIS)

    Roop, B.

    1992-01-01

    An x-ray photoemission spectroscopic study was undertaken to compare the cleaning of the Advanced Photon Source (APS) aluminum extrusion storage ring vacuum chambers after machining with and without water soluble coolants. While there was significant contamination left by the coolants, the cleaning process was capable of removing the residue. The variation of the surface and near surface composition of samples machined either dry or with coolants was negligible after cleaning. The use of such coolants in the machining process is therefore recommended

  5. Long-term recovery of pressurized water reactors following a large break loss-of-coolant accident

    International Nuclear Information System (INIS)

    Fletcher, C.D.; Callow, R.A.

    1989-01-01

    The USNRC recently identified a possible safety concern for PWR's. Following the reflood phase of a large break loss-of-coolant accident, long-term cooling of the reactor core may not be ensured. Specifically, the concern is that, for a pump discharge cold leg break, the loop seals in the reactor coolant pump suction piping will refill with liquid and the post-reflood steam production may depress the liquid levels in the downflow sides of the loop seals. A loop seal depression would cause a corresponding depression of the core liquid levels and possibly a fuel rod heatup in the upper core region. This paper is intended as an introduction of the safety issue that: 1) describes the important aspects of the problem, 2) provides an initial analysis of the consequences, and 3) discusses ongoing work in this area. Because the elevation of the loop seals is near the mid-core elevation in plants of WE design, the concern is greatest for those plants. There is less concern for most plants of CE design, and likely no concern for plants of BW design. This issue was addressed by employing both steady-state and transient systems analysis approaches. Two approaches were used because of uncertainties regarding actual reactor coolant system behavior during the post-reflood period. The steady-state approach involved the development and application of a simple computer program to investigate reactor coolant system behavior assuming quiescent post-reflood conditions. The transient systems approach involved investigating this behavior using the RELAP5/MOD2 computer code and a comprehensive RELAP5 model of a WE PWR. The steady-state analysis indicated only a moderate fuel rod heatup is possible. The transient systems analysis indicated boiling and condensation-induced flow oscillations are sufficient to prevent fuel rod heatup. Analysis uncertainties are discussed. (orig./HP)

  6. Density dependence of the radiolysis yields of primary species from fast neutron-irradiated supercritical water at 400 {sup o}C

    Energy Technology Data Exchange (ETDEWEB)

    Butarbutar, S.L. [Universite de Sherbrooke, Sherbrooke, QC (Canada); National Nuclear Energy Agency, BATAN, Tangerang Selatan, Banten (Indonesia); Meesungnoen, J. [Universite de Sherbrooke, Sherbrooke, QC (Canada); Guzonas, D.A.; Stuart, C.R. [Atomic Energy of Canada Limited, Chalk River, ON (Canada); Jay-Gerin, J-P [Universite de Sherbrooke, Sherbrooke, QC (Canada)

    2014-07-01

    A reliable understanding of radiolysis processes in supercritical water (SCW)-cooled reactors is crucial to developing chemistry control strategies that minimize corrosion and the transport of both corrosion products and radionuclides. Chemistry control is one of the most important factors to keep the integrity of materials from degradation processes and also to reduce out-of-core radiation fields and worker dose. However, directly measuring the chemistry in reactor cores is difficult due to the extreme conditions of high temperature and pressure and mixed neutron and γ-radiation fields that are not compatible with normal chemical instrumentation. Thus, chemical models and computer simulations are an important route of investigation for predicting the detailed radiation chemistry of the coolant in a SCW reactor and the consequences for materials. Surprisingly, there is only limited information on the fast neutron radiolysis of water at high temperatures, and no experimental data are yet available on the radiolysis yields for fast neutron irradiation of SCW. In this work, Monte Carlo simulations were used to predict the G-values for the primary species e{sup -}{sub aq}, H{sup •}, H{sub 2}, {sup •}OH, and H{sub 2}O{sub 2} formed from the radiolysis of pure, deaerated SCW (H{sub 2}O) by 2-MeV mono-energetic neutrons at 400 {sup o}C as a function of water density in the range of ~0.15-0.6 g/cm{sup 3}. The 2-MeV neutron was taken as representative of a fast neutron flux in a reactor. For light water, the moderation of these neutrons after knock-on collisions with water molecules generated mostly recoil protons of 1.264, 0.465, 0.171, and 0.063 MeV having linear energy transfer (LET) values of ~3.3, 6.5, 10.4, and 11.4 keV/μm at 0.15 g/cm{sup 3}, and ~13.3, 26, 42, and 46 keV/μm at 0.6 g/cm{sup 3}, respectively. Neglecting oxygen ion recoils and assuming that the most significant contribution to the radiolysis came from these first four recoil protons, the fast

  7. Metal corrosion in a supercritical carbon dioxide - liquid sodium power cycle.

    Energy Technology Data Exchange (ETDEWEB)

    Moore, Robert Charles; Conboy, Thomas M.

    2012-02-01

    A liquid sodium cooled fast reactor coupled to a supercritical carbon dioxide Brayton power cycle is a promising combination for the next generation nuclear power production process. For optimum efficiency, a microchannel heat exchanger, constructed by diffusion bonding, can be used for heat transfer from the liquid sodium reactor coolant to the supercritical carbon dioxide. In this work, we have reviewed the literature on corrosion of metals in liquid sodium and carbon dioxide. The main conclusions are (1) pure, dry CO{sub 2} is virtually inert but can be highly corrosive in the presence of even ppm concentrations of water, (2) carburization and decarburization are very significant mechanism for corrosion in liquid sodium especially at high temperature and the mechanism is not well understood, and (3) very little information could be located on corrosion of diffusion bonded metals. Significantly more research is needed in all of these areas.

