n-Heptane cool flame chemistry: Unraveling intermediate species measured in a stirred reactor and motored engine

KAUST Repository

Wang, Zhandong

2017-10-03

This work identifies classes of cool flame intermediates from n-heptane low-temperature oxidation in a jet-stirred reactor (JSR) and a motored cooperative fuel research (CFR) engine. The sampled species from the JSR oxidation of a mixture of n-heptane/O2/Ar (0.01/0.11/0.88) were analyzed using a synchrotron vacuum ultraviolet radiation photoionization (SVUV-PI) time-of-flight molecular-beam mass spectrometer (MBMS) and an atmospheric pressure chemical ionization (APCI) Orbitrap mass spectrometer (OTMS). The OTMS was also used to analyze the sampled species from a CFR engine exhaust. Approximately 70 intermediates were detected by the SVUV-PI-MBMS, and their assigned molecular formulae are in good agreement with those detected by the APCI-OTMS, which has ultra-high mass resolving power and provides an accurate elemental C/H/O composition of the intermediate species. Furthermore, the results show that the species formed during the partial oxidation of n-heptane in the CFR engine are very similar to those produced in an ideal reactor, i.e., a JSR.The products can be classified by species with molecular formulae of C7H14Ox (x = 0–5), C7H12Ox (x = 0–4), C7H10Ox (x = 0–4), CnH2n (n = 2–6), CnH2n−2 (n = 4–6), CnH2n+2O (n = 1–4), CnH2nO (n = 1–6), CnH2n−2O (n = 2–6), CnH2n−4O (n = 4–6), CnH2n+2O2 (n = 0–4, 7), CnH2nO2 (n = 1–6), CnH2n−2O2 (n = 2–6), CnH2n−4O2 (n = 4–6), and CnH2nO3 (n = 3–6). The identified intermediate species include alkenes, dienes, aldehyde/keto compounds, olefinic aldehyde/keto compounds, diones, cyclic ethers, peroxides, acids, and alcohols/ethers. Reaction pathways forming these intermediates are proposed and discussed herein. These experimental results are important in the development of more accurate kinetic models for n-heptane and longer-chain alkanes.

Chemical conditions in the repository for low- and intermediate-level reactor waste

International Nuclear Information System (INIS)

Snellman, M.; Uotila, H.

1984-01-01

The chemical conditions in the proposed repositories for low- and intermediate-level reactor waste at Haestholmen (IVO) and Olkiluoto (TVO) have been discussed with respect to materials introduced into the repository, their possible long-term changes and interaction with groundwater flowing into the repository. The main possible groundwater-rock interactions have been discussed, as well as the role of micro-organisms, organic acids and colloids in the estimation of the barrier integrity. Experimental and theoretical studies have been performed on the basis of the natural groundwater compositions expected at the repository sites. Main emphasis is put on the chemical parameters which might influence the integrity of the different barriers in the repository as well as on the parameters which might effect the release and transport of radionuclides from the repository

Nuclear reactor control column

International Nuclear Information System (INIS)

Bachovchin, D.M.

1982-01-01

The nuclear reactor control column comprises a column disposed within the nuclear reactor core having a variable cross-section hollow channel and containing balls whose vertical location is determined by the flow of the reactor coolant through the column. The control column is divided into three basic sections wherein each of the sections has a different cross-sectional area. The uppermost section of the control column has the greatest crosssectional area, the intermediate section of the control column has the smallest cross-sectional area, and the lowermost section of the control column has the intermediate cross-sectional area. In this manner, the area of the uppermost section can be established such that when the reactor coolant is flowing under normal conditions therethrough, the absorber balls will be lifted and suspended in a fluidized bed manner in the upper section. However, when the reactor coolant flow falls below a predetermined value, the absorber balls will fall through the intermediate section and into the lowermost section, thereby reducing the reactivity of the reactor core and shutting down the reactor

Intermediate-energy neutron beam for NCT at MURR

International Nuclear Information System (INIS)

Brugger, R.M.; Less, T.J.; Passmore, G.G.

1986-01-01

The University of Missouri Research Reactor (MURR) is one of the high-flux reactors in the USA and it can be used to produce an intense beam of intermediate-energy neutrons for neutron capture therapy. Two methods are being evaluated at MURR to produce such a beam. The first uses a moderator of Al 2 O 3 replacing part of the graphite and water on one side of the core of the reactor to produce a source of predominantly intermediate-energy neutrons. The second method is a filter of 238 U between the core and the patient position to pass only intermediate-energy neutrons. The results of these evaluations are presented in this paper along with an outline of the other resources at the University of Missouri-Columbia that are available to support an NCT program. 4 references, 7 figures, 1 table

An evaluation of the fluid-elastic instability for Intermediate Heat Exchanger of Prototype Sodium-cooled fast Reactor

International Nuclear Information System (INIS)

Cho, Jaehun; Kim, Sungkyun; Koo, Gyeonghoi

2014-01-01

The sodium-cooled fast reactor (SFR) module consists of the vessel, containment vessel, head, rotating plug (RP), upper internal structure (UIS), intermediate heat exchanger (IHX), decay heat exchanger (DHX), primary pump, internal structure, internal components and reactor core. The IHXs transfer heat from the radioactive sodium coolant (primary sodium) in the primary heat transport system to the nonradioactive sodium coolant (secondary sodium) in the intermediate heat transport system. Each sodium flows like Fig. 1. Primary sodium flows inside of tube and secondary sodium flows outside. During transferring heat two sodium to sodium, the fluid-elastic instability is occurred among tube bundle by cross flow. Large amplitude vibration occurred by the fluid-elastic instability is caused such as crack and wear of tube. Thus it is important to decrease the fluid-elastic instability in terms of a safety. The purpose of this paper is to evaluate the fluid-elastic instability for tube bundle in the IHX following ASME code. This paper evaluated the fluid-elastic instability of tube bundle in the SFR IHX. According evaluation results, the fluid-elastic instability of IHX tube bundle is occurred. A installing an additional TSP under the upper tubesheet can decrease a probability of fluid-elastic instability. If a location of an additional TSP does not exceed tube length to become a 750 mm, tube bundle of IHX is safety from the fluid-elastic instability

Preliminary thermal sizing of intermediate heat exchanger for NHDD system

International Nuclear Information System (INIS)

Kim, Chan Soo; Hong, Sung Deok; Kim, Yong Wan; Chang, Jongh Wa

2009-01-01

Nuclear Hydrogen Development and Demonstration (NHDD) system is a Very High Temperature gascooled Reactor (VHTR) coupled with hydrogen production systems. Intermediate heat exchanger transfers heat from the nuclear reactor to the hydrogen production system. This study presented the sensitivity analysis on a preliminary thermal sizing of the intermediate heat exchanger. Printed Circuit Heat Exchanger (PCHE) was selected for the thermal sizing because the printed circuit heat exchanger has the largest compactness among the heat exchanger types. The analysis was performed to estimate the effect of key parameters including the operating condition of the intermediate system, the geometrical factors of the PCHE, and the working fluid of the intermediate system.

Control of reactor coolant flow path during reactor decay heat removal

Science.gov (United States)

Hunsbedt, Anstein N.

1988-01-01

An improved reactor vessel auxiliary cooling system for a sodium cooled nuclear reactor is disclosed. The sodium cooled nuclear reactor is of the type having a reactor vessel liner separating the reactor hot pool on the upstream side of an intermediate heat exchanger and the reactor cold pool on the downstream side of the intermediate heat exchanger. The improvement includes a flow path across the reactor vessel liner flow gap which dissipates core heat across the reactor vessel and containment vessel responsive to a casualty including the loss of normal heat removal paths and associated shutdown of the main coolant liquid sodium pumps. In normal operation, the reactor vessel cold pool is inlet to the suction side of coolant liquid sodium pumps, these pumps being of the electromagnetic variety. The pumps discharge through the core into the reactor hot pool and then through an intermediate heat exchanger where the heat generated in the reactor core is discharged. Upon outlet from the heat exchanger, the sodium is returned to the reactor cold pool. The improvement includes placing a jet pump across the reactor vessel liner flow gap, pumping a small flow of liquid sodium from the lower pressure cold pool into the hot pool. The jet pump has a small high pressure driving stream diverted from the high pressure side of the reactor pumps. During normal operation, the jet pumps supplement the normal reactor pressure differential from the lower pressure cold pool to the hot pool. Upon the occurrence of a casualty involving loss of coolant pump pressure, and immediate cooling circuit is established by the back flow of sodium through the jet pumps from the reactor vessel hot pool to the reactor vessel cold pool. The cooling circuit includes flow into the reactor vessel liner flow gap immediate the reactor vessel wall and containment vessel where optimum and immediate discharge of residual reactor heat occurs.

Adoption of nitrogen power conversion system for small scale ultra-long cycle fast reactor eliminating intermediate sodium loop

International Nuclear Information System (INIS)

Seo, Seok Bin; Seo, Han; Bang, In Cheol

2016-01-01

Highlights: • N 2 power conversion system for both safety and thermal performance aspects. • Sensitivity studies of several controlled parameters on N 2 power conversion system. • The elimination of the intermediate loop increased the cycle thermal efficiency. • The elimination of the intermediate loop expects economic advantages. - Abstract: As one of SFRs, the ultra-long cycle fast reactor with a power rating of 100 MW e (UCFR-100) was introduced for a 60-year operation. As an alternative to the traditional steam Rankine cycle for the power conversion system, gas based Brayton cycle has been considered for UCFR-100. Among Supercritical CO 2 (S-CO 2 ), Helium (He), Nitrogen (N 2 ) as candidates for the power conversion system for UCFR-100, an N 2 power conversion system was chosen considering both safety and thermal performance aspects. The elimination of the intermediate sodium loop could be achieved due to the safety and stable characteristics of nitrogen working fluid. In this paper, sensitivity studies with respect to several controlled parameters on N 2 power conversion system were performed to optimize the system. Furthermore, the elimination of the intermediate loop was evaluated with respect to its impact on the thermodynamic performance and other aspects.

FBR type reactor

International Nuclear Information System (INIS)

Kimura, Kimitaka; Fukuie, Ken; Iijima, Tooru; Shimpo, Masakazu.

1994-01-01

In an FBR type reactor for exchanging fuels by pulling up reactor core upper mechanisms, a connection mechanism is disposed for connecting the top of the reactor core and the lower end of the reactor core upper mechanisms. In addition, a cylindrical body is disposed surrounding the reactor core upper mechanisms, and a support member is disposed to the cylindrical body for supporting an intermediate portion of the reactor core upper mechanisms. Then, the lower end of the reactor core upper mechanisms is connected to the top of the reactor core. Same displacements are caused to both of them upon occurrence of earthquakes and, as a result, it is possible to eliminate mutual horizontal displacement between a control rod guide hole of the reactor core upper mechanisms and a control rod insertion hole of the reactor core. In addition, since the intermediate portion of the reactor core upper mechanisms is supported by the support member disposed to the cylindrical body surrounding the reactor core upper mechanisms, deformation caused to the lower end of the reactor core upper mechanisms is reduced, so that the mutual horizontal displacement with respect to the control rod insertion hole of the reactor core can be reduced. As a result, performance of control rod insertion upon occurrence of the earthquakes is improved, so that reactor shutdown is conducted more reliably to improve reactor safety. (N.H.)

Sustainable thorium nuclear fuel cycles: A comparison of intermediate and fast neutron spectrum systems

International Nuclear Information System (INIS)

Brown, N.R.; Powers, J.J.; Feng, B.; Heidet, F.; Stauff, N.E.; Zhang, G.; Todosow, M.; Worrall, A.; Gehin, J.C.; Kim, T.K.; Taiwo, T.A.

2015-01-01

Highlights: • Comparison of intermediate and fast spectrum thorium-fueled reactors. • Variety of reactor technology options enables self-sustaining thorium fuel cycles. • Fuel cycle analyses indicate similar performance for fast and intermediate systems. • Reproduction factor plays a significant role in breeding and burn-up performance. - Abstract: This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1 eV and 10 5 eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight lattice heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this self-sustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems

Sustainable thorium nuclear fuel cycles: A comparison of intermediate and fast neutron spectrum systems

Energy Technology Data Exchange (ETDEWEB)

Brown, N.R., E-mail: nbrown@bnl.gov [Brookhaven National Laboratory, Upton, NY (United States); Powers, J.J. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Feng, B.; Heidet, F.; Stauff, N.E.; Zhang, G. [Argonne National Laboratory, Argonne, IL (United States); Todosow, M. [Brookhaven National Laboratory, Upton, NY (United States); Worrall, A.; Gehin, J.C. [Oak Ridge National Laboratory, Oak Ridge, TN (United States); Kim, T.K.; Taiwo, T.A. [Argonne National Laboratory, Argonne, IL (United States)

2015-08-15

Highlights: • Comparison of intermediate and fast spectrum thorium-fueled reactors. • Variety of reactor technology options enables self-sustaining thorium fuel cycles. • Fuel cycle analyses indicate similar performance for fast and intermediate systems. • Reproduction factor plays a significant role in breeding and burn-up performance. - Abstract: This paper presents analyses of possible reactor representations of a nuclear fuel cycle with continuous recycling of thorium and produced uranium (mostly U-233) with thorium-only feed. The analysis was performed in the context of a U.S. Department of Energy effort to develop a compendium of informative nuclear fuel cycle performance data. The objective of this paper is to determine whether intermediate spectrum systems, having a majority of fission events occurring with incident neutron energies between 1 eV and 10{sup 5} eV, perform as well as fast spectrum systems in this fuel cycle. The intermediate spectrum options analyzed include tight lattice heavy or light water-cooled reactors, continuously refueled molten salt reactors, and a sodium-cooled reactor with hydride fuel. All options were modeled in reactor physics codes to calculate their lattice physics, spectrum characteristics, and fuel compositions over time. Based on these results, detailed metrics were calculated to compare the fuel cycle performance. These metrics include waste management and resource utilization, and are binned to accommodate uncertainties. The performance of the intermediate systems for this self-sustaining thorium fuel cycle was similar to a representative fast spectrum system. However, the number of fission neutrons emitted per neutron absorbed limits performance in intermediate spectrum systems.

Reactor container

International Nuclear Information System (INIS)

Oyamada, Osamu; Furukawa, Hideyasu; Uozumi, Hiroto.

1979-01-01

Purpose: To lower the position of an intermediate slab within a reactor container and fitting a heat insulating material to the inner wall of said intermediate slab, whereby a space for a control rod exchanging device and thermal stresses of the inner peripheral wall are lowered. Constitution: In the pedestal at the lower part of a reactor pressure vessel there is formed an intermediate slab at a position lower than diaphragm floor slab of the outer periphery of the pedestal thereby to secure a space for providing automatic exchanging device of a control rod driving device. Futhermore, a heat insulating material is fitted to the inner peripheral wall at the upper side of the intermediate slab part, and the temperature gradient in the wall thickness direction at the time of a piping rupture trouble is made gentle, and thermal stresses at the inner peripheral wall are lowered. (Sekiya, K.)

Behavior of the future LHC magnet protection diodes irradiated in a nuclear reactor at 4.6 K with intermediate annealing

International Nuclear Information System (INIS)

Berland, V.; Hagedorn, D.; Gerstenberg, H.

1996-01-01

In the framework of the LHC project at CERN, the effects of radiation on the electrical characteristics of epitaxial diodes for superconducting magnet protection were studied. The diodes were exposed to an irradiation dose up to 50 kGy and a neutron fluence of 10 15 n/cm 2 with intermediate thermal annealing each 10 kGy dose steps in the Technical University of Munich reactor at 4.6 K

The management of intermediate level wastes in Sweden

International Nuclear Information System (INIS)

Hultgren, Aa.; Thegerstroem, C.

1980-01-01

A brief overview of current practices and research in Sweden on the management of intermediate level wastes is given. Intermediate level wastes include spent resins, filters and core components from the six power reactors in operation; radioactive wastes from nuclear fuel development at Studsvik and from non-nuclear applications are a minor contribution. (Auth.)

PNGMDR - Characterisation of intermediate-level long-lived wastes

International Nuclear Information System (INIS)

2014-12-01

This document presents the status of the characterization of intermediate-level long-lived wastes which are warehoused on exploited EDF sites or which will be produced during the deconstruction of first-generation reactors. It addresses aspects related to characterisation and packaging of wastes produced before 2015. More specifically, it addresses aspects related to contamination and to activation. Contamination is assessed by measurements whereas activation assessment is based on numerical simulations associated with measurements performed during parcel production. After having mentioned the concerned reactors, the document presents the methodology adopted for these assessments, and reports the progress status of the characterization process for these intermediate-level long-lived wastes

Analysis of calculated neutron flux response at detectors of G.A. Siwabessy multipurpose reactor (RSG-GAS Reactor)

International Nuclear Information System (INIS)

Taryo, Taswanda

2002-01-01

Multi Purpose Reactor G.A. Siwabessy (RSG-GAS) reactor core possesses 4 fission-chamber detectors to measure intermediate power level of RSG-GAS reactor. Another detector, also fission-chamber detector, is intended to measure power level of RSG-GAS reactor. To investigate influence of space to the neutron flux values for each detector measuring intermediate and power levels has been carried out. The calculation was carried out using combination of WIMS/D4 and CITATION-3D code and focused on calculation of neutron flux at different detector location of RSG-GAS typical working core various scenarios. For different scenarios, all calculation results showed that each detector, located at different location in the RSG-GAS reactor core, causes different neutron flux occurred in the reactor core due to spatial time effect

Thorium Fuel Performance in a Tight-Pitch Light Water Reactor Lattice

International Nuclear Information System (INIS)

Kim, Taek Kyum; Downar, Thomas J.

2002-01-01

Research on the utilization of thorium-based fuels in the intermediate neutron spectrum of a tight-pitch light water reactor (LWR) lattice is reported. The analysis was performed using the Studsvik/Scandpower lattice physics code HELIOS. The results show that thorium-based fuels in the intermediate spectrum of tight-pitch LWRs have considerable advantages in terms of conversion ratio, reactivity control, nonproliferation characteristics, and a reduced production of long-lived radiotoxic wastes. Because of the high conversion ratio of thorium-based fuels in intermediate spectrum reactors, the total fissile inventory required to achieve a given fuel burnup is only 11 to 17% higher than that of 238 U fertile fuels. However, unlike 238 U fertile fuels, the void reactivity coefficient with thorium-based fuels is negative in an intermediate spectrum reactor. This provides motivation for replacing 238 U with 232 Th in advanced high-conversion intermediate spectrum LWRs, such as the reduced-moderator reactor or the supercritical reactor

Method for controlling FBR type reactor

International Nuclear Information System (INIS)

Tamano, Toyomi; Iwashita, Tsuyoshi; Sakuragi, Masanori

1991-01-01

The present invention provides a controlling method for moderating thermal transient upon trip in an FBR type reactor. A flow channel for bypassing an intermediate heat exchanger is disposed in a secondary Na system. Then, bypassing flow rate is controlled so as to suppress fluctuations of temperature at a primary exit of the intermediate heat exchanger. Bypassing operation by using the bypassing flow channel is started at the same time with plant trip, to reduce the flow rate of secondary Na flown to the intermediate heat exchanger, so that the imbalance between the primary and the secondary Na flowrates is reduced. Accordingly, fluctuations of the temperature at the primary exit of the intermediate heat exchanger upon trip is suppressed. In view of the above, thermal transient applied to the reactor container upon plant trip can be moderated. As a result, the working life of the reactor can be extended, to improve plant integrity and safety. (I.S.)

VHTR engineering design study: intermediate heat exchanger program. Final report

International Nuclear Information System (INIS)

1976-11-01

The work reported is the result of a follow-on program to earlier Very High Temperature Reactor (VHTR) studies. The primary use of the VHTR is to provide heat for various industrial processes, such as hydrocarbon reforming and coal gasification. For many processes the use of an intermediate heat transfer barrier between the reactor coolant and the process is desirable; for some processes it is mandatory. Various intermediate heat exchanger (IHX) concepts for the VHTR were investigated with respect to safety, cost, and engineering design considerations. The reference processes chosen were steam-hydrocarbon reforming, with emphasis on the chemical heat pipe, and steam gasification of coal. The study investigates the critically important area of heat transfer between the reactor coolant, helium, and the various chemical processes

Study of nuclear reactions involving heavy nuclei and intermediate- and high-energy protons and an application in nuclear reactor physics (ADS)

International Nuclear Information System (INIS)

Matuoka, Paula Fernanda Toledo

2016-01-01

In the present work, intermediate- and high-energy nuclear reactions involving heavy nuclei and protons were studied with the Monte Carlo CRISP (Rio - Ilheus - Sao Paulo Collaboration) model. The most relevant nuclear processes studied were intranuclear cascade and fission-evaporation competition. Preliminary studies showed fair agreement between CRISP model calculation and experimental data of multiplicity of evaporated neutrons (E 20 MeV) were emitted mostly in the intranuclear cascade stage, while evaporation presented larger neutron multiplicity. Fission cross section of 209 mb and spallation cross section of 1788 mb were calculated { both in agreement with experimental data. The fission process resulted in a symmetric mass distribution. Another Monte Carlo code, MCNP, was used for radiation transport in order to understand the role of a spallation neutron source in a ADS (Accelerator Driven System) nuclear reactor. Initially, a PWR reactor was simulated to study the isotopic compositions in spent nuclear fuel. As a rst attempt, a spallation neutron source was adapted to an industrial size nuclear reactor. The results showed no evidence of incineration of transuranic elements and modifications were suggested. (author)

Containment vessel construction for nuclear power reactors

International Nuclear Information System (INIS)

Sulzer, H.D.; Coletti, J.L.

1975-01-01

A nuclear containment vessel houses an inner reactor housing structure whose outer wall is closely spaced from the inner wall of the containment vessel. The inner reactor housing structure is divided by an intermediate floor providing an upper chamber for housing the reactor and associated steam generators and a lower chamber directly therebeneath containing a pressure suppression pool. Communication between the upper chamber and the pressure suppression pool is established by conduits extending through the intermediate floor which terminate beneath the level of the pressure suppression pool and by inlet openings in the reactor housing wall beneath the level of the pressure suppression pool which communicate with the annulus formed between the outer wall of the reactor housing structure and the inner wall of the containment vessel. (Official Gazette)

A pump/intermediate heat exchanger assembly for a liquid metal reactor

International Nuclear Information System (INIS)

Nathenson, R.D.; Alexion, C.C.; Sumpman, W.C.

1987-01-01

A heat exchanger and electromagnetic pump assembly is disclosed comprising a heat exchanger housing defining an annularly shaped cavity and supporting therein a plurality of heat transfer tubes. An electromagnetic pump disposed beneath the heat exchanger comprises a circular array of flow couplers. Each flow coupler comprises a pump duct receiving primary liquid metal and a generator duct receiving a pumped intermediate liquid metal. A first plenum chamber is in communication with the generator ducts of all the flow couplers and receives intermediate liquid metal from inlet duct. The generator ducts exit their flows of intermediate liquid metal to a second plenum chamber in communication with the heat exchanger annularly shaped cavity to permit the flow of the intermediate liquid metal therethrough. A third plenum chamber receives collectively the flows of the primary liquid metal from the tubes and directs the primary liquid metal to the pump ducts of the flow couplers. The annular magnetic field of the electromagnetic pump is produced by a circular array of electromagnets having hollow windings cooled by a flow of intermediate liquid metal via tubes and manifolds. The leads to the electromagnets pass through an annular space around the inlet duct. (author)

Exploring the negative temperature coefficient behavior of acetaldehyde based on detailed intermediate measurements in a jet-stirred reactor

KAUST Repository

Tao, Tao; Sun, Wenyu; Hansen, Nils; Jasper, Ahren W.; Moshammer, Kai; Chen, Bingjie; Wang, Zhandong; Huang, Can; Dagaut, Philippe; Yang, Bin

2018-01-01

Acetaldehyde is an observed emission species and a key intermediate produced during the combustion and low-temperature oxidation of fossil and bio-derived fuels. Investigations into the low-temperature oxidation chemistry of acetaldehyde are essential to develop a better core mechanism and to better understand auto-ignition and cool flame phenomena. Here, the oxidation of acetaldehyde was studied at low-temperatures (528–946 K) in a jet-stirred reactor (JSR) with the corrected residence time of 2.7 s at 700 Torr. This work describes a detailed set of experimental results that capture the negative temperature coefficient (NTC) behavior in the low-temperature oxidation of acetaldehyde. The mole fractions of 28 species were measured as functions of the temperature by employing a vacuum ultra-violet photoionization molecular-beam mass spectrometer. To explain the observed NTC behavior, an updated mechanism was proposed, which well reproduces the concentration profiles of many observed peroxide intermediates. The kinetic analysis based on the updated mechanism reveals that the NTC behavior of acetaldehyde oxidation is caused by the competition between the O-addition to and the decomposition of the CHCO radical.

Exploring the negative temperature coefficient behavior of acetaldehyde based on detailed intermediate measurements in a jet-stirred reactor

KAUST Repository

Tao, Tao

2018-03-20

Acetaldehyde is an observed emission species and a key intermediate produced during the combustion and low-temperature oxidation of fossil and bio-derived fuels. Investigations into the low-temperature oxidation chemistry of acetaldehyde are essential to develop a better core mechanism and to better understand auto-ignition and cool flame phenomena. Here, the oxidation of acetaldehyde was studied at low-temperatures (528–946 K) in a jet-stirred reactor (JSR) with the corrected residence time of 2.7 s at 700 Torr. This work describes a detailed set of experimental results that capture the negative temperature coefficient (NTC) behavior in the low-temperature oxidation of acetaldehyde. The mole fractions of 28 species were measured as functions of the temperature by employing a vacuum ultra-violet photoionization molecular-beam mass spectrometer. To explain the observed NTC behavior, an updated mechanism was proposed, which well reproduces the concentration profiles of many observed peroxide intermediates. The kinetic analysis based on the updated mechanism reveals that the NTC behavior of acetaldehyde oxidation is caused by the competition between the O-addition to and the decomposition of the CHCO radical.

