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)

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)

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)

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

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

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.

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)

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)

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)

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

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

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)

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

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

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)

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.

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.

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)

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.

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)

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)

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

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

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

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

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

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.

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

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.

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.

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.

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

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.)

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

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.

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

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

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

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

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

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

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

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.

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.

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)

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

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

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.

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)

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

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)

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.

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

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...

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

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.

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

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.)

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

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

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

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 ...

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)

A Simplified Supercritical Fast Reactor with Thorium Fuel

Directory of Open Access Journals (Sweden)

Peng Zhang

2014-01-01

Full Text Available 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 and keeping negative coolant void reactivity during the whole core life. A core burnup simulation scheme based on Monte Carlo lattice homogenization is adopted in this study, and the reactor physics analysis has been performed with DU-MOX and Th-MOX fuel. The main issues discussed include the fuel conversion ratio and the coolant void reactivity. The analysis shows that thorium-based fuel can provide inherent safety for SCFR without use of blanket, which is favorable for the mechanical design of SCFR.

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

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

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

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.

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.

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

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

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

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.

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

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

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

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

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)

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.)

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.

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

Specifics of forced-convective heat transfer in supercritical carbon dioxide

Energy Technology Data Exchange (ETDEWEB)

Saltanov, A.E.; Mann, B.D.; Harvel, C.G.; Pioro, D.I., E-mail: Eugene.saltanov@hotmail.com [University of Ontario Institute of Technology, Oshawa, ON (Canada)

2015-07-01

The appropriate description of heat-transfer to coolants at supercritical state is one of the main challenges in development of supercritical-fluids applications for the Generation-IV reactors. In this paper the basis for comparison of relatively recent experimental data on supercritical carbon dioxide (CO{sub 2}) obtained at facilities of the Korea Atomic Energy Research Institute (KAERI) and Chalk River Laboratories (CRL) of Atomic Energy of Canada Limited (AECL) is discussed, and a preliminary heat-transfer correlation for joint CRL and KAERI datasets is presented. (author)

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)

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)

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

Stability analysis of fluid at supercritical pressure in a heated channel

International Nuclear Information System (INIS)

Gallaway, T.; Podowski, M. Z.

2010-01-01

The Supercritical Water Reactor (SCWR) is one of several reactor design concepts included in the Generation IV International Advanced Reactor Design Program. This reactor design is based upon current light water reactors and supercritical fossil-fuel power plants. Water at supercritical pressures is used as the reactor coolant. At these conditions, there is no phase change in the coolant; however the fluid properties undergo significant variation, particularly in the pseudo-critical region. The fluid density may decrease by a factor of six with increasing temperature. It has been seen before that variations in fluid density can lead to density-wave oscillations in two-phase flow systems in general and boiling water reactors in particular. Such instabilities may cause many undesired problems for reactor operation and safety. Similar issues must be addressed in the design and safety analysis of SCWRs. The objective of the present work has been the development of a detailed one-dimensional model of instabilities in a heated channel corresponding to the geometry and flow conditions in the proposed typical SCWRs. The new model is capable of analyzing in detail transient effects of local property variations in parallel channels subject to a constant pressure drop boundary condition. In particular, such a model can be used to establish SCWR power limits imposed by the onset of instabilities in the hot channel of the reactor. Both time and frequency-domain methods of stability analysis have been developed. The latter method is particularly important since it is not associated with any numerical issues, is very accurate, and allows for establishing general stability boundaries in a computationally effective manner. Model testing has included a study of dependence of the proposed spatial discretization scheme on the accuracy of calculations. A parametric study has also been performed on the effect of channel operating conditions on flow oscillations. Finally, a stability map

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...

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

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)

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)

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

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

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.

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.

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

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.)

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

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.

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

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)

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

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

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)

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)

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.

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

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.)

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)

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)

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)

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)

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)

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.

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)

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)

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.

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

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

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

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...

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)

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

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.

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)

Thermal circuit and supercritical steam generator of the BGR-300 nuclear power plant

International Nuclear Information System (INIS)

Afanas'ev, B.P.; Godik, I.B.; Komarov, N.F.; Kurochnkin, Yu.P.

1979-01-01

Secondary coolant circuit and a steam generator for supercritical steam parameters of the BGR-300 reactor plant are described. The BGR-300 plant with a 300 MW(e) high-temperature gas-cooled fast reactor is developed as a pilot commercial plant. It is shown that the use of a supercritical pressure steam increases the thermal efficiency of the plant and descreases thermal releases to the environment, permits to use home-made commercial turbine plants of large unit power. The proposed supercritical pressure steam generator has considerable advantages from the viewpoint of heat transfer and hydrodynamical processes

Impact of neutron thermal scattering laws on the burn-up analysis of supercritical LWR's fuel assemblies

International Nuclear Information System (INIS)

Conti, Andrea

2011-10-01

This work is a contribution to the HPLWR2 (High Performance Light Water Reactor Phase 2), a research project having the goal to investigate the technical feasibility of the High Performance Light Water Reactor. The basic idea of the HPLWR is that of an LWR working at supercritical pressure, which would allow heating up the coolant to a temperature of about 500 C without having phase transition and sending the coolant directly to the turbine. One issue aroused by this design, deserving to be addressed by research, is the behaviour of thermal neutrons in supercritical water. At thermal energies, the De Broglie wavelength associated with the neutron is comparable to the interatomic distances in crystals and molecules and the scattering is fully governed by the laws of quantum mechanics, according to which the geometry of the aggregates the nuclei are bound to and their intra- and intermolecular dynamics are of crucial importance. It can be shown that there is a certain mathematical relation between the Fourier-transform of the hydrogen atoms' velocity autocorrelation function and their double-differential scattering cross section. This Fourier-transform, called ''generalized frequency distribution'', can be derived from experimental measurements and, effectively, Bernnat et al. of the Institut fuer Kernenergetik und Energiesysteme of the University of Stuttgart derived the generalized frequency distribution for liquid water on the basis of experimental results of Page and Haywood. Unfortunately there exists no experimental facility nowadays to support a thorough work of this type on supercritical water and therefore the scattering kernel for thermal neutrons in supercritical water is unknown. In criticality calculations involving supercritical water one can turn to one of the thermal scattering kernels available nowadays for hydrogen bound to the H 2 O molecule: for liquid water, for vapour or considering the nuclei of hydrogen as unbound. The third, most naive option

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

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.)

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

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.

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

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)

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

CFD study on the supercritical carbon dioxide cooled pebble bed reactor

Energy Technology Data Exchange (ETDEWEB)

Yu, Dali, E-mail: ydlmitd@outlook.com; Peng, Minjun; Wang, Zhongyi

2015-01-15

Highlights: • An innovation concept of supercritical carbon dioxide cooled pebble bed reactor is proposed. • Body-centered cuboid (BCCa) arrangement is adopted for the pebbles. • S-CO{sub 2} would be a good candidate coolant for using in pebble bed reactor. - Abstract: The thermal hydraulic study of using supercritical carbon dioxide (S-CO{sub 2}), a superior fluid state brayton cycle medium, in pebble bed type nuclear reactor is assessed through computational fluid dynamics (CFD) methodology. Preliminary concept design of this S-CO{sub 2} cooled pebble bed reactor (PBR) is implemented by the well-known KTA heat transfer correlation and Ergun pressure drop equation. Eddy viscosity transport turbulence model is adopted and verified by KTA calculated results. Distributions of the temperature, velocity, pressure and Nusselt (Nu) number of the coolant near the surface of the middle spherical fuel element are obtained and analyzed. The conclusion of the assessment is that S-CO{sub 2} would be a good candidate coolant for using in pebble bed reactor due primarily to its good heat transfer characteristic and large mass density, which could lead to achieve lower pressure drop and higher power density.

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

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.

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

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

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)

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.)

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.

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.

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

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)

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

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.

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

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.

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)

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.

Subchannel analysis of Al{sub 2}O{sub 3} nanofluid as a coolant in VMHWR

Energy Technology Data Exchange (ETDEWEB)

Zarifi, Ehsan; Tashakor, Saman [Nuclear Science and Technology Research Institute (NSTRI), Tehran (Iran, Islamic Republic of). Reactor Research School

2015-11-15

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{sub 2}O{sub 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.

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)

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.

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.

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.

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

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).

Theoretical and Numerical Study of Heat Transfer Deterioration in High Performance Light Water Reactor

Directory of Open Access Journals (Sweden)

David Palko

2008-01-01

Full Text Available A numerical investigation of the heat transfer deterioration (HTD phenomena is performed using the low-Re k-ω turbulence model. Steady-state Reynolds-averaged Navier-Stokes equations are solved together with equations for the transport of enthalpy and turbulence. Equations are solved for the supercritical water flow at different pressures, using water properties from the standard IAPWS (International Association for the Properties of Water and Steam tables. All cases are extensively validated against experimental data. The influence of buoyancy on the HTD is demonstrated for different mass flow rates in the heated pipes. Numerical results prove that the RANS low-Re turbulence modeling approach is fully capable of simulating the heat transfer in pipes with the water flow at supercritical pressures. A study of buoyancy influence shows that for the low-mass flow rates of coolant, the influence of buoyancy forces on the heat transfer in heated pipes is significant. For the high flow rates, buoyancy influence could be neglected and there are clearly other mechanisms causing the decrease in heat transfer at high coolant flow rates.

Evaluation and optimization of a supercritical carbon dioxide power conversion cycle for nuclear applications

International Nuclear Information System (INIS)

Harvego, Edwin A.; McKellar, Michael G.

2011-01-01

There have been a number of studies involving the use of gases operating in the supercritical mode for power production and process heat applications. Supercritical carbon dioxide (CO 2 ) is particularly attractive because it is capable of achieving relatively high power conversion cycle efficiencies in the temperature range between 550degC and 750degC. Therefore, it has the potential for use with any type of high-temperature nuclear reactor concept, assuming reactor core outlet temperatures of at least 550degC. The particular power cycle investigated in this paper is a supercritical CO 2 recompression Brayton Cycle. The CO 2 recompression Brayton Cycle can be used as either a direct or indirect power conversion cycle, depending on the reactor type and reactor outlet temperature. The advantage of this cycle when compared to the helium Brayton Cycle is the lower required operating temperature; 550degC versus 750degC. However, the supercritical CO 2 recompression Brayton Cycle requires a high end operating pressure in the range of 20 MPa, which is considerably higher than the required helium Brayton cycle high end operating pressure of 7 MPa. This paper presents results of analyses performed using the UniSim process analyses software to evaluate the performance of the supercritical CO 2 recompression Brayton cycle for different reactor coolant outlet temperatures and mass flow rates. The UniSim model assumed a 600 MWt reactor power source, which provides heat to the power cycle at a maximum temperature of between 550degC and 850degC. Sensitivity calculations were also performed to determine the affect of reactor coolant mass flow rates for a reference reactor coolant outlet temperature of 750degC. The UniSim model used realistic component parameters and operating conditions to model the complete power conversion system. CO 2 properties were evaluated, and the operating range for the cycle was adjusted to take advantage of the rapidly changing conditions near the

Coolant make-up device for BWR type reactor

International Nuclear Information System (INIS)

Sasagawa, Hiroshi.

1994-01-01

In a coolant make-up device, an opening of a pressure equalizing pipeline in a pressure vessel is disposed in coolants above a reactor core and below a usual fluctuation range of a reactor vessel water level. Further, a float check valve is disposed to the pressure equalizing pipeline for preventing coolants in the pressure vessel flowing into the pipeline. If the water level in the pressure vessel is lowered than the setting position for the float check valve, the float drops by its own weight to open the opening of the pressure equalizing pipeline. Then, steams in the pressure vessel are flown into the pipeline, to equalize the pressure between a coolant storage tank and the pressure vessel of the reactor. Coolants in the coolant storage tank is injected to the pressure vessel by way of the water injection pipeline due to the difference of the pressure head between the water level in the coolants storage tank and the water level in the pressure vessel. If the coolants are lowered than the setting position for the float check value, the float check valve does not close unless the water level is recovered to the setting position for the float valve and, accordingly, the coolant make-up is continued. (N.H.)

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

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.)