  8. Development of Nordic Standard for analysis of oil and fat in water based on supercritical fluid extraction. Preliminary study, part 2

    International Nuclear Information System (INIS)

    Jenssen, L.

    1994-06-01

    This report describes a preliminary study of a method of determining oil in water. The method is based on solid phase extraction and supercritical fluid extraction (SPE-SFE). The oil is extracted from the water by absorption to extraction disks from which it is then desorbed by supercritical carbon dioxide and detected by means of infrared spectrophotometry or gas chromatography. The results of the study will indicate if the method is suitable as a future substitute for the present Norwegian Standard, NS 9803 (Swedish Standard, SS 02 8145). The method has been validated using water samples with addition of real oil to 1-100 ppm. The accuracy is almost 70%, and the method has good repeatability and is linear in the 1-100 ppm range. 5 refs., 6 figs., 10 tabs

  9. Analysis of a hot-leg small break loss-of-coolant accident in a three-loop westinghouse pressurized water reactor plant

    International Nuclear Information System (INIS)

    Peterson, C.E.; Chexal, V.K.; Clements, T.B.

    1985-01-01

    The RETRAN-02 computer code was used to perform a best-estimate analysis of a 7.52-cm-diam hotleg break in a three-loop Westinghouse pressurized water reactor. This break size produced a net primary coolant mass depletion through the early portion of the transient. The primary system started to refill only after the accumulator valves opened. As the primary system refilled, there were extreme temperature differentials around the system with cold, denser fluid collecting at the lower elevations and two-phase fluid at higher elevations

  10. Operation and Performance of the Supercritical Fluids Reactor (SFR)

    National Research Council Canada - National Science Library

    Hanush, R

    1996-01-01

    The Supercritical Fluids Reactor (SFR) at Sandia National Laboratories, CA has been developed to examine and solve engineering, process, and fundamental chemistry issues regarding the development of supercritical water oxidation (SCWO...

  11. Continuous production of phosphor YAG:Tb nanoparticles by hydrothermal synthesis in supercritical water

    International Nuclear Information System (INIS)

    Hakuta, Yukiya; Haganuma, Tsukasa; Sue, Kiwamu; Adschiri, Tadafumi; Arai, Kunio

    2003-01-01

    Phosphor YAG:Tb ((Y 2.7 Tb 0.3 )Al 5 O 12 ) nano particles were synthesized by a hydrothermal method at supercritical conditions (400 deg. C and 30 MPa) using a flow reactor. Hydroxide sol solutions formed by stoichiometric aluminum nitrate, yttrium nitrate, terbium nitrate and potassium hydroxide solutions. The relationship between particle size and experimental variables including pH, concentration of coexistent ions and hydroxide sol were investigated. Particles were characterized by XRD, TEM and photo-luminescence measurements. Particle size of YAG:Tb became finer as pH was increased or potassium nitrate concentration of the starting metal salt solution was increased. By removing the coexisting ions (NO 3 - , K + ) from the metal salt solution, single phase YAG:Tb particles with 20 nm particle size were obtained. The emission spectra of YAG:Tb particles of 14 nm shows a blue shift

  12. CFD analysis of the dynamic behaviour of a fuel rod subchannel in a supercritical water reactor with point kinetics

    International Nuclear Information System (INIS)

    Ampomah-Amoako, Emmanuel; Akaho, Edward H.K.; Nyarko, Benjamin J.B.; Ambrosini, Walter

    2013-01-01

    Highlights: • The analysis of flow stability of nuclear fuel subchannels with supercritical water is presented. • The results obtained by a CFD code are compared with those of a system code. • The model includes also heat conduction in the fuel rod and point neutron kinetics. - Abstract: The paper presents the analysis by a CFD code of coupled neutronic–thermal hydraulic instabilities in a subchannel slice belonging to a square lattice assembly. The work represents a further phase in the assessment of the suitability of CFD codes for studies of flow stability of supercritical fluids; the research started in previous work with the analysis of bare 2D circular pipes and already addressed 3D subchannel slices with no allowance for heat conduction or neutronic effects. In the present phase, a more realistic system is considered, dealing with a slice of a fuel assembly subchannel containing the regions of the pellet, the gap and the cladding and including also the effect of inlet and outlet throttling. The adopted neutronic model is a point kinetics one, including six delayed neutron groups with global Doppler and fluid density feedbacks. The response of the model to perturbations applied starting from a steady-state condition at the rated power is compared with that of a similar model developed for a 1D system code. Upward, horizontal and downward flow orientations are addressed making use of a uniform power profile and changing relevant parameters as the gap equivalent conductance and the density reactivity coefficient. A bottom-peaked power profile is also considered in the case of vertical upward flow. Though the adopted model can still be considered simple in comparison with actual details of fuel assemblies, the obtained results demonstrate the potential of the adopted methodology for more accurate analyses to be made with larger computational resources

  13. A Simplified Supercritical Fast Reactor with Thorium Fuel