A Design of He-Molten Salt Intermediate Heat Exchanger for VHTR

International Nuclear Information System (INIS)

Jeong, Hui Seong; Bang, Kwang Hyun

2010-01-01

The Very High Temperature Reactor (VHTR), one of the most challenging next generation nuclear reactors, has recently drawn an international interest due to its higher efficiency and the operating conditions adequate for supplying process heat to the hydrogen production facilities. To make the design of VHTR complete and plausible, the designs of the Intermediate Heat Transport Loop (IHTL) as well as the Intermediate Heat Exchanger (IHX) are known to be one of the difficult engineering tasks due to its high temperature operating condition (up to 950 .deg. C). A type of compact heat exchangers such as printed circuit heat exchanger (PCHE) has been recommended for the IHX in the technical and economical respects. Selection of the heat transporting fluid for the intermediate heat transport loop is important in consideration of safety and economical aspects. Although helium is currently the primary interest for the intermediate loop fluid, several safety concerns of gas fluids have been expressed. If the pressure boundary of the intermediate loop fails, the blowdown of the gas may overcool the reactor core and then the heat sink is lost after the blowdown. Also the large inventory of gas in the intermediate loop may leak into the primary side. There is also a recommendation that the nuclear plant and the hydrogen production plant be separated by a certain distance to ensure the safety of the nuclear plant in case of accidental heavy gas release from the chemical plant. In this circumstance, the pumping power of gas fluid in the intermediate loop will be large enough to degrade the economics of nuclear hydrogen.An alternative candidate for the intermediate loop fluid in consideration of these safety and economical problems of gas fluid can be molten salts. The Flinak molten salt, a eutectic mixture of LiF, NaF and KF (46.5:11.5:42.0 mole %) is considered to be a potential candidate for the heat transporting fluid in the IHTL due to its chemical stability against the

Nuclear reactor

International Nuclear Information System (INIS)

Gilroy, J.E.

1980-01-01

An improved cover structure for liquid metal cooled fast breeder type reactors is described which it is claimed reduces the temperature differential across the intermediate grid plate of the core cover structure and thereby reduces its subjection to thermal stresses. (UK)

Intermediate temperature embrittlement of one new Ni-26W-6Cr based superalloy for molten salt reactors

Energy Technology Data Exchange (ETDEWEB)

Jiang, Li [Thorium Molten Salts Reactor Center, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); University of Chinese Academy of Science, Beijing 100049 (China); Ye, Xiangxi [University of Chinese Academy of Science, Beijing 100049 (China); Cui, Chuanyong [Institute of Metal Research, Chinese Academy of Sciences, Shenyang 110016 (China); Huang, Hefei; Leng, Bin [Thorium Molten Salts Reactor Center, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Li, Zhijun, E-mail: lizhijun@sinap.ac.cn [Thorium Molten Salts Reactor Center, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China); Zhou, Xingtai [Thorium Molten Salts Reactor Center, Shanghai Institute of Applied Physics, Chinese Academy of Sciences, Shanghai 201800 (China)

2016-06-21

Ni-26W-6Cr based superalloy is considered a potential structure material for the molten salt reactors due to its high strength and good compatibility with the fluoride salt. In the present work, the temperature dependence of the tensile behavior of the alloy was studied by tensile tests in the temperature range of 25–850 °C. This alloy exhibited a good ductility at RT and 450 °C, a ductility minimum from 650 to 750 °C and an intermediate ductility at 850 °C. TEM and EBSD characterization was performed on specimens tested at three typical temperature points (RT, 650 °C and 850 °C) to determine the deformation and fracture mechanisms accounting for the intermediate temperature embrittlement. At RT, the grain boundaries can accommodate enough dislocations to provide compatibility of the sliding between adjacent grains, then M{sub 6}C carbides act as crack origins and cause the fracture. In case of 650 °C, the grain boundaries cannot withstand the local stress even if only a small number of dislocation pile-ups exist. The premature cracks at grain boundaries impede the development of plastic deformation from single slips to multiple ones and cause the low ductility. If tested at 850 °C, the fracture process is retarded by the dynamic recovery and local dynamic recrystallization at crack tips.

Structural assessment of intermediate printed circuit heat exchanger for sodium-cooled fast reactor with supercritical CO2 cycle

International Nuclear Information System (INIS)

Lee, Youho; Lee, Jeong Ik

2014-01-01

Highlights: • We numerically model PCHE stress arising from pressure, and thermal loadings. • Stress levels are the highest around S-CO 2 channels, due to high pressure of S-CO 2 . • The conventional analytic models for PCHE underestimate actual stress levels. • Plasticity sufficiently lowers stress levels at channel tips. • PCHE for SFR-SCO 2 is anticipated to assure compliance with ASME design standards. - Abstract: Structural integrity of intermediate Printed Circuit Heat Exchanger (PCHE) for Sodium-cooled Fast Reactor (SFR) attached to Supercritical CO 2 (S-CO 2 ) is investigated. ANSYS-Mechanical was used to simulate stress fields of representative PCHE channels, with temperature fields imported from FLUENT simulation. Mechanical stress induced by pressure loading is found to be the primary source of stress. As plasticity sufficiently lowers local stress concentration at PCHE channel tips, PCHE type intermediate heat exchangers made of SS316 are anticipated to reliably assure compliance with design standards prescribed in the ASME standards, thanks to the structure temperature that is below the effective creep inducing point. The actual life time of PCHE for SFR-SCO 2 is likely to be affected by mechanical behavior change of SS316 with reactions with S-CO 2 and fatigue

The Intermediate Neutrino Program

CERN Document Server

Adams, C.; Ankowski, A.M.; Asaadi, J.A.; Ashenfelter, J.; Axani, S.N.; Babu, K.; Backhouse, C.; Band, H.R.; Barbeau, P.S.; Barros, N.; Bernstein, A.; Betancourt, M.; Bishai, M.; Blucher, E.; Bouffard, J.; Bowden, N.; Brice, S.; Bryan, C.; Camilleri, L.; Cao, J.; Carlson, J.; Carr, R.E.; Chatterjee, A.; Chen, M.; Chen, S.; Chiu, M.; Church, E.D.; Collar, J.I.; Collin, G.; Conrad, J.M.; Convery, M.R.; Cooper, R.L.; Cowen, D.; Davoudiasl, H.; de Gouvea, A.; Dean, D.J.; Deichert, G.; Descamps, F.; DeYoung, T.; Diwan, M.V.; Djurcic, Z.; Dolinski, M.J.; Dolph, J.; Donnelly, B.; Dwyer, D.A.; Dytman, S.; Efremenko, Y.; Everett, L.L.; Fava, A.; Figueroa-Feliciano, E.; Fleming, B.; Friedland, A.; Fujikawa, B.K.; Gaisser, T.K.; Galeazzi, M.; Galehouse, D.C.; Galindo-Uribarri, A.; Garvey, G.T.; Gautam, S.; Gilje, K.E.; Gonzalez-Garcia, M.; Goodman, M.C.; Gordon, H.; Gramellini, E.; Green, M.P.; Guglielmi, A.; Hackenburg, R.W.; Hackenburg, A.; Halzen, F.; Han, K.; Hans, S.; Harris, D.; Heeger, K.M.; Herman, M.; Hill, R.; Holin, A.; Huber, P.; Jaffe, D.E.; Johnson, R.A.; Joshi, J.; Karagiorgi, G.; Kaufman, L.J.; Kayser, B.; Kettell, S.H.; Kirby, B.J.; Klein, J.R.; Kolomensky, Yu. G.; Kriske, R.M.; Lane, C.E.; Langford, T.J.; Lankford, A.; Lau, K.; Learned, J.G.; Ling, J.; Link, J.M.; Lissauer, D.; Littenberg, L.; Littlejohn, B.R.; Lockwitz, S.; Lokajicek, M.; Louis, W.C.; Luk, K.; Lykken, J.; Marciano, W.J.; Maricic, J.; Markoff, D.M.; Martinez Caicedo, D.A.; Mauger, C.; Mavrokoridis, K.; McCluskey, E.; McKeen, D.; McKeown, R.; Mills, G.; Mocioiu, I.; Monreal, B.; Mooney, M.R.; Morfin, J.G.; Mumm, P.; Napolitano, J.; Neilson, R.; Nelson, J.K.; Nessi, M.; Norcini, D.; Nova, F.; Nygren, D.R.; Orebi Gann, G.D.; Palamara, O.; Parsa, Z.; Patterson, R.; Paul, P.; Pocar, A.; Qian, X.; Raaf, J.L.; Rameika, R.; Ranucci, G.; Ray, H.; Reyna, D.; Rich, G.C.; Rodrigues, P.; Romero, E.Romero; Rosero, R.; Rountree, S.D.; Rybolt, B.; Sanchez, M.C.; Santucci, G.; Schmitz, D.; Scholberg, K.; Seckel, D.; Shaevitz, M.; Shrock, R.; Smy, M.B.; Soderberg, M.; Sonzogni, A.; Sousa, A.B.; Spitz, J.; St. John, J.M.; Stewart, J.; Strait, J.B.; Sullivan, G.; Svoboda, R.; Szelc, A.M.; Tayloe, R.; Thomson, M.A.; Toups, M.; Vacheret, A.; Vagins, M.; Van de Water, R.G.; Vogelaar, R.B.; Weber, M.; Weng, W.; Wetstein, M.; White, C.; White, B.R.; Whitehead, L.; Whittington, D.W.; Wilking, M.J.; Wilson, R.J.; Wilson, P.; Winklehner, D.; Winn, D.R.; Worcester, E.; Yang, L.; Yeh, M.; Yokley, Z.W.; Yoo, J.; Yu, B.; Yu, J.; Zhang, C.

2015-01-01

The US neutrino community gathered at the Workshop on the Intermediate Neutrino Program (WINP) at Brookhaven National Laboratory February 4-6, 2015 to explore opportunities in neutrino physics over the next five to ten years. Scientists from particle, astroparticle and nuclear physics participated in the workshop. The workshop examined promising opportunities for neutrino physics in the intermediate term, including possible new small to mid-scale experiments, US contributions to large experiments, upgrades to existing experiments, R&D plans and theory. The workshop was organized into two sets of parallel working group sessions, divided by physics topics and technology. Physics working groups covered topics on Sterile Neutrinos, Neutrino Mixing, Neutrino Interactions, Neutrino Properties and Astrophysical Neutrinos. Technology sessions were organized into Theory, Short-Baseline Accelerator Neutrinos, Reactor Neutrinos, Detector R&D and Source, Cyclotron and Meson Decay at Rest sessions.This report summ...

Analysis and solution of spike current of intermediate range for nuclear instrumentation system

International Nuclear Information System (INIS)

Li Xingqiang; Xiao Yu; Xue Bin; Wang Yinli

2015-01-01

During the initial start-up of HYH NPP unit 1, spike currents occurred in intermediate range channel of RPN system and reactor trip was triggered. After analyzing the operation principle of intermediate range channel and site inspection, and doing simulating test, the root cause was fixed on the bug of range switching of intermediate range channel. Then a solution based on parameters optimizing was made and executed on site. (authors)

Peaking cladding temperature and break equivalent size of intermediate break loss of coolant accident

International Nuclear Information System (INIS)

Luo Bangqi

2012-01-01

The analysis results of intermediate break loss of coolant accident for the nuclear power plant of million kw level showed to be as following: (1) At the begin of life, the break occur simultaneity reactor shutdown with L(X)P. it's equivalent break size and peaking cladding temperature is respectively 20 cm and 849℃. (2) At the begin of life, the break occur simultaneity reactor shutdown without loop. the reactor coolant pumps will be stop after reactor shutdown 10 minutes, it's equivalent break size and peaking cladding temperature is respectively 10.5 cm and 921℃. (3) At the bur up of 31 GWd/t(EOC1). the break occur simultaneity reactor shutdown without loop, the reactor coolant pumps will be stop after reactor shutdown 20 minutes, it's equivalent break size and peaking cladding temperature is respectively 8 cm and 1145℃. The above analysis results showed that the peaking cladding temperature of intermediate break loss of coolant accident is not only related with the break equivalent size and core bur up, and is closely related with the stop time of coolant pumps because the coolant pumps would drive the coolant from safety system to produce the seal loop in break loop and affect the core coolant flow, results in the fuel cladding temperature increasing or damaging. Therefore, the break spectrum, burn up spectrum, the stop time of coolant pumps and operator action time will need to detail analysis and provide appropriate operating procedure, otherwise the peaking cladding temperature will exceed 1204℃ and threaten the safety of the reactor core when the intermediate break loss of coolant accident occur in some break equivalent size, burn up, stop pumps time and operator action not appropriate. The pressurizer pressure low signal simultaneity containment pressure higher signal were used as the operator manual close the signal of reactor coolant pumps after reactor shutdown of 20 minutes. have successful solved the operator intervention time from 10 minutes

Reactor core

International Nuclear Information System (INIS)

Azekura, Kazuo; Kurihara, Kunitoshi.

1992-01-01

In a BWR type reactor, a great number of pipes (spectral shift pipes) are disposed in the reactor core. Moderators having a small moderating cross section (heavy water) are circulated in the spectral shift pipes to suppress the excess reactivity while increasing the conversion ratio at an initial stage of the operation cycle. After the intermediate stage of the operation cycle in which the reactor core reactivity is lowered, reactivity is increased by circulating moderators having a great moderating cross section (light water) to extend the taken up burnup degree. Further, neutron absorbers such as boron are mixed to the moderator in the spectral shift pipe to control the concentration thereof. With such a constitution, control rods and driving mechanisms are no more necessary, to simplify the structure of the reactor core. This can increase the fuel conversion ratio and control great excess reactivity. Accordingly, a nuclear reactor core of high conversion and high burnup degree can be attained. (I.N.)

The Intermediate Neutrino Program

Energy Technology Data Exchange (ETDEWEB)

Adams, C.; et al.

2015-03-23

The US neutrino community gathered at the Workshop on the Intermediate Neutrino Program (WINP) at Brookhaven National Laboratory February 4-6, 2015 to explore opportunities in neutrino physics over the next five to ten years. Scientists from particle, astroparticle and nuclear physics participated in the workshop. The workshop examined promising opportunities for neutrino physics in the intermediate term, including possible new small to mid-scale experiments, US contributions to large experiments, upgrades to existing experiments, R&D plans and theory. The workshop was organized into two sets of parallel working group sessions, divided by physics topics and technology. Physics working groups covered topics on Sterile Neutrinos, Neutrino Mixing, Neutrino Interactions, Neutrino Properties and Astrophysical Neutrinos. Technology sessions were organized into Theory, Short-Baseline Accelerator Neutrinos, Reactor Neutrinos, Detector R&D and Source, Cyclotron and Meson Decay at Rest sessions.This report summarizes discussion and conclusions from the workshop.

The Intermediate Neutrino Program

Energy Technology Data Exchange (ETDEWEB)

Adams, C. [Yale Univ., New Haven, CT (United States); Alonso, J. R. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Ankowski, A. M. [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States); Asaadi, J. A. [Syracuse Univ., NY (United States); Ashenfelter, J. [Yale Univ., New Haven, CT (United States); Axani, S. N. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Babu, K [Oklahoma State Univ., Stillwater, OK (United States); Backhouse, C. [California Inst. of Technology (CalTech), Pasadena, CA (United States); Band, H. R. [Yale Univ., New Haven, CT (United States); Barbeau, P. S. [Duke Univ., Durham, NC (United States); Barros, N. [Univ. of Pennsylvania, Philadelphia, PA (United States); Bernstein, A. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Betancourt, M. [Illinois Inst. of Technology, Chicago, IL (United States); Bishai, M. [Brookhaven National Lab. (BNL), Upton, NY (United States); Blucher, E. [Univ. of Chicago, IL (United States); Bouffard, J. [State Univ. of New York (SUNY), Albany, NY (United States); Bowden, N. [Lawrence Livermore National Lab. (LLNL), Livermore, CA (United States); Brice, S. [Illinois Inst. of Technology, Chicago, IL (United States); Bryan, C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Camilleri, L. [Columbia Univ., New York, NY (United States); Cao, J. [Inst. of High Energy Physics, Beijing (China); Carlson, J. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Carr, R. E. [Columbia Univ., New York, NY (United States); Chatterjee, A. [Univ. of Texas, Arlington, TX (United States); Chen, M. [Univ. of California, Irvine, CA (United States); Chen, S. [Tsinghua Univ., Beijing (China); Chiu, M. [Brookhaven National Lab. (BNL), Upton, NY (United States); Church, E. D. [Illinois Inst. of Technology, Chicago, IL (United States); Pacific Northwest National Lab. (PNNL), Richland, WA (United States); Collar, J. I. [Univ. of Chicago, IL (United States); Collin, G. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Conrad, J. M. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Convery, M. R. [SLAC National Accelerator Lab., Menlo Park, CA (United States); Cooper, R. L. [Indiana Univ., Bloomington, IN (United States); Cowen, D. [Pennsylvania State Univ., University Park, PA (United States); Davoudiasl, H. [Brookhaven National Lab. (BNL), Upton, NY (United States); Gouvea, A. D. [Northwestern Univ., Evanston, IL (United States); Dean, D. J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Deichert, G. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Descamps, F. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); DeYoung, T. [Michigan State Univ., East Lansing, MI (United States); Diwan, M. V. [Brookhaven National Lab. (BNL), Upton, NY (United States); Djurcic, Z. [Argonne National Lab. (ANL), Argonne, IL (United States); Dolinski, M. J. [Drexel Univ., Philadelphia, PA (United States); Dolph, J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Donnelly, B. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Dwyer, D. A. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Dytman, S. [Univ. of Pittsburgh, PA (United States); Efremenko, Y. [Univ. of Tennessee, Knoxville, TN (United States); Everett, L. L. [Univ. of Wisconsin, Madison, WI (United States); Fava, A. [University of Padua, Padova (Italy); Figueroa-Feliciano, E. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Fleming, B. [Yale Univ., New Haven, CT (United States); Friedland, A. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Fujikawa, B. K. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Gaisser, T. K. [Univ. of Delaware, Newark, DE (United States); Galeazzi, M. [Univ. of Miami, FL (United States); Galehouse, DC [Univ. of Akron, OH (United States); Galindo-Uribarri, A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Garvey, G. T. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Gautam, S. [Tribhuvan Univ., Kirtipur (Nepal); Gilje, K. E. [Illinois Inst. of Technology, Chicago, IL (United States); Gonzalez-Garcia, M. [Stony Brook Univ., NY (United States); Goodman, M. C. [Argonne National Lab. (ANL), Argonne, IL (United States); Gordon, H. [Brookhaven National Lab. (BNL), Upton, NY (United States); Gramellini, E. [Yale Univ., New Haven, CT (United States); Green, M. P. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Guglielmi, A. [University of Padua, Padova (Italy); Hackenburg, R. W. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hackenburg, A. [Yale Univ., New Haven, CT (United States); Halzen, F. [Univ. of Wisconsin, Madison, WI (United States); Han, K. [Yale Univ., New Haven, CT (United States); Hans, S. [Brookhaven National Lab. (BNL), Upton, NY (United States); Harris, D. [Illinois Inst. of Technology, Chicago, IL (United States); Heeger, K. M. [Yale Univ., New Haven, CT (United States); Herman, M. [Brookhaven National Lab. (BNL), Upton, NY (United States); Hill, R. [Univ. of Chicago, IL (United States); Holin, A. [Univ. College London, Bloomsbury (United Kingdom); Huber, P. [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States); Jaffe, D. E. [Brookhaven National Lab. (BNL), Upton, NY (United States); Johnson, R. A. [Univ. of Cincinnati, OH (United States); Joshi, J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Karagiorgi, G. [Univ. of Manchester (United Kingdom); Kaufman, L. J. [Indiana Univ., Bloomington, IN (United States); Kayser, B. [Illinois Inst. of Technology, Chicago, IL (United States); Kettell, S. H. [Brookhaven National Lab. (BNL), Upton, NY (United States); Kirby, B. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Klein, J. R. [Univ. of Texas, Arlington, TX (United States); Kolomensky, Y. G. [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Univ. of California, Berkeley, CA (United States); Kriske, R. M. [Univ. of Minnesota, Minneapolis, MN (United States); Lane, C. E. [Drexel Univ., Philadelphia, PA (United States); Langford, T. J. [Yale Univ., New Haven, CT (United States); Lankford, A. [Univ. of California, Irvine, CA (United States); Lau, K. [Univ. of Houston, TX (United States); Learned, J. G. [Univ. of Hawaii, Honolulu, HI (United States); Ling, J. [Univ. of Illinois, Urbana-Champaign, IL (United States); Link, J. M. [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States); Lissauer, D. [Brookhaven National Lab. (BNL), Upton, NY (United States); Littenberg, L. [Brookhaven National Lab. (BNL), Upton, NY (United States); Littlejohn, B. R. [Illinois Inst. of Technology, Chicago, IL (United States); Lockwitz, S. [Illinois Inst. of Technology, Chicago, IL (United States); Lokajicek, M. [Inst. of Physics of the Academy of Sciences of Czech Republic, Prague (Czech Republic); Louis, W. C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Luk, K. [Univ. of California, Berkeley, CA (United States); Lykken, J. [Illinois Inst. of Technology, Chicago, IL (United States); Marciano, W. J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Maricic, J. [Univ. of Hawaii, Honolulu, HI (United States); Markoff, D. M. [North Carolina Central Univ., Durham, NC (United States); Caicedo, D. A. M. [Illinois Inst. of Technology, Chicago, IL (United States); Mauger, C. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mavrokoridis, K. [Univ. of Liverpool (United Kingdom); McCluskey, E. [Illinois Inst. of Technology, Chicago, IL (United States); McKeen, D. [Univ. of Washington, Seattle, WA (United States); McKeown, R. [Thomas Jefferson National Accelerator Facility (TJNAF), Newport News, VA (United States); Mills, G. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Mocioiu, I. [Pennsylvania State Univ., University Park, PA (United States); Monreal, B. [Univ. of California, Santa Barbara, CA (United States); Mooney, M. R. [Brookhaven National Lab. (BNL), Upton, NY (United States); Morfin, J. G. [Illinois Inst. of Technology, Chicago, IL (United States); Mumm, P. [National Inst. of Standards and Technology (NIST), Boulder, CO (United States); Napolitano, J. [Temple Univ., Philadelphia, PA (United States); Neilson, R. [Drexel Univ., Philadelphia, PA (United States); Nelson, J. K. [College of William and Mary, Williamsburg, VA (United States); Nessi, M. [European Organization for Nuclear Research (CERN), Geneva (Switzerland); Norcini, D. [Yale Univ., New Haven, CT (United States); Nova, F. [Univ. of Texas, Austin, TX (United States); Nygren, D. R. [Univ. of Texas, Arlington, TX (United States); Gann, GDO [Lawrence Berkeley National Lab. (LBNL), Berkeley, CA (United States); Univ. of California, Berkeley, CA (United States); Palamara, O. [Illinois Inst. of Technology, Chicago, IL (United States); Parsa, Z. [Brookhaven National Lab. (BNL), Upton, NY (United States); Patterson, R. [California Inst. of Technology (CalTech), Pasadena, CA (United States); Paul, P. [Stony Brook Univ., NY (United States); Pocar, A. [Univ. of Massachusetts, Amherst, MA (United States); Qian, X. [Brookhaven National Lab. (BNL), Upton, NY (United States); Raaf, J. L. [Illinois Inst. of Technology, Chicago, IL (United States); Rameika, R. [Illinois Inst. of Technology, Chicago, IL (United States); Ranucci, G. [National Inst. of Nuclear Physics, Milano (Italy); Ray, H. [Univ. of Florida, Gainesville, FL (United States); Reyna, D. [Sandia National Lab. (SNL-NM), Albuquerque, NM (United States); Rich, G. C. [Triangle Universities Nuclear Lab., Durham, NC (United States); Rodrigues, P. [Univ. of Rochester, NY (United States); Romero, E. R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Univ. of Tennessee, Knoxville, TN (United States); Rosero, R. [Brookhaven National Lab. (BNL), Upton, NY (United States); Rountree, S. D. [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States); Rybolt, B. [Univ. of Tennessee, Knoxville, TN (United States); Sanchez, M. C. [Iowa State Univ., Ames, IA (United States); Santucci, G. [Stony Brook Univ., NY (United States); Schmitz, D. [Univ. of Chicago, IL (United States); Scholberg, K. [Duke Univ., Durham, NC (United States); Seckel, D. [Univ. of Delaware, Newark, DE (United States); Shaevitz, M. [Columbia Univ., New York, NY (United States); Shrock, R. [Stony Brook Univ., NY (United States); Smy, M. B. [Univ. of California, Irvine, CA (United States); Soderberg, M. [Syracuse Univ., NY (United States); Sonzogni, A. [Brookhaven National Lab. (BNL), Upton, NY (United States); Sousa, A. B. [Univ. of Cincinnati, OH (United States); Spitz, J. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); John, J. M. S. [Univ. of Cincinnati, OH (United States); Stewart, J. [Brookhaven National Lab. (BNL), Upton, NY (United States); Strait, J. B. [Illinois Inst. of Technology, Chicago, IL (United States); Sullivan, G. [Univ. of Maryland, College Park, MD (United States); Svoboda, R. [Univ. of California, Davis, CA (United States); Szelc, A. M. [Yale Univ., New Haven, CT (United States); Tayloe, R. [Indiana Univ., Bloomington, IN (United States); Thomson, M. A. [Univ. of Cambridge (United Kingdom); Toups, M. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Vacheret, A. [Univ. of Oxford (United Kingdom); Vagins, M. [Univ. of California, Irvine, CA (United States); Water, R. G. V. D. [Los Alamos National Lab. (LANL), Los Alamos, NM (United States); Vogelaar, R. B. [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States); Weber, M. [Bern (Switzerland); Weng, W. [Brookhaven National Lab. (BNL), Upton, NY (United States); Wetstein, M. [Univ. of Chicago, IL (United States); White, C. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); White, B. R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Whitehead, L. [Univ. of Houston, TX (United States); Whittington, D. W. [Indiana Univ., Bloomington, IN (United States); Wilking, M. J. [Stony Brook Univ., NY (United States); Wilson, R. J. [Colorado State Univ., Fort Collins, CO (United States); Wilson, P. [Illinois Inst. of Technology, Chicago, IL (United States); Winklehner, D. [Massachusetts Inst. of Technology (MIT), Cambridge, MA (United States); Winn, D. R. [Fairfield Univ., CT (United States); Worcester, E. [Brookhaven National Lab. (BNL), Upton, NY (United States); Yang, L. [Univ. of Illinois, Urbana-Champaign, IL (United States); Yeh, M [Brookhaven National Lab. (BNL), Upton, NY (United States); Yokley, Z. W. [Virginia Polytechnic Inst. and State Univ. (Virginia Tech), Blacksburg, VA (United States); Yoo, J. [Illinois Inst. of Technology, Chicago, IL (United States); Yu, B. [Brookhaven National Lab. (BNL), Upton, NY (United States); Yu, J. [Univ. of Texas, Arlington, TX (United States); Zhang, C. [Brookhaven National Lab. (BNL), Upton, NY (United States)

2017-04-03

The US neutrino community gathered at the Workshop on the Intermediate Neutrino Program (WINP) at Brookhaven National Laboratory February 4-6, 2015 to explore opportunities in neutrino physics over the next five to ten years. Scientists from particle, astroparticle and nuclear physics participated in the workshop. The workshop examined promising opportunities for neutrino physics in the intermediate term, including possible new small to mid-scale experiments, US contributions to large experiments, upgrades to existing experiments, R&D plans and theory. The workshop was organized into two sets of parallel working group sessions, divided by physics topics and technology. Physics working groups covered topics on Sterile Neutrinos, Neutrino Mixing, Neutrino Interactions, Neutrino Properties and Astrophysical Neutrinos. Technology sessions were organized into Theory, Short-Baseline Accelerator Neutrinos, Reactor Neutrinos, Detector R&D and Source, Cyclotron and Meson Decay at Rest sessions.This report summarizes discussion and conclusions from the workshop.