Optimized, Competitive Supercritical-CO2 Cycle GFR for Gen IV Service

International Nuclear Information System (INIS)

M.J. Driscoll; P. Hejzlar; G. Apostolakis

2008-01-01

An overall plant design was developed for a gas-cooled fast reactor employing a direct supercritical Brayton power conversion system. The most important findings were that (1) the concept could be capital-cost competitive, but startup fuel cycle costs are penalized by the low core power density, specified in large part to satisfy the goal of significant post-accident passive natural convection cooling; (2) active decay heat removal is preferable as the first line of defense, with passive performance in a backup role; (3) an innovative tube-in-duct fuel assembly, vented to the primary coolant, appears to be practicable; and (4) use of the S-Co2 GFR to support hydrogen production is a synergistic application, since sufficient energy can be recuperated from the product H2 and 02 to allow the electrolysis cell to run 250 C hotter than the reactor coolant, and the water boilers can be used for reactor decay heat removal. Increasing core power density is identified as the top priority for future work on GFRs of this type

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.

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.

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)

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)

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

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.

Pre-conceptual core design of a small modular fast reactor cooled by supercritical CO{sub 2}

Energy Technology Data Exchange (ETDEWEB)

Liu, Baolin; Cao, Liangzhi; Wu, Hongchun [School of Nuclear Science and Technology, Xi’an Jiaotong University, No 28, Xianning West Road, Xi’an 710049, Shaanxi (China); Yuan, Xianbao, E-mail: ztsbaby@163.com [School of Nuclear Science and Technology, Xi’an Jiaotong University, No 28, Xianning West Road, Xi’an 710049, Shaanxi (China); College of Mechanical & Power Engineering, China Three Gorges University, No 8, Daxue Road, Yichang 443002, Hubei (China); Wang, Kunpeng [Nuclear and Radiation Safety Center, PO Box 8088, Beijing 100082 (China)

2016-04-15

Abstracts: A Small Modular fast reactor cooled by Supercritical CO{sub 2} (SMoSC) is pre-conceptually designed through three-dimensional coupled neutronics/thermal-hydraulics analysis. The power rating of the SMoSC is designed to be 300 MW{sub th} to meet the energy demand of small electrical grids. The excellent thermal properties of supercritical CO{sub 2} (S-CO{sub 2}) are employed to obtain a high thermal efficiency of about 40% with an electric output of 120 MWe. MOX fuel is utilized in the core design to improve fuel efficiency. The tube-in-duct (TID) assembly is applied to get lower coolant volume fraction and reduce the positive coolant void reactivity. According to the coupled neutronics/thermal-hydraulics calculations, the coolant void reactivity is kept negative throughout the whole core life. With a specific power density of 9.6 kW/kg and an average discharge burnup of 70.1 GWd/tHM, the SmoSC can be operated for 20 Effective Full Power Years (EFPYs) without refueling.

Coolant processing device for nuclear reactor

International Nuclear Information System (INIS)

Kizawa, Hideo; Funakoshi, Toshio; Izumoji, Yoshiaki

1981-01-01

Purpose: To reduce an entire facility cost by concentrating and isolating tritium accumulated in coolants, removing the tritium out of the system, and returning hydrogen gas generated at a reactor accident to a recombiner in a closed loop by the switching of a valve. Constitution: Coolant from a reactor cooling system processed by a chemical volume control system facility (CVCS) and coolant drain from various devices processed by a liquid waste disposing system facility (LWDS) are fed to a tritium isolating facility, in which they are isolated into concentrated tritium water and dilute tritium water. The concentrated tritium water is removed out of the system and stored. The dilute tritium water is reused as supply water for coolant. If an accident occurs to cause hydrogen to be generated, a closed loop is formed between the containment vessel and the recombiner, the hydrogen is recombined with oxygen in the air of the closed loop to be thus returned to water. (Kamimura, M.)

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.)

Coupled neutronics/thermal-hydraulics analysis of a high-performance light-water reactor fuel assembly

International Nuclear Information System (INIS)

Waata, C.L.

2006-07-01

The use of water at supercritical pressure as coolant and moderator introduces a challenge in the design of a High-Performance Light-Water Reactor (HPLWR) fuel assembly. At supercritical pressure condition (P=25 MPa), the thermal-hydraulics behaviour of water differs strongly from that at sub-critical pressure due to a rapid variation of the thermal-physical properties across the pseudo-critical line. Due of the strong link between the water (moderation) and the neutron spectrum and subsequently the power distribution, a coupling of neutronics and thermal-hydraulics has become a necessity for reactor concepts operating at supercritical pressure condition. The effect of neutron moderation on the local parameters of thermal-hydraulics and vice-verse in a fuel assembly has to be considered for an accurate design analysis. In this study, the Monte Carlo N-Particle code (MCNP) and the sub-channel code STAFAS (Sub-channel Thermal-hydraulics Analysis of a Fuel Assembly under Supercritical conditions) have been coupled for the design analysis of a fuel assembly with supercritical water as coolant and moderator. Both codes are well known for complex geometry modelling. The MCNP code is used for neutronics analyses and for the prediction of power profiles of individual fuel rods. The sub-channel code STAFAS for the thermal-hydraulics analyses takes into account the coolant properties beyond the critical point as well as separate moderator channels. The coupling procedure is realized automatically. MCNP calculates the power distribution in each fuel rod, which is then transferred into STAFAS to obtain the corresponding thermal-hydraulic conditions in each sub-channel. The new thermal-hydraulic conditions are used to generate a new input deck for the next MCNP calculation. This procedure is repeated until a converged state is achieved. The coupled code system was tested on a proposed fuel assembly design of a HPLWR. An under-relaxation was introduced to achieve convergence

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.

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

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

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.

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)

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

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)

CFD analyses of coolant channel flowfields

Science.gov (United States)

Yagley, Jennifer A.; Feng, Jinzhang; Merkle, Charles L.

1993-01-01

The flowfield characteristics in rocket engine coolant channels are analyzed by means of a numerical model. The channels are characterized by large length to diameter ratios, high Reynolds numbers, and asymmetrical heating. At representative flow conditions, the channel length is approximately twice the hydraulic entrance length so that fully developed conditions would be reached for a constant property fluid. For the supercritical hydrogen that is used as the coolant, the strong property variations create significant secondary flows in the cross-plane which have a major influence on the flow and the resulting heat transfer. Comparison of constant and variable property solutions show substantial differences. In addition, the property variations prevent fully developed flow. The density variation accelerates the fluid in the channels increasing the pressure drop without an accompanying increase in heat flux. Analyses of the inlet configuration suggest that side entry from a manifold can affect the development of the velocity profile because of vortices generated as the flow enters the channel. Current work is focused on studying the effects of channel bifurcation on the flow field and the heat transfer characteristics.

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)

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.)

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)

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

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)

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)

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

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

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.

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.

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

Supercritical Fluids Processing of Biomass to Chemicals and Fuels

Energy Technology Data Exchange (ETDEWEB)

Olson, Norman K. [Iowa State Univ., Ames, IA (United States)

2011-09-28

The main objective of this project is to develop and/or enhance cost-effective methodologies for converting biomass into a wide variety of chemicals, fuels, and products using supercritical fluids. Supercritical fluids will be used both to perform reactions of biomass to chemicals and products as well as to perform extractions/separations of bio-based chemicals from non-homogeneous mixtures. This work supports the Biomass Program’s Thermochemical Platform Goals. Supercritical fluids are a thermochemical approach to processing biomass that, while aligned with the Biomass Program’s interests in gasification and pyrolysis, offer the potential for more precise and controllable reactions. Indeed, the literature with respect to the use of water as a supercritical fluid frequently refers to “supercritical water gasification” or “supercritical water pyrolysis.”

The effect of coolant quantity on local fuel–coolant interactions in a molten pool

International Nuclear Information System (INIS)

Cheng, Songbai; Matsuba, Ken-ichi; Isozaki, Mikio; Kamiyama, Kenji; Suzuki, Tohru; Tobita, Yoshiharu

2015-01-01

Highlights: • We investigate local fuel–coolant interactions in a molten pool. • As water volume increases, limited pressurization and mechanical energy observed. • Only a part of water is evaporated and responsible for the pressurization. - Abstract: Studies on local fuel–coolant interactions (FCI) in a molten pool are important for severe accident analyses of sodium-cooled fast reactors (SFRs). Motivated by providing some evidence for understanding this interaction, in this study several experimental tests, with comparatively larger difference in coolant volumes, were conducted by delivering a given quantity of water into a simulated molten fuel pool (formed with a low-melting-point alloy). Interaction characteristics including the pressure-buildup as well as mechanical energy release and its conversion efficiency are evaluated and compared. It is found that as water quantity increases, a limited pressure-buildup and the resultant mechanical energy release are observable. The performed analyses also suggest that only a part of water is probably vaporized during local FCIs and responsible for the pressurization and mechanical energy release, especially for those cases with much larger water volumes

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.)

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...

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)

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)

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

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

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

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)

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

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)

Linear predictions of supercritical flow instability in two parallel channels

International Nuclear Information System (INIS)

Shah, M.

2008-01-01

A steady state linear code that can predict thermo-hydraulic instability boundaries in a two parallel channel system under supercritical conditions has been developed. Linear and non-linear solutions of the instability boundary in a two parallel channel system are also compared. The effect of gravity on the instability boundary in a two parallel channel system, by changing the orientation of the system flow from horizontal flow to vertical up-flow and vertical down-flow has been analyzed. Vertical up-flow is found to be more unstable than horizontal flow and vertical down flow is found to be the most unstable configuration. The type of instability present in each flow-orientation of a parallel channel system has been checked and the density wave oscillation type is observed in horizontal flow and vertical up-flow, while the static type of instability is observed in a vertical down-flow for the cases studied here. The parameters affecting the instability boundary, such as the heating power, inlet temperature, inlet and outlet K-factors are varied to assess their effects. This study is important for the design of future Generation IV nuclear reactors in which supercritical light water is proposed as the primary coolant. (author)

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

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

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

Mechanical design of core components for a high performance light water reactor with a three pass core

International Nuclear Information System (INIS)

Fischer, Kai; Schneider, Tobias; Redon, Thomas; Schulenberg, Thomas; Starflinger, Joerg

2007-01-01

Nuclear reactors using supercritical water as coolant can achieve more than 500 deg. C core outlet temperature, if the coolant is heated up in three steps with intermediate mixing to avoid hot streaks. This method reduces the peak cladding temperatures significantly compared with a single heat up. The paper presents an innovative mechanical design which has been developed recently for such a High Performance Light Water Reactor. The core is built with square assemblies of 40 fuel pins each, using wire wraps as grid spacers. Nine of these assemblies are combined to a cluster having a common head piece and a common foot piece. A downward flow of additional moderator water, separated from the coolant, is provided in gaps between the assemblies and in a water box inside each assembly. The cluster head and foot pieces and mixing chambers, which are key components for this design, are explained in detail. (authors)

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.)

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.

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

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.

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.

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

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).

Reactor auxiliary cooling facility and coolant supplying method therefor

International Nuclear Information System (INIS)

Ando, Koji; Kinoshita, Shoichiro.

1996-01-01

A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

Reactor auxiliary cooling facility and coolant supplying method therefor

Energy Technology Data Exchange (ETDEWEB)

Ando, Koji; Kinoshita, Shoichiro

1996-06-07

A reactor auxiliary cooling facility of the present invention comprises a coolant recycling line for recycling coolants by way of a reactor auxiliary coolant pump and a cooling load, a gravitational surge tank for supplying coolants to the coolant recycling line and a supplemental water supplying line for supplying a supply the supplemental water to the tank. Then, a pressurization-type supply water surge tank is disposed for operating the coolant recycling line upon performing an initial system performance test in parallel with the gravitational surge tank. With such a constitution, the period of time required from the start of the installation of reactor auxiliary cooling facilities to the completion of the system performance test can be shortened at a reduced cost without enlarging the scale of the facility. (T.M.)

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)

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.

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.

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.

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

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

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)

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

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.

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.

Knock-limited performance of several internal coolants

Science.gov (United States)

Bellman, Donald R; Evvard, John C

1945-01-01

The effect of internal cooling on the knock-limited performance of an-f-28 fuel was investigated in a CFR engine, and the following internal coolants were used: (1) water, (2), methyl alcohol-water mixture, (3) ammonia-methyl alcohol-water mixture, (4) monomethylamine-water mixture, (5) dimethylamine-water mixture, and (6) trimethylamine-water mixture. Tests were run at inlet-air temperatures of 150 degrees and 250 degrees F. to indicate the temperature sensitivity of the internal-coolant solutions.