Propane Oxidation at High Pressure and Intermediate Temperatures

DEFF Research Database (Denmark)

Hashemi, Hamid; Christensen, Jakob Munkholt; Glarborg, Peter

Propane oxidation at intermediate temperatures (500—900 K) and high pressure (100 bar) has been characterized by conducting experiments in a laminar flow reactor over a wide range of stoichiometries. The onset of fuel oxidation was found to be 600—725 K, depending on mixture stoichiometry...

Current issues in the management of low- and intermediate-level radioactive wastes from Ontario Hydro's CANDU reactors

International Nuclear Information System (INIS)

Krasznai, J.P.; Vaughan, B.R.; Williamson, A.S.

1990-01-01

Nuclear generating stations (NGSs) in Canada are operated by utilities in Ontario, Quebec, and New Brunswick. Ontario Hydro, with a committed nuclear program of 13,600 MW(electric) is the major producer of CANDU pressurized heavy-water reactor (PHWR) low- and intermediate-level radioactive wastes. All radioactive wastes with the exception of irradiated fuel are processed and retrievably stored at a centralized facility at the Bruce Nuclear Power Development site. Solid-waste classifications and annual production levels are given. Solid-waste management practices at the site as well as the physical, chemical, and radiochemical characteristics of the wastes are well documented. The paper summarizes types, current inventory, and estimated annual production rate of liquid waste. Operation of the tritium recovery facility at Darlington NGS, which removes tritium from heavy water and produces tritium gas in the process, gives rise to secondary streams of tritiated solid and liquid wastes, which will receive special treatment and packaging. In addition to the treatment of radioactive liquid wastes, there are a number of other important issues in low- and intermediate-level radioactive waste management that Ontario Hydro will be addressing over the next few years. The most pressing of these is the reduction of radioactive wastes through in-station material control, employee awareness, and improved waste characterization and segregation programs. Since Ontario Hydro intends to store retrievable wastes for > 50 yr, it is necessary to determine the behavior of wastes under long-term storage conditions

Plenum separator system for pool-type nuclear reactors

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1983-01-01

This invention provides a plenum separator system for pool-type nuclear reactors which substantially lessens undesirable thermal effects on major components. A primary feature of the invention is the addition of one or more intermediate plena, containing substantially stagnant and stratified coolant, which separate the hot and cold plena and particularly the hot plena from critical reactor components. This plenum separator system also includes a plurality of components which together form a dual pass flow path annular region spaced from the reactor vessel wall by an annular gas space. The bypass flow through the flow path is relatively small and is drawn from the main coolant pumps and discharged to an intermediate plenum

Full-Scale Continuous Mini-Reactor Setup for Heterogeneous Grignard Alkylation of a Pharmaceutical Intermediate

DEFF Research Database (Denmark)

Pedersen, Michael Jønch; Holm, Thomas; Rahbek, Jesper P.

2013-01-01

A reactor setup consisting of two reactors in series has been implemented for a full-scale, heterogeneous Grignard alkylation. Solutions pass from a small filter reactor into a static mixer reactor with multiple side entries, thus combining continuous stirred tank reactor (CSTR) and plug flow...

Nuclear power plant with several reactors

Energy Technology Data Exchange (ETDEWEB)

Grishanin, E I; Ilyunin, V G; Kuznetsov, I A; Murogov, V M; Shmelev, A N

1972-05-10

A design of a nuclear power plant suggested involves several reactors consequently transmitting heat to a gaseous coolant in the joint thermodynamical circuit. In order to increase the power and the rate of fuel reproduction the low temperature section of the thermodynamical circuit involves a fast nuclear reactor, whereas a thermal nuclear reactor is employed in the high temperature section of the circuit for intermediate heating and for over-heating of the working body. Between the fast nuclear and the thermal nuclear reactors there is a turbine providing for the necessary ratio between pressures in the reactors. Each reactor may employ its own coolant.

Nuclear Capacity Building through Research Reactors

International Nuclear Information System (INIS)

2017-01-01

Four Instruments: •The IAEA has recently developed a specific scheme of services for Nuclear Capacity Building in support of the Member States cooperating research reactors (RR) willing to use RRs as a primary facility to develop nuclear competences as a supporting step to embark into a national nuclear programme. •The scheme is composed of four complementary instruments, each of them being targeted to specific objective and audience: Distance Training: Internet Reactor Laboratory (IRL); Basic Training: Regional Research Reactor Schools; Intermediate Training: East European Research Reactor Initiative (EERRI); Group Fellowship Course Advanced Training: International Centres based on Research Reactors (ICERR)

Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

Energy Technology Data Exchange (ETDEWEB)

Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Greenwood, Michael Scott [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2015-12-01

Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

Preliminary Demonstration Reactor Point Design for the Fluoride Salt-Cooled High-Temperature Reactor

International Nuclear Information System (INIS)

Qualls, A. L.; Betzler, Benjamin R.; Brown, Nicholas R.; Carbajo, Juan; Greenwood, Michael Scott; Hale, Richard Edward; Harrison, Thomas J.; Powers, Jeffrey J.; Robb, Kevin R.; Terrell, Jerry W.

2015-01-01

Development of the Fluoride Salt-Cooled High-Temperature Reactor (FHR) Demonstration Reactor (DR) is a necessary intermediate step to enable commercial FHR deployment through disruptive and rapid technology development and demonstration. The FHR DR will utilize known, mature technology to close remaining gaps to commercial viability. Lower risk technologies are included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include tristructural-isotropic (TRISO) particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell heat exchangers. This report provides an update on the development of the FHR DR. At this writing, the core neutronics and thermal hydraulics have been developed and analyzed. The mechanical design details are still under development and are described to their current level of fidelity. It is anticipated that the FHR DR can be operational within 10 years because of the use of low-risk, near-term technology options.

Advanced High-Temperature Reactor Dynamic System Model Development: April 2012 Status

Energy Technology Data Exchange (ETDEWEB)

Qualls, A L; Cetiner, M S; Wilson, Jr, T L

2012-04-30

The Advanced High-Temperature Reactor (AHTR) is a large-output fluoride-salt-cooled high-temperature reactor (FHR). An early-phase preconceptual design of a 1500 MW(e) power plant was developed in 2011 [Refs. 1 and 2]. An updated version of this plant is shown as Fig. 1. FHRs feature low-pressure liquid fluoride salt cooling, coated-particle fuel, a high-temperature power cycle, and fully passive decay heat rejection. The AHTR is designed to be a “walk away” reactor that requires no action to prevent large off-site releases following even severe reactor accidents. This report describes the development of dynamic system models used to further the AHTR design toward that goal. These models predict system response during warmup, startup, normal operation, and limited off-normal operating conditions. Severe accidents that include a loss-of-fluid inventory are not currently modeled. The scope of the models is limited to the plant power system, including the reactor, the primary and intermediate heat transport systems, the power conversion system, and safety-related or auxiliary heat removal systems. The primary coolant system, the intermediate heat transport system and the reactor building structure surrounding them are shown in Fig. 2. These systems are modeled in the most detail because the passive interaction of the primary system with the surrounding structure and heat removal systems, and ultimately the environment, protects the reactor fuel and the vessel from damage during severe reactor transients. The reactor silo also plays an important role during system warmup. The dynamic system modeling tools predict system performance and response. The goal is to accurately predict temperatures and pressures within the primary, intermediate, and power conversion systems and to study the impacts of design changes on those responses. The models are design tools and are not intended to be used in reactor qualification. The important details to capture in the primary

Loop-type FBR reactor

International Nuclear Information System (INIS)

Ogura, Kenji; Kimura, Kimitaka; Jinbo, Masaichi; Hirayama, Hiroshi; Taguchi, Junzo; Hirata, Noriaki; Ozaki, Kenji; Maruyama, Shigeki.

1996-01-01

The inside of a vessel of an intermediate heat exchanger is divided vertically by a partition wall into a high temperature plenum region and a low temperature plenum region, a perforated horizontal plate is disposed in a horizontal direction at the upper portion and a flow shroud is disposed so as to surround the upper outside of the intermediate heat exchanger while passing through a lid from a perforated hole of the perforated horizontal plate. In addition, there is disposed a cylinder passing through the partition wall and the horizontal perforated plate for inserting a liquid surface penetrating equipment. The cylinder has an upper end opened above the liquid level of a liquid metal during normal operation and below the liquid level of the liquid metal during shut down of the reactor, and the lower end is opened in a lower plenum region. Vibrations of liquid level due to the high temperature liquid metal inflown from a hot leg pipeline to the inside of the vessel of the intermediate heat exchanger are suppressed by the perforated horizontal plate during reactor operation. On the other hand, upon shut down of the reactor, since the liquid level rises up to the upper portion of the cylinder, the liquid metal at low temperature inflows into the lower plenum region, and the liquid metal at high temperature above the horizontal perforated plate is eliminated in an early stage. (N.H.)

The UK commercial demonstration fast reactor design

International Nuclear Information System (INIS)

Holmes, J.A.G.

1987-01-01

The paper on the UK Commercial Demonstration Fast Reactor design was presented to the seminar on 'European Commercial Fast Reactor Programme, London 1987. The design is discussed under the topic headings:- primary circuit, intermediate heat exchangers and pumps, fuel and core, refuelling, steam generators, and nuclear island layout. (U.K.)

Applications of a lead pile coupled with fast reactor core of Yayoi as an intermediate energy neutron standard field

International Nuclear Information System (INIS)

Kosako, Toshiso; Nakazawa, Masaharu; Sekiguchi, Akira; Wakabayashi, Hiroaki.

1976-10-01

Intermediate neutron column of YAYOI reactor is here evaluated as an intermediate energy neutron standard field which provides a base of the measurements of various reaction rates in that energy region, including detector calibration and Doppler coefficient determination. The experiments were performed using YAYOI's core as a fast neutron source by coupling with the large lead pile, which is a 160 ton's octagon of 2.5 m high and with a thickness of about 2.5 m face to face distance. Spatial variation of the neutron flux in the lead pile was estimated by gold activation foils, and the neutron spectrum by sandwich foils, a helium-3 proportional counter and a proton recoil counter. The calculated results were obtained using one and two- dimensional discrete ordinate code, ANISN and TWOTRAN II. Through comparison of experiment with calculation, it became clear that the neutron field at the central block has simple energy spectrum and stable spatial distribution of the neutron flux, the absolute of which was 5.0 x 10 4 (n/cm 2 /sec/Watt) at the representative energy of 1 KeV. The energy spectrum of the position and the spatial dependent neutron flux in the lead pile are both represented by the semiempirical formula, which must be useful both for evaluation of experimental data and for future applications. (auth.)

Modeling the behavior of metallic fast reactor fuels during extended transients

International Nuclear Information System (INIS)

Kramer, J.M.; Liu, Y.Y.; Billone, M.C.; Tsai, H.C.

1993-01-01

Passive safety features in metal-fueled reactors utilizing the Integral Fast Reactor (IFR) fuel system make it possible to avoid core damage for extended time periods even when automatic scram system fail to operate or heat removal systems are severely degraded. The time scale for these transients are intermediate between those that have traditionally been analyzed in fast reactor safety assessments and those of normal operation. Consequently, it has been necessary to validate models and computer codes (FPIN2 and LIFE-METAL) for application to this intermediate time regime. Results from out-of-reactor Whole Pin Furnace tests are being used for this purpose. Pretest predictions for tests FM-1 through FM-6 have been performed and calculations have been compared with the experimental measurements. (orig.)

Heat exchanger with intermediate evaporating and condensing fluid

International Nuclear Information System (INIS)

Fraas, A.P.

1978-01-01

A shell and tube-type heat exchanger, such as a liquid sodium-operated steam generator for use in nuclear reactors, comprises a shell containing a primary fluid tube bundle, a secondary fluid tube bundle at higher elevation, and an intermediate fluid vaporizing at the surface of the primary fluid tubes and condensing at the surface of the secondary fluid tubes

Control of Advanced Reactor-Coupled Heat Exchanger System: Incorporation of Reactor Dynamics in System Response to Load Disturbances

Directory of Open Access Journals (Sweden)

Isaac Skavdahl

2016-12-01

Full Text Available Alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX are presented in this paper. One scheme is designed to control the cold outlet temperature of the SHX (Tco and the hot outlet temperature of the intermediate heat exchanger (Tho2 by manipulating the hot-side flow rates of the heat exchangers (Fh/Fh2 responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the control of the cold outlet temperature of the SHX (Tco only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1 flow rate manipulation; (2 reactor power manipulation; or (3 a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The third option appears preferable as it has an intermediate response time and requires only a minimal flow rate change.

Control of advanced reactor-coupled heat exchanger system: Incorporation of reactor dynamics in system response to load disturbances

Energy Technology Data Exchange (ETDEWEB)

Skavdahi, Isaac; Utgikar, Vivek [Dept. of Chemical and Materials Engineering, University of Idaho, Moscow (United States); Christensen, Richard [Nuclear Engineering Program, University of Idaho, Idaho Falls (United States); Chen, Ming Hui; Sun, Xiao Dong [Nuclear Engineering Program, Department of Mechanical and Aerospace Engineering, The Ohio State University, Columbus (United States); Sabharwall, Piyush [Idaho National Laboratory, Idaho Falls (United States)

2016-12-15

Alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX) are presented in this paper. One scheme is designed to control the cold outlet temperature of the SHX (T{sub co}) and the hot outlet temperature of the intermediate heat exchanger (Th{sub o2}) by manipulating the hot-side flow rates of the heat exchangers (F{sub h}/F{sub h2}) responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the control of the cold outlet temperature of the SHX (T{sub co}) only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1) flow rate manipulation; (2) reactor power manipulation; or (3) a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The third option appears preferable as it has an intermediate response time and requires only a minimal flow rate change.

Control of advanced reactor-coupled heat exchanger system: Incorporation of reactor dynamics in system response to load disturbances

International Nuclear Information System (INIS)

Skavdahi, Isaac; Utgikar, Vivek; Christensen, Richard; Chen, Ming Hui; Sun, Xiao Dong; Sabharwall, Piyush

2016-01-01

Alternative control schemes for an Advanced High Temperature Reactor system consisting of a reactor, an intermediate heat exchanger, and a secondary heat exchanger (SHX) are presented in this paper. One scheme is designed to control the cold outlet temperature of the SHX (T_c_o) and the hot outlet temperature of the intermediate heat exchanger (Th_o_2) by manipulating the hot-side flow rates of the heat exchangers (F_h/F_h_2) responding to the flow rate and temperature disturbances. The flow rate disturbances typically require a larger manipulation of the flow rates than temperature disturbances. An alternate strategy examines the control of the cold outlet temperature of the SHX (T_c_o) only, since this temperature provides the driving force for energy production in the power conversion unit or the process application. The control can be achieved by three options: (1) flow rate manipulation; (2) reactor power manipulation; or (3) a combination of the two. The first option has a quicker response but requires a large flow rate change. The second option is the slowest but does not involve any change in the flow rates of streams. The third option appears preferable as it has an intermediate response time and requires only a minimal flow rate change

Method of reactor fueling

International Nuclear Information System (INIS)

Saito, Toshiro.

1983-01-01

Purpose: To decrease the cost and shorten the working time by saving fueling neutron detectors and their components. Method: Incore drive tubes for the neutron source range monitor (SRM) and intermediate range monitor (IRM) are disposed respectively within in a reactor core and a SRM detector assembly is inserted to the IRM incore drive tube which is most nearest to the neutron source upon reactor fueling. The reactor core reactivity is monitored by the SRM detector assembly. The SRM detector asesembly inserted into the IRM drive tube is extracted at the time of charging fuels up to the frame connecting the SRM and, thereafter, IRM detection assembly is inserted into the IRM drive tube and the SRM detector assembly is inserted into the SRM drive tube respectively for monitoring the reactor core. (Sekiya, K.)

FBR type reactors

International Nuclear Information System (INIS)

Maemoto, Junko.

1985-01-01

Purpose: To moderate abrupt temperature change near the inner walls of a suspended body thereby prevent thermal shocks and thermal deformations to structural materials. Constitution: High temperature coolants during ordinary operation of the nuclear reactor flow from the reactor core through the flow holes of the suspended body and from the upper plenum into an intermediate heat exchanger. The temperature of the coolants is lowered with heat exchanging effect with secondary coolants in the heat exchange and the coolants are then flow through the lower plenum into the reactor core and heated again. Upon generation of reactor scram, the temperature of the coolants at the exit of the reactor core is reduced abruptly and the flow rate is lowered due to the pump coast down. However, mixing of the coolants in the suspended body is accelerated by the coolants at high temperature flowing out of the flow holes and the coolants at the low temperature flowing from the flow hole group, to reduce the temperature difference and moderate the stratification flow forming an abrupt temperature slope. (Yoshihara, H.)

CONDITIONING OF INTERMEDIATE-LEVEL WASTE AT FORSCHUNGSZENTRUM JUELICH GMBH

International Nuclear Information System (INIS)

Krumbach, H.

2003-01-01

This contribution to the group of low-level, intermediate, mixed and hazardous waste describes the conditioning of intermediate-level mixed waste (dose rate above 10 mSv/h at the surface) from Research Centre Juelich (FZJ). Conditioning of the waste by supercompaction is performed at Research Centre Karlsruhe (FZK). The waste described is radioactive waste arising from research at Juelich. This waste includes specimens and objects from irradiation experiments in the research reactors Merlin (FRJ-1) and Dido (FRJ-2) at FZJ. In principle, radioactive waste at Forschungszentrum Juelich GmbH is differentiated by the surface dose rate at the waste package. Up to a surface dose rate of 10 mSv/h, the waste is regarded as low-level. The radioactive waste described here has a surface dose rate above 10 mSv/h. Waste up to 10 mSv/h is conditioned at the Juelich site according to different conditioning methods. The intermediate-level waste can only be conditioned by supercompaction in the processing facility for intermediate-level waste from plant operation at Research Centre Karlsruhe. Research Centre Juelich also uses this waste cell to condition its intermediate-level waste from plant operation

Creep-Data Analysis of Alloy 617 for High Temperature Reactor Intermediate Heat Exchanger

International Nuclear Information System (INIS)

Kim, Woo Gon; Ryu, Woo Seog; Kim, Yong Wan; Yin, Song Nan

2006-01-01

The design of the metallic components such as hot gas ducts, intermediate heat exchanger (IHX) tube, and steam reformer tubes of very high temperature reactor (VHTR) is principally determined by the creep properties, because an integrity of the components should be preserved during a design life over 30 year life at the maximum operating temperature up to 1000 .deg. C. For designing the time dependent creep of the components, a material database is needed, and an allowable design stress at temperature should be determined by using the material database. Alloy 617, a nicked based superalloy with chromium, molybdenum and cobalt additions, is considered as a prospective candidate material for the IHX because it has the highest design temperature. The alloy 617 is approved to 982 .deg. C (1800 .deg. F) and other alloys approved to 898 .deg. C (1650 .deg. C), such as alloy 556, alloy 230, alloy HX, alloy 800. Also, the alloy 617 exhibits the highest level of creep strength at high temperatures. Therefore, it is needed to collect the creep data for the alloy 617 and the creep-rupture life at the given conditions of temperature and stress should be predicted for the IHX construction. In this paper, the creep data for the alloy 617 was collected through literature survey. Using the collected data, the creep life for the alloy 617 was predicted based on the Larson-Miller parameter. Creep master curves with standard deviations were presented for a safety design, and failure probability for the alloy 617 was obtained with a time coefficient

Assessment of the Capability of Molten Salt Reactors as a Next Generation High Temperature Reactors

International Nuclear Information System (INIS)

Elsheikh, B.M.

2017-01-01

Molten Salt Reactor according to Aircraft Reactor Experiment (ARE) and the Molten Salt Reactor Experiment (MSRE) programs, was designed to be the first full-scale, commercial nuclear power plant utilizing molten salt liquid fuels that can be used for producing electricity, and producing fissile fuels (breeding)burning actinides. The high temperature in the primary cycle enables the realization of efficient thermal conversion cycles with net thermal efficiencies reach in some of the designs of nuclear reactors greater than 45%. Molten salts and liquid salt because of their low vapor pressure are excellent candidates for meeting most of the requirements of these high temperature reactors. There is renewed interest in MSRs because of changing goals and new technologies in the use of high-temperature reactors. Molten Salt Reactors for high temperature create substantial technical challenges to have high effectiveness intermediate heat transfer loop components. This paper will discuss and investigate the capability and compatibility of molten salt reactors, toward next generation high temperature energy system and its technical challenges

LMFBR intermediate-heat-exchanger experience

International Nuclear Information System (INIS)

Cho, S.M.; Beaver, T.R.

1983-01-01

This paper presents developmental and operating experience of large Intermediate Heat Exchangers (IHX's) in US from the Fast Flux Test Facility (FFTF) to the Clinch River Breeder Reactor Plant (CRBRP) to the Large Development Plant (LDP). Design commonalities and deviations among these IHX's are synopsized. Various developmental tests that were conducted in the areas of hydraulic, structural and mechanical design are also presented. The FFTF is currently operating. Performance data of the FFTF IHXs are reviewed, and comparisons between actual and predicted performances are made. The results are used to assess the adequacy of IHX designs

The radiation safety assessment of the heating loop of district heating reactors

International Nuclear Information System (INIS)

Liu Yuanzhong

1993-01-01

The district heating reactors are used to supply heating to the houses in cities. The concerned problems are whether the radioactive materials reach the heated houses through heating loop, and whether the safety of the dwellers can be ensured. In order to prevent radioactive materials getting into the heated houses, the district heating reactors have three loops, namely, primary loop, intermediate loop, and heating loop. In the paper, the measures of preventing radioactive materials getting into the heating loop are presented, and the possible sources of the radioactivity in the water of the intermediate loop and the heating loop are given. The regulatory aim limit of radioactive concentration in the water of the intermediate loop is put forward, which is 18.5 Bq/l. Assuming that specific radioactivity of the water of contaminated intermediate loop is up to 18.5 Bq/l, the maximum concentration of radionuclides in water of the heating loop is calculated for the normal operation and the accident of district heating reactor. The results show that the maximum possible concentration is 5.7 x 10 -3 Bq/l. The radiation safety assessment of the heating loop is made out. The conclusions are that the district heating reactors do not bring any harmful impact to the dwellers, and the safety of the dwellers can be safeguarded completely

Practices and developments in the management of low and intermediate level radioactive waste in Sweden

International Nuclear Information System (INIS)

Hultgren, Aa.

1983-06-01

In the Swedish nuclear power program ten reactors are in operation and two more under construction. About 100000 m 3 of low and intermediate level radioactive waste will be produced from the operation of these reactors until the year 2010 and about 150000 m 3 from their decommissioning. All burnable radioactive wastes are sent to the Studsvik incineration plant for incineration. Spent resins are incorporated into cement or bitumen. The volume of non-combustible solid waste is reduced by compaction where possible. At the Studsvik research centre a substantial program for improved management of accumulated and future radioactive waste is at the beginning of its implementation. This includes advanced treatment and intermediate storage in a rock cavity. An R and D program on volume reduction of spent resins has reached the point of process verification and equipment design. All low and intermediate radioactive waste will be disposed in a rock cavity planned for commissioning by 1988. The paper reviews actual management experience and development efforts for low and intermediate level radioactive waste in Sweden. Contribution to the Seminar on the Management of Radioactive Waste, Taipei, Taiwan, 25-26 June, 1983. (Author)

Nuclear reactors

International Nuclear Information System (INIS)

Humphreys, P.; Davidson, D.F.; Thatcher, G.

1980-01-01

The cooling system of a liquid metal cooled fast breeder nuclear reactor of the pool kind is described. It has an intermediate heat exchange module comprising a tube-in-shell heat exchanger and an electromagnetic flow coupler in the base region of the module. Primary coolant is flowed through the heat exchanger being driven by electromagnetic interaction with secondary liquid metal coolant flow effected by a mechanical pump. (author)

Multicomponent activation detector measurements of reactor neutron spectra

International Nuclear Information System (INIS)

Sandberg, J.; Aarnio, P. A.; Routti, J. T.

1984-01-01

Information on the neutron flux is required in many applications of research reactors, such as activation analysis or radiation damage measurements. Flux spectrum measurements are commonly carried out with activation foils. The reaction types used are threshold reactions in the fast energy region, resonance reactions in the intermediate region and neutron capture reactions with l/v-cross section in the thermal region. It has been shown that it is possible to combine several detector elements into homogeneous multicomponent detectors. The activities of all detector reaction products can be determined with a single gamma spectrum measurement. The multicomponent principle sets some restrictions on the choice of detector reactions, for example, each product nuclide may be produced in one reaction only. Separate multicomponent threshold and resonance detectors were designed for the fast and intermediate regions, respectively. The detectors were fabricated in polyethylene irradiation capsules or quartz glass ampoules, and they were irradiated in a cadmium cover. The detectors were succesfully used in the irradiation ring and in the core of a Triga reactor. The intermediate and fast neutron spectrum was unfolded with the least-squares unfolding program LOUHI. According to the preliminary results multicomponent activation detectors might constitute a convenient means for carrying out routine neutron spectrum measurements in research reactors. (orig.)

Tritium Mitigation/Control for Advanced Reactor System

Energy Technology Data Exchange (ETDEWEB)

Sun, Xiaodong; Christensen, Richard; Saving, John P

2018-03-31

A tritium removal facility, which is similar to the design used for tritium recovery in fusion reactors, is proposed in this study for fluoride-salt-cooled high-temperature reactors (FHRs) to result in a two-loop FHR design with the elimination of an intermediate loop. Using this approach, an economic benefit can potentially be obtained by removing the intermediate loop, while the safety concern of tritium release can be mitigated. In addition, an intermediate heat exchanger (IHX) that can yield a similar tritium permeation rate to the production rate of 1.9 Ci/day in a 1,000 MWe PWR needs to be designed to prevent the residual tritium that is not captured in the tritium removal system from escaping into the power cycle and ultimately the environment. The main focus of this study is to aid the mitigation of tritium permeation issue from the FHR primary side to significantly reduce the concentration of tritium in the secondary side and the process heat application side (if applicable). The goal of the research is to propose a baseline FHR system without the intermediate loop. The specific objectives to accomplish the goals are: 1. To estimate tritium permeation behavior in FHRs; 2. To design a tritium removal system for FHRs; 3. To meet the same tritium permeation level in FHRs as the tritium production rate of 1.9 Ci/day in 1,000 MWe PWRs; 4. To demonstrate economic benefits of the proposed FHR system via comparing with the three-loop FHR system. The objectives were accomplished by designing tritium removal facilities, developing a tritium analysis code, and conducting an economic analysis. In the fusion reactor community, tritium extraction has been widely investigated and researched. Borrowing the experiences from the fusion reactor community, a tritium control and mitigation system was proposed. Based on mass transport theories, a tritium analysis code was developed, and the tritium behaviors were analyzed using the developed code. Tritium removal facilities

Conditioning of intermediate-level waste at Forschungszentrum Juelich GmbH

International Nuclear Information System (INIS)

Krumbach, H.

2003-01-01

This contribution to the group of low-level, intermediate, mixed and hazardous waste describes the conditioning of intermediate-level mixed waste (dose rate above 10 mSv/h at the surface) from Research Centre Juelich (FZJ). Conditioning of the waste by supercompaction is performed at Research Centre Karlsruhe (FZK). The waste described is radioactive waste arising from research at Juelich. This waste includes specimens and objects from irradiation experiments in the research reactors Merlin (FRJ-1) and Dido (FRJ-2) at FZJ. In principle, radioactive waste at Forschungszentrum Juelich GmbH is differentiated by the surface dose rate at the waste package. Up to a surface dose rate of 10 mSv/h, the waste is regarded as low-level. The radioactive waste described here has a surface dose rate above 10 mSv/h. Waste up to 10 mSv/h is conditioned at the Juelich site according to different conditioning methods. The intermediate-level waste can only be conditioned by supercompaction in the processing facility for intermediate-level waste from plant operation at Research Centre Karlsruhe. Research Centre Juelich also uses this waste cell to condition its intermediate-level waste from plant operation. (orig.)

Intermediate Leg SBLOCA - Long Lasting Pressure Transient

International Nuclear Information System (INIS)

Konjarek, D.; Bajs, T.; Vukovic, J.

2010-01-01

The basic phenomenology of Small Break Loss of Coolant Accident (SBLOCA) for PWR plant is described with focus on analysis of scenario in which reactor coolant pressure decreases below secondary system pressure. Best estimate light water reactor transient analysis code RELAP5/mod3.3 was used in calculation. Rather detailed model of the plant was used. The break occurs in intermediate leg on lowest elevation near pump suction. The size of the break is chosen to be small enough to cause cycling of safety valves (SVs) on steam generators (SGs) for some time, but, afterwards, it is large enough to remove decay heat through the break, causing cooling the secondary side. In this case of SBLOCA, when primary pressure decreases below secondary pressure, long lasting pressure transients with significant amplitude occur. Reasons for such behavior are explained.(author).

Design, construction and operation experience of the He-He intermediate heat exchanger

International Nuclear Information System (INIS)

Itoh, M.

1980-01-01

The conditions required for the primary helium, which is the cooling medium for the high temperature gas cooled reactor, are prescribed to be 1,000 0 C in temperature and 40 kgf/cm 2 in pressure at the outlet of the reactor, while the conditions required for the secondary helium at the outlet of the intermediate heat exchanger are prescribed to be 925 0 C in temperature and 45 kgf/cm 2 in pressure. This means that relatively high temperatures and high pressure are required for the system. The purpose of the present research and development project is to establish a design method, safety evaluation techniques, and safety-securing measures to be applied to an intermediate heat exchanger and to the overall heat exchanging system, which will satisfy those strict operating conditions as mentioned above. Research and development work on the high temperature heat exchanger has been and is being carried out

Sensitivity Studies of Advanced Reactors Coupled to High Temperature Electrolysis (HTE) Hydrogen Production Processes

International Nuclear Information System (INIS)

Edwin A. Harvego; Michael G. McKellar; James E. O'Brien; J. Stephen Herring

2007-01-01

High Temperature Electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800 C to 950 C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an Intermediate Heat Exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the steam or air sweep loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies. The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO2 reactor concepts). Sensitivity studies were then performed to evaluate the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycle producing the highest efficiencies varied depending on the temperature range considered

Back-end of the research reactor fuel cycle

International Nuclear Information System (INIS)

Gruber, Gehard J.

1996-01-01

This paper outlines the status of topics and issues related to: (1) Research Reactor Spent Nuclear Fuel Return to the U.S., including policy, shipments and ports of entry, management sites, fees, storage technologies, contracts, actual shipment, and legal process, (2) UKAEA: MTR Spent Nuclear Fuel Reprocessing, (3) COGEMA: MTR Spent Nuclear Fuel Reprocessing, and (4) Intermediate Storage + Direct Disposal for Research Reactors. (author)

Fuel assembly for a nuclear reactor

International Nuclear Information System (INIS)

Gjertsen, R.K.; Tower, S.N.; Huckestein, E.A.

1982-01-01

A fuel assembly for a nuclear reactor comprises a 5x5 array of guide tubes in a generally 20x20 array of fuel elements, the guide tubes being arranged to accommodate either control rods or water displacer rods. The fuel assembly has top and bottom Inconel (Registered Trade Mark) grids and intermediate Zircaloy grids in engagement with the guide tubes and supporting the fuel elements and guide tubes while allowing flow of reactor coolant through the assembly. (author)

Aerosol core nuclear reactor for space-based high energy/power nuclear-pumped lasers

International Nuclear Information System (INIS)

Prelas, M.A.; Boody, F.P.; Zediker, M.S.

1987-01-01

An aerosol core reactor concept can overcome the efficiency and/or chemical activity problems of other fuel-reactant interface concepts. In the design of a laser using the nuclear energy for a photon-intermediate pumping scheme, several features of the aerosol core reactor concept are attractive. First, the photon-intermediate pumping concept coupled with photon concentration methods and the aerosol fuel can provide the high power densities required to drive high energy/power lasers efficiently (about 25 to 100 kW/cu cm). Secondly, the intermediate photons should have relatively large mean free paths in the aerosol fuel which will allow the concept to scale more favorably. Finally, the aerosol core reactor concept can use materials which should allow the system to operate at high temperatures. An excimer laser pumped by the photons created in the fluorescer driven by a self-critical aerosol core reactor would have reasonable dimensions (finite cylinder of height 245 cm and radius of 245 cm), reasonable laser energy (1 MJ in approximately a 1 millisecond pulse), and reasonable mass (21 kg uranium, 8280 kg moderator, 460 kg fluorescer, 450 kg laser medium, and 3233 kg reflector). 12 references

Sodium-sodium intermediate heat exchangers design problems

International Nuclear Information System (INIS)

Chandramohan, R.

1975-01-01

This paper deals briefly with the calculation methods adapted, in working-out the stresses due to fluid pressures (normal as well as transient), weights, piping-reactions, vibration in the tube-bundle and also the thermal stresses during normal and transient conditions, for the mechanical design of intermediate heat-exchanger. The thermal stress evaluation of the tube-sheet is given particular emphasis. A brief outline of the design problems connected with the Na-Na exchangers of large size sodium cooled fast reactor plants is also given. (author)

Pump/heat exchanger assembly for pool-type reactor

International Nuclear Information System (INIS)

Nathenson, R.D.; Slepian, R.M.

1987-01-01

A heat exchanger and pump assembly comprising a heat exchanger including a housing for defining an annularly shaped cavity and supporting therein a plurality of heat transfer tubes. A pump is disposed beneath the heat exchanger and is comprised of a plurality of flow couplers disposed in a circular array. Each flow coupler is comprised of a pump duct for receiving a first electrically conductive fluid, i.e. the primary liquid metal, from a pool thereof, and a generator duct for receiving a second electrically conductive fluid, i.e. the intermediate liquid metal. The primary liquid metal is introduced from the reactor pool into the top, inlet ends of the tubes, flowing downward therethrough to be discharged from the tubes' bottom ends directly into the reactor pool. The primary liquid metal is variously introduced into the pump ducts directly from the reactor pool, either from the bottom or top end of the flow coupler. The intermediate fluid introduced into the generator ducts via the inlet duct and inlet plenum and after leaving the generator ducts passes through the annular cavity of the exchanger to cool the primary liquid in the tubes. The annular magnetic field of the pump is produced by a circular array of electromagnets having hollow windings cooled by a flow of the intermediate metal. (author)

Inherently safe shutdown of a liquid metal reactor upon a loss of intermediate cooling

International Nuclear Information System (INIS)

Feldman, E.E.; Mohr, D.

1986-01-01

Two unprotected (i.e., no scram or plant protection system action) loss-of-heat sink transients are scheduled to be performed on the Experimental Breeder Reactor-II in the Spring of 1986. One is to be initiated from full power (about 60 MW) and the other from half power. The loss-of-heat sink in each test is to be accomplished by essentially stopping the secondary-loop sodium coolant flow in about 20 s. Pretest predictions for the two tests, provided herein, show the reactor to passively shut down with average reactor outlet coolant temperatures rising no more that 5/degree/C before dropping and asymptotically approaching a quenching (or ''smothering'') temperature. This mild behavior is seen to be a direct consequence of the substantial negative reactivity feedback of the reactor coupled with the heat capacitance of the massive sodium pool in the primary tank. 14 refs., 11 figs., 1 tab

Reactor vessel nozzle cracks: a photoelastic study

International Nuclear Information System (INIS)

Smith, C.W.

1979-01-01

A method consisting of a marriage between the ''frozen stress'' photoelastic approach and the local stress field equations of linear elastic fracture mechanics for estimating stress intensity factor distributions in three dimensional, finite cracked body problems is reviewed and extensions of the method are indicated. The method is then applied to the nuclear reactor vessel nozzle corner crack problem for both Intermediate Test Vessel and Boiling Water Reactor geometries. Results are compared with those of other investigators. 35 refs

Mirror Fusion Test Facility: an intermediate device to a mirror fusion reactor

International Nuclear Information System (INIS)

Karpenko, V.N.

1983-01-01

The Mirror Fusion Test Facility (MFTF-B) now under construction at Lawrence Livermore National Laboratory represents more than an order-of-magnitude step from earlier magnetic-mirror experiments toward a future mirror fusion reactor. In fact, when the device begins operating in 1986, the Lawson criteria of ntau = 10 14 cm -3 .s will almost be achieved for D-T equivalent operation, thus signifying scientific breakeven. Major steps have been taken to develop MFTF-B technologies for tandem mirrors. Steady-state, high-field, superconducting magnets at reactor-revelant scales are used in the machine. The 30-s beam pulses, ECRH, and ICRH will also introduce steady-state technologies in those systems

Containment for small pressurized water reactors

International Nuclear Information System (INIS)

Siler, W.C.; Marda, R.S.; Smith, W.R.

1977-01-01

Babcock and Wilcox Company has prepared studies under ERDA contract of small and intermediate size (313, 365 and 1200 MWt) PWR reactor plants, for industrial cogeneration or electric power generation. Studies and experience with nuclear plants in this size range indicate unfavorable economics. To offset this disadvantage, modular characteristics of an integral reactor and close-coupled vapor suppression containment have been exploited to shorten construction schedules and reduce construction costs. The resulting compact reactor/containment complex is illustrated. Economic studies to date indicate that the containment design and the innovative construction techniques developed to shorten erection schedules have been important factors in reducing estimated project costs, thus potentially making such smaller plants competetive with competing energy sources

Neutronic calculation of reactor cells

International Nuclear Information System (INIS)

Jaliff, J.O.

1981-01-01

Multigroup calculations of cylindrical pin cells were programmed, in Fortran IV, upon the basis of collision probabilities in each energy group. A rational approximation to the fuel-to-fuel collision probability in resonance groups was used. Together with the intermediate resonance theory, cross sections corrected for heterogeneity and absorber interactions were found. For the optimization of the program, the cell of a BWR reactor was taken as reference. Data for such a cell and the reactor's operating conditions are presented. PINCEL is a fast and flexible program, with checked results, around a 69-group library. (M.E.L.) [es

Utilization of large electromagnetic pumps in the fast breeder reactors

International Nuclear Information System (INIS)

Deverge, C.; Lefrere, J.P.; Peturaud, P.; Sauvage, M.

1984-04-01

After an overview concerning the induction annular electromagnetic pumps and the dimensioning methods usually utilized, development of these components for a fast breeder integrated reactor is considered: - utilization of cooled EMP in the intermediate circuit, - utilization of immersed pumps, coupled with the intermediate exchanger, for the primary pumping; dimensioning, energetic aspects, and effects on the power plant geometrical configurations [fr

Design study of an IHX support structure for a POOL-TYPE Sodium-cooled fast reactor

International Nuclear Information System (INIS)

Park, Chang Gyu; Kim, Jong Bum; Lee, Jae Han

2009-01-01

The IHX (Intermediate Heat eXchanger) for a pool-type SFR (Sodium-cooled Fast Reactor) system transfers heat from the primary high temperature sodium to the intermediate cold temperature sodium. The upper structure of the IHX is a coaxial structure designed to form a flow path for both the secondary high temperature and low temperature sodium. The coaxial structure of the IHX consists of a central downcomer and riser for the incoming and outgoing intermediate sodium, respectively. The IHX of a pool-type SFR is supported at the upper surface of the reactor head with an IHX support structure that connects the IHX riser cylinder to the reactor head. The reactor head is generally maintained at the low temperature regime, but the riser cylinder is exposed in the elevated temperature region. The resultant complicated temperature distribution of the co-axial structure including the IHX support structure may induce a severe thermal stress distribution. In this study, the structural feasibility of the current upper support structure concept is investigated through a preliminary stress analysis and an alternative design concept to accommodate the IHTS (Intermediate Heat Transport System) piping expansion loads and severe thermal stress is proposed. Through the structural analysis it is found that the alternative design concept is effective in reducing the thermal stress and acquiring structural integrity

Fluoride Salt-Cooled High-Temperature Demonstration Reactor Point Design

Energy Technology Data Exchange (ETDEWEB)

Qualls, A. L. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Betzler, Benjamin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Carbajo, Juan [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hale, Richard Edward [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Harrison, Thomas J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, Jeffrey J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Robb, Kevin R. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Terrell, Jerry W. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Wysocki, Aaron J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States)

2016-02-01

The fluoride salt-cooled high-temperature reactor (FHR) demonstration reactor (DR) is a concept for a salt-cooled reactor with 100 megawatts of thermal output (MWt). It would use tristructural-isotropic (TRISO) particle fuel within prismatic graphite blocks. FLiBe (2 LiF-BeF₂) is the reference primary coolant. The FHR DR is designed to be small, simple, and affordable. Development of the FHR DR is a necessary intermediate step to enable near-term commercial FHRs. Lower risk technologies are purposely included in the initial FHR DR design to ensure that the reactor can be built, licensed, and operated within an acceptable budget and schedule. These technologies include TRISO particle fuel, replaceable core structural material, the use of that same material for the primary and intermediate loops, and tube-and-shell primary-to-intermediate heat exchangers. Several preconceptual and conceptual design efforts that have been conducted on FHR concepts bear a significant influence on the FHR DR design. Specific designs include the Oak Ridge National Laboratory (ORNL) advanced high-temperature reactor (AHTR) with 3400/1500 MWt/megawatts of electric output (MWe), as well as a 125 MWt small modular AHTR (SmAHTR) from ORNL. Other important examples are the Mk1 pebble bed FHR (PB-FHR) concept from the University of California, Berkeley (UCB), and an FHR test reactor design developed at the Massachusetts Institute of Technology (MIT). The MIT FHR test reactor is based on a prismatic fuel platform and is directly relevant to the present FHR DR design effort. These FHR concepts are based on reasonable assumptions for credible commercial prototypes. The FHR DR concept also directly benefits from the operating experience of the Molten Salt Reactor Experiment (MSRE), as well as the detailed design efforts for a large molten salt reactor concept and its breeder variant, the Molten Salt Breeder Reactor. The FHR DR technology is most representative of the 3400 MWt AHTR

Investigation/evaluation of water cooled fast reactor in the feasibility study on commercialized fast reactor cycle systems. Intermediate evaluation of phase-II study

International Nuclear Information System (INIS)

Kotake, Syoji; Nishikawa, Akira

2005-01-01

Feasibility study on commercialized fast reactor cycle systems aims at investigation and evaluation of FBR design requirement's attainability, operation and maintenance, and technical feasibility of the candidate system. Development targets are 1) ensuring safety, 2) economic competitiveness, 3) efficient utilization of resources, 4) reduction of environmental load and 5) enhancement of nuclear non-proliferation. Based on the selection of the promising concepts in the first phase, conceptual design for the plant system has proceeded with the following plant system: a) sodium cooled reactors at large size and medium size module reactors, b) a lead-bismuth cooled medium size reactor, c) a helium gas cooled large size reactor and d) a BWR type large size FBR. Technical development and feasibility has been assessed and the study considers the need of respective key technology development for the confirmation of the feasibility study. (T. Tanaka)

Performance of the prism reactor's passive decay heat removal system

International Nuclear Information System (INIS)

Magee, P.M.; Hunsbedt, A.

1989-01-01

The PRISM modular reactor concept has a totally passive safety-grade decay heat removal system referred to as the Reactor Vessel Auxiliary Cooling System (RVACS) that rejects heat from the reactor by radiation and natural convection of air. The system is inherently reliable and is not subject to the failure modes commonly associated with active cooling systems. The thermal performance of RVACS exceeds requirements and significant thermal margins exist. RVACS has been shown to perform its function under many postulated accident conditions. The PRISM power plant is equipped with three methods for shutdown: condenser cooling in conjunction with intermediate sodium and steam generator systems, and auxiliary cooling system (ACS) which removes heat from the steam generator by natural convection of air and transport of heat from the core by natural convection in the primary and intermediate systems, and a safety- grade reactor vessel auxiliary cooling system (RVACS) which removes heat passively from the reactor containment vessel by natural convection of air. The combination of one active and two passive systems provides a highly reliable and economical shutdown heat removal system. This paper provides a summary of the RVACS thermal performance for expected operating conditions and postulated accident events. The supporting experimental work, which substantiates the performance predictions, is also summarized

Historical Developments of Pyrolysis Reactors : A Review

NARCIS (Netherlands)

Garcia-Nunez, J. A.; Pelaez-Samaniego, M.R.; Garcia-Perez, M. E.; Fonts, I.; Abrego, J.; Westerhof, R. J.M.; Garcia Perez, M.

2017-01-01

This paper provides a review of pyrolysis technologies, focusing on reactor designs and companies commercializing these technologies. The renewed interest in pyrolysis is driven by the potential to convert lignocellulosic materials into bio-oil and biochar and the use of these intermediates for the

Reactor application of an improved bundle divertor

International Nuclear Information System (INIS)

Yang, T.F.; Ruck, G.W.; Lee, A.Y.; Smeltzer, G.; Prevenslik, T.

1978-11-01

A Bundle Divertor was chosen as the impurity control and plasma exhaust system for the beam driven Demonstration Tokamak Hybrid Reactor - DTHR. In the context of a preconceptual design study of the reactor and associated facility a bundle divertor concept was developed and integrated into the reactor system. The overall system was found feasible and scalable for reactors with intermediate torodial field strengths on axis. The important design characteristics are: the overall average current density of the divertor coils is 0.73 kA for each tesla of toroidal field on axis; the divertor windings are made from super-conducting cables supported by steel structures and are designed to be maintainable; the particle collection assembly and auxiliary cryosorption vacuum pump are dual systems designed such that they can be reactivated alterntively to allow for continuous reactor operation; and the power requirement for energizing and operating the divertor is about 5 MW

Mathematical modeling of a fast-breeder-reactor generating unit

International Nuclear Information System (INIS)

Kim, V.E.; Golovach, E.A.; Senkin, V.I.

1984-01-01

Dynamics equations are given for a reactor, intermediate heat exchanger, steam generator, and turbogenerator. The dynamic characteristics of the generating unit are described when perturbations occur in grid frequency, turbine valves, and feedwater consumption

CHF enhancement through Pressurized Intermediate Layer in IVR-ERVC Strategy

International Nuclear Information System (INIS)

Park, Seong Dae; Bang, In Cheol

2014-01-01

The molten fuel is sequentially relocated to bottom of reactor vessel. In-vessel retention through the external reactor vessel cooling (IVR-ERVC) strategy has been adapted to some reactors at this situation in order to prevent the progression of an accident. The limitation of IVR-ERVC strategy is CHF phenomenon on the outer wall of reactor vessel. The boiling is main heat transfer mode to remove decay heat between the reactor vessel and the coolant surrounding the reactor vessel. Heated molten radioactive material is leaked. The fuel coolant interaction (FCI) phenomenon could cause the steam explosion in a state of fully flooding condition. Therefore, the CHF should be enhanced in order to be a successful IVR-ERVC strategy. Related studies were performed to confirm the CHF limit with UPLU, SBLB, KAIST and UNIST test facilities The recommendations to increase CHF include coating some materials on the vessel outer surface, increasing the reactor cavity flood level and streamlining the gap between the vessel and the vessel insulation. Recently, flooding the liquid metal is proposed to prevent the boiling itself. In this work, the effects of pressurized liquid layer inserted between the reactor vessel and flooded coolant was studied. Suitable reactor geometry was also presented to apple this concept. Generally, CHF is increased as high pressure was applied until about 1/3 of critical pressure. The limit of IVR-ERVC strategy could overcome by using pressurized intermediate layer. The CFD analysis was performed to confirm the feasibility of pressurized IVR-ERVC system. There are enough thermal margins for due to the enlarged heat transfer area and the convection heat transfer

CHF enhancement through Pressurized Intermediate Layer in IVR-ERVC Strategy

Energy Technology Data Exchange (ETDEWEB)

Park, Seong Dae; Bang, In Cheol [Ulsan National Institute of Science and Technology, Ulsan (Korea, Republic of)

2014-05-15

The molten fuel is sequentially relocated to bottom of reactor vessel. In-vessel retention through the external reactor vessel cooling (IVR-ERVC) strategy has been adapted to some reactors at this situation in order to prevent the progression of an accident. The limitation of IVR-ERVC strategy is CHF phenomenon on the outer wall of reactor vessel. The boiling is main heat transfer mode to remove decay heat between the reactor vessel and the coolant surrounding the reactor vessel. Heated molten radioactive material is leaked. The fuel coolant interaction (FCI) phenomenon could cause the steam explosion in a state of fully flooding condition. Therefore, the CHF should be enhanced in order to be a successful IVR-ERVC strategy. Related studies were performed to confirm the CHF limit with UPLU, SBLB, KAIST and UNIST test facilities The recommendations to increase CHF include coating some materials on the vessel outer surface, increasing the reactor cavity flood level and streamlining the gap between the vessel and the vessel insulation. Recently, flooding the liquid metal is proposed to prevent the boiling itself. In this work, the effects of pressurized liquid layer inserted between the reactor vessel and flooded coolant was studied. Suitable reactor geometry was also presented to apple this concept. Generally, CHF is increased as high pressure was applied until about 1/3 of critical pressure. The limit of IVR-ERVC strategy could overcome by using pressurized intermediate layer. The CFD analysis was performed to confirm the feasibility of pressurized IVR-ERVC system. There are enough thermal margins for due to the enlarged heat transfer area and the convection heat transfer.

In-core fuel management for nuclear reactor

International Nuclear Information System (INIS)

Ross, M.F.; Visner, S.

1986-01-01

This patent describes in-core fuel management for nuclear reactor in which the first cycle of a pressurized water nuclear power reactor has a multiplicity of elongated, square fuel assemblies supported side-by-side to form a generally cylindrical, stationary core consisting entirely of fresh fuel assemblies. Each assembly of the first type has a substantially similar low average fissile enrichment of at least about 1.8 weight percent U-235, each assembly of the second type having a substantially similar intermediate average fissile enrichment at least about 0.4 weight percent greater than that of the first type, and each assembly of the third type having a substantially similar high average fissile enrichment at least about 0.4 weight percent greater than that of the intermediate type, the arrangement of the low, intermediate, and high enrichment assembly types which consists of: a generally cylindrical inner core region consisting of approximately two-thirds the total assemblies in the core and forming a figurative checkerboard array having a first checkerboard component at least two-thirds of which consists of high enrichment and intermediate enrichment assemblies, at least some of the high enrichment assemblies containing fixed burnable poison shims, and a second checkerboard component consisting of assemblies other than the high enrichment type; and a generally annular outer region consisting of the remaining assemblies and including at least some but less than two-thirds of the high enrichment type assemblies

An updated overview of low and intermediate level waste disposal facilities around the world

International Nuclear Information System (INIS)

Cuccia, Valeria; Uemura, George; Ferreira, Vinicius Verna M.; Tello, Cledola Cassia O. de; Malta, Ricardo Scott V.