Supercritical Fluid Chromatographic Separation of Dimethylpolysiloxane Polymer

Energy Technology Data Exchange (ETDEWEB)

Pyo, Dong Jin; Lim, Chang Hyun [Kangwon National University, Chuncheon (Korea, Republic of)

2005-02-15

Water was used as a polar modifier and a μ-porasil column as a saturator column. The μ-porasil column was inserted between the pump outlet and the injection valve. During the passage of the supercritical fluid mobile phase through the silica column, a polar modifier (water) can be dissolved in the pressurized supercritical fluid. Dimethylpolysiloxane polymer has been known as more polar polymer than polystyrene polymer. Dimethylpolysiloxane polymer has never been separated using water modified mobile phase. In this paper, using a μ-porasil column as a saturator column, excellent supercritical fluid chromatograms of dimethylpolysiloxane oligomers were obtained. The use of compressed (dense) gases and supercritical fluids as chromatographic mobile phases in conjunction with liquid chromatographic (LC)-type packed columns was first reported by Klesper et al. in 1962. During its relatively short history, supercritical fluid chromatography (SFC) has become an attractive alternative to GC and LC in certain industrially important applications. SFC gives the advantage of high efficiency and allows the analysis of nonvolatile or thermally labile mixtures.

Supercritical Fluid Chromatographic Separation of Dimethylpolysiloxane Polymer

International Nuclear Information System (INIS)

Pyo, Dong Jin; Lim, Chang Hyun

2005-01-01

Water was used as a polar modifier and a μ-porasil column as a saturator column. The μ-porasil column was inserted between the pump outlet and the injection valve. During the passage of the supercritical fluid mobile phase through the silica column, a polar modifier (water) can be dissolved in the pressurized supercritical fluid. Dimethylpolysiloxane polymer has been known as more polar polymer than polystyrene polymer. Dimethylpolysiloxane polymer has never been separated using water modified mobile phase. In this paper, using a μ-porasil column as a saturator column, excellent supercritical fluid chromatograms of dimethylpolysiloxane oligomers were obtained. The use of compressed (dense) gases and supercritical fluids as chromatographic mobile phases in conjunction with liquid chromatographic (LC)-type packed columns was first reported by Klesper et al. in 1962. During its relatively short history, supercritical fluid chromatography (SFC) has become an attractive alternative to GC and LC in certain industrially important applications. SFC gives the advantage of high efficiency and allows the analysis of nonvolatile or thermally labile mixtures

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.

Modelling of heat transfer to fluids at a supercritical pressure

International Nuclear Information System (INIS)

Shuisheng, He

2014-01-01

A key feature of Supercritical Water-cooled Reactor (SCWR) is that, by raising the pressure of the reactor coolant fluid above the critical value, a phase change crisis is avoided. However, the changes in water density as it flows through the core of an SCWR are actually much higher than in the current water-cooled reactors. In a typical design, the ratio of the density of water at the core inlet to that at exit is as high as 7:1. Other fluid properties also vary significantly, especially around the pseudo-critical temperature (at which the specific heat capacity peaks). As a result, turbulent flow and heat transfer behaviour in the core is extremely complex and under certain conditions, significant heat transfer deterioration can potentially occur. Consequently, understanding and being able to predict flow and heat transfer phenomena under normal steady operation conditions and in start-up and hypothetical fault conditions are fundamental to the design of SCWR. There have been intensive studies on flow and heat transfer to fluids at supercritical pressure recently and several excellent review papers have been published. In the talk, we will focus on some turbulence modelling issues encountered in CFD simulations. The talk will first discuss some flow and heat transfer issues related to fluids at supercritical pressures and their potential implications in SCWR, and some recent developments in the understanding and modelling techniques of such problems, which will be followed by an outlook for some future developments.Factors which have a major influence on the flow and will be discussed are buoyancy and flow acceleration due to thermal expansion (both are due to density variations but involve different mechanisms) and the nonuniformity of other fluid properties. In addition, laminar-turbulent flow transition coupled with buoyancy and flow acceleration plays an important role in heat transfer effectiveness and wall temperature in the entrance region but such

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).

Heat Transfer Phenomena of Supercritical Fluids

Energy Technology Data Exchange (ETDEWEB)

Krau, Carmen Isabella; Kuhn, Dietmar; Schulenberg, Thomas [Forschungszentrum Karlsruhe, Institute for Nuclear and Energy Technologies, 76021 Karlsruhe (Germany)

2008-07-01

In concepts for supercritical water cooled reactors, the reactor core is cooled and moderated by water at supercritical pressures. The significant temperature dependence of the fluid properties of water requires an exact knowledge of the heat transfer mechanism to avoid fuel pin damages. Near the pseudo-critical point a deterioration of heat transfer might happen. Processes, that take place in this case, are not fully understood and are due to be examined systematically. In this paper a general overview on the properties of supercritical water is given, experimental observations of different authors will be reviewed in order to identify heat transfer phenomena and onset of occurrence. The conceptional design of a test rig to investigate heat transfer in the boundary layer will be discussed. Both, water and carbon dioxide, may serve as operating fluids. The loop, including instrumentation and safety devices, is shown and suitable measuring methods are described. (authors)

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)

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.

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

Development of a Convective Heat Transfer Correlation of a Supercritical CO2 with Vertical Downward Flow in Circular Tubes

International Nuclear Information System (INIS)

Yoo, Tae Ho; Kim, Hwan Yeol; Bae, Yoon Yeong

2009-01-01

Pressure of coolant flowing through a SCWR core subchannel is supercritical and the heat transfer behavior is known to be quite different from those at a subcritical pressure. Therefore the heat transfer study in a supercritical pressure is required for the acquisition of a reliable heat transfer correlation. A downward flow as well as an upward flow occurs in a multi-pass reactor core. The heat transfer at a supercritical pressure in downward channel has been known to result in a quite different behavior from an upward flow. An experiment for a supercritical CO 2 flowing vertically downward in circular tubes with inner diameters of 6.32 mm and 9 mm was performed by using SPHINX(Supercritical Pressure Heat transfer Investigation for NeXt generation) at KAERI. The obtained test results are compared with the estimations from the existing correlations and an empirical formula for a downward flow is suggested

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

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.

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

Development of a test facility for analyzing supercritical fluid blowdown

International Nuclear Information System (INIS)

Roberto, Thiago D.; Alvim, Antonio C.M.

2015-01-01

The generation IV nuclear reactors under development mostly use supercritical fluids as the working fluid because higher temperatures improve the thermal efficiency. Supercritical fluids are used by modern nuclear power plants to achieve thermal efficiencies of around 45%. With water as the supercritical working fluid, these plants operate at a high temperature and pressure. However, experiments on supercritical water are limited by technical and financial difficulties. These difficulties can be overcome by using model fluids, which have more feasible supercritical conditions and exhibit a lower critical pressure and temperature. Experimental research is normally used to determine the conditions under which model fluids represent supercritical fluids under steady-state conditions. A fluid-to-fluid scaling approach has been proposed to determine model fluids that can represent supercritical fluids in a transient state. This paper presents an application of fractional scale analysis to determine the simulation parameters for a depressurization test facility. Carbon dioxide (CO 2 ) and R134a gas were considered as the model fluids because their critical point conditions are more feasible than those of water. The similarities of water (prototype), CO 2 (model) and R134a (model) for depressurization in a pressure vessel were analyzed. (author)

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.

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

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.

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

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

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)

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)

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.

Fuel-coolant interactions: preliminary experiments on the effect of gases dissolved in the 'coolant'

International Nuclear Information System (INIS)

Asher, R.C.; Davies, D.; Jones, P.G.

1976-12-01

A simple apparatus has been used to study fuel-coolant interactions under reasonably well controlled conditions. Preliminary experiments have used water as the 'coolant' and molten tin at 800 0 C as the 'fuel' and have investigated how the violence of the interaction is affected by dissolving gases (oxygen, nitrogen, carbon dioxide and nitrous oxide) in the water. It was found that saturating the water with carbon dioxide or nitrous oxide completely suppresses the violent interaction. Experiments in which the concentrations of these gases were varied showed that a certain critical concentration was needed; below this concentration the dissolved gas has no significant effect but above it the suppression is

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)

Simulation of steam explosion in stratified melt-coolant configuration

International Nuclear Information System (INIS)

Leskovar, Matjaž; Centrih, Vasilij; Uršič, Mitja

2016-01-01

Highlights: • Strong steam explosions may develop spontaneously in stratified configurations. • Considerable melt-coolant premixed layer formed in subcooled water with hot melts. • Analysis with MC3D code provided insight into stratified steam explosion phenomenon. • Up to 25% of poured melt was mixed with water and available for steam explosion. • Better instrumented experiments needed to determine dominant mixing process. - Abstract: A steam explosion is an energetic fuel coolant interaction process, which may occur during a severe reactor accident when the molten core comes into contact with the coolant water. In nuclear reactor safety analyses steam explosions are primarily considered in melt jet-coolant pool configurations where sufficiently deep coolant pool conditions provide complete jet breakup and efficient premixture formation. Stratified melt-coolant configurations, i.e. a molten melt layer below a coolant layer, were up to now believed as being unable to generate strong explosive interactions. Based on the hypothesis that there are no interfacial instabilities in a stratified configuration it was assumed that the amount of melt in the premixture is insufficient to produce strong explosions. However, the recently performed experiments in the PULiMS and SES (KTH, Sweden) facilities with oxidic corium simulants revealed that strong steam explosions may develop spontaneously also in stratified melt-coolant configurations, where with high temperature melts and subcooled water conditions a considerable melt-coolant premixed layer is formed. In the article, the performed study of steam explosions in a stratified melt-coolant configuration in PULiMS like conditions is presented. The goal of this analytical work is to supplement the experimental activities within the PULiMS research program by addressing the key questions, especially regarding the explosivity of the formed premixed layer and the mechanisms responsible for the melt-water mixing. To

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

Development of a Convective Heat Transfer Correlation of a Supercritical CO{sub 2} with Vertical Downward Flow in Circular Tubes

Energy Technology Data Exchange (ETDEWEB)

Yoo, Tae Ho; Kim, Hwan Yeol; Bae, Yoon Yeong [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2009-05-15

Pressure of coolant flowing through a SCWR core subchannel is supercritical and the heat transfer behavior is known to be quite different from those at a subcritical pressure. Therefore the heat transfer study in a supercritical pressure is required for the acquisition of a reliable heat transfer correlation. A downward flow as well as an upward flow occurs in a multi-pass reactor core. The heat transfer at a supercritical pressure in downward channel has been known to result in a quite different behavior from an upward flow. An experiment for a supercritical CO{sub 2} flowing vertically downward in circular tubes with inner diameters of 6.32 mm and 9 mm was performed by using SPHINX(Supercritical Pressure Heat transfer Investigation for NeXt generation) at KAERI. The obtained test results are compared with the estimations from the existing correlations and an empirical formula for a downward flow is suggested.

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

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

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

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

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

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.

Coolant cleanup system for BWR type reactor

International Nuclear Information System (INIS)

Kinoshita, Shoichiro; Araki, Hidefumi.

1993-01-01

The cleanup system of the present invention removes impurity ions and floating materials accumulated in a reactor during evaporation of coolants in the nuclear reactor. That is, coolants pass pipelines from a pressure vessel using pressure difference between a high pressure in the pressure vessel and a low pressure at the upstream of a condensate filtration/desalting device of a condensate/feed water system as a driving source, during which cations and floating materials are removed in a high temperature filtration/desalting device and coolants flow into the condensate/feedwater system. Impurities containing anions are removed here by the condensates filtration/desalting device. Then, they return to the pressure vessel while pressurized and heated by a condensate pump, a feed water pump and a feed water heater. At least pumps, a heat exchanger for heating, a filtration/desalting device for removing anions and pipelines connecting them used exclusively for the coolant cleanup system are no more necessary. (I.S.)

Pulse radiolysis study on temperature and pressure dependence of the yield of solvated electron in methanol from room temperature to supercritical condition

International Nuclear Information System (INIS)

Han, Zhenhui; He, Hui; Lin, Mingzhang; Muroya, Yusa; Katsumura, Yosuke

2012-09-01

A new concept of nuclear reactor, supercritical water-cooled reactor (SCWR), has been proposed, which is based on the success of the use of supercritical water (SCW) in fossil fuel power plants for more than three decades. This new concept reactor has advantages of higher thermal conversion efficiency, simplicity in structure, safety, etc, and it has been selected as one of the reactor concepts for the next generation nuclear reactor systems. In these reactors, the same as in boiling water reactors (BWR) and pressurized water reactors (PWR), water is used not only as a coolant but also as a moderator. It is very important to understand the behavior of the radiolysis products of water under the supercritical condition, since the water is exposed to a strong radiation field under very high temperature condition. Usually, in order to predict the concentrations of water decomposition products with carrying out some kinds of computer simulations, knowledge of the temperature and/or pressure dependent G-values (denoting the experimentally measured radiolytic yields) as well as of the rate constants of a set of reactions becomes very important. Therefore, in recent years, two groups from Argonne National Laboratory and The University of Tokyo, simultaneously conducted two projects aimed at obtaining basic data on radiolysis of SCW. However, it is still lack of reliable radiolytic yields of water decomposition products in very high temperature region. As we known, the properties of solvated electrons in polar liquid are very helpful for our understanding how they play a central role in many processes, such as solvation and reducing reactions. The solvated electron can also be used as a probe to determine the dynamic nature of the polar liquid systems. Comparing to water, the primary alcohols have much milder critical points, for example, for water and methanol, the critical temperature and pressure are 374 deg. C and 22.1 MPa and 239.5 deg. C and 8.1 MPa, respectively

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)

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

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.)

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)

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

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.

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

Neutronic and Thermal-hydraulic Modelling of High Performance Light Water Reactor

Energy Technology Data Exchange (ETDEWEB)

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

2008-07-01

High Performance Light Water Reactor (HPLWR), which is studied in EU project 'HPLWR2', uses water at supercritical pressures as coolant and moderator to achieve higher core outlet temperature and thus higher efficiency compared to present reactors. At VTT Technical Research Centre of Finland, functionality of the thermal-hydraulics in the coupled reactor dynamics code TRAB3D/ SMABRE was extended to supercritical pressures for the analyses of HPLWR. Input models for neutronics and thermal-hydraulics were made for TRAB3D/ SMABRE according to the latest HPLWR design. A preliminary analysis was performed in which the capability of SMABRE in the transition from supercritical pressures to subcritical pressures was demonstrated. Parameterized two-group cross sections for TRAB3D neutronics were received from Hungarian Academy of Sciences KFKI Atomic Energy Research Institute together with a subroutine for handling them. PSG, a new Monte Carlo transport code developed at VTT, was also used to generate two-group constants for HPLWR and comparisons were made with the KFKI cross sections and MCNP calculations. (author)

Neutronic and Thermal-hydraulic Modelling of High Performance Light Water Reactor

International Nuclear Information System (INIS)

Seppaelae, Malla

2008-01-01

High Performance Light Water Reactor (HPLWR), which is studied in EU project 'HPLWR2', uses water at supercritical pressures as coolant and moderator to achieve higher core outlet temperature and thus higher efficiency compared to present reactors. At VTT Technical Research Centre of Finland, functionality of the thermal-hydraulics in the coupled reactor dynamics code TRAB3D/ SMABRE was extended to supercritical pressures for the analyses of HPLWR. Input models for neutronics and thermal-hydraulics were made for TRAB3D/ SMABRE according to the latest HPLWR design. A preliminary analysis was performed in which the capability of SMABRE in the transition from supercritical pressures to subcritical pressures was demonstrated. Parameterized two-group cross sections for TRAB3D neutronics were received from Hungarian Academy of Sciences KFKI Atomic Energy Research Institute together with a subroutine for handling them. PSG, a new Monte Carlo transport code developed at VTT, was also used to generate two-group constants for HPLWR and comparisons were made with the KFKI cross sections and MCNP calculations. (author)

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

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)

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)

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.

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.

Organic coolant in Winnipeg riverbed sediments

International Nuclear Information System (INIS)

Guthrie, J.E.; Acres, O.E.

1979-03-01

Between January and May 1977 a prolonged leak of organic coolant occurred from the Whiteshell Nuclear Research Establishment's nuclear reactor, and a minimum of 1450 kg of coolant entered the Winnipeg River and was deposited on the riverbed. The level of radioactivity associated with this coolant was low, contributing less than 0.2 μGy (0.02 mrad) a year to the natural background gamma radiation field from the riverbed. The concentration of coolant in the water samples never exceeded 0.02 mg/L, the lower limit of detection. The mortality of crayfish, held in cages where the riverbed was covered with the largest deposits of coolant, was not significantly different from that in the control cages upstream of the outfall. No evidence of fish kill was found. (author)

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)

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)

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.

Cryogenic recovery analysis of forced flow supercritical helium cooled superconductors

International Nuclear Information System (INIS)

Lee, A.Y.

1977-08-01

A coupled heat conduction and fluid flow method of solution was presented for cryogenic stability analysis of cabled composite superconductors of large scale magnetic coils. The coils are cooled by forced flow supercritical helium in parallel flow channels. The coolant flow reduction in one of the channels during the spontaneous recovery transient, after the conductor undergoes a transition from superconducting to resistive, necessitates a parallel channel analysis. A way to simulate the parallel channel analysis is described to calculate the initial channel inlet flow rate required for recovery after a given amount of heat is deposited. The recovery capability of a NbTi plus copper composite superconductor design is analyzed and the results presented. If the hydraulics of the coolant flow is neglected in the recovery analysis, the recovery capability of the superconductor will be over-predicted

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.

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)

Analytical approximations for thermophysical properties of supercritical nitrogen (SCN) to be used in futuristic high temperature superconducting (HTS) cables

Energy Technology Data Exchange (ETDEWEB)

Dondapati, Raja Sekhar, E-mail: drsekhar@ieee.org [School of Mechanical Engineering, Lovely Professional University, Phagwara, Punjab 144401 (India); Ravula, Jeswanth [School of Mechanical Engineering, Lovely Professional University, Phagwara, Punjab 144401 (India); Thadela, S. [Department of Mechanical Engineering, Andhra University, Visakhapatnam, Andhra Pradesh (India); Usurumarti, Preeti Rao [Department of Mechanical Engineering, P.V.K. Institute of Technology, Anantapur, Andhra Pradesh (India)

2015-12-15

Future power transmission applications demand higher efficiency due to the limited resources of energy. In order to meet such demand, a novel method of transmission is being developed using High Temperature Superconducting (HTS) cables. However, these HTS cables need to be cooled below the critical temperature of superconductors used in constructing the cable to retain the superconductivity. With the advent of new superconductors whose critical temperatures having reached up to 134 K (Hg based), a need arises to find a suitable coolant which can accommodate the heating loads on the superconductors. The present work proposes, Supercritical Nitrogen (SCN) to be a feasible coolant to achieve the required cooling. Further, the feasibility of proposed coolant to be used in futuristic HTS cables is investigated by studying the thermophysical properties such as density, viscosity, specific heat and thermal conductivity with respect to temperature (T{sub C} + 10 K) and pressure (P{sub C} + 10 bar). In addition, few temperature dependent analytical functions are developed for thermophysical properties of SCN which are useful in predicting thermohydraulic performance (pressure drop, pumping power and cooling capacity) using numerical or computational techniques. Also, the developed analytical functions are used to calculate the pumping power and the temperature difference between inlet and outlet of HTS cable. These results are compared with those of liquid nitrogen (LN2) and found that the circulating pumping power required to pump SCN is significantly smaller than that to pump LN2. Further, it is found that the temperature difference between the inlet and outlet is smaller as compared to that when LN2 is used, SCN can be preferred to cool long length Hg based HTS cables. - Highlights: • Analytical functions are developed for thermophysical properties of Supercritical Nitrogen. • Error analysis shows extremely low errors in the developed analytical functions. �

Analytical approximations for thermophysical properties of supercritical nitrogen (SCN) to be used in futuristic high temperature superconducting (HTS) cables

International Nuclear Information System (INIS)

Dondapati, Raja Sekhar; Ravula, Jeswanth; Thadela, S.; Usurumarti, Preeti Rao

2015-01-01

Future power transmission applications demand higher efficiency due to the limited resources of energy. In order to meet such demand, a novel method of transmission is being developed using High Temperature Superconducting (HTS) cables. However, these HTS cables need to be cooled below the critical temperature of superconductors used in constructing the cable to retain the superconductivity. With the advent of new superconductors whose critical temperatures having reached up to 134 K (Hg based), a need arises to find a suitable coolant which can accommodate the heating loads on the superconductors. The present work proposes, Supercritical Nitrogen (SCN) to be a feasible coolant to achieve the required cooling. Further, the feasibility of proposed coolant to be used in futuristic HTS cables is investigated by studying the thermophysical properties such as density, viscosity, specific heat and thermal conductivity with respect to temperature (T_C + 10 K) and pressure (P_C + 10 bar). In addition, few temperature dependent analytical functions are developed for thermophysical properties of SCN which are useful in predicting thermohydraulic performance (pressure drop, pumping power and cooling capacity) using numerical or computational techniques. Also, the developed analytical functions are used to calculate the pumping power and the temperature difference between inlet and outlet of HTS cable. These results are compared with those of liquid nitrogen (LN2) and found that the circulating pumping power required to pump SCN is significantly smaller than that to pump LN2. Further, it is found that the temperature difference between the inlet and outlet is smaller as compared to that when LN2 is used, SCN can be preferred to cool long length Hg based HTS cables. - Highlights: • Analytical functions are developed for thermophysical properties of Supercritical Nitrogen. • Error analysis shows extremely low errors in the developed analytical functions. �

Study on specifics of thermophysical properties of supercritical fluids in power engineering applications

International Nuclear Information System (INIS)

Mann, David; Pioro, Igor

2015-01-01

SuperCritical Pressures (SCPs) and SuperCritical Fluids (SFCs) are widely used in many industries worldwide. The largest application of SCPs is in the power industry in advanced coal-fired power plants. It is well-known that moving from subcritical-pressure power plants to SCP power plants increases gross thermal efficiency from 38-42% to about 50-55%. Despite all advances in thermal power-plants design and operation worldwide, they are still considered as not “environmentally friendly” due to significant carbon-dioxide emissions and air pollution as a result of the combustion process. In addition, coal-fired power-plants also produce virtual mountains of slag and ash, and other gas emissions that may contribute to acid rains. Therefore, the demand for clean, non-fossil-based electricity is growing. Due to this, nuclear power is considered as a basis for future electricity generation in the world. One of the major problems with current fleet of Nuclear Power Plants (NPPs) is their relatively low thermal efficiencies, especially, of water-cooled-reactor NPPs (the vast majority of NPPs) (30-36%), compared to those of advanced thermal power plants (55-62%). Based on that, next generation or Generation-IV reactors corresponding to those NPPs should definitely be more efficient. Higher level of thermal efficiencies can be reached only due to higher temperatures and, in some cases, higher pressures inside reactors and, especially, in power cycles of Generation-IV NPPs. Analysis of the six concepts of Generation-IV reactors and NPPs shows that three reactor concepts will use SCFs as reactor coolants (helium and water) and all concepts can be linked to SCFs as working fluids in power cycles (SC helium and /or carbon dioxide in the Brayton gas-turbine cycle, and SC water in the Rankine steam-turbine cycle). Therefore, the exact knowledge of specifics of thermophysical properties of SC helium, water and carbon dioxide is very important for any advances in these new

Thermodynamic Optimization of Supercritical CO{sub 2} Brayton Cycles

Energy Technology Data Exchange (ETDEWEB)

Rhim, Dong-Ryul; Park, Sung-Ho; Kim, Su-Hyun; Yeom, Choong-Sub [Institute for Advanced Engineering, Yongin (Korea, Republic of)

2015-05-15

The supercritical CO{sub 2} Brayton cycle has been studied for nuclear applications, mainly for one of the alternative power conversion systems of the sodium cooled fast reactor, since 1960's. Although the supercritical CO{sub 2} Brayton cycle has not been expected to show higher efficiency at lower turbine inlet temperature over the conventional steam Rankine cycle, the higher density of supercritical CO{sub 2} like a liquid in the supercritical region could reduce turbo-machinery sizes, and the potential problem of sodium-water reaction with the sodium cooled fast reactor might be solved with the use of CO{sub 2} instead of water. The supercritical CO{sub 2} recompression Brayton cycle was proposed for the better thermodynamic efficiency than for the simple supercritical CO{sub 2} Brayton cycle. Thus this paper presents the efficiencies of the supercritical CO{sub 2} recompression Brayton cycle along with several decision variables for the thermodynamic optimization of the supercritical CO{sub 2} recompression Brayton cycle. The analytic results in this study show that the system efficiency reaches its maximum value at a compressor outlet pressure of 200 bars and a recycle fraction of 30 %, and the lower minimum temperature approach at the two heat exchangers shows higher system efficiency as expected.

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

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.

Supercritical fluid technologies for ceramic-processing applications

International Nuclear Information System (INIS)

Matson, D.W.; Smith, R.D.