2011-01-01

Low and intermediate level radioactive waste should be disposed off in proper disposal facilities. Some countries already have these facilities and others are planning theirs. Information about disposal facilities around the world is useful and necessary; however, data on this matter are usually scattered in official reports per country. In order to allow an easier access to this information, this paper aims to provide an overview of disposal facilities for low and intermediate level radioactive waste around the world, as updated as possible. Also, characteristics of the facilities are provided, when possible. Considering that the main source of radioactive waste are the activities of nuclear reactors in research or power generation, the paper will also provide a summarized overview of these reactors around the world, updated until April, 2011. This data collection may be an important tool for researchers, and other professionals in this field. Also, it might provide an overview about the final disposal of radioactive waste. (author)

Is light water reactor technology sustainable?

International Nuclear Information System (INIS)

Rothwell, G.; Van der Zwaan, B.

2001-01-01

This paper proposes criteria for determining ''intermediate sustainability'' over a 500-year horizon. We apply these criteria to Light Water Reactor (LWR) technology and the LWR industry. We conclude that LWR technology does not violate intermediate sustainability criteria for (1) environmental externalities, (2) worker and public health and safety, or (3) accidental radioactive release. However, it does not meet criteria to (1) efficiently use depleted uranium and (2) avoid uranium enrichment technologies that can lead to nuclear weapons proliferation. Finally, current and future global demand for LWR technology might be below the minimum needed to sustain the current global LWR industry. (author)

Safety Research Experiment Facility Project. Conceptual design report. Volume VII. Reactor cooling

International Nuclear Information System (INIS)

1975-12-01

The Reactor Cooling System (RCS) will provide the required cooling during test operations of the Safety Research Experiment Facility (SAREF) reactor. The RCS transfers the reactor energy generated in the core to a closed-loop water storage system located completely inside the reactor containment building. After the reactor core has cooled to a safe level, the stored heat is rejected through intermediate heat exchangers to a common forced-draft evaporative cooling tower. The RCS is comprised of three independent cooling loops of which any two can remove sufficient heat from the core to prevent structural damage to the system components

The MHD intermediate shock interaction with an intermediate wave: Are intermediate shocks physical?

International Nuclear Information System (INIS)

Wu, C.C.

1988-01-01

Contrary to the usual belief that MHD intermediate shocks are extraneous, the authors have recently shown by numerical solutions of dissipative MHD equations that intermediate shocks are admissible and can be formed through nonlinear steepening from a continuous wave. In this paper, he clarifies the differences between the conventional view and the results by studying the interaction of an MHD intermediate shock with an intermediate wave. The study reaffirms his results. In addition, the study shows that there exists a larger class of shocklike solutions in the time-dependent dissiaptive MHD equations than are given by the MHD Rankine-Hugoniot relations. it also suggests a mechanism for forming rotational discontinuities through the interaction of an intermediate shock with an intermediate wave. The results are of importance not only to the MHD shock theory but also to studies such as magnetic field reconnection models

Parametric Evaluation of Large-Scale High-Temperature Electrolysis Hydrogen Production Using Different Advanced Nuclear Reactor Heat Sources

International Nuclear Information System (INIS)

Harvego, Edwin A.; McKellar, Michael G.; O'Brien, James E.; Herring, J. Stephen

2009-01-01

High Temperature Electrolysis (HTE), when coupled to an advanced nuclear reactor capable of operating at reactor outlet temperatures of 800 C to 950 C, has the potential to efficiently produce the large quantities of hydrogen needed to meet future energy and transportation needs. To evaluate the potential benefits of nuclear-driven hydrogen production, the UniSim process analysis software was used to evaluate different reactor concepts coupled to a reference HTE process design concept. The reference HTE concept included an Intermediate Heat Exchanger and intermediate helium loop to separate the reactor primary system from the HTE process loops and additional heat exchangers to transfer reactor heat from the intermediate loop to the HTE process loops. The two process loops consisted of the water/steam loop feeding the cathode side of a HTE electrolysis stack, and the sweep gas loop used to remove oxygen from the anode side. The UniSim model of the process loops included pumps to circulate the working fluids and heat exchangers to recover heat from the oxygen and hydrogen product streams to improve the overall hydrogen production efficiencies. The reference HTE process loop model was coupled to separate UniSim models developed for three different advanced reactor concepts (a high-temperature helium cooled reactor concept and two different supercritical CO2 reactor concepts). Sensitivity studies were then performed to evaluate the affect of reactor outlet temperature on the power cycle efficiency and overall hydrogen production efficiency for each of the reactor power cycles. The results of these sensitivity studies showed that overall power cycle and hydrogen production efficiencies increased with reactor outlet temperature, but the power cycles producing the highest efficiencies varied depending on the temperature range considered

The use of nuclear reactor in radiation biology

International Nuclear Information System (INIS)

Ujeno, Yowri

1991-01-01

The Kyoto University Reactor (KUR) is widely used not only in biology, but also in applied biology, today. These studies were surveyed in the present paper and the future possibility to use KUR in radiation biology was discussed. The researches on the effects of thermal neutrons on various normal tissues, the biological effects of neutrons except thermal neutrons, especially intermediate neutrons between thermal and high speed neutrons or cold neutrons, the adaptive response of cells to thermal neutron radiation, the application of nuclear reactor-produced radionuclides including 195m Pt to biology, and the mutation in botanical science and so on, should be continued using nuclear reactor. The necessity of nuclear reactor in biology and applied biology is emphasized. (author)

Conceptual design study of small lead-bismuth cooled reactor

International Nuclear Information System (INIS)

Chikazawa, Yoshitaka; Hori, Toru; Kida, Masanori; Konomura, Mamoru

2004-11-01

In phase 2 of the feasibility study of commercialized fast reactor cycle systems of JNC, we make a concept of a small sodium cooled reactor for a power source of a city with various requirements, such as, safety and economical competitiveness. various reactor concepts are surveyed and a tank type reactor whose intermediate heat exchanger and primary main pumps are arranged in series is selected. In this study, a compact long life core and a simple reactor structure designs are pursued. The core type is three regional Zr concentration with one Pu enrichment core, the reactor outlet temperature achieves 550degC and the reactor electric output increases from 150 MWe to 165 MWe. The construction cost is much higher than the economical goal in the case of FOAK. But the construction cost in the case of NOAK is estimated to be 85.6% achieving the economical goal. (author)

Nuclear reactor sealing system

International Nuclear Information System (INIS)

McEdwards, J.A.

1983-01-01

A liquid metal-cooled nuclear reactor sealing system is disclosed. The nuclear reactor includes a vessel sealed at its upper end by a closure head. The closure head comprises at least two components, one of which is rotatable; and the two components define an annulus therebetween. The sealing system includes at least a first and second inflatable seal disposed in series in an upper portion of the annulus. The system further includes a dip seal extending into a body of insulation located adjacent a bottom portion of the closure head. The dip seal comprises a trough formed by a lower portion of one of the components, and a seal blade pendently supported from the other component and extending downwardly into the trough. A body of liquid metal is contained in the trough which submerges a portion of the seal blade. The seal blade is provided with at least one aperture located above the body of liquid metal for providing fluid communication between the annulus intermediate the dip seal and the inflatable seals, and a body of cover gas located inside the vessel. There also is provided means for introducing a purge gas into the annulus intermediate the inflatable seals and the seal blade. The purge gas is introduced in an amount sufficient to substantially reduce diffusion of radioactive cover gas or sodium vapor up to the inflatable seals. The purge gas mixes with the cover gas in the reactor vessel where it can be withdrawn from the vessel for treatment and recycle to the vessel

Slow and fast pyrolysis of Douglas-fir lignin: Importance of liquid-intermediate formation on the distribution of products

International Nuclear Information System (INIS)

Zhou, Shuai; Pecha, Brennan; Kuppevelt, Michiel van; McDonald, Armando G.; Garcia-Perez, Manuel

2014-01-01

The formation of liquid intermediates and the distribution of products were studied under slow and fast pyrolysis conditions. Results indicate that monomers are formed from lignin oligomeric products during secondary reactions, rather than directly from the native lignin. Lignin from Douglas-fir (Pseudotsuga menziesii) wood was extracted using the milled wood enzyme lignin isolation method. Slow pyrolysis using a microscope with hot-stage captured the liquid formation (>150 °C), shrinking, swelling (foaming), and evaporation behavior of lignin intermediates. The activation energy (E a ) for 5–80% conversions was 213 kJ mol −1 , and the pre-exponential factor (log A) was 24.34. Fast pyrolysis tests in a wire mesh reactor were conducted (300–650 °C). The formation of the liquid intermediate was visualized with a fast speed camera (250 Hz), showing the existence of three well defined steps: formation of lignin liquid intermediates, foaming and liquid intermediate swelling, and evaporation and droplet shrinking. GC/MS and UV-Fluorescence of the mesh reactor condensate revealed lignin oligomer formation but no mono-phenols were seen. An increase in pyrolytic lignin yield was observed as temperature increased. The molar mass determined by ESI-MS was not affected by pyrolysis temperature. SEM of the char showed a smooth surface with holes, evidence of a liquid intermediate with foaming; bursting from these foams could be responsible for the removal of lignin oligomers. Py-GC/MS studies showed the highest yield of guaiacol compounds at 450–550 °C. - Highlights: • The formation of a liquid intermediate phase is a critical step during lignin pyrolysis. • The lignin oligomers are thermally ejected from the liquid intermediate phase. • The mono-phenols are formed mainly from the secondary reactions of lignin oligomers

Low and intermediate radioactive waste management at OPG's western waste management facility

International Nuclear Information System (INIS)

Ellsworth, M.

2006-01-01

'Full text:' This paper will discuss low and intermediate level radioactive waste operations at Ontario Power Generation's Western Waste Management Facility. The facility has been in operation since 1974 and receives about 5000 - 7000 m 3 of low and intermediate level radioactive waste per year from Ontario's nuclear power plants. Low-level radioactive waste is received at the Waste Volume Reduction Building for possible volume reduction before it is placed into storage. Waste may be volume reduced by one of two methods at the WWMF, through either compaction or incineration. The Compactor is capable of reducing the volume of waste by a factor up to 5:1 for most waste. The Radioactive Incinerator is capable of volume reducing incinerable material by a factor up to 70:1. After processing, low-level waste is stored in above ground concrete warehouse-like structures called Low Level Storage Buildings. Low-level waste that cannot be volume reduced is placed into steel containers and stored in the Low Level Storage Buildings. Intermediate level waste is stored mainly in steel lined concrete storage structures. WWMF has both above ground and in-ground storage structures for intermediate level waste. Intermediate level waste consists primarily of resin and filters used to keep reactor water systems clean, and some used reactor core components. All low and intermediate level waste storage at the WWMF is considered interim storage and the material can be retrieved for future disposal or permanent storage. Current improvement initiatives include the installation of a new radioactive incinerator and a shredder/bagger. The new incinerator is a continuous feed system that is expected to achieve volume reduction rates up to 70:1, while incinerating higher volumes of waste than its predecessor. The shredder will break down large/bulky items into a form, which can be processed for further volume reduction. A Refurbishment Waste Storage Project is underway in anticipation of the

Optimization of intermediate heat exchangers for sodium cooled fast breeder reactors

International Nuclear Information System (INIS)

Barratt, R.O.; Cox, J.; Beith, R.M.V.

1978-01-01

Design evolution of the Intermediate Heat Exchanger for the U.S. Fast Breeder Programme is traced from the initial FFTF through to the Clinch River Designs and anticipated onward Commercial Plant Concepts. Supporting development work is outlined. Technical merits of certain features, such as sine wave tube banks and replaceable tube banks (in-situ), are weighed up against quantified cost advantages with simpler arrangements. Future design trends are outlined which will lead eventually to fully optimised designs. (author)

Modelling solid-convective flash pyrolysis of straw and wood in the Pyrolysis Centrifuge Reactor

DEFF Research Database (Denmark)

Bech, Niels; Larsen, Morten Boberg; Jensen, Peter Arendt

2009-01-01

in the Pyrolysis Centrifuge Reactor, a novel solid-convective flash pyrolysis reactor. The model relies on the original concept for ablative pyrolysis of particles being pyrolysed through the formation of an intermediate liquid compound which is further degraded to form liquid organics, char, and gas. To describe...

Status report of the program on neutron beam utilization at the Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Vuong Huu Tan

1996-08-01

The thermal reactor is an intense source not only of thermal neutron, but also intermediate as well as fast neutrons. Using the filtered neutron beam technique at steady state atomic reactor allows receiving the neutrons in the intermediate energy region with the most available intense flux at present. In the near time at the Dalat reactor the filtered neutron beam technique has been applied. Utilization of the filtered neutron beams in basic and applied researches has been a important activity of the Dalat Nuclear Research Institute (DNRI). This report presents some relevant characteristics of the filtered neutron beams and their utilization in nuclear data measurements, neutron capture gamma ray spectroscopy, neutron radiography, neutron dose calibration and other applications. (author). 3 refs, 2 figs

A conceptual design of neutron tumor therapy reactor facility with a YAYOI based fast neutron source reactor

International Nuclear Information System (INIS)

Wakabayashi, Hiroaki; An, Shigehiro.

1983-01-01

Fast neutron is known as one of useful radiations for radiation therapy of tumors. Boron neutron capture therapy (BNCT) of tumors which makes use of 10 B(n, α) 7 Li reaction of 10 B compounds selectively attached to tumor cells with thermal and intermediate neutrons is another way of neutron based radiation therapy which is, above all, attractive enough to kill tumor cells selectively sparing normal tissue. In Japan, BNCT has already been applied and leaned to be effective. After more than a decade operational experiences and the specific experiments designed for therapeutical purposes, in this paper, a conceptual design of a special neutron therapy reactor facility based on YAYOI - fast neutron source reactor of Nuclear Engineering Research Laboratory, Faculty of Engineering, the University of Tokyo - modified to provide an upward beam of fast and intermediate neutrons is presented. Emphasis is placed on the in-house nature of facility and on the coordinating capability of biological and physical researches as well as maintenances of the facility. (author)

Cement-based processes for the immobilization of intermediate level radioactive waste

International Nuclear Information System (INIS)

Brown, D.J.; Lee, D.J.; Price, M.S.T.; Smith, D.L.G.

1985-01-01

Increasing attention is being paid to the use of cement-based materials for the immobilisation of intermediate level wastes. Various cementitious materials are surveyed and the use of blast furnace slag is shown to be advantageous. The properties of cemented wastes are surveyed both during processing and as solid products. The application of Winfrith Cementation Laboratory technology to plant and flowsheet development for Winfrith Reactor sludge immobilisation is described. (author)

Loop type LMFBR reactor

International Nuclear Information System (INIS)

Ito, Hiroyuki

1989-01-01

In conventional FBR type reactors, primary coolants at high temperature uprise at a great flow rate and, due to the dynamic pressure thereof, the free surface is raised or sodium is partially jetted upwardly and then fallen again. Then, a wave killing plate comprising a buffer plate and a deflection plate is disposed to the liquid surface of coolants. Most of primary sodium uprising from the reactor core along the side of the upper mechanism during operation collide against the buffer plate of the wave killing plate to moderate the dynamic pressure and, further, disperse radially of the reactor vessel. On the other hand, primary sodium passing through flowing apertures collides against the deflection plate opposed to the flowing apertures to moderate the dynamic pressure, by which the force of raising the free surface is reduced. Thus, uprising and waving of the free surface can effectively be suppressed to reduce the incorporation of cover gases into the primary sodium, so that it is possible to prevent in injury of the recycling pump, abrupt increase of the reactor core reactivity and reduction of the heat efficiency of intermediate heat exchangers. (N.H.)

Investigation of molten salt fast reactor

International Nuclear Information System (INIS)

Kubota, Kenichi; Konomura, Mamoru

2002-01-01

On survey research for practicability strategy of fast reactor (FR) (phase 1), to extract future practicability image candidates of FR from wide options, in addition to their survey and investigation objects of not only solid fuel reactors of conventional research object but also molten salt reactor as a flowing fuel reactor, investigation on concept of molten salt FR plant was carried out. As a part of the first step of the survey research for practicability strategy, a basic concept on plant centered at nuclear reactor facility using chloride molten salt reactor capable of carrying out U-Pu cycle was examined, to perform a base construction to evaluate economical potential for a practical FBR. As a result, a result could be obtained that because of inferior fuel inventory and heat transmission to those in Na cooling reactor in present knowledge, mass of reactor vessel and intermediate heat exchanger were to widely increased to expect reduction of power generation unit price even on considering cheapness of its fuel cycle cost. Therefore, at present step further investigation on concept design of the chloride molten salt reactor plant system is too early in time, and it is at a condition where basic and elementary researches aiming at upgrading of economical efficiency such as wide reduction of fuel inventory, a measure expectable for remarkable rationalization effect of reprocessing system integrating a reactor to a processing facility, and so on. (G.K.)

MIT pebble bed reactor project

Energy Technology Data Exchange (ETDEWEB)

Kadak, Andrew C. [Massachusetts Institute of Technology, Cambridge (United States)

2007-03-15

The conceptual design of the MIT modular pebble bed reactor is described. This reactor plant is a 250 Mwth, 120 Mwe indirect cycle plant that is designed to be deployed in the near term using demonstrated helium system components. The primary system is a conventional pebble bed reactor with a dynamic central column with an outlet temperature of 900 C providing helium to an intermediate helium to helium heat exchanger (IHX). The outlet of the IHX is input to a three shaft horizontal Brayton Cycle power conversion system. The design constraint used in sizing the plant is based on a factory modularity principle which allows the plant to be assembled 'Lego' style instead of constructed piece by piece. This principle employs space frames which contain the power conversion system that permits the Lego-like modules to be shipped by truck or train to sites. This paper also describes the research that has been conducted at MIT since 1998 on fuel modeling, silver leakage from coated fuel particles, dynamic simulation, MCNP reactor physics modeling and air ingress analysis.

MIT pebble bed reactor project

International Nuclear Information System (INIS)

Kadak, Andrew C.

2007-01-01

The conceptual design of the MIT modular pebble bed reactor is described. This reactor plant is a 250 Mwth, 120 Mwe indirect cycle plant that is designed to be deployed in the near term using demonstrated helium system components. The primary system is a conventional pebble bed reactor with a dynamic central column with an outlet temperature of 900 C providing helium to an intermediate helium to helium heat exchanger (IHX). The outlet of the IHX is input to a three shaft horizontal Brayton Cycle power conversion system. The design constraint used in sizing the plant is based on a factory modularity principle which allows the plant to be assembled 'Lego' style instead of constructed piece by piece. This principle employs space frames which contain the power conversion system that permits the Lego-like modules to be shipped by truck or train to sites. This paper also describes the research that has been conducted at MIT since 1998 on fuel modeling, silver leakage from coated fuel particles, dynamic simulation, MCNP reactor physics modeling and air ingress analysis

Flexible Conversion Ratio Fast Reactor Systems Evaluation

Energy Technology Data Exchange (ETDEWEB)

Neil Todreas; Pavel Hejzlar

2008-06-30

Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores reated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcme the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better themal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor.

Flexible Conversion Ratio Fast Reactor Systems Evaluation

International Nuclear Information System (INIS)

Neil Todreas; Pavel Hejzlar

2008-01-01

Conceptual designs of lead-cooled and liquid salt-cooled fast flexible conversion ratio reactors were developed. Both concepts have cores treated at 2400 MWt placed in a large-pool-type vessel with dual-free level, which also contains four intermediate heat exchanges coupling a primary coolant to a compact and efficient supercritical CO2 Brayton cycle power conversion system. Decay heat is removed passively using an enhanced Reactor Vessel Auxiliary Cooling System and a Passive Secondary Auxiliary Cooling System. The most important findings were that (1) it is feasible to design the lead-cooled and salt-cooled reactor with the flexible conversion ratio (CR) in the range of CR=0 and CR=1 n a manner that achieves inherent reactor shutdown in unprotected accidents, (2) the salt-cooled reactor requires Lithium thermal Expansion Modules to overcome the inherent salt coolant's large positive coolant temperature reactivity coefficient, (3) the preferable salt for fast spectrum high power density cores is NaCl-Kcl-MgCl2 as opposed to fluoride salts due to its better thermal-hydraulic and neutronic characteristics, and (4) both reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor

The Carem reactor: Bridging the gap to nuclear power generation

International Nuclear Information System (INIS)

Ordonez, J.P.

1998-01-01

An idea is presented as an alternative for the introduction of nuclear power in presently non-nuclear countries. This idea involves going through an intermediate step between the traditional research reactor and the first commercial nuclear power plant. This intermediate step would consist of a very small nuclear power plant, with the principal goal of gaining in experience in the country on all the processes involved in introducing commercial nuclear generation. (author)

Is light water reactor technology sustainable?

Energy Technology Data Exchange (ETDEWEB)

Rothwell, G. [Stanford Univ., Dept. of Economics, CA (United States); Van der Zwaan, B. [Vrije Univ., Amsterdam, Inst. for Environmental Studies (Netherlands)

2001-07-01

This paper proposes criteria for determining ''intermediate sustainability'' over a 500-year horizon. We apply these criteria to Light Water Reactor (LWR) technology and the LWR industry. We conclude that LWR technology does not violate intermediate sustainability criteria for (1) environmental externalities, (2) worker and public health and safety, or (3) accidental radioactive release. However, it does not meet criteria to (1) efficiently use depleted uranium and (2) avoid uranium enrichment technologies that can lead to nuclear weapons proliferation. Finally, current and future global demand for LWR technology might be below the minimum needed to sustain the current global LWR industry. (author)

Studies on components for a molten salt reactor

International Nuclear Information System (INIS)

Nejedly, M.; Matal, O.

2003-01-01

The aim is contribute to a design of selected components of molten salt reactors with fuel in the molten fluoride salt matrix. Molten salt reactors (MSRs) permit the utilization of plutonium and minor actinides as new nuclear fuel from a traditional nuclear power station with production of electric energy. Results of preliminary feasibility studies of an intermediate heat exchanger, a small power molten salt pump and a modular conception of a steam generator for a demonstration unit of the MSR (30 MW) are summarized. (author)

Core design concepts for high performance light water reactors

International Nuclear Information System (INIS)

Schulenberg, T.; Starflinger, J.

2007-01-01

Light water reactors operated under supercritical pressure conditions have been selected as one of the promising future reactor concepts to be studied by the Generation IV International Forum. Whereas the steam cycle of such reactors can be derived from modern fossil fired power plants, the reactor itself, and in particular the reactor core, still need to be developed. Different core design concepts shall be described here to outline the strategy. A first option for near future applications is a pressurized water reactor with 380 .deg. C core exit temperature, having a closed primary loop and achieving 2% pts. higher net efficiency and 24% higher specific turbine power than latest pressurized water reactors. More efficiency and turbine power can be gained from core exit temperatures around 500 .deg. C, which require a multi step heat up process in the core with intermediate coolant mixing, achieving up to 44% net efficiency. The paper summarizes different core and assembly design approaches which have been studied recently for such High Performance Light Water Reactors

Decommissioning of the ICI TRIGA Mark I reactor

International Nuclear Information System (INIS)

Parry, D.R.; England, M.R.; Ward, A.; Green, D.

2000-01-01

This paper considers the fuel removal, transportation and subsequent decommissioning of the ICI TRIGA Mark I Reactor at Billingham, UK. BNFL Waste Management and Decommissioning carried out this work on behalf of ICI. The decommissioning methodology was considered in the four stages to be described, namely Preparatory Works, Reactor Defueling, Intermediate Level Waste Removal and Low Level Waste Removal. This paper describes the principal methodologies involved in the defueling of the reactor and subsequent decommissioning operations, highlighting in particular the design and safety case methodologies used in order to achieve a solution which was completed without incident or accident and resulted in a cumulative radiation dose to personnel of only 1.57 mSv. (author)

Mechanical Design Features of the KALIMER-600 Sodium-Cooled Reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Jae Han; Park, Chang Gyu; Kim, Jong Bum [Korea Atomic Energy Research Institute, Taejon (Korea, Republic of)

2005-07-01

KALIMER-600 is a sodium cooled reactor with a fast spectrum neutron reactor core. The NSSS design has three heat transport systems of a PHTS (Primary Heat Transport System), a IHTS (Intermediate Heat Transport System) and a SGS (Steam Generation System). PHTS is a pool type and has a large amount of sodium in the pool. The mechanical design targets are maintaining the enough structural integrity for a seismic load of SSE 0.3g and the thermal and mechanical loads by the high temperature environments and an economical competitiveness when compared with other reactor types.

Mechanical Design Features of the KALIMER-600 Sodium-Cooled Reactor

International Nuclear Information System (INIS)

Lee, Jae Han; Park, Chang Gyu; Kim, Jong Bum

2005-01-01

KALIMER-600 is a sodium cooled reactor with a fast spectrum neutron reactor core. The NSSS design has three heat transport systems of a PHTS (Primary Heat Transport System), a IHTS (Intermediate Heat Transport System) and a SGS (Steam Generation System). PHTS is a pool type and has a large amount of sodium in the pool. The mechanical design targets are maintaining the enough structural integrity for a seismic load of SSE 0.3g and the thermal and mechanical loads by the high temperature environments and an economical competitiveness when compared with other reactor types

Operation characteristics and conditions of training reactor VR-1

International Nuclear Information System (INIS)

Matejka, K.; Kolros, A.; Polach, S.; Sklenka, L.

1994-01-01

The first 3 years of operation of the VR-1 training reactor are reviewed. This period includes its physical start-up (preparation, implementation, results) and operation development as far as the current operating configuration of the reactor core. The physical start-up was commenced using a reactor core referred to as AZ A1, whose physical parameters had been verified by calculation and whose configuration was based on data tested experimentally on the SR-0 reactor at Vochov. The next operating core, labelled AZ A2, was already prepared during the test operation of the VR-1 reactor. Its configuration was such that both of the main horizontal channels, radial and tangential, could be employed. The configuration that followed, AZ A3, was an intermediate step before testing the graphite side reflector. The current reactor core, labelled AZ A3 G, was obtained by supplementing the previous core with a one-sided graphite side reflector. (Z.S.). 2 tabs., 11 figs., 2 refs

Review of microscopic integral cross section data in fundamental reactor dosimetry benchmark neutron fields

International Nuclear Information System (INIS)

Fabry, A.; McElroy, W.N.; Kellogg, L.S.; Lippincott, E.P.; Grundl, J.A.; Gilliam, D.M.; Hansen, G.E.