1989-01-01

This paper reports on the applications of supercritical fluid technologies for ceramic processing. The physical and chemical properties of these densified gases are summarized and related to their use as solvents and processing media. Several areas are identified in which specific ceramic processes benefit from the unique properties of supercritical fluids. The rapid expansion of supercritical fluid solutions provides a technique for producing fine uniform powders and thin films of widely varying materials. Supercritical drying technologies allow the formation of highly porous aerogel products with potentially wide application. Hydrothermal processes leading to the formation of large single crystals and microcrystalline powders can also be extended into the supercritical regime of water. Additional applications and potential applications are identified in the areas of extraction of binders and other additives from ceramic compacts, densification of porous ceramics, the formation of powders in supercritical micro-emulsions, and in preceramic polymer processing

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)

Analysis of supercritical vapor explosions using thermal detonation wave theory

Energy Technology Data Exchange (ETDEWEB)

Shamoun, B.I.; Corradini, M.L. [Univ. of Wisconsin, Madison, WI (United States)

1995-09-01

The interaction of certain materials such as Al{sub 2}O{sub 3} with water results in vapor explosions with very high (supercritical) pressures and propagation velocities. A quasi-steady state analysis of supercritical detonation in one-dimensional multiphase flow was applied to analyze experimental data of the KROTOS (26-30) set of experiments conducted at the Joint Research Center at Ispra, Italy. In this work we have applied a new method of solution which allows for partial fragmentation of the fuel in the shock adiabatic thermodynamic model. This method uses known experiment values of the shock pressure and propagation velocity to estimate the initial mixing conditions of the experiment. The fuel and coolant were both considered compressible in this analysis. In KROTOS 26, 28, 29, and 30 the measured values of the shock pressure by the experiment were found to be higher than 25, 50, 100, and 100 Mpa respectively. Using the above data for the wave velocity and our best estimate for the values of the pressure, the predicted minimum values of the fragmented mass of the fuel were found to be 0.026. 0.04, 0.057, and 0.068 kg respectively. The predicted values of the work output corresponding to the above fragmented masses of the fuel were found to be 40, 84, 126, and 150 kJ respectively, with predicted initial void fractions of 112%, 12.5%, 8%, and 6% respectively.

Can supercritical oxidation of sewage sludge be an alternative for supercritical gasification?; Kan superkritische oxidatie van zuiveringsslib een alternatief zijn voor superkritische vergassing?

Energy Technology Data Exchange (ETDEWEB)

Rulkens, W. [Wageningen UR, Wageningen (Netherlands); Wentink, J. [Horizon Solutions, Leiden (Netherlands)

2013-05-15

In the context of the development of The Energy Factory a number of technologies has been identified that may be interesting to develop further. Two of these techniques relate to the conversion of sludge in supercritical water: supercritical gasification of sludge and supercritical oxidation of sludge [Dutch] In het kader van de ontwikkeling van De Energiefabriek is een aantal technologieen geidentificeerd die mogelijk interessant zijn om verder te ontwikkelen. Twee van deze technieken hebben betrekking op de conversie van slib in superkritisch water: superkritische slibvergassing en superkritische sliboxidatie.

Fuel coolant interaction experiment by direct electrical heating method

International Nuclear Information System (INIS)

Takeda, Tsuneo; Hirano, Kenmei

1979-01-01

In the PCM (Power Cooling Mismatch) experiments, the FCI (Fuel Coolant Interaction) test is one of necessary tests in order to predict various phenomena that occur during PCM in the core. A direct electrical heating method is used for the FCI tests for fuel pellet temperature of over 1000 0 C. Therefore, preheating is required before initiating the direct electrical heating. The fuel pin used in the FCI tests is typical LWR fuel element, which is surrounded by coolant water. It is undersirable to heat up the coolant water during preheating of the fuel pin. Therefore, a zirconia (ZrO 2 ) pellet which is similar to a UO 2 pellet in physical and chemical properties is used. Electric property (electric conductivity) of ZrO 2 is particularly suitable for direct electrical heating as in the case of UO 2 . In this experiment, ZrO 2 pellet (melting point 2500 0 C) melting was achieved by use of both preheating and direct electrical heating. Temperature changes of coolant and fuel surface, as well as the pressure change of coolant water, were measured. The molten fuel interacted with the coolant and generated shock waves. A portion of this molten fuel fragmented into small particles during this interaction. The peak pressure of the observed shock wave was about 35 bars. The damaged fuel pin was photographed after disassembly. This report shows the measured coolant pressure changes and the coolant temperature changes, as well as photographs of damaged fuel pin and fuel fragments. (author)

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...

Device for preventing coolant in a reactor from being lost

International Nuclear Information System (INIS)

Maruyama, Hiromi; Matsumoto, Tomoyuki.

1975-01-01

Object: To prevent all of coolant from being lost from the core at the time of failure in rupture of pipe in a recirculation system to cool the core with the coolant remained within the reactor. Structure: A valve, which will be closed when a water level of the coolant within the core is in a level less than a predetermined level, is provided on a recirculating water outlet nozzle in a pressure vessel to thereby prevent the coolant from being lost when the pipe is broken, thus cooling the core by means of reduced-pressure boiling of coolant remained within the core and boiling due to heat, and restraining core reactivity by means of void produced at that time. (Kamimura, M.)

Limits to fuel/coolant mixing

International Nuclear Information System (INIS)

Corradini, M.L.; Moses, G.A.

1985-01-01

The vapor explosion process involves the mixing of fuel with coolant prior to the explosion. A number of analysts have identified limits to the amount of fuel/coolant mixing that could occur within the reactor vessel following a core melt accident. Past models are reviewed and a sim plified approach is suggested to estimate the upper limit on the amount of fuel/coolant mixing pos sible. The approach uses concepts first advanced by Fauske in a different way. The results indicat that water depth is an important parameter as well as the mixing length scale D /SUB mix/ , and for large values of D /SUB mix/ the fuel mass mixed is limited to <7% of the core mass

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

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

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

A Conceptual Study on a Supercritical CO_2-cooled Micro Modular Reactor

International Nuclear Information System (INIS)

Yu, Hwanyeal; Hartanto, Donny; Kim, Yonghee

2014-01-01

A Micro Modular Reactor (MMR) using Supercritical-CO_2 (S-CO_2) as coolant has been investigated from the neutronics perspective. The MMR is designed to be transportable so it can reach the remote areas. The thermal power of the reactor is 36.2 M Wth. The size of the active core is limited to 1.2 m length and 93.16 cm width. The size of whole core is 2.8 m length and 166.9 cm width. The reactor lifetime design target is 20 years. To maximize the fuel volume fraction in the core, high density uranium nitride UN"1"5 was used. The PbO/MgO reflector was also utilized to improve the neutron economy. The S-CO_2 is chosen as the coolant because it offers a higher thermal efficiency. In this study, neutronics calculations and depletion using McCARD Monte Carlo code has been done to determine the lifetime and behavior of the core. Several important safety parameters such as Control Rod worth, Doppler reactivity coefficients and coolant void reactivity coefficient have also been analyzed. (author)

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.

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.

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.

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)

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.

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...

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

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)

Stress Analysis of Fuel Rod under Axial Coolant Flow

Energy Technology Data Exchange (ETDEWEB)

Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung [Chungnam National University, Daejeon (Korea, Republic of); Park, Num Kyu; Jeon, Kyung Rok [Kerea Nuclear Fuel., Daejeon (Korea, Republic of)

2010-05-15

A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

Stress Analysis of Fuel Rod under Axial Coolant Flow

International Nuclear Information System (INIS)

Jin, Hai Lan; Lee, Young Shin; Lee, Hyun Seung; Park, Num Kyu; Jeon, Kyung Rok

2010-01-01

A pressurized water reactor(PWR) fuel assembly, is a typical bundle structure, which uses light water as a coolant in most commercial nuclear power plants. Fuel rods that have a very slender and long clad are supported by fuel assembly which consists of several spacer grids. A coolant is a fluid which flows through device to prevent its overheating, transferring the heat produced by the device to other devices that use or dissipate it. But at the same time, the coolant flow will bring out the fluid induced vibration(FIV) of fuel rods and even damaged the fuel rod. This study has been conducted to investigate the flow characteristics and nuclear reactor fuel rod stress under effect of coolant. Fluid structure interaction(FSI) analysis on nuclear reactor fuel rod was performed. Fluid analysis of the coolant which flow along the axial direction and structural analysis under effect of flow velocity were carried out under different output flow velocity conditions

Reactor coolant pump transportation incident

International Nuclear Information System (INIS)

Noce, D.

1992-01-01

This paper reports on an incident, which occurred on August 27, 1991, in which a Reactor Coolant Pump motor en route from Surry Power Station to Westinghouse repair facilities struck the overpass at the junction of Interstate 64 and Jefferson Avenue in Newport News, Virginia. The transport container that housed the reactor coolant pump motor failed to clear the overpass. The force of the impact dislodged the container and motor from the truck bed, and it landed on the acceleration land and road shoulder. Upon impact, the container broke open and exposed the reactor coolant pump motor. Incidental radioactively contaminated water that remained in the motor coolers drained onto the road, contaminating the aggregate as well as the underlying gravel

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)

Analysis of molten fuel-coolant interaction during a reactivity-initiated accident experiment

International Nuclear Information System (INIS)

El-Genk, M.S.; Hobbins, R.R.

1981-01-01

The results of a reactivity-initiated accident experiment, designated RIA-ST-4, are discussed and analyzed with regard to molten fuel-coolant interaction (MFCI). In this experiment, extensive amounts of molten UO 2 fuel and zircaloy cladding were produced and fragmented upon mixing with the coolant. Coolant pressurization up to 35 MPa and coolant overheating in excess of 940 K occurred after fuel rod failure. The initial coolant conditions were similar to those in boiling water reactors during a hot startup (that is, coolant pressure of 6.45 MPa, coolant temperature of 538 K, and coolant flow rate of 85 cm 3 /s). It is concluded that the high coolant pressure recorded in the RIA-ST-4 experiment was caused by an energetic MFCI and was not due to gas release from the test rod at failure, Zr/water reaction, or to UO 2 fuel vapor pressure. The high coolant temperature indicated the presence of superheated steam, which may have formed during the expansion of the working fluid back to the initial coolant pressure; yet, the thermal-to-mechanical energy conversion ratio is estimated to be only 0.3%

Thorium Fuel Utilization Analysis on Small Long Life Reactor for Different Coolant Types

Science.gov (United States)

Permana, Sidik

2017-07-01

A small power reactor and long operation which can be deployed for less population and remote area has been proposed by the IAEA as a small and medium reactor (SMR) program. Beside uranium utilization, it can be used also thorium fuel resources for SMR as a part of optimalization of nuclear fuel as a “partner” fuel with uranium fuel. A small long-life reactor based on thorium fuel cycle for several reactor coolant types and several power output has been evaluated in the present study for 10 years period of reactor operation. Several key parameters are used to evaluate its effect to the reactor performances such as reactor criticality, excess reactivity, reactor burnup achievement and power density profile. Water-cooled types give higher criticality than liquid metal coolants. Liquid metal coolant for fast reactor system gives less criticality especially at beginning of cycle (BOC), which shows liquid metal coolant system obtains almost stable criticality condition. Liquid metal coolants are relatively less excess reactivity to maintain longer reactor operation than water coolants. In addition, liquid metal coolant gives higher achievable burnup than water coolant types as well as higher power density for liquid metal coolants.