1976-01-01

This paper is intended to review and critically discuss microscopic integral cross section measurement and calculation data for fundamental reactor dosimetry benchmark neutron fields. Specifically the review covers the following fundamental benchmarks: the spontaneous californium-252 fission neutron spectrum standard field; the thermal-neutron induced uranium-235 fission neutron spectrum standard field; the (secondary) intermediate-energy standard neutron field at the center of the Mol-ΣΣ, NISUS, and ITN-ΣΣ facilities; the reference neutron field at the center of the Coupled Fast Reactor Measurement Facility; the reference neutron field at the center of the 10% enriched uranium metal, cylindrical, fast critical; the (primary) Intermediate-Energy Standard Neutron Field

Progress of the DUPIC fuel compatibility analysis (I) - reactor physics

Energy Technology Data Exchange (ETDEWEB)

Choi, Hang Bok; Jeong, Chang Joon; Roh, Gyu Hong; Rhee, Bo Wook; Park, Jee Won

2003-12-01

Since 1992, the direct use of spent pressurized water reactor fuel in CANada Deuterium Uranium (CANDU) reactors (DUPIC) has been studied as an alternative to the once-through fuel cycle. The DUPIC fuel cycle study is focused on the technical feasibility analysis, the fabrication of DUPIC fuels for irradiation tests and the demonstration of the DUPIC fuel performance. The feasibility analysis was conducted for the compatibility of the DUPIC fuel with existing CANDU-6 reactors from the viewpoints of reactor physics, reactor safety, fuel cycle economics, etc. This study has summarized the intermediate results of the DUPIC fuel compatibility analysis, which includes the CANDU reactor physics design requirements, DUPIC fuel core physics design method, performance of the DUPIC fuel core, regional overpower trip setpoint, and the CANDU primary shielding. The physics analysis showed that the CANDU-6 reactor can accommodate the DUPIC fuel without deteriorating the physics design requirements by adjusting the fuel management scheme if the fissile content of the DUPIC fuel is tightly controlled.

A resting bottom sodium cooled fast reactor

International Nuclear Information System (INIS)

Costes, D.

2012-01-01

This follows ICAPP 2011 paper 11059 'Fast Reactor with a Cold Bottom Vessel', on sodium cooled reactor vessels in thermal gradient, resting on soil. Sodium is frozen on vessel bottom plate, temperature increasing to the top. The vault cover rests on the safety vessel, the core diagrid welded to a toric collector forms a slab, supported by skirts resting on the bottom plate. Intermediate exchangers and pumps, fixed on the cover, plunge on the collector. At the vessel top, a skirt hanging from the cover plunges into sodium, leaving a thin circular slit partially filled by sodium covered by argon, providing leak-tightness and allowing vessel dilatation, as well as a radial relative holding due to sodium inertia. No 'air conditioning' at 400 deg. C is needed as for hanging vessels, and this allows a large economy. The sodium volume below the slab contains isolating refractory elements, stopping a hypothetical corium flow. The small gas volume around the vessel limits any LOCA. The liner cooling system of the concrete safety vessel may contribute to reactor cooling. The cold resting bottom vessel, proposed by the author for many years, could avoid the complete visual inspection required for hanging vessels. However, a double vessel, containing support skirts, would allow introduction of inspecting devices. Stress limiting thermal gradient is obtained by filling secondary sodium in the intermediate space. (authors)

Melting of metallic intermediate level waste

Energy Technology Data Exchange (ETDEWEB)

Huutoniemi, Tommi; Larsson, Arne; Blank, Eva [Studsvik Nuclear AB, Nykoeping (Sweden)

2013-08-15

This report presents a feasibility study of a melting facility for core components and reactor internals. An overview is given of how such a facility for treatment of intermediate level waste might be designed, constructed and operated and highlights both the possibilities and challenges. A cost estimate and a risk analysis are presented in order to make a conclusion of the technical feasibility of such a facility. Based on the authors' experience in operating a low level waste melting facility, their conclusion is that without technical improvements such a facility is not feasible today. This is based on the cost of constructing and operating such a facility, in conjunction with the radiological risks associated with operation and the uncertain benefits to disposal and long term safety.

Melting of metallic intermediate level waste

International Nuclear Information System (INIS)

Huutoniemi, Tommi; Larsson, Arne; Blank, Eva

2013-08-01

This report presents a feasibility study of a melting facility for core components and reactor internals. An overview is given of how such a facility for treatment of intermediate level waste might be designed, constructed and operated and highlights both the possibilities and challenges. A cost estimate and a risk analysis are presented in order to make a conclusion of the technical feasibility of such a facility. Based on the authors' experience in operating a low level waste melting facility, their conclusion is that without technical improvements such a facility is not feasible today. This is based on the cost of constructing and operating such a facility, in conjunction with the radiological risks associated with operation and the uncertain benefits to disposal and long term safety

System for cooling the upper wall of a nuclear reactor vessel

International Nuclear Information System (INIS)

Pailla, Henri; Schaller, Karl; Vidard, Michel.

1974-01-01

A system for cooling the upper wall of the main vessel of a fast neutron reactor is described. This vessel is suspended from an upper shield by the upper wall. It includes coils carrying a coolant which are immersed in an intermediate liquid bathing the wall and contained in a tank integral with the vessel. At least one of the two cooling and intermediate liquids is a liquid metal. The main vessel is contained in a safety vessel, the space between the main and safety vessels is occluded in its upper part by an insulating shield placed under the tank. There is a liquid metal seal between the upper wall and the upper shield under the tank. This system has been specially designed for sodium cooled fast neutron reactors [fr

Multiple microprocessor based nuclear reactor power monitor

International Nuclear Information System (INIS)

Lewis, P.S.; Ethridge, C.D.

1979-01-01

The reactor power monitor is a portable multiple-microprocessor controlled data acquisition device being built for the International Atomic Energy Association. Its function is to measure and record the hourly integrated operating thermal power level of a nuclear reactor for the purpose of detecting unannounced plutonium production. The monitor consists of a 3 He proportional neutron detector, a write-only cassette tape drive and control electronics based on two INTEL 8748 microprocessors. The reactor power monitor operates from house power supplied by the plant operator, but has eight hours of battery backup to cover power interruptions. Both the hourly power levels and any line power interruptions are recorded on tape and in memory. Intermediate dumps from the memory to a data terminal or strip chart recorder can be performed without interrupting data collection

AERE contracts with DoE on the treatment and disposal of intermediate level wastes

International Nuclear Information System (INIS)

Partridge, B.A.

1984-06-01

This document reports work carried out in 1983/84 under 10 contracts between DoE and AERE on the treatment and disposal of intermediate level wastes. Individual summaries are provided for each contract report within the document, under the headings: comparative evaluation of α and βγ irradiated medium level waste forms; modelling and characterisation of intermediate level waste forms based on polymers; optimisation of processing parameters for polymer and bitumen modified cements; ceramic waste forms; radionuclide release during leaching; ion exchange processes; electrical processes for the treatment of medium active liquid wastes; fast reactor fuel element cladding; dissolver residues; flowsheeting/systems study. (U.K.)

Heat transfer in intermediate heat exchanger under low flow rate conditions

International Nuclear Information System (INIS)

Mochizuki, H.

2008-01-01

The present paper describes the heat transfer in intermediate heat exchangers (IHXs) of liquid metal cooled fast reactors when flow rate is low such as a natural circulation condition. Although empirical correlations of heat transfer coefficients for IHX were derived using test data at the fast reactor 'Monju' and 'Joyo' and also at the 50 MW steam generator facility, the heat transfer coefficient was very low compared to the well known correlation for liquid metals proposed by Seban-Shimazaki. The heat conduction in IHX was discussed as a possible cause of the low Nusselt number. As a result, the heat conduction is not significant under the natural circulation condition, and the heat conduction term in the energy equation can be neglected in the one-dimensional plant dynamics calculation. (authors)

An analysis of system pressure and temperature distribution in self-pressurizer of SMART considering thermal stratification at intermediate cavity

Energy Technology Data Exchange (ETDEWEB)

Kang, Yeon Moon; Lee, Doo Jeong; Yoon, Ju Hyun; Kim, Hwan Yeol [Korea Atomic Energy Research Institute, Taejon (Korea)

1999-03-01

Because the pressurizer is in reactor vessel, the heat transfer from primary water would increase the temperatures of fluids in pressurizer to same temperature of hotleg, if no cooling equipment were supplied. Thus, heat exchanger and thermal insulator are needed to minimize heat transferred from primary water and to remove heat in pressurizer. The temperatures in cavities of pressurizer for normal operation are 70 deg C and 74 deg C for intermediate and end cavity, respectively, which considers the solubility of nitrogen gas in water. Natural convection is the mechanism of heat balance in pressurizer of SMART. In SMART, the heat exchanger in pressurizer is placed in lower part of intermediate cavity, so the heat in upper part of intermediate cavity can't be removed adequately and it can cause thermal stratification. If thermal stratification occurred, it increases heat transfers to nitrogen gas and system pressure increases as the result. Thus, proper evaluation of those effects on system pressure and ways to mitigate thermal stratification should be established. This report estimates the system pressure and temperatures in cavities of pressurizer with considering thermal stratification in intermediate cavity. The system pressure and temperatures for each cavities considered size of wet thermal insulator, temperature of upper plate of reactor vessel, parameters of heat exchanger in intermediate cavity such as flow rate and temperature of cooling water, heat transfer area, effective tube height, and location of cooling tube. In addition to the consideration of thermal stratification thermal mixing of all water in intermediate cavity also considered and compared in this report. (author). 6 refs., 60 figs., 2 tabs.

VVANTAGE 6 - an advanced fuel assembly design for VVER reactors

International Nuclear Information System (INIS)

Doshi, P.K.; DeMario, E.E.; Knott, R.P.

1993-01-01

Over the last 25 years, Westinghouse fuel assemblies for pressurized water reactors (PWR's) have undergone significant changes to the current VANTAGE 5. VANTAGE 5 PWR fuel includes features such as removable top nozzles, debris filter bottom nozzles, low-pressure-drop zircaloy grids, zircaloy intermediate flow mixing grids, optimized fuel rods, in-fuel burnable absorbers, and increased burnup capability to region average values of 48000 MWD/MTU. These features have now been adopted to the VVER reactors. Westinghouse has completed conceptual designs for an advanced fuel assembly and other core components for VVER-1000 reactors known as VANTAGE 6. This report describes the VVANTAGE 6 fuel assembly design

Fast reactor fuel reprocessing. An Indian perspective

International Nuclear Information System (INIS)

Natarajan, R.; Raj, Baldev

2005-01-01

The Department of Atomic Energy (DAE) envisioned the introduction of Plutonium fuelled fast reactors as the intermediate stage, between Pressurized Heavy Water Reactors and Thorium-Uranium-233 based reactors for the Indian Nuclear Power Programme. This necessitated the closing of the fast reactor fuel cycle with Plutonium rich fuel. Aiming to develop a Fast Reactor Fuel Reprocessing (FRFR) technology with low out of pile inventory, the DAE, with over four decades of operating experience in Thermal Reactor Fuel Reprocessing (TRFR), had set up at the India Gandhi Center for Atomic Research (IGCAR), Kalpakkam, R and D facilities for fast reactor fuel reprocessing. After two decades of R and D in all the facets, a Pilot Plant for demonstrating FRFR had been set up for reprocessing the FBTR (Fast Breeder Test Reactor) spent mixed carbide fuel. Recently in this plant, mixed carbide fuel with 100 GWd/t burnup fuel with short cooling period had been successfully reprocessed for the first time in the world. All the challenging problems encountered had been successfully overcome. This experience helped in fine tuning the designs of various equipments and processes for the future plants which are under construction and design, namely, the DFRP (Demonstration Fast reactor fuel Reprocessing Plant) and the FRP (Fast reactor fuel Reprocessing Plant). In this paper, a comprehensive review of the experiences in reprocessing the fast reactor fuel of different burnup is presented. Also a brief account of the various developmental activities and strategies for the DFRP and FRP are given. (author)

Review of microscopic integral cross section data in fundamental reactor dosimetry benchmark neutron fields

International Nuclear Information System (INIS)

Fabry, A.; McElroy, W.N.; Kellogg, L.S.; Lippincott, E.P.; Grundl, J.A.; Gilliam, D.M.; Hansen, G.E.

1976-10-01

The paper is intended to review and critically discuss microscopic integral cross section measurement and calculation data for fundamental reactor dosimetry benchmark neutron fields. Specifically the review covers the following fundamental benchmarks: (1) the spontaneous californium-252 fission neutron spectrum standard field; (2) the thermal-neutron induced uranium-235 fission neutron spectrum standard field; (3) the (secondary) intermediate-energy standard neutron field at the center of the Mol-ΣΣ, NISUS, and ITN--ΣΣ facilities; (4) the reference neutron field at the center of the Coupled Fast Reactor Measurement Facility (CFRMF); (5) the reference neutron field at the center of the 10 percent enriched uranium metal, cylindrical, fast critical; and (6) the (primary) Intermediate-Energy Standard Neutron Field

Design considerations for economically competitive sodium cooled fast reactors

International Nuclear Information System (INIS)

Zhang, Hongbin; Zhao, Haihua; Mousseau, Vincent; Szilard, Ronaldo

2009-01-01

The technological viability of sodium cooled fast reactors (SFR) has been established by various experimental and prototype (demonstration) reactors such as EBR-II, FFTF, Phenix, JOYO, BN-600 etc. However, the economic competitiveness of SFR has not been proven yet. The perceived high cost premium of SFRs over LWRs has been the primary impediment to the commercial expansion of SFR technologies. In this paper, cost reduction options are discussed for advanced SFR designs. These include a hybrid loop-pool design to optimize the primary system, multiple reheat and intercooling helium Brayton cycle for the power conversion system and the potential for suppression of intermediate heat transport system. The design options for the fully passive decay heat removal systems are also thoroughly examined. These include direct reactor auxiliary cooling system (DRACS), reactor vessel auxiliary cooling system (RVACS) and the newly proposed pool reactor auxiliary cooling system (PRACS) in the context of the hybrid loop-pool design. (author)

Evaluation Metrics for Intermediate Heat Exchangers for Next Generation Nuclear Reactors

International Nuclear Information System (INIS)

Sabharwall, Piyush; Kim, Eung Soo; Anderson, Nolan

2011-01-01

The Department of Energy (DOE) is working with industry to develop a next generation, high-temperature gas-cooled reactor (HTGR) as a part of the effort to supply the United States with abundant, clean, and secure energy as initiated by the Energy Policy Act of 2005 (EPAct; Public Law 109-58,2005). The NGNP Project, led by the Idaho National Laboratory (INL), will demonstrate the ability of the HTGR to generate hydrogen, electricity, and/or high-quality process heat for a wide range of industrial applications.

Multifrequency tests in the EBR-II reactor plant

International Nuclear Information System (INIS)

Feldman, E.E.; Mohr, D.; Gross, K.C.

1989-01-01

A series of eight multifrequency tests was conducted on the Experimental Breeder Reactor II. In half of the tests a control rod was oscillated and in the other half the controller input voltage to the intermediate-loop-sodium pump was perturbed. In each test the input disturbance consisted of several superimposed single-frequency sinusoidal harmonics of the same fundamental. The tests are described along with the theoretical and practical aspects of their development and design. Samples of measured frequency responses are also provided for both the reactor and the power plant. 22 refs., 5 figs., 2 tabs

Heavy Section Steel Technology Program. Part II. Intermediate vessel testing

International Nuclear Information System (INIS)

Whitman, G.D.

1975-01-01

The testing of the intermediate pressure vessels is a major activity under the Heavy Section Steel Technology Program. A primary objective of these tests is to develop or verify methods of fracture prediction, through the testing of selected structures and materials, in order that a valid basis can be established for evaluating the serviceability and safety of light-water reactor pressure vessels. These vessel tests were planned with sufficiently specific objectives that substantial quantitative weight could be given to the results. Each set of testing conditions was chosen so as to provide specific data by which analytical methods of predicting flaw growth, and in some cases crack arrest, could be evaluated. Every practical effort was made to assure that results would be relevant to some aspect of real reactor pressure vessel performance through careful control of material properties, selection of test temperatures, and design of prepared flaws. 5 references

Philosophy of safety evaluation on fast breeder reactor

International Nuclear Information System (INIS)

1981-01-01

This is the report submitted from the special subcommittee on reactor safety standard to the Nuclear Safety Commission on October 14, 1980, and it was decided to temporarily apply this concept to the safety examination on fast breeder reactors. The examination and discussion of this report were performed by taking the prototype reactor ''Monju'' into consideration, which is to be the present target, referring to the philosophy of the safety evaluation on fast breeder reactors in foreign countries and based on the experiences in the fast experimental reactor ''Joyo''. The items applicable to the safety evaluation for liquid metal-cooled fast breeder reactors (LMFBR) as they are among the existing safety examination guidelines are applied. In addition to the existing guidelines, the report describes the matters to be considered specifically for core, fuel, sodium, sodium void, reactor shut-down system, reactor coolant boundary, cover gas boundary and others, intermediate cooling system, removal of decay heat, containment vessels, high temperature structures, and aseismatic property in the safety design of LMFBR's. For the safety evaluation for LMFBR's, the abnormal transient changes in operation and the phenomena to be evaluated as accidents are enumerated. In order to judge the propriety of the criteria of locating LMFBR facilities, the serious and hypothetical accidents are decided to be evaluated in accordance with the guideline for reactor location investigation. (Wakatsuki, Y.)

The applications of research reactors. Report of an advisory group meeting

International Nuclear Information System (INIS)

2001-08-01

Owners and operators of many research reactors are finding that their facilities are not being utilized as fully as they might wish. Perhaps the original mission of the reactor has been accomplished or a particular analysis is now performed better in other ways. In addition, the fact that a research reactor exists and is available does not guarantee that users will come seeking to take advantage of the facility. Therefore, many research reactor owners and operators recognize that there is a need to develop a strategic plan for long term sustainability, including the 'marketing' of their facilities. An important first element in writing a strategic plan is to evaluate the current and potential capabilities of the reactor. The purpose of this document is to assist in such an evaluation by providing some factual and advisory information with respect to all of the current applications of research reactors. By reference to this text, each facility owner and operator will be able to assess whether or not a new application is feasible with the reactor, and what will be required to develop capability in that application. Applications fall into four broad categories: human resource development, irradiations, extracted beam work and testing. The human resource category includes public information, training and education and can be accomplished by any reactor. Irradiation applications involves inserting material into the reactor to induce radioactivity for analytical purposes, to produce radioisotopes or to induce radiation damage effects. Almost all reactors can be utilized for some irradiation applications, but as the reactor flux gets higher the range of potential uses gets larger. Beam work usually includes using neutron beams outside of the reactor for a variety of analytical purposes. Because of the magnitude of the fluxes needed at some distance from the core, most beam work can only be performed by the intermediate and higher powered research reactors. Testing nuclear

Operation of the High Flux Reactor. Annual report 1985

International Nuclear Information System (INIS)

1985-01-01

This year was characterized by the end of a major rebuilding of the installation during which the reactor vessel and its peripheral components were replaced by new and redesigned equipment. Both operational safety and experimental use were largely improved by the replacement. The reactor went back to routine operation on February 14, 1985, and has been operating without problem since then. All performance parameters were met. Other upgrading actions started during the year concerned new heat exchangers and improvements to the reactor building complex. The experimental load of the High Flux Reactor reached a satisfactory level with an average of 57%. New developments aimed at future safety related irradiation tests and at novel applications of neutrons from the horizontal beam tubes. A unique remote encapsulation hot cell facility became available adding new possibilities for fast breeder fuel testing and for intermediate specimen examination. The HFR Programme hosted an international meeting on development and use of reduced enrichment fuel for research reactors. All aspects of core physics, manufacture technology, and licensing of novel, proliferation-free, research reactor fuel were debated

Safety and environmental advantages of breeding blanketless fusion reactors

International Nuclear Information System (INIS)

Zucchetti, M.; Merola, M.; Matera, R.

1994-01-01

Next-step reactors will use DT cycle. However, environmental advantage will be the main chance for fusion to compete with other energy sources. The environmental problems of DT cycle due to tritium and neutron activation, are examined. Fusion commercial reactors could be based on alternative fuel cycles like D-He3. Advantages and disadvantages of this fuel cycle are outlined. All the technologies related with the self-breeding of tritium and the concept of breeding blanket itself may be not reactor relevant. In the frame of the Next-step studies, the potential advantages of intermediate DT devices without breeding blanket are discussed. Simplified design, lower cost, higher safety are the main ones. The problem of the source of tritium is examined. (author)

Method of reactor operation

International Nuclear Information System (INIS)

Nakajima, Takeshi

1988-01-01

Purpose: To minimize the power change due to the increase in xenone and power distribution after reaching the rated power in the case of using fresh fuels no requiring conditioning operation thereby starting the nuclear reactor in a short period of time and stably. Method: When control rods are entirely inserted only with a purpose for the compensation of the reactivity in a xenon-unsaturated state such as upon starting of the nuclear reactor, peaking is generated in the lower portion of the reactor core. Therefore, it is necessary to insert control rods for additionally suppressing the peaking in the lower portion of the reactor core to a relatively shallow level. In view of the above, a plurality of control rods are divided into a first control rod group finally inserted in the rated power state and a second control rod group other than the above. Then, the power is once elevated to the rated power level by means of such an intermediate control rod pattern that the ratio of the total extraction amount between the first control rod group and the second control rod group is made constant. Then, the control rods are extracted stepwise while setting the ratio of the total extraction amount constant in accordance with the change of the accumulating amount of xenone, to thereby obtain the purpose. (kamimura, M.)

Solid-Core Heat-Pipe Nuclear Batterly Type Reactor

International Nuclear Information System (INIS)

Ehud Greenspan

2008-01-01

This project was devoted to a preliminary assessment of the feasibility of designing an Encapsulated Nuclear Heat Source (ENHS) reactor to have a solid core from which heat is removed by liquid-metal heat pipes (HP). Like the SAFE 400 space nuclear reactor core, the HPENHS core is comprised of fuel rods and HPs embedded in a solid structure arranged in a hexagonal lattice in a 3:1 ratio. The core is oriented horizontally and has a square rather cylindrical cross section for effective heat transfer. The HPs extend from the two axial reflectors in which the fission gas plena are embedded and transfer heat to an intermediate coolant that flows by natural-circulation. The HP-ENHS is designed to preserve many features of the ENHS including 20-year operation without refueling, very small excess reactivity throughout life, natural circulation cooling, walkaway passive safety, and robust proliferation resistance. The target power level and specific power of the HP-ENHS reactor are those of the reference ENHS reactor. Compared to previous ENHS reactor designs utilizing a lead or lead-bismuth alloy natural circulation cooling system, the HP-ENHS reactor offers a number of advantageous features including: (1) significantly enhanced passive decay heat removal capability; (2) no positive void reactivity coefficients; (3) relatively lower corrosion of the cladding (4) a core that is more robust for transportation; (5) higher temperature potentially offering higher efficiency and hydrogen production capability. This preliminary study focuses on five areas: material compatibility analysis, HP performance analysis, neutronic analysis, thermal-hydraulic analysis and safety analysis. Of the four high-temperature structural materials evaluated, Mo TZM alloy is the preferred choice; its upper estimated feasible operating temperature is 1350 K. HP performance is evaluated as a function of working fluid type, operating temperature, wick design and HP diameter and length. Sodium is the

Treatment of low and intermediate level wastes

International Nuclear Information System (INIS)

Hoehlein, G.

1978-05-01

The methods described of low and intermediate level waste treatment are based exclusively on operating experience gathered with the KfK facilities for waste management, the Karlsruhe Reprocessing Plant (WAK), the ALKEM fuel element fabrication plant, the MZFR, KNK and FR 2 reactors as well as at the Karlsruhe Nuclear Research Center and at the state collecting depot of Baden-Wuerttemberg. The processing capacities and technical status are similar to that in 1976. With an annual throughput of 10000 m 3 of solid and liquid raw wastes, an aggregate activity of 85000 Ci, 500 kg of U and 2 kg of Pu, final waste in the amount of 500 m 3 was produced which was stored in the ASSE II salt mine. (orig.) [de

Large packages for reactor decommissioning waste

International Nuclear Information System (INIS)

Price, M.S.T.

1991-01-01

This study was carried out jointly by the Atomic Energy Establishment at Winfrith (now called the Winfrith Technology Centre), Windscale Laboratory and Ove Arup and Partners. The work involved the investigation of the design of large transport containers for intermediate level reactor decommissioning waste, ie waste which requires shielding, and is aimed at European requirements (ie for both LWR and gas cooled reactors). It proposes a design methodology for such containers covering the whole lifetime of a waste disposal package. The design methodology presented takes account of various relevant constraints. Both large self shielded and returnable shielded concepts were developed. The work was generic, rather than specific; the results obtained, and the lessons learned, remain to be applied in practice

85,000-GPM, single-stage, single-suction LMFBR intermediate centrifugal pump

International Nuclear Information System (INIS)

Fair, C.E.; Cook, M.E.; Huber, K.A.; Rohde, R.

1983-01-01

The mechanical and hydraulic design features of the 85,000-gpm, single-stage, single-suction pump test article, which is designed to circulate liquid-sodium coolant in the intermediate heat-transport system of a Large-Scale Liquid Metal Fast Breeder Reactor (LS-LMFBR), are described. The design and analytical considerations used to satisfy the pump performance and operability requirements are presented. The validation of pump hydraulic performance using a hydraulic scale-model pump is discussed, as is the featute test for the mechanical-shaft seal system

Fast Flux Test Facility: a unique resource for breeder reactor development

International Nuclear Information System (INIS)

Peckinpaugh, C.L.; Patterson, J.R.

1984-08-01

Based on encouraging data from the first two years of operation, the FFTF program has been redirected to close out testing on the intermediate lifetime fuel system and move directly to an early demonstration of a fuel system with a projected life of three to five years. Full advantage of favorable reactor physics effects will be taken to enhance inherent shutdown features and maximize reactor safety while reducing capital costs. Plant reliability will be demonstrated by achievement of high plant capacity factors

The story of fission reactors: from Chicago Pile to advanced energy systems

International Nuclear Information System (INIS)

Kannan, Umasankari

2017-01-01

Nuclear reactors have been designed which cater to different applications from small research reactors of a few watts to power reactors of several Giga Watts. Based on the neutron energy, there are thermal, intermediate and fast reactors operating are being designed. On the fuel utilization front, there are designs ranging from reactors using natural uranium fuel to enriched uranium to more efficient thorium based reactors. Reactors have also been designed which are neutron eaters, minor actinide burners and breeders. There have been variety of coolant and moderating materials used for different applications from water, gas cooled, liquid sodium cooled to molten salt cooled reactors. Several new reactor designs have been developed using innovative concepts in high temperature reactors, nuclear power packs and compact reactors for special purposes. The design challenges are many from modest designs to complicated hybrid reactors. The GEN-IV forum of IAEA has selected a few of these reactor designs for commercial power production in the coming years based on several quantified indicators. The evolutionary and revolutionary design approaches have been made over the years catering to different need of energy generation. A glimpse of some of the reactors being currently developed and the design modifications done in existing reactors have been given in this paper

Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor

International Nuclear Information System (INIS)

Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A.

1994-02-01

This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC's ''Statement of Policy for the Regulation of Advanced Nuclear Power Plants'' (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC's preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant's research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified

High heat flux testing of TiC coated molybdenum with a tungsten intermediate layer

International Nuclear Information System (INIS)

Fujitsuka, Masakazu; Fukutomi, Masao; Okada, Masatoshi

1988-01-01

The use of low atomic number (Z) material coatings for fusion reactor first-wall components has proved to be a valuable technique to reduce the plasma radiation losses. Molybdenum coated with titanium carbide is considered very promising since it has a good capability of receiving heat from the plasma. An interfacial reaction between the TiC film and the molybdenum substrate, however, causes a severe deterioration of the film at elevated temperatures. In order to solve this problem a TiC coated molybdenum with an intermediate tungsten layer was developed. High temperature properties of this material was evaluated by a newly devised electron beam heating apparatus. TiC coatings prepared on a vacuum-heat-treated molybdenum with a tungsten intermediate layer showed good high temperature stability and survived 2.0 s pulses of heating at a power density as high as 53 MW/m 2 . The melt area of the TiC coatings in high heat flux testings also markedly decreased when a tungsten intermediate layer was applied. The melting mechanism of the TiC coatings with and without a tungsten intermediate layer was discussed by EPMA measurements. (author)

German research reactor back-end provisions

International Nuclear Information System (INIS)

Koester, Siegfried; Gruber, Gerhard

2002-01-01

Germany has several types of Research Reactors in operation. These reactors use fuel containing uranium of U.S. origin. Basically all the fuel which will be spent until May 2006 will be returned to the U.S. under existing contracts with the U.S. Department of Energy. The contracts are based on the U.S. FRR SNF (Foreign Research Reactor Spent Nuclear Fuel) Program which started in May 1996 and which will last for 10 years. In 1990, the German Federal Government started a program to long-term store (approx. 40 years) and finally dispose of spent fuel in Germany after the so-called U.S. fuel return window will be closed. In order to long-term store the fuel, a special container was designed which covers all different types of spent fuel from the Research Reactors. The container called 'CASTOR MTR 2' is basically licensed and is already in use for the spent fuel of Russian origin from the 'Research Reactor Rossendorf' in the eastern part of Germany. All that fuel is expected to be stored in the existing intermediate storage facility, the so-called BZA (Brennelemente Zwischenlager Ahaus). BZA already accomodates spent fuel from the former THTR-300 high temperature reactor. A final repository does not yet exist in Germany. Alternative provisions to close the back-end of the Research Reactor fuel cycle are reprocessing at COGEMA (France) or in Russian facilities, perspectively. Waste return in a form to be agreed will be mandatory, at least in France. (author)

Problems specific to the piping of sodium-cooled fast breeder reactors

International Nuclear Information System (INIS)

Vrillon, B.; Befre, J.; Schaller, K.

1975-01-01

A certain number of specific problems arising in connection with the sodium pipes in fast neutron reactors, especially those of large diameter, are presented. The supporting system must be designed to achieve the best compromise among stresses due to weight and various stresses of thermal origin. Large-scale experimental studies carried out on actual elements of the intermediate circuit of the Phenix reactor showed that the circuits can withstand considerable deformation collapse of the walls without danger of leakage. Protection studies against earthquakes are mentionned [fr

Removal of linear alkylbenzene sulfonate (LAS) and its intermediate, sulfophenylalkanoates (spa) using a bioreactor

Energy Technology Data Exchange (ETDEWEB)

Lee, In Ku [Seoul National Univ., Seoul (Korea, Republic of); Kim, Ji Hyeon; Yoo, Young Je [Korea Advanced Institute of Science and Technology, Taejon (Korea, Republic of); Kim, Hwa Yong [Seoul National Univ., Seoul (Korea, Republic of)

1995-12-01

A fluidized-bed bioreactor was used to treat the industrial wastewater containing linear alkylbenzene sulfonate(LAS). Microorganism obtained by acclimation showed low growth rate and substrate inhibition at relatively high substrate concentration. It was found from the experiment using fluidized-bed bioreactor that 64 {mu}m-thick biofilm was made after 15 days. The 93.4% LAS removal efficiency in the reactor was obtained at 120 mg/L of initial LAS concentration, while the removal efficiency of sulfophenylalkannoates(SPA) which is a representative intermediate of biodegradation of LAS was 60-88% depending on the initial concentration of LAS. When sodium benzoate was added to the second reactor, the remaining SPA became lowered by 40% compared to the control experiment. And when activated carbon was added to the reactor, the removal efficiency of SPA was 96.2% and that of LAS was 100% when initial inflowing concentration of LAS was 350 mg/L. 13 refs., 4 tabs., 8 figs.

Mechanical behaviour of PWR fuel rods during intermediate storage

International Nuclear Information System (INIS)

Bouffioux, P.; Dalmas, R.; Bernaudat, C.

2000-01-01

EDF, which owns the irradiated fuel coming from its NPPs, has initiated studies regarding the mechanical behaviour of a fuel rod and the integrity of its cladding, in the case where the spent fuel is stored for a significant duration. During the phases following in-reactor irradiation (ageing in a water-pool, transport and intermediate storage), many phenomena, which are strongly coupled, may influence the cladding integrity: - residual power and temperature decay; - helium production and release in the free volume of the rod (especially for MOX fuel); - fuel column swelling; - cladding creep-out under the inner gas pressure of the fuel rod; - metallurgical changes due to high temperatures during transportation. In parallel, the quantification of the radiological risk is based on the definition of a cladding integrity criterion. Up to now, this criterion requires that the clad hoop strain due to creep-out does not exceed 1%. A more accurate criterion is being investigated. The study and modelling of all the phenomena mentioned above are included in a R and D programme. This programme also aims at redefining the cladding integrity criterion, which is assumed to be too conservative. The R and D programme will be presented. In order to predict the overall behaviour of the rod during the intermediate storage phases, the AVACYC code has been developed. It includes the models developed in the R and D programme. The input data of the AVACYC code are provided by the results of in-reactor rod behaviour simulations, using the thermal-mechanical CYRANO3 code. Its main results are the evolution vs. time of hoop stresses in the cladding, rod internal pressure and cladding hoop strains. Chained CYRANO-AVACYC calculations have been used to simulate the behaviour of MOX fuel rods irradiated up to 40 GWd/t and stored under air during 100 years, or under water during 50 years. For such fuels, where the residual power remains high, we show that a large part of the cladding strain

Design of the prestressed concrete reactor vessel for gas-cooled heating reactors

International Nuclear Information System (INIS)

Becker, G.; Notheisen, C.; Steffen, G.

1987-01-01

The GHR pebble bed reactor offers a simple, safe and economic possibility of heat generation. An essential component of this concept is the prestressed concrete reactor vessel. A system of cooling pipes welded to the outer surface of the liner is used to transfer the heat from the reactor to the intermediate circuit. The high safety of this vessel concept results from the clear separation of the functions of the individual components and from the design principle of the prestressed conncrete. The prestressed concrete structure is so designed that failure can be reliably ruled out under all operating and accident conditions. Even in the extremely improbable event of failure of all decay heat removal systems when decay heat and accumulated heat are transferred passively by natural convection only, the integrity of the vessel remains intact. For reasons of plant availability the liner and the liner cooling system shall be designed so as to ensure safe elimination of failure over the total operating life. The calculations which were peformed partly on the basis of extremely adverse assumption, also resulted in very low loads. The prestressed concrete vessel is prefabricated to the greatest possible extent. Thus a high quality and optimized fabrication technology can be achieved especially for the liner and the liner cooling system. (orig./HP)

A review of calculation methods for fast and intermediate reactors; Expose des methodes pour le calcul de reacteurs a neutrons rapides et intermediaires; Obzor metodov rascheta reaktorov na promezhutochnykh i bystrykh nejtronakh; Estudio panoramico de los metodos de calculo de los reactores rapidos e intermedios

Energy Technology Data Exchange (ETDEWEB)

Marchuk, G I [Akademiya Nauk, Moskva, Union of Soviet Socialist Republics (Russian Federation)

1962-03-15

This paper discusses the development of methods for calculating intermediate and fast reactors. It deals with various approaches to the problems of physical calculation. The calculation of resonance effects is discussed. Consideration is given to multi-group systems of fundamental and conjugate equations, various applications of perturbation theory to the problems of physical reactor calculation, and numerical methods of solving fundamental and conjugate reactor equations, which approximate the method of spherical harmonics. The paper describes an application of the response method to the solution of critical-mass problems, and methods of calculating reactors with hydrogeneous moderators. The fundamental features of an effective one-group reactor model are described. (author) [French] L'auteur examine la mise au point de methodes pour le calcul de reacteurs a neutrons rapides et intermediaires . Il decrit diverses manieres d'aborder les problemes des calculs sur la physique des reacteurs, notamment le calcul des effets de resonance. Il s'attache particulierement aux points suivants: systemes d'equations fondamentales et conjuguees a plusieurs groupes; diverses applications de la theorie des perturbations aux problemes de calculs sur la physique des reacteurs; methodes numeriques pour resoudre les equations fondamentales et conjuguees, voisines de la methode des harmoniques spheriques. L'auteur decrit ensuite une maniere d'appliquer la methode de la reponse aux problemes de la masse critique ainsi que des methodes pour le calcul de reacteurs ralentis a l'hydrogene. Il decrit les caracteristique s fondamentale s d'un modele de reacteur a un groupe effectif. (author) [Spanish] El autor analiza el desarrollo de los metodos de calculo de los reactores nucleares que trabajan con neutrones rapidos y con neutrones intermedios. Examina diversos planteos de los problemas del calculo fisico. Indica la forma de tomar en cuenta los efectos de resonancia y menciona los sistemas

Compatibility of steels for fast breeder reactor in high temperature sodium

International Nuclear Information System (INIS)

Yuhara, Shunichi

1981-01-01

In recent years, considerable progress has been made and experience has been obtained for material applicability in sodium-cooled fast breeder reactors. In this report, materials, principal dimensions and sodium conditions for the reactor system components, which include fuel pin cladding, intermediate heat exchangers, steam generators and pipings, are reviewed with emphasis on the thin-walled, high temperature and high strength components. The corrosion, mechanical and tribological behavior in sodium of important materials used for the reactor components, such as Types 304 and 316 stainless steel and 2 1/4Cr-1Mo steel, are discussed on the basis of characteristic testing results. Furthermore, material requirements concerned with compatibility in sodium are summarized from this review and discussion. (author)

Selection of engineering materials and fabrication of liquid metal fast breeder reactors

International Nuclear Information System (INIS)

Patriarca, P.

1975-01-01

Information is presented graphically and pictorially concerning the need for nuclear power; basic nuclear concepts including BWR, PWR, HTGR, and LMFBR; the fissioning process; nuclear reactor fuel; fabrication of reactor vessels for LMFBR's; fabrication of intermediate heat exchangers for LMFBR's; piping fabrication for LMFBR's; transition welds; steam generators for LMFBR demonstration plants worldwide; stress corrosion cracking of steam generator materials and weldments; post--test examination of the Alco/BLH sodium-heated steam generator; alternate steam generator designs; and alternate structural materials. (DCC)

The Chemistry of iodine in reactor safety: summary and conclusions: OECD Workshop

International Nuclear Information System (INIS)

1996-01-01

About seventy experts from fourteen OECD member countries attended this Fourth OECD Workshop on the chemistry of iodine in reactor safety, as well as experts from Latvia and the Commission of the European Communities. Thirty four papers were presented, in five sessions: national and international programmes (integral and intermediate-scale experiments), experimental homogeneous phase chemistry, surface processes, thermodynamic and kinetic studies, safety applications. A final session is devoted to a general discussion on remaining research studies relevant to reactor safety

Results of fission product release from intermediate-scale MCCI [molten core-concrete interaction] tests

International Nuclear Information System (INIS)

Spencer, B.W.; Thompson, D.H.; Fink, J.K.; Gunther, W.H.; Sehgal, B.R.

1988-01-01

A program of reactor-material molten core-concrete interaction (MCCI) tests and related analyses are under way at Argonne National Laboratory under sponsorship of the Electric Power Research Institute (EPRI). The particular objective of these tests is to provide data pertaining to the release of nonvolatile fission products such as La, Ba, and Sr, plus other aerosol materials, from the coupled thermal-hydraulic and chemical processes of the MCCI. The first stages of the program involving small and intermediate-scale tests have been completed. Three small-scale tests (/approximately/5 kg corium) and nine intermediate-scale tests (/approximately/30 kg corium) were performed between September 1985 and September 1987. Real reactor materials were used in these tests. Sustained internal heat generation at nominally 1 kW per kg of melt was provided by direct electrical heating of the corium mixture. MCCI tests were performed with both fully and partially oxidized corium mixtures that contained a variety of nonradioactive materials such as La 2 O 3 , BaO, and SrO to represent fission products. Both limestone/common sand and basaltic concrete basemats were used. The system was instrumented for characterization of the thermal hydraulic, chemical, gas release, and aerosol release processes

Refueling system for a nuclear reactor

International Nuclear Information System (INIS)

Koschkin, J.N.; Ordynskij, G.V.; Schchijan, C.G.; Schapkin, A.F.; Fadeev, A.I.; Laptev, F.V.; Batjukov, V.I.; Korolkov, K.I.; Borodin, I.V.; Tschernomordik, E.N.

1979-01-01

With the refueling system fuel elements are transferred from the intermediate distributing chamber within the fast breeder reactor vessel to the storage tanks for new and irradiated fuel elements outside of the reactor vessel and vice versa. It consists of a hermetic chamber, filled with inert gas, within which the refueling machine, having got a vertical swing pipe, is placed. On the swing pipe there is mounted by means of a bracket a hanging support tube for a tube manipulator that can be moved over the openings to the fuel elements. At the end of the tube manipulator there is a gripping device whose drive mechanism is arranged within the support tube. The swing pipe is moved by means of a drive mechanism outside of the chamber. (DG) [de

FBR type reactor

International Nuclear Information System (INIS)

Jinbo, Masakazu; Kawakami, Hiroto; Nagaoka, Kazuhito.

1996-01-01

In a LMFBR type reactor, a liquid level control means is disposed for lowering a level of liquid metal present in an annular gap along with temperature elevation of the liquid metal after the level is once elevated upon start-up of the reactor. In addition, a liquid level measuring means is disposed for measuring the level of the liquid metal present in the annular gap so as to intermittently lower the liquid level. Thus, temperature gradient in the vertical direction of the container can be moderated compared with the case where the liquid level is not changed or the case where temperature is changed together with the elevation of the liquid level. As a result, the change of difference of thermal expansion is decreased to reduce stresses generated in the circumferential direction thereby preventing occurrence of a liquid level heat ratchet phenomenon. Even if the liquid level control means should stop during operation, the liquid level lowers and does not cause a sharp heat gradient as in the case where the liquid level is elevated, and since the temperature of the liquid level is lowered even after shut down of the reactor, generated stresses are not increased. Safety of an intermediate heat exchanger vessel is ensured and observation from a control chamber is enabled. (N.H.)

Intermediate treatments

Science.gov (United States)

John R. Jones; Wayne D. Shepperd

1985-01-01

Intermediate treatments are those applied after a new stand is successfully established and before the final harvest. These include not only intermediate cuttings - primarily thinning - but also fertilization, irrigation, and protection of the stand from damaging agents.

Micro reactor and flow chemistry for industrial applications in drug discovery and development

NARCIS (Netherlands)

Tambarussi Baraldi, P.; Hessel, V.

2012-01-01

In this review, case studies focused on syntheses of active pharmaceutical ingredients, intermediates and lead compounds are reported employing micro reactors and continuous flow technology in areas such as medicinal chemistry, chemical development and manufacturing. The advantages of flow

Simplified simulation of an experimental fast reactor plant

International Nuclear Information System (INIS)

Fujii, Masaaki; Fujita, Minoru.

1978-01-01

Purposes of the simulation are to study the dynamic behavior of a liquid metal-cooled experimental fast breeder reactor plant and to design the control system of the reactor plant by modified-RAPID (Reactor and Plant Integrated Dynamics) computer program. As for the plant model, the Japan Experimental Fast Reactor ''Joyo'' was referred to approximately. This computer program is designed for the calculation of steady-state and transient temperatures in a FBR plant; which is described by a model consisting of the core, upper and lower plenums, an intermediate heat exchanger, an air dump heat exchanger, primary-secondary and tertiary coolant systems and connecting pipes. The basic equations are solved numerically by finite difference approximation. The mathematical model for an experimental FBR plant is useful for the design of the control system of FBR plants. The results of numerical simulation showed that the proportional change in the flow rates of the primary and secondary coolant loops provides good performance in relation to the stepped change in the power level. (J.P.N.)

Feasibility of a single-purpose reactor plant for district heating in Finland

International Nuclear Information System (INIS)

Tarjanne, R.; Vuori, S.; Eerikaeinen, L.; Saukkoriipi, L.

A feasibility study of a single-purpose reactor for district heating is presented. The reactor chosen is of an ordinary pressurized water reactor type with a thermal output of 100 to 200 MW. Primary circuit steam generators employed in ordinary PWR's are replaced by water-water heat exchangers. For safety reasons an intermediate circuit separates the primary from the network water. The conditions of the district heating systems in Finland were taken into account, which led to the choice of an average temperature of 160 0 C for the reactor coolant and a pressure of 13.5 bar. This, coupled with minimal control requirements helped design a considerably simple reactor plant. On condition, the reactor satisfies the basic heat demand in a district heating system, the effective annual full-power operation time of the heating reactor is from 5000 h to 7000 h. Economic comparisons indicated that the heating reactor may be competitive if the operation period is of this order. As the reactor has to be sited near the heat consumption area for reasons of economy, the safety aspects are of major importance and may in themselves preclude the realization of the heating idea. (author)

Neutron spectrum determination by activation method in fast neutron fields at the RB reactor

International Nuclear Information System (INIS)

Sokcic-Kostic, M.; Pesic, M.; Antic, D.

1994-01-01

The fast neutron fields of the RB reactor are presented in this paper. The activation method for spectrum determination is described and explained. The obtained results for intermediate and fast spectrum are given and discussed. (author)

Status of fast breeder reactor development in the United States of America - April 1985

International Nuclear Information System (INIS)

Horton, K.E.

1986-01-01

In order to provide a continuum of development for liquid metal reactors, the U.S. program has been reshaped into two portions -- advanced converter reactor technology including the Liquid Metal Reactor for the near and intermediate term, and the Liquid Metal Fast Breeder Reactor for deployment in the twenty-first century. The focused research and development program is directed at innovative means to improve economics, provide inherent safety, and to meet the needs of the ultimate user, the utilities. Work in the fuels and materials, and reprocessing areas is being continued to support eventual deployment. A major factor in successful deployment will be the effectiveness of international collaboration in reducing costs and duplication of efforts

Thermal-structural response of EBR-II major components under reactor operational transients

International Nuclear Information System (INIS)

Chang, L.K.; Lee, M.J.

1983-01-01

Until recently, the LMFBR safety research has been focused primarily on severe but highly unlikely accident, such as hypothetical-core-disruptive accidents (HCDA's), and not enough attention has been given to accident prevention, which is less severe but more likely sequence. The objective of the EBR-II operational reliability testing (ORT) is to demonstrate that the reactor can be designed and operated to prevent accident. A series of mild duty cycles and overpower transients were designed for accident prevention tests. An assessment of the EBR-II major plant components has been performed to assure structural integrity of the reactor plant for the ORT program. In this paper, the thermal-structural response and structural evaluation of the reactor vessel, the reactor-vessel cover, the intermediate heat exchanger (IHX) and the superheater are presented

Evaluation of alternative fluids for SFR intermediate loops

International Nuclear Information System (INIS)

Brissonneau, L.; Simon, N.; Baque, F.

2009-01-01

Among the Generation IV systems, Sodium Fast Reactors (SFR) are promising and benefit of considerable technological experience, but improvements are researched on safety approach and capital cost reduction. One of the main drawback to be solved by the standard SFR design is the proper management of the risk of leakage between the intermediate circuit filled with sodium and the energy conversion system using a water Rankine cycle. The limitation of this risk requires notably an early detection of water leakage to prevent a water-sodium reaction. One innovative solution consists in the replacement of the sodium in the secondary loops by an alternative liquid fluid, not or less reactive with water. This alternative fluid might also allow innovative designs, e.g. intermediate heat exchanger and steam generator grouped in the same component. CEA, Areva NP and EdF have joined in a working group in order to evaluate different 'alternative fluids' that might replace sodium. A first selection retained seven fluids on the basis of 'required properties' as large operating range (low melting point, high boiling point ...), fluid cost and availability, acceptable corrosion at SFR working temperature. These are three bismuth alloys, two nitrate salts, one hydroxide melt and sodium with nanoparticles of nickel. Then, it was decided to evaluate these fluids through a multi-criteria analysis in order to quantify advantages and drawbacks of each fluid and to compare them with sodium. Lack of knowledge, impact on materials, design, working conditions and reactor availability should be emphasized by this analysis, in order to provide sound arguments for a research program on one or two promising fluids. A global note is given to each fluid by evaluating them with respect to 'grand criteria', weighted differently according to their importance. The grand criteria are : thermal properties, reactivity with structures, reactivity with other fluids (air, water, sodium), chemistry control

Reactors with pressure vessel in pre-stressed concrete

International Nuclear Information System (INIS)

Devillers, Christian; Lafore, Pierre

1964-12-01

After having proposed a general description of the evolution of the general design of reactors with a vessel in pre-stressed concrete, this report outlines the interest of this technical solution of a vessel in pre-stressed concrete with integrated exchangers, which is to replace steel vessel. This solution is presented as much safer. The authors discuss the various issues related to protection: inner and outer biological protection of the vessel, material protection (against heating, steel irradiation, Wigner effect, and moderator radiolytic corrosion). They report the application of calculation methods: calculation of vessel concrete heating, study of the intermediate zone in integrated reactors, neutron spectrum and flows in the core of a graphite pile

AERE contracts with DoE on the treatment and disposal of intermediate level wastes

International Nuclear Information System (INIS)

Partridge, B.A.

1984-11-01

Reports are presented on work on the following topics concerned with the treatment and disposal of intermediate-level radioactive wastes: comparative evaluation of α and β γ irradiated medium level waste forms; modelling and characterisation of intermediate level waste forms based on polymers; optimisation of processing parameters for polymer and bitumen modified cements; α damage in non-reference waste form matrix materials; leaching mechanisms and modelling; inorganic ion exchange treatment of medium active effluents; electrical processes for the treatment of medium active liquid waste; fast reactor fuel element cladding; dissolver residues; effects of radiation on the properties of cemented MTR waste forms; equilibrium leach testing of cemented MTR waste forms; radiolytic oxidation of radionuclides; immobilisation of liquid organic waste; quality control, non-conformances and corrective action. (U.K.)

Preapplication safety evaluation report for the Power Reactor Innovative Small Module (PRISM) liquid-metal reactor. Final report

Energy Technology Data Exchange (ETDEWEB)

Donoghue, J.E.; Donohew, J.N.; Golub, G.R.; Kenneally, R.M.; Moore, P.B.; Sands, S.P.; Throm, E.D.; Wetzel, B.A. [Nuclear Regulatory Commission, Washington, DC (United States). Associate Directorate for Advanced Reactors and License Renewal

1994-02-01

This preapplication safety evaluation report (PSER) presents the results of the preapplication desip review for die Power Reactor Innovative Small Module (PRISM) liquid-mew (sodium)-cooled reactor, Nuclear Regulatory Commission (NRC) Project No. 674. The PRISM conceptual desip was submitted by the US Department of Energy in accordance with the NRC`s ``Statement of Policy for the Regulation of Advanced Nuclear Power Plants`` (51 Federal Register 24643). This policy provides for the early Commission review and interaction with designers and licensees. The PRISM reactor desip is a small, modular, pool-type, liquid-mew (sodium)-cooled reactor. The standard plant design consists of dim identical power blocks with a total electrical output rating of 1395 MWe- Each power block comprises three reactor modules, each with a thermal rating of 471 MWt. Each module is located in its own below-grade silo and is co to its own intermediate heat transport system and steam generator system. The reactors utilize a metallic-type fuel, a ternary alloy of U-Pu-Zr. The design includes passive reactor shutdown and passive decay heat removal features. The PSER is the NRC`s preliminary evaluation of the safety features in the PRISM design, including the projected research and development programs required to support the design and the proposed testing needs. Because the NRC review was based on a conceptual design, the PSER did not result in an approval of the design. Instead it identified certain key safety issues, provided some guidance on applicable licensing criteria, assessed the adequacy of the preapplicant`s research and development programs, and concluded that no obvious impediments to licensing the PRISM design had been identified.