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

Experimental interaction of magma and “dirty” coolants

Science.gov (United States)

Schipper, C. Ian; White, James D. L.; Zimanowski, Bernd; Büttner, Ralf; Sonder, Ingo; Schmid, Andrea

2011-03-01

The presence of water at volcanic vents can have dramatic effects on fragmentation and eruption dynamics, but little is known about how the presence of particulate matter in external water will further alter eruptions. Volcanic edifices are inherently “dirty” places, where particulate matter of multiple origins and grainsizes typically abounds. We present the results of experiments designed to simulate non-explosive interactions between molten basalt and various “coolants,” ranging from homogeneous suspensions of 0 to 30 mass% bentonite clay in pure water, to heterogeneous and/or stratified suspensions including bentonite, sand, synthetic glass beads and/or naturally-sorted pumice. Four types of data are used to characterise the interactions: (1) visual/video observations; (2) grainsize and morphology of resulting particles; (3) heat-transfer data from a network of eight thermocouples; and (4) acoustic data from three force sensors. In homogeneous coolants with ~20% sediment, heat transfer is by forced convection and conduction, and thermal granulation is less efficient, resulting in fewer blocky particles, larger grainsizes, and weaker acoustic signals. Many particles are droplet-shaped or/and “vesicular,” containing bubbles filled with coolant. Both of these particle types indicate significant hydrodynamic magma-coolant mingling, and many of them are rewelded into compound particles. The addition of coarse material to heterogeneous suspensions further slows heat transfer thus reducing thermal granulation, and variable interlocking of large particles prevents efficient hydrodynamic mingling. This results primarily in rewelded melt piles and inefficient distribution of melt and heat throughout the coolant volume. Our results indicate that even modest concentrations of sediment in water will significantly limit heat transfer during non-explosive magma-water interactions. At high concentrations, the dramatic reduction in cooling efficiency and increase in

Minimizing secondary coolant blowdown in HANARO

International Nuclear Information System (INIS)

Park, Y. C.; Woo, J. S.; Ryu, J. S.; Cho, Y. G.; Lim, N. Y.

2000-01-01

There is about 80m 3 /h loss of the secondary cooling water by evaporation, windage and blowdown during the operation of HANARO, 30MW research reactor. The evaporation and the windage is necessary loss to maintain the performance of cooling tower, but the blowdown is artificial lose to get rid of the foreign material and to maintain the quality of the secondary cooling water. Therefore, minimizing the blowdown loss was studied. It was confirmed, through the relation of the number of cycle and the loss rate of secondary coolant, that the number of cycle is saturated to 12 without blowdown because of the windage loss. When the secondary coolant is treated by high Ca-hardness treatment program (the number of cycle > 10) to maintain the number of cycle around 12 without blowdown, only the turbidity exceeds the limit. By adding filtering system it was confirmed, through the relation of turbidity and filtering rate of secondary cooling water, that the turbidity is reduced below the limit (5 deg.) by 2% of filtering rate without blowdown. And it was verified, through the performance test of back-flow filtering unit, that this unit gets rid of foreign material up to 95% of the back-flow and that the water can be reused as coolant. Therefore, the secondary cooling water can be treated by the high Ca-hardness program and filter system without blowdown

Radioactivity analysis of KAMINI reactor coolant from regulatory perspectives

International Nuclear Information System (INIS)

Srinivasan, T.K.; Sulthan, Bajeer; Sarangapani, R.; Jose, M.T.; Venkatraman, B.; Thilagam, L.

2016-01-01

KAMINI (a 30kWt) research reactor is operated for neutron radiography of fuel subassemblies and pyro devices and activation analysis of various samples. The reactor is fueled by 233 U and DM water is used as the coolant. During reactor operation, fission product noble gasses (FPNGs) such as 85m Kr, 87 Kr, 88 Kr, 135 Xe, 135m Xe and 138 Xe are detected in the coolant water. In order to detect clad failure, the water is sampled during reactor operation at regular intervals as per the technical specifications. In the present work, analysis of measured activities in coolant samples collected during reactor operation at 25 kWt are presented and compared with computed values obtained using ORIGEN (Isotope Generation) code

Fuel-Coolant Interactions: Visualization and Mixing Measurements

International Nuclear Information System (INIS)

Loewen, Eric P.; Bonazza, Riccardo; Corradini, Michael L.; Johannesen, Robert E.

2002-01-01

Dynamic X-ray imaging of fuel-coolant interactions (FCI), including quantitative measurement of fuel-coolant volume fractions and length scales, has been accomplished with a novel imaging system at the Nuclear Safety Research Center at the University of Wisconsin, Madison. The imaging system consists of visible-light high-speed digital video, low-energy X-ray digital imaging, and high-energy X-ray digital imaging subsystems. The data provide information concerning the melt jet velocity, melt jet configuration, melt volume fractions, void fractions, and spatial and temporal quantification of premixing length scales for a model fuel-coolant system of molten lead poured into a water pool (fuel temperatures 500 to 1000 K; jet diameters 10 to 30 mm; coolant temperatures 20 to 90 deg. C). Overall results indicate that the FCI has three general regions of behavior, with the high fuel-coolant temperature region similar to what might be expected under severe accident conditions. It was observed that the melt jet leading edge has the highest void fraction and readily fragments into discrete masses, which then subsequently subdivide into smaller masses of length scales <10 mm. The intact jet penetrates <3 to 5 jet length/jet diameter before this breakup occurs into discrete masses, which continue to subdivide. Hydrodynamic instabilities can be visually identified at the leading edge and along the jet column with an interfacial region that consists of melt, vapor, and water. This interface region was observed to grow in size as the water pool temperature was increased, indicating mixing enhancement by boiling processes

Experimental Investigation of Heat Transfer Characteristics of Automobile Radiator using TiO2-Nanofluid Coolant

Science.gov (United States)

Salamon, V.; Senthil kumar, D.; Thirumalini, S.

2017-08-01

The use of nanoparticle dispersed coolants in automobile radiators improves the heat transfer rate and facilitates overall reduction in size of the radiators. In this study, the heat transfer characteristics of water/propylene glycol based TiO2 nanofluid was analyzed experimentally and compared with pure water and water/propylene glycol mixture. Two different concentrations of nanofluids were prepared by adding 0.1 vol. % and 0.3 vol. % of TiO2 nanoparticles into water/propylene glycol mixture (70:30). The experiments were conducted by varying the coolant flow rate between 3 to 6 lit/min for various coolant temperatures (50°C, 60°C, 70°C, and 80°C) to understand the effect of coolant flow rate on heat transfer. The results showed that the Nusselt number of the nanofluid coolant increases with increase in flow rate. At low inlet coolant temperature the water/propylene glycol mixture showed higher heat transfer rate when compared with nanofluid coolant. However at higher operating temperature and higher coolant flow rate, 0.3 vol. % of TiO2 nanofluid enhances the heat transfer rate by 8.5% when compared to base fluids.

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

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.'

Supercritical carbon dioxide for textile applications and recent developments

Science.gov (United States)

Eren, H. A.; Avinc, O.; Eren, S.

2017-10-01

In textile industry, supercritical carbon dioxide (scCO2), possessing liquid-like densities, mostly find an application on textile dyeing processes such as providing hydrophobic dyes an advantage on dissolving. Their gas-like low viscosities and diffusion properties can result in shorter dyeing periods in comparison with the conventional water dyeing process. Supercritical carbon dioxide dyeing is an anhydrous dyeing and this process comprises the usage of less energy and chemicals when compared to conventional water dyeing processes leading to a potential of up to 50% lower operation costs. The advantages of supercritical carbon dioxide dyeing method especially on synthetic fiber fabrics hearten leading textile companies to alter their dyeing method to this privileged waterless dyeing technology. Supercritical carbon dioxide (scCO2) waterless dyeing is widely known and applied green method for sustainable and eco-friendly textile industry. However, not only the dyeing but also scouring, desizing and different finishing applications take the advantage of supercritical carbon dioxide (scCO2). In this review, not only the principle, advantages and disadvantages of dyeing in supercritical carbon dioxide but also recent developments of scCO2 usage in different textile processing steps such as scouring, desizing and finishing are explained and commercial developments are stated and summed up.

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)

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.

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)

Heat transfer and fluid flow aspects of fuel--coolant interactions

International Nuclear Information System (INIS)

Corradini, M.L.

1978-09-01

A major portion of the safety analysis effort for the LMFBR is involved in assessing the consequences of a Hypothetical Core Disruptive Accident (HCDA). The thermal interaction of the hot fuel and the sodium coolant during the HCDA is investigated in two areas. A postulated loss of flow transient may produce a two-phase fuel at high pressures. The thermal interaction phenomena between fuel and coolant as the fuel is ejected into the upper plenum are investigated. A postulated transient overpower accident may produce molten fuel being released into sodium coolant in the core region. An energetic coolant vapor explosion for these reactor materials does not seem likely. However, experiments using other materials (e.g., Freon/water, tin/water) have demonstrated the possibility of this phenomenon

Introduction to supercritical fluids a spreadsheet-based approach

CERN Document Server

Smith, Richard; Peters, Cor

2013-01-01

This text provides an introduction to supercritical fluids with easy-to-use Excel spreadsheets suitable for both specialized-discipline (chemistry or chemical engineering student) and mixed-discipline (engineering/economic student) classes. Each chapter contains worked examples, tip boxes and end-of-the-chapter problems and projects. Part I covers web-based chemical information resources, applications and simplified theory presented in a way that allows students of all disciplines to delve into the properties of supercritical fluids and to design energy, extraction and materials formation systems for real-world processes that use supercritical water or supercritical carbon dioxide. Part II takes a practical approach and addresses the thermodynamic framework, equations of state, fluid phase equilibria, heat and mass transfer, chemical equilibria and reaction kinetics of supercritical fluids. Spreadsheets are arranged as Visual Basic for Applications (VBA) functions and macros that are completely (source code) ...

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)

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)

Thermal hydraulic studies for passive heat transport systems relevant to advanced reactors

International Nuclear Information System (INIS)

Vijayan, P.K.; Sharma, M.; Borgohain, A.; Srivastava, A.K.; Pilkhwal, D.S.; Maheshwari, N.K.

2014-01-01

Nuclear is the only non-green house gas generating power source that can replace fossil fuels and can be commercially deployed in large scale. However, the enormous developmental efforts and safety upgrades during the past six decades have somewhat eroded the economic competitiveness of water-cooled reactors which form the mainstay of the current nuclear power programme. Further, the introduction of the supercritical Rankine cycle and the gas turbine based advanced fuel cycles have enhanced the efficiency of fossil fired power plants (FPP) thereby reducing its greenhouse gas emissions. The ongoing development of ultra-supercritical and advanced ultra-supercritical turbines aims to further reduce the greenhouse gas emissions and economic competitiveness of FPPs. In the backdrop of these developments, the nuclear industry also initiated development of advanced nuclear power plants (NPP) with improved efficiency, sustainability and enhanced safety as the main goals. A review of the advanced reactor concepts being investigated currently reveals that excepting the SCWR, all other concepts use coolants other than water. The coolants used are lead, lead bismuth eutectic, liquid sodium, molten salts, helium and supercritical water. Besides, some of these are employing passive systems to transport heat from the core under normal operating conditions. In view of this, a study is in progress at BARC to examine the performance of simple passive systems using SC CO 2 , SCW, LBE and molten salts as the coolant. This paper deals with some of the recent results of these studies. The study focuses on the steady state, transient and stability behaviour of the passive systems with these coolants. (author)

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.

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.

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.

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)

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.

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)

Thermal performance and efficiency of supercritical nuclear reactors

International Nuclear Information System (INIS)

Romney Duffey; Tracy Zhou; Hussam Khartabil

2009-01-01

The paper reviews the major advances and innovative aspects of the thermal performance of recent concepts for super-critical water-cooled nuclear reactors (SCWR). The concepts are based on the extensive experience in the thermal power industry with super and ultra-supercritical boilers and turbines. The challenges and goals of increased efficiency, reduced cost, enhanced safety and co-generation have been pursued over the last ten years, and have resulted both in viable concepts and a vibrant defined R and D effort. The supercritical concept has wide acceptance among industry, as it reflects standard engineering practices and current thermal plant technology that is being already deployed. The SCWR concept represents a continuous development of water-cooled reactor technology, which utilizes the best and latest advances made in the thermal power industry. (author)

On line monitoring of temperatures of coolant channels by thermal imaging in a laboratory set-up fabricated for the detection of leakage of coolants

Energy Technology Data Exchange (ETDEWEB)

Mukherjee, S; Ghosh, J K [Bhabha Atomic Research Centre, Bombay (India). Radiometallurgy Div.; Patel, R J [Bhabha Atomic Research Centre, Mumbai (India). Refuelling Technology Division

1994-12-31

Leakage from coolant channels in Pressurised Heavy Water Reactors (PHWR) increases the temperatures of the faulty channels. Measurement of temperatures of the coolant channels is, therefore, one way to detect the leaking channel. Thermal imaging technique offers a unique means for this detection providing a fast, non-contact, on-line measurement. An experiment was carried out for the detection of leakage of coolants through the seal plugs of the coolant channels in PHWR using an experimental setup under the simulated conditions of temperature and pressure of the coolant channels inside the reactor and using an infrared imaging system. The experimental details and the observations have been presented. 7 figs.

On line monitoring of temperatures of coolant channels by thermal imaging in a laboratory set-up fabricated for the detection of leakage of coolants

International Nuclear Information System (INIS)

Mukherjee, S.; Ghosh, J.K.; Patel, R.J.

1994-01-01

Leakage from coolant channels in Pressurised Heavy Water Reactors (PHWR) increases the temperatures of the faulty channels. Measurement of temperatures of the coolant channels is, therefore, one way to detect the leaking channel. Thermal imaging technique offers a unique means for this detection providing a fast, non-contact, on-line measurement. An experiment was carried out for the detection of leakage of coolants through the seal plugs of the coolant channels in PHWR using an experimental setup under the simulated conditions of temperature and pressure of the coolant channels inside the reactor and using an infrared imaging system. The experimental details and the observations have been presented. 7 figs

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.

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

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

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...

Fatigue management considering LWR coolant environments

International Nuclear Information System (INIS)

Park, Heung Bae; Jin, Tae eun

2000-01-01

Design fatigue curve for structural material in the ASME Boiler and Pressure Vessel Code do not explicitly address the effects of reactor coolant environments on fatigue life. Environmentally assisted cracking (EAC) of low-alloy steels in light water reactor (LWR) coolant environments has been a concern ever since the early 1970's. And, recent fatigue test data indicate a significant decrease in fatigue lives of carbon steels, low-alloy steels and austenitic stainless steels in LWR coolant environments. For these reasons, fatigue of major components has been identified as a technical issue remaining to be resolved for life management and license renewal of nuclear power plants. In the present paper, results of recent investigations by many organizations are reviewed to provide technical justification to support the development of utility approach regarding the management of fatigue considering LWR coolant environments for the purpose of life management and license renewal of nuclear power plants. (author)

Primary Coolant pH Control for Soluble Boron-Free PWRs

International Nuclear Information System (INIS)

Cheon, Yang Ho; Lee, Nam Yeong; Park, Byeong Ho; Park, Seong Chan; Kim, Eun Kee

2015-01-01

These should be considered when evaluating and designing the operating pH program for nuclear power plants. This paper discusses the advanced water chemistry strategies to keep pace with the recent global trends related to pH control in the primary water system for soluble boron pressurized water reactor (PWR) plants. Finally, the objective of this work is to study primary coolant pH control for soluble boron-free PWR plants. This paper reviewed the advanced water chemistry strategies to keep pace with the recent global trends related to pH control in the primary water chemistry system for soluble boron PWR plants. The new chemistry trend for the primary coolant is towards adaption of the constant and elevated chemistry. Finally, this work studied primary coolant pH control for soluble boron-free PWR plants. The ammonia-based water chemistry related to pH control for boron-free PWR plants was discussed. The ammonia-based water chemistry is not recommended to avoid fluctuation of the pH value by ammonia radiolysis and to reduce C-14 production in reactor coolant from reaction with dissolved nitrogen. Also, the potassium-based water chemistry related to pH control for boron-free PWR plants was discussed. KOH has a potential as an alternative pH control agent for soluble boron-free PWR plants. The potassium-based water chemistry related to pH control is recommended for boron-free operation as follows

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

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.

Supercritical heat transfer phenomena in nuclear system

International Nuclear Information System (INIS)

Seo, Kyoung Woo; Kim, Moo Hwan; Anderson, Mark H.; Corradini, Michael L.

2005-01-01

A supercritical water (SCW) power cycle has been considered as one of the viable candidates for advanced fission reactor designs. However, the dramatic variation of thermo-physical properties with a modest change of temperature near the pseudo-critical point make existing heat transfer correlations such as the Dittus-Boelter correlation not suitably accurate to calculate the heat transfer in supercritical fluid. Several other correlations have also been suggested but none of them are able to predict the heat transfer over a parameter range, needed for reactor thermal-hydraulics simulation and design. This has prompted additional research to understand the characteristic of supercritical fluid heat transfer

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

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.

Fusion-reactor blanket and coolant material compatibility

International Nuclear Information System (INIS)

Jeppson, D.W.; Keough, R.F.

1981-01-01

Fusion reactor blanket and coolant compatibility tests are being conducted to aid in the selection and design of safe blanket and coolant systems for future fusion reactors. Results of scoping compatibility tests to date are reported for blanket material and water interactions at near operating temperatures. These tests indicate the quantitative hydrogen release, the maximum temperature and pressures produced and the rates of interactions for selected blanket materials

Iron crud supply device to reactor coolant

International Nuclear Information System (INIS)

Baba, Takao.

1993-01-01

In a device for supplying iron cruds into reactor coolants in a BWR type power plant, a system in which feed water containing iron cruds is supplied to the reactor coolants after once passing through an ion exchange resin is disposed. As a result, iron cruds having characteristics similar with those of naturally occurring iron cruds in the plant are obtained and they react with ionic radioactivity, to form composite oxides. Then, iron cruds having high performance of being secured to the surface of a fuel cladding tube can be supplied to the reactor coolants, thereby enabling to greatly reduce the density of reactor water ionic radioactivity. In its turn, dose rate on the surface of pipelines can be reduced, thereby enabling to reduce operators' radiation exposure dose in the plant. Further, contamination of a condensate desalting device due to iron cruds can be prevented, and further, the density of the iron cruds supplied can easily be controlled. (N.H.)

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.

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.

Biodiesel from waste cooking oil via base-catalytic and supercritical methanol transesterification

International Nuclear Information System (INIS)

Demirbas, Ayhan

2009-01-01

In this study, waste cooking oil has subjected to transesterification reaction by potassium hydroxide (KOH) catalytic and supercritical methanol methods obtaining for biodiesel. In catalyzed methods, the presence of water has negative effects on the yields of methyl esters. In the catalytic transesterification free fatty acids and water always produce negative effects since the presence of free fatty acids and water causes soap formation, consumes catalyst, and reduces catalyst effectiveness. Free fatty acids in the waste cooking oil are transesterified simultaneously in supercritical methanol method. Since waste cooking oil contains water and free fatty acids, supercritical transesterification offers great advantage to eliminate the pre-treatment and operating costs. The effects of methanol/waste cooking oils ratio, potassium hydroxide concentration and temperature on the biodiesel conversion were investigated

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

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)

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)

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

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

Reactivity of dolomite in water-saturated supercritical carbon dioxide: Significance for carbon capture and storage and for enhanced oil and gas recovery

International Nuclear Information System (INIS)

Wang Xiuyu; Alvarado, Vladimir; Swoboda-Colberg, Norbert; Kaszuba, John P.

2013-01-01

Highlights: ► Dolomite reactivity with wet and dry supercritical CO 2 were evaluated. ► Dolomite does not react with dry CO 2 . ► H 2 O-saturated supercritical CO 2 dissolves dolomite and precipitates carbonate mineral. ► Temperature/reaction time control morphology and extent of carbonate mineralization. ► Reaction with wet CO 2 may impact trapping, caprock integrity, and CCS/EOR injectivity. - Abstract: Carbon dioxide injection in porous reservoirs is the basis for carbon capture and storage, enhanced oil and gas recovery. Injected carbon dioxide is stored at multiple scales in porous media, from the pore-level as a residual phase to large scales as macroscopic accumulations by the injection site, under the caprock and at reservoir internal capillary pressure barriers. These carbon dioxide saturation zones create regions across which the full spectrum of mutual CO 2 –H 2 O solubility may occur. Most studies assume that geochemical reaction is restricted to rocks and carbon dioxide-saturated formation waters, but this paradigm ignores injection of anhydrous carbon dioxide against brine and water-alternating-gas flooding for enhanced oil recovery. A series of laboratory experiments was performed to evaluate the reactivity of the common reservoir mineral dolomite with water-saturated supercritical carbon dioxide. Experiments were conducted at reservoir conditions (55 and 110 °C, 25 MPa) and elevated temperature (220 °C, 25 MPa) for approximately 96 and 164 h (4 and 7 days). Dolomite dissolves and new carbonate mineral precipitates by reaction with water-saturated supercritical carbon dioxide. Dolomite does not react with anhydrous supercritical carbon dioxide. Temperature and reaction time control the composition, morphology, and extent of formation of new carbonate minerals. Mineral dissolution and re-precipitation due to reaction with water-saturated carbon dioxide may affect the contact line between phases, the carbon dioxide contact angle, and the

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.

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

Reactor coolant pump seals: improving their performance

International Nuclear Information System (INIS)

Pothier, N.E.; Metcalfe, R.

1986-06-01

Large CANDU plants are benefitting from transient-resistant four-year reliable reactor coolant pump seal lifetimes, a direct result of AECL's 20-year comprehensive seal improvement program involving R and D staff, manufacturers, and plant designers and operators. An overview of this program is presented, which covers seal modification design, testing, post-service examination, specialized maintenance and quality control. The relevancy of this technology to Light Water Reactor Coolant Pump Seals is also discussed

Condition monitoring of main coolant pumps, Dhruva

International Nuclear Information System (INIS)

Prasad, V.; Satheesh, C.; Acharya, V.N.; Tikku, A.C.; Mishra, S.K.

2002-01-01

Full text: Dhruva is a 100 MW research reactor with natural uranium fuel, heavy water as moderator and primary coolant. Three Centrifugal pumps circulate the primary coolant across the core and the heat exchangers. Each pump is coupled to a flywheel (FW) assembly in order to meet operational safety requirements. All the 3 main coolant pump (MCP) sets are required to operate during operation of the reactor. The pump-sets are in operation since the year 1984 and have logged more than 1,00,000 hrs. Frequent breakdowns of its FW bearings were experienced during initial years of operation. Condition monitoring of these pumps, largely on vibration based parameters, was initiated on regular basis. Break-downs of main coolant pumps reduced considerably due to the fair accurate predictions of incipient break-downs and timely maintenance efforts. An effort is made in this paper to share the experience

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)

SUPERCRITICAL FLUID EXTRACTION OF POLYCYCLIC AROMATIC HYDROCARBON MIXTURES FROM CONTAMINATED SOILS

Science.gov (United States)

Highly contaminated (with PAHs) topsoils were extracted with supercritical CO2 to determine the feasibility and mechanism of supercritical fluid extraction (SFE). Effect of SCF density, temperature, cosolvent type and amount, and of slurrying the soil with water were ...

Flow rate control systems for coolants for BWR type reactors

International Nuclear Information System (INIS)

Igarashi, Yoko; Kato, Naoyoshi.

1981-01-01

Purpose: To increase spontaneous recycling flow rate of coolants in BWR type reactors when the water level in the reactor decreases, by communicating a downcomer with a lower plenum. Constitution: An opening is provided to the back plate disposed at the lower end of a reactor core shroud for communicating a downcomer with a lower plenum, and an ON-OFF valve actuated by an operation rod is provided to the opening. When abnormal water level or pressure in the reactor is detected by a level metal or pressure meter, the operation rod is driven to open the ON-OFF valve, whereby coolants fed from a jet pump partially flows through the opening to increase the spontaneous recycling flow rate of the coolants. This can increase the spontaneous recycling flow rate of the coolants upon spontaneous recycling operation, thereby maintaining the reactor safety and the fuel soundness. (Moriyama, K.)

Dynamic response of INTOR/NET blankets after coolant tube rupture

International Nuclear Information System (INIS)

Klippel, H.T.

1985-01-01

The dynamic response of different water-cooled liquid Li 17 Pb 83 breeder blanket modules has been calculated to study the potential of these modules in case of coolant tube rupture. Numerical calculations with the code PISCES have been carried out taking into account the fluid-structure interaction and the elasto-plastic behaviour of the structural material. The results show that for inert coolant characteristics the proposed conceptual designs for NET and INTOR have sufficient resistance against coolant tube rupture but when taking into account energy release due to chemical reaction of water with LiPb-alloy up to doubling of the wall thickness has to be envisaged to guarantee structural reliability. (orig.)

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

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)

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.

Recent developments in coolant systems for Indus Accelerator Complex at RRCAT, Indore

International Nuclear Information System (INIS)

Nanda, Dipankar; Tiwari, Bablu; Pandey, R.M.

2015-01-01

Scarcity of fresh water forces mankind to explore other possible water sources that can meet the increasing demand of coolants in industries, R and D sectors and other establishments where water is used as coolant. It also becomes a challenge for water chemist to control water chemistry to keep the equipments/devices intact during its operation using water as coolant. Deionised (DI) and soft water have been used as coolants for Indus Accelerator Complex, RRCAT, Indore. DI water is produced and its quality is maintained either by conventional ion exchange method or a hybrid method of membrane separation and ion exchange technique. This requires handling of corrosive chemicals, manpower, space for plant installation, and a long array of water treatment units. CSL has implemented the idea of rain water harvesting to produce DI water after systematic studies in laboratory. The concerning issues are reduced to almost one-fourth by using rain water to produce DI water. The harvesting system has been in use for last three years. Heat is dissipated into air by evaporation of soft water in cooling tower. Requirement of soft water makeup has been estimated to be about 40,000 ltrs. / day (max.) if the machine is operated at its designed specifications. Non-availability of soft water (which circulates in open loop) may lead to shut down like situation and looking for alternate source becomes quite essential. Laboratory studies (water analysis and treatment) on sewage water (available 1,00,000 ltrs/day) from RRCAT colony as a possible source of producing soft water show promising result. (author)

High Density Thermal Energy Storage with Supercritical Fluids

Science.gov (United States)

Ganapathi, Gani B.; Wirz, Richard

2012-01-01

A novel approach to storing thermal energy with supercritical fluids is being investigated, which if successful, promises to transform the way thermal energy is captured and utilized. The use of supercritical fluids allows cost-affordable high-density storage with a combination of latent heat and sensible heat in the two-phase as well as the supercritical state. This technology will enhance penetration of several thermal power generation applications and high temperature water for commercial use if the overall cost of the technology can be demonstrated to be lower than the current state-of-the-art molten salt using sodium nitrate and potassium nitrate eutectic mixtures.

Review on research of small break loss of coolant accident

International Nuclear Information System (INIS)

Bo Jinhai; Wang Fei

1998-01-01

The Small Break Loss of Coolant Accident (SBLOCA) and its research art-of -work are reviewed. A typical SBLOCA process in Pressurized Water Reactor (PWR) and Boiling Water Reactor (BWR) and the influence of break size, break location and reactor coolant pump on the process are described. The existing papers are classified in two categories: experimental and numerical modeling, with the primary experimental apparatuses in the world listed and the research works on SBLOCA summarized

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)

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.

Experimental analysis on a novel solar collector system achieved by supercritical CO2 natural convection

International Nuclear Information System (INIS)

Chen, Lin; Zhang, Xin-Rong

2014-01-01

Highlights: • Supercritical CO 2 flow is proposed for natural circulation solar water heater system. • Experimental system established and consists of supercritical fluid high pressure side and water side. • Stable supercritical CO 2 natural convective flow is well induced and water heating process achieved. • Seasonal solar collector system efficiency above 60% achieved and optimization discussed. - Abstract: Solar collector has become a hot topic both in scientific research and engineering applications. Among the various applications, the hot water supply demand accounts for a large part of social energy consumption and has become one promising field. The present study deals with a novel solar thermal conversion and water heater system achieved by supercritical CO 2 natural circulation. Experimental systems are established and tested in Zhejiang Province (around N 30.0°, E 120.6°) of southeast China. The current system is designed to operate in the supercritical region, thus the system can be compactly made and achieve smooth high rate natural convective flow. During the tests, supercritical CO 2 pipe flow with Reynolds number higher than 6700 is found. The CO 2 fluid temperature in the heat exchanger can be as high as 80 °C and a stable supply of hot water above 45 °C is achieved. In the seasonal tests, relative high collector efficiency generally above 60.0% is obtained. Thermal and performance analysis is carried out with the experiment data. Comparisons between the present system and previous solar water heaters are also made in this paper

Feeding and purge systems of coolant primary circuit and coolant secondary circuit control of the I sup(123) target

International Nuclear Information System (INIS)

Almeida, G.L. de.

1986-01-01

The Radiation Protection Service of IEN (Brazilian-CNEN) detected three faults in sup(123)I target cooling system during operation process for producing sup(123)I: a) non hermetic vessel containing contaminated water from primary coolant circuit; possibility of increasing radioactivity in the vessel due to accumulation of contaminators in cooling water and; situation in region used for personnels to arrange and adjust equipments in nuclear physics area, to carried out maintenance of cyclotron and target coupling in irradiation room. The primary circuit was changed by secondary circuit for target coolant circulating through coil of tank, which receive weater from secondary circuit. This solution solved the three problems simultaneously. (M.C.K.)

Coolant controls of a PEM fuel cell system

Science.gov (United States)

Ahn, Jong-Woo; Choe, Song-Yul

When operating the polymer electrolyte membrane (PEM) fuel cell stack, temperatures in the stack continuously change as the load current varies. The temperature directly affects the rate of chemical reactions and transport of water and reactants. Elevated temperature increases the mobility of water vapor, which reduces the ohmic over-potential in the membrane and eases removal of water produced. Adversely, the high temperature might impose thermal stress on the membrane and cathode catalyst and cause degradation. Conversely, excessive supply of coolants lowers the temperature in the stack and reduces the rate of the chemical reactions and water activity. Corresponding parasitic power dissipated at the electrical coolant pump increases and overall efficiency of the power system drops. Therefore, proper design of a control for the coolant flow plays an important role in ensuring highly reliable and efficient operations of the fuel cell system. Herein, we propose a new temperature control strategy based on a thermal circuit. The proposed thermal circuit consists of a bypass valve, a radiator with a fan, a reservoir and a coolant pump, while a blower and inlet and outlet manifolds are components of the air supply system. Classic proportional and integral (PI) controllers and a state feedback control for the thermal circuit were used in the design. In addition, the heat source term, which is dependent upon the load current, was feed-forwarded to the closed loop and the temperature effects on the air flow rate were minimized. The dynamics and performance of the designed controllers were evaluated and analyzed by computer simulations using developed dynamic fuel cell system models, where a multi-step current and an experimental current profile measured at the federal urban driving schedule (FUDS) were applied. The results show that the proposed control strategy cannot only suppress a temperature rise in the catalyst layer and prevent oxygen starvation, but also reduce the

Coolant clean up system in nuclear reactor

International Nuclear Information System (INIS)

Tajima, Fumio; Iwami, Hiroshi.

1981-01-01

Purpose: To decrease the amount of main steams and improve the plant heat efficiency by the use of condensated water as coolants for not-regenerative heat exchangers in a coolant clean up system of a nuclear reactor. Constitution: In a coolant clean up system of a nuclear reactor, a portion of condensates is transferred to the shell of a non-regenerative heat exchanger by way of a condensate pump for non-regenerative heat exchanger through a branched pipeway provided to the outlet of a condensate desalter for using the condensates as the coolants for the shell of the heat exchanger and the condensates are then returned to the inlet of a feedwater heater after the heat exchange. The branched flow rate of the condensates is controlled by the flow rate control valve mounted in the pipeway. Condensates passed through the heat exchanger and the condensates not passed through the heat exchanger are mixed and heated in a heater and then fed to the nuclear reactor. In a case where no feedwater is necessary to the nuclear reactor such as upon shutdown of the reactor, the condensates are returned by way of feedwater bypass pipeway to the condensator. By the use of the condensates as the coolants for the heat exchanger, the main steam loss can be decreased and the thermal load for the auxiliary coolant facility can be reduced. (Kawakami, Y.)

A Technique for Dynamic Corrosion Testing in Supercritical CO2

International Nuclear Information System (INIS)

Loewen, Eric P.; Davis, Cliff B.; Shropshire, David E.; Weaver, Kevan

2004-01-01

An experimental apparatus for the investigation of the flow-assisted corrosion of potential fuel cladding and structural materials to be used on a fast reactor cooled by supercritical carbon dioxide has been designed. This experimental project is part of a larger research at the Department of Energy being lead by the Idaho National Engineering and Environmental Laboratory (INEEL) to investigate the suitability of supercritical carbon dioxide for cooling a fast reactor designed to produce low-cost electricity as well as for actinide burning. The INEEL once-through corrosion apparatus consists of two syringe pumps, a pre-heat furnace, a 1.3 meter long heated corrosion test section, and a gas measuring system. The gas flow rates, heat input, and operating pressure can be adjusted so that a controlled coolant flow rate, temperature, and oxygen potential are created within each of six test sections. The corrosion cell will test tubing that is commercially available in the U.S. and specialty coupons to temperatures up to 600 deg. C and a pressure of 20 MPa. The ATHENA computer code was used to estimate the fluid conditions in each of the six test sections during normal operation. (authors)

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.

Method of injecting iron ion into reactor coolant

International Nuclear Information System (INIS)

Ito, Kazuyuki; Sawa, Toshio; Nishino, Yoshitaka; Adachi, Tetsuro; Osumi, Katsumi.

1988-01-01

Purpose: To form iron ions stably and inject them into nuclear reactor coolants with no substantial degradation of the severe water quality conditions for reactor coolants. Method: Iron ions are formed by spontaneous corrosion of iron type materials and electroconductivity is increased with the iron ions. Then, the liquids are introduced into an electrolysis vessel using iron type material as electrodes and, thereafter, incorporation of newly added ions other than the iron ions are prevented by supplying electric current. Further, by retaining the iron type material in the packing vessel by the magnetic force therein, only the iron ions are flow out substantially from the packing vessel while preventing the discharge of iron type materials per se or solid corrosion products and then introduced into the electrolysis vessel. Powdery or granular pure iron or carbon steel is used as the iron type material. Thus, iron ions and hydroxides thereof can be injected into coolants by using reactor water at low electroconductivity and incapable of electrolysis. (Kamimura, M.)

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

Using muonium to probe the kinetics of the reaction between the H atom and OH"- in superheated water

International Nuclear Information System (INIS)

Ghandi, K.; Alcorn, C.; Brodovitch, J.-C.; Driedger, E.; Mozafari, M.; Percival, P.W.

2011-01-01

Operation of a supercritical-water-cooled nuclear reactor requires knowledge of water chemistry over a wide range of conditions. The considerable knowledge gap for conditions above the operating temperature of current-generation CANDU reactors is the target of this study. Since the H atom is difficult to probe at the desired temperatures and pressures, muonium is used as an alternative. In the current CANDU reactors, coolant pH is controlled using LiOH. We are studying how the rate constants for the reaction of muonium with LiOH and NaOH change with temperature and pressure. (author)

Fuel-Coolant Interactions - some Basic Studies at the UKAEA Culham Laboratory

International Nuclear Information System (INIS)

Reynolds, J.A.; Dullforce, T.A.; Peckover, R.S.; Vaughan, G.J.

1976-01-01

In a hypothetical fault sequence important effects of fuel-coolant interactions include voiding and dispersion of core debris as well as the pressure damage usually discussed. The development of the fuel-coolant interaction probably depends on any pre-mixing Weber break-up that may occur, and is therefore a function of the way the fuel and coolant come together. Four contact modes are identified: jetting, shock tube, drops and static, and Culham's experiments have been mainly concerned with simulating the falling drop mode by using molten tin in water. It was observed that the fuel-coolant interaction is a short series of violent coolant oscillations centred at a localized position on the drop, generating a spray of submillimeter sized debris. The interaction started spontaneously at a specific time after the drop first contacted the water. There was a definite limited fuel-coolant interaction zone on a plot of initial coolant temperature versus initial fuel temperature outside which interactions never occurred. The. interaction time was a function of the initial temperatures. Theoretical scaling formulae are given which describe the fuel-coolant interaction zone and dwell time. Bounds of fuel and coolant temperature below which fuel-coolant interactions do not occur are explained by freezing. Upper bounds of fuel and coolant temperatures above which there were no fuel-coolant interactions are interpreted in terms of heat transfer through vapour films of various thicknesses. In conclusion: We have considered the effects of fuel-coolant interactions in a hypothetical fault sequence, emphasising that debris and vapour production as well as the pressure pulse can be important factors. The fuel-coolant interaction has been classified into types, according to possible modes of mixing in the fault sequence. Culham has been studying one type, the self-triggering of falling drops, by simulant experiments. It is found that there is a definite zone of interaction on a plot

Supercritical Regeneration of an Activated Carbon Fiber Exhausted with Phenol

OpenAIRE

M. Jesus Sanchez-Montero; Jennifer Pelaz; Nicolas Martin-Sanchez; Carmen Izquierdo; Francisco Salvador

2018-01-01

The properties of supercritical CO2 (SCCO2) and supercritical water (SCW) turn them into fluids with a great ability to remove organic adsorbates retained on solids. These properties were used herein to regenerate an activated carbon fiber (ACF) saturated with a pollutant usually contained in wastewater and drinking water, phenol. Severe regeneration conditions, up to 225 bar and 400 °C, had to be employed in SCCO2 regeneration to break the strong interaction established between phenol and th...

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

Supercritical water oxidation of dioxins and furans in waste incinerator fly ash, sewage sludge and industrial soil.

Science.gov (United States)