Safety and international development of small modular reactors (SMR). A study of GRS

Energy Technology Data Exchange (ETDEWEB)

Buchholz, Sebastian; Kruessenberg, Anne; Schaffrath, Andreas [Gesellschaft fuer Anlagen- und Reaktorsicherheit mbH (GRS), Garching (Germany). Bereich Reaktorsicherheitsforschung

2015-11-15

The abbreviation SMR stands for Small Modular Reactor and describes reactors with low power output. One reactor module, composed of primary, secondary and, where necessary, intermediate circuit and auxiliary systems, may be transported to the construction site as a whole or in few parts only and can therefore be connected quickly to the grid. Various modules can form a larger nuclear power plant and additional modules may be added one by one, while the others are in operation. Designers develop SMR for the deployment mainly in remote, sparsely populated areas or near cities respectively. SMR may provide in both cases electricity, district heating and potable water.

Intermediate heat exchanger project for Super Phenix

International Nuclear Information System (INIS)

Roumailhac, J.; Desir, D.

1975-01-01

The Super Phenix (1200 MWe) intermediate heat exchangers are derived directly from those of Phenix (250 MWe). The intermediate exchangers are housed in the reactor vessel annulus: as this annulus must be of the smallest volume possible, these IHX are required to work at a high specific rating. The exchange surface is calculated for nominal conditions. A range is then defined, consistent with the above requirements and throughout which the ratio between bundle thickness and bundle length remains acceptable. Experimental technics and calculations were used to determine the number of tube constraint systems required to keep the vibration amplitude within permissible limits. From a knowledge of this number, the pressure drop produced by the primary flow can be calculated. The bundle geometry is determined together with the design of the corresponding tube plates and the way in which these plates should be joined to the body of the IHX. The experience (technical and financial) acquired in the construction of Phenix is then used to optimize the design of the Super Phenix project. An approximate definition of the structure of the IHX is obtained by assuming a simplified load distribution in the calculations. More sophisticated calculations (e.g. finite element method) are then used to determine the behaviour of the different points of the IHX, under nominal and transient conditions

Modeling the behavior of metallic fast reactor fuels during extended transients

International Nuclear Information System (INIS)

Kramer, J.M.; Liu, Y.Y.; Billone, M.C.; Tsai, H.C.

1992-01-01

Passive safety features in the metal-fueled Integral Fast Reactor (IFR) make it possible to avoid core damage for extended time periods even when automatic scram systems fail to operate or heat removal systems are severely degraded. The time scale for these transients are intermediate between those that have traditionally been analyzed in fast reactor safety assessments and those of normal operation. Consequently, it has been necessary to validate models and computer codes (FPIN2 and LIFE-METAL) for application to this time regime. Results from out-of-reactor Whole Pin Furnace tests are being used for this purpose. Pretest predictions for tests FM-1 through FM-6 have been performed and calculations have been compared with the experimental measurements

Steam reforming of heptane in a fluidized bed membrane reactor

Science.gov (United States)

Rakib, Mohammad A.; Grace, John R.; Lim, C. Jim; Elnashaie, Said S. E. H.

n-Heptane served as a model compound to study steam reforming of naphtha as an alternative feedstock to natural gas for production of pure hydrogen in a fluidized bed membrane reactor. Selective removal of hydrogen using Pd 77Ag 23 membrane panels shifted the equilibrium-limited reactions to greater conversion of the hydrocarbons and lower yields of methane, an intermediate product. Experiments were conducted with no membranes, with one membrane panel, and with six panels along the height of the reactor to understand the performance improvement due to hydrogen removal in a reactor where catalyst particles were fluidized. Results indicate that a fluidized bed membrane reactor (FBMR) can provide a compact reformer for pure hydrogen production from a liquid hydrocarbon feedstock at moderate temperatures (475-550 °C). Under the experimental conditions investigated, the maximum achieved yield of pure hydrogen was 14.7 moles of pure hydrogen per mole of heptane fed.

Research nuclear reactor RA, Annual Report 2001

International Nuclear Information System (INIS)

Sotic, O.

2002-01-01

During 2001, activities at the RA research nuclear reactor in were performed according to the Contract about financing of the RA reactor for the period January-December 2001, signed by the Ministry of Science, technology and development of the Republic of Serbia. RA reactor was not operated since shutdown in August 1984. Although, the most of the planned reconstruction activities were finished until 1991, the most important, which was concerned with exchange of the reactor instrumentation, financed by the IAEA, was interrupted due to international sanctions imposed on the country. Since 1992, all the renewal and reconstruction activities were ceased. Continuous aging and degradation of the equipment and facilities demand decision making about the future status of the Ra reactor. Until this decision is made it is an obligation to maintain control and maintenance of ventilation system, power supply, internal transportation system, spent fuel storage, hot cells, electronic fuel surveillance system, and part of the stationary dosimetry system. In 2001, apart from the mentioned activities, actions were undertaken related to maintenance of the reactor building and installations. The most important tasks fulfilled were: protection of the roof of the ventilation system building, purchase and installing the fire protection system and twelve new battery cells in the reactor building. There were no actions concerned with improvement of the conditions for intermediate spent fuel storage. With the support of IAEA, actions were initiated for possible transport of the spent fuel tu Russia. At the end of 2001, preparations were started for possible future decommissioning of the RA reactor. After, renewal of the membership of our country in the IAEA, the Government of Yugoslavia has declared its attitude about the intention of RA reactor decommissioning at the General Conference in September 2001 [sr

Assessment of high temperature nuclear energy storage systems for the production of intermediate and peak-load electric power

International Nuclear Information System (INIS)

Fox, E.C.; Fuller, L.C.; Silverman, M.D.

1977-01-01

Increased cost of energy, depletion of domestic supplies of oil and natural gas, and dependence on foreign suppliers, have led to an investigation of energy storage as a means to displace the use of oil and gas presently being used to generate intermediate and peak-load electricity. Dedicated nuclear thermal energy storage is investigated as a possible alternative. An evaluation of thermal storage systems is made for several reactor concepts and economic comparisons are presented with conventional storage and peak power producing systems. It is concluded that dedicated nuclear storage has a small but possible useful role in providing intermediate and peak-load electric power

Gas cooled reactor decommissioning. Packaging of waste for disposal in the United Kingdom deep repository

International Nuclear Information System (INIS)

Barlow, S.V.; Wisbey, S.J.; Wood, P.

1998-01-01

United Kingdom Nirex Limited has been established to develop and operate a deep underground repository for the disposal of the UK's intermediate and certain low level radioactive waste. The UK has a significant Gas Cooled Reactor (GCR) programme, including both Magnox and AGR (Advanced Gas-cooled Reactor) capacity, amounting to 26 Magnox reactors, 15 AGR reactors as well as research and prototype reactor units such as the Windscale AGR and the Windscale Piles. Some of these units are already undergoing decommissioning and Nirex has estimated that some 15,000 m 3 (conditioned volume) will come forward for disposal from GCR decommissioning before 2060. This volume does not include final stage (Stage 3) decommissioning arisings from commercial reactors since the generating utilities in the UK are proposing to adopt a deferred safe store strategy for these units. Intermediate level wastes arising from GCR decommissioning needs to be packaged in a form suitable for on-site interim storage and eventual deep disposal in the planned repository. In the absence of Conditions for Acceptance for a repository in the UK, the dimensions, key features and minimum performance requirements for waste packages are defined in Waste Package Specifications. These form the basis for all assessments of the suitability of wastes for disposal, including GCR wastes. This paper will describe the nature and characteristics of GCR decommissioning wastes which are intended for disposal in a UK repository. The Nirex Waste Package Specifications and the key technical issues, which have been identified when considering GCR decommissioning waste against the performance requirements within the specifications, are discussed. (author)

Intermediate-Size Inducer Pump design report. [LMFBR

Energy Technology Data Exchange (ETDEWEB)

Boardman, T.J.

1979-06-15

This report summarizes the mechanical, structural, and hydrodynamic design of the Intermediate-Size Inducer Pump (ISIP). The design was performed under Atomics International's DOE Base Technology Program by the Atomics International and Rocketdyne Divisions of Rockwell International. The pump was designed to utilize the FFTF prototype pump frame as a test vehicle to test the inducer, impeller, and diffuser plus necessary adapter hardware under simulated Large Scale Liquid Metal Fast Breeder Reactor service conditions. The report describes the design requirements including the purpose and objectives, and discusses those design efforts and considerations made to meet the requirements. Included in the report are appendices showing calculative methods and results. Also included are overall assembly and layout drawings plus some details used as illustrations for discussion of the design results and the results of water tests performed on a model of the inducer.

Basic philosophy of the safety design of the Toshiba boiling water reactor

International Nuclear Information System (INIS)

Sato, T.

1992-01-01

This paper discusses the safety design of the Toshiba Boiling Water Reactor (TOSBWR) which was created ∼8 years ago. The design concept is intermediate between conventional boiling water reactors (BWRs) and the advanced BWR (ABWR). It utilizes internal pumps and fine motion control rod drive, but the emergency core cooling system (ECCS) configuration is different from both conventional BWRs and the ABWR. The plant output is 1350 MW (electric). The design is based on two important philosophies: the positive cost reduction philosophy and the constant risk philosophy

Neutron spectrum determination by activation method in fast neutron fields at the RB reactors

International Nuclear Information System (INIS)

Sokcic-Kostic, M.S.; Pesic, M.P.; Antic, D.P.

1994-01-01

The fast neutron fields of the RB reactor are presented in this paper. The activation method for spectrum determination is described and explained. The obtained results for intermediate and fast spectrum are given and discussed. (authors). 7 refs., 3 tabs

Vessel core seismic interaction for a fast reactor

International Nuclear Information System (INIS)

Martelli, A.; Maresca, G.

1984-01-01

This report deals with the analysis carried out in collaboration between ENEA and NIRA for optimizing the iterative procedure applied for the evaluation of the effects of the vessel core dynamic interaction for a fast reactor in the case of a earthquake. In fact, as shown in a previous report the convergence of such procedure was very slow for the design solution adopted for the PEC reactor, i.e. with a core restraint plate located close to the top of the core elements. This study, although performed making use of preliminary data (the same of the cited previous report) demonstrates that the convergence is fast if a suitable linear core model is applied in the first iteration linear calculations carried out by NIRA, with an intermediate stiffness with respect to those corresponding to the two limit models previously assumed and increased damping coefficients. Thus, the optimized iterative procedures is now applied in the PEC reactor block seismic verification analysis

Nuclear reactor with scrammable part length rod

International Nuclear Information System (INIS)

Bevilacqua, F.

1979-01-01

A new part length rod is provided. It may be used to control xenon induced power oscillations but to contribute to shutdown reactivity when a rapid shutdown of the reactor is required. The part length rod consists of a control rod with three regions. The lower control region is a longer weaker active portion separated from an upper stronger shorter poison section by an intermediate section which is a relative non-absorber of neutrons. The combination of the longer weaker control section with the upper high worth poison section permits the part length rod of this to be scrammed into the core when a reactor shutdown is required but also permits the control rod to be used as a tool to control power distribution in both the axial and radial directions during normal operation

Acoustic emission results obtained from testing the ZB-1 intermediate scale pressure vessel

International Nuclear Information System (INIS)

Hutton, P.H.; Kurtz, R.J.; Pappas, R.A.; Dawson, J.F.; Dake, L.S.; Skorpik, J.R.

1985-09-01

Acoustic emission (AE) monitoring of flaw growth in an intermediate scale vessel during cyclic loading at 65 0 C and 288 0 C is described in this report. The report deals with background, methodology, and results. The work discussed is of major significance in a program supported by NRC to develop and demonstrate application of AE monitoring for continuous surveillance of reactor pressure boundaries to detect and evaluate growing flaws. Several areas of technical concern are addressed. Results support the feasibility of effective continuous monitoring

Projection to 2035 for the radioactive wastes of low and intermediate level in Mexico

International Nuclear Information System (INIS)

Paredes G, L.C.; Sanchez U, S.

2004-01-01

It is necessary to establish in few years a definitive warehouse for the radioactive waste of low and intermediate level, generated in the country and to satisfy the necessities of their confinement in the next ones 50 to 80 years. Therefore, it is required to be considered those volumes produced annually, those stored at the present and those estimated to medium and long term. The results of the simulation of 4 cases are presented, considering the operation from the 2 nuclear power reactors to 40 and 60 years, the use of the technology of current treatment and the use of super compaction of solids, as well as the importance in the taking of decision of the methodology for the dismantlement of each reactor to the finish of their useful life. At the moment the Nuclear Power Plant of Laguna Verde, produces an average of 250 m 3 /year of radioactive waste of low and intermediate level, constituted by solid dry wastes, humid solids and liquids. In the last 3 years, the power plant has reached an effectiveness of re utilization of effluents of 95%. On the other hand, in Mexico the non energetic applications of the radioisotopes, produce annually of the order of 20 m 3 /year of solid wastes, 280 m 3 /year of liquid wastes and 300 worn out radioactive sources. (Author)

Desk top calculation strategies for reactor analysis using Mathematica

International Nuclear Information System (INIS)

Kullberg, C.

1991-01-01

Mathematics is one of several recently developed equations analysis programs that is particularly well suited for solving a broad range of intermediate engineering problems. The objective of this paper is to demonstrate, using a couple of reactor-related examples, how Mathematica can be exploited as a user-friendly analysis tool to symbolically and numerically handle systems of algebraic and differential equations. 7 refs., 3 figs

Improved performance of parallel surface/packed-bed discharge reactor for indoor VOCs decomposition: optimization of the reactor structure

International Nuclear Information System (INIS)

Jiang, Nan; Hui, Chun-Xue; Li, Jie; Lu, Na; Shang, Ke-Feng; Wu, Yan; Mizuno, Akira

2015-01-01

The purpose of this paper is to develop a high-efficiency air-cleaning system for volatile organic compounds (VOCs) existing in the workshop of a chemical factory. A novel parallel surface/packed-bed discharge (PSPBD) reactor, which utilized a combination of surface discharge (SD) plasma with packed-bed discharge (PBD) plasma, was designed and employed for VOCs removal in a closed vessel. In order to optimize the structure of the PSPBD reactor, the discharge characteristic, benzene removal efficiency, and energy yield were compared for different discharge lengths, quartz tube diameters, shapes of external high-voltage electrode, packed-bed discharge gaps, and packing pellet sizes, respectively. In the circulation test, 52.8% of benzene was removed and the energy yield achieved 0.79 mg kJ −1 after a 210 min discharge treatment in the PSPBD reactor, which was 10.3% and 0.18 mg kJ −1 higher, respectively, than in the SD reactor, 21.8% and 0.34 mg kJ −1 higher, respectively, than in the PBD reactor at 53 J l −1 . The improved performance in benzene removal and energy yield can be attributed to the plasma chemistry effect of the sequential processing in the PSPBD reactor. The VOCs mineralization and organic intermediates generated during discharge treatment were followed by CO x selectivity and FT-IR analyses. The experimental results indicate that the PSPBD plasma process is an effective and energy-efficient approach for VOCs removal in an indoor environment. (paper)

The role of a small teaching reactor in education and training

International Nuclear Information System (INIS)

Bobek, L.M.; Mayer, J.A. Jr.

1992-01-01

It cannot be simply concluded that because an undergraduate nuclear engineering program has access to a higher power research reactor that the number of BS graduates will be proportionately larger than a program whose reactor operates at a much lower power level. What can be concluded is that although smaller in size and capability, low-power research reactors and the nuclear engineering programs they serve provide an important role in producing much-needed nuclear engineers and scientists at the undergraduate level. Designed and built by General Electric primarily as a teaching tool for nuclear engineering education, the nuclear reactor at Worcester Polytechnic Institute (WPI) first began operation in 1959. The reactor power level was upgraded from 1 to 10 kW in 1969, and its 20-yr operating license was renewed in 1983. With the support of DOE funds, the reactor was converted to low-enriched fuel in 1988. Under partial funding from the DOE University Reactor Instrumentation Program, the reactor control console will soon be replaced. Since a small research reactor is an ideal tool for providing basic and intermediate nuclear training, the incorporation of nuclear subjects into traditional disciplines will consequently enhance reactor facility usage. With its continued modernization, the WPI nuclear reactor facility will play a key role in meeting nuclear manpower needs while providing excellent and rewarding career opportunities for students in all disciplines for many years to come

Relationships between chemical oxygen demand (COD) components and toxicity in a sequential anaerobic baffled reactor/aerobic completely stirred reactor system treating Kemicetine

Energy Technology Data Exchange (ETDEWEB)

Sponza, Delia Teresa, E-mail: delya.sponza@deu.edu.tr [Department of Environmental Engineering, Faculty of Engineering, Dokuz Eyluel University, Buca Kaynaklar Campus, Tinaztepe, 35160 Izmir (Turkey); Demirden, Pinar, E-mail: pinar.demirden@kozagold.com [Environmental Engineer, Koza Gold Company, Environmental Department, Ovacik, Bergama Izmir (Turkey)

2010-04-15

In this study the interactions between toxicity removals and Kemicetine, COD removals, intermediate products of Kemicetine and COD components (CODs originating from slowly degradable organics, readily degradable organics, inert microbial products and from the inert compounds) were investigated in a sequential anaerobic baffled reactor (ABR)/aerobic completely stirred tank reactor (CSTR) system with a real pharmaceutical wastewater. The total COD and Kemicetine removal efficiencies were 98% and 100%, respectively, in the sequential ABR/CSTR systems. 2-Amino-1 (p-nitrophenil)-1,3 propanediol, l-p-amino phenyl, p-amino phenol and phenol were detected in the ABR as the main readily degradable inter-metabolites. In the anaerobic ABR reactor, the Kemicetin was converted to corresponding inter-metabolites and a substantial part of the COD was removed. In the aerobic CSTR reactor the inter-metabolites produced in the anaerobic reactor were completely removed and the COD remaining from the anerobic reactor was biodegraded. It was found that the COD originating from the readily degradable organics did not limit the anaerobic degradation process, while the CODs originating from the slowly degradable organics and from the inert microbial products significantly decreased the anaerobic ABR reactor performance. The acute toxicity test results indicated that the toxicity decreased from the influent to the effluent of the aerobic CSTR reactor. The ANOVA test statistics showed that there was a strong linear correlation between acute toxicity, CODs originating from the slowly degradable organics and inert microbial products. A weak correlation between acute toxicity and CODs originating from the inert compounds was detected.

Relationships between chemical oxygen demand (COD) components and toxicity in a sequential anaerobic baffled reactor/aerobic completely stirred reactor system treating Kemicetine

International Nuclear Information System (INIS)

Sponza, Delia Teresa; Demirden, Pinar

2010-01-01

In this study the interactions between toxicity removals and Kemicetine, COD removals, intermediate products of Kemicetine and COD components (CODs originating from slowly degradable organics, readily degradable organics, inert microbial products and from the inert compounds) were investigated in a sequential anaerobic baffled reactor (ABR)/aerobic completely stirred tank reactor (CSTR) system with a real pharmaceutical wastewater. The total COD and Kemicetine removal efficiencies were 98% and 100%, respectively, in the sequential ABR/CSTR systems. 2-Amino-1 (p-nitrophenil)-1,3 propanediol, l-p-amino phenyl, p-amino phenol and phenol were detected in the ABR as the main readily degradable inter-metabolites. In the anaerobic ABR reactor, the Kemicetin was converted to corresponding inter-metabolites and a substantial part of the COD was removed. In the aerobic CSTR reactor the inter-metabolites produced in the anaerobic reactor were completely removed and the COD remaining from the anerobic reactor was biodegraded. It was found that the COD originating from the readily degradable organics did not limit the anaerobic degradation process, while the CODs originating from the slowly degradable organics and from the inert microbial products significantly decreased the anaerobic ABR reactor performance. The acute toxicity test results indicated that the toxicity decreased from the influent to the effluent of the aerobic CSTR reactor. The ANOVA test statistics showed that there was a strong linear correlation between acute toxicity, CODs originating from the slowly degradable organics and inert microbial products. A weak correlation between acute toxicity and CODs originating from the inert compounds was detected.

n-Heptane cool flame chemistry: Unraveling intermediate species measured in a stirred reactor and motored engine

KAUST Repository

Wang, Zhandong; Chen, Bingjie; Moshammer, Kai; Popolan-Vaida, Denisia M.; Sioud, Salim; Shankar, Vijai; Vuilleumier, David; Tao, Tao; Ruwe, Lena; Brä uer, Eike; Hansen, Nils; Dagaut, Philippe; Kohse-Hö inghaus, Katharina; Raji, Misjudeen; Sarathy, Mani

2017-01-01

-OTMS, which has ultra-high mass resolving power and provides an accurate elemental C/H/O composition of the intermediate species. Furthermore, the results show that the species formed during the partial oxidation of n-heptane in the CFR engine are very similar

Lead-cooled flexible conversion ratio fast reactor

International Nuclear Information System (INIS)

Nikiforova, Anna; Hejzlar, Pavel; Todreas, Neil E.

2009-01-01

Lead-cooled reactor systems capable of accepting either zero or unity conversion ratio cores depending on the need to burn actinides or operate in a sustained cycle are presented. This flexible conversion ratio reactor is a pool-type 2400 MWt reactor coupled to four 600 MWt supercritical CO 2 (S-CO 2 ) power conversion system (PCS) trains through intermediate heat exchangers. The cores which achieve a power density of 112 kW/l adopt transuranic metallic fuel and reactivity feedbacks to achieve inherent shutdown in anticipated transients without scram, and lead coolant in a pool vessel arrangement. Decay heat removal is accomplished using a reactor vessel auxiliary cooling system (RVACS) complemented by a passive secondary auxiliary cooling system (PSACS). The transient simulation of station blackout (SBO) using the RELAP5-3D/ATHENA code shows that inherent shutdown without scram can be accommodated within the cladding temperature limit by the enhanced RVACS and a minimum (two) number of PSACS trains. The design of the passive safety systems also prevents coolant freezing in case all four of the PSACS trains are in operation. Both cores are also shown able to accommodate unprotected loss of flow (ULOF) and unprotected transient overpower (UTOP) accidents using the S-CO 2 PCS.

Large-signal, dynamic simulation of the slowpoke-3 nuclear heating reactor

International Nuclear Information System (INIS)

Tseng, C.M.; Lepp, R.M.

1983-07-01

A 2 MWt nuclear reactor, called SLOWPOKE-3, is being developed at the Chalk River Nuclear Laboratories (CRNL). This reactor, which is cooled by natural circulation, is designed to produce hot water for commercial space heating and perhaps generate some electricity in remote locations where the costs of alternate forms of energy are high. A large-signal, dynamic simulation of this reactor, without closed-loop control, was developed and implemented on a hybrid computer, using the basic equations of conservation of mass, energy and momentum. The natural circulation of downcomer flow in the pool was simulated using a special filter, capable of modelling various flow conditions. The simulation was then used to study the intermediate and long-term transient response of SLOWPOKE-3 to large disturbances, such as loss of heat sink, loss of regulation, daily load following, and overcooling of the reactor coolant. Results of the simulation show that none of these disturbances produce hazardous transients

Monte Carlo calculations for intermediate-energy standard neutron field

International Nuclear Information System (INIS)

Joneja, O.P.; Subbukutty, K.; Iyengar, S.B.D.; Navalkar, M.P.

Intermediate-Energy Standard Neutron Field (ISNF) which produces a well characterised spectrum in the energy range of interest for fast reactors including breeders, has been set up at NBS using thin enriched 235 U fission sources. A proposal has been made for setting up a similar facility at BARC using however, easily available natural U instead of enriched U sources, to start with. In order to simulate the neutronics of such a facility Monte Carlo method of calculations has been adopted and developed. The results of these calculations have been compared with those of NBS and it is found that there may be a maximum difference of 10% in spectrum characteristics for the two cases of using thick and thin fission sources. (K.B.)

Design of a power conversion system for an indirect cycle, helium cooled pebble bed reactor system

International Nuclear Information System (INIS)

Wang, C.; Ballinger, R.G.; Stahle, P.W.; Demetri, E.; Koronowski, M.

2002-01-01

A design is presented for the turbomachinery for an indirect cycle, closed, helium cooled modular pebble bed reactor system. The design makes use of current technology and will operate with an overall efficiency of 45%. The design uses an intermediate heat exchanger which isolated the reactor cycle from the turbomachinery. This design excludes radioactive fission products from the turbomachinery. This minimizes the probability of an air ingress accident and greatly simplifies maintenance. (author)

The integration of fast reactor to the fuel cycle in Slovakia

International Nuclear Information System (INIS)

Zajac, R.; Darilek, P.; Necas, V.

2009-01-01

A very topical problem of nuclear power is the fuel cycle back-end. One of the options is a LWR spent fuel reprocessing and a fissile nuclides re-use in the fast reactor. A large amount of spent fuel has been stored in the power plant intermediate storage during the operation of WWER-440 reactors in Slovakia. This paper is based on an analysis of Pu and minor actinides content in actual WWER-440 spent fuel stored in Slovakia. The next part presents the possibilities of reprocessing and Pu re-use in fast reactor under Slovak conditions. The fuel cycle consisting of the WWER-440 reactor, PUREX reprocessing plant and a sodium fast reactor was designed. The last section compares two parts of this fuel cycle: one is UOX cycle in WWER-440 reactor and the other is cycle in the fast reactor - SUPER PHENIX loaded with MOX fuel (Pu + Minor Actinides). The starting point is a single recycling of Pu from WWER-440 in the fission products. The next step is multi recycling of Pu in the fission products to obtain equilibrium cycle. This article is dealing with the solution of power production and fuel cycle indicators. All kinds of calculations were performed by computer code HELIOS 1.10. (Authors)

The integration of fast reactor to the fuel cycle in Slovakia

International Nuclear Information System (INIS)

Zajac, R.; Darilek, P.; Necas, V.

2009-01-01

A very topical problem of nuclear power is the fuel cycle back-end. One of the options is a LWR spent fuel reprocessing and a fissile nuclides re-use in the fast reactor. A large amount of spent fuel has been stored in the power plant intermediate storage during the operation of VVER-440 reactors in Slovakia. This paper is based on an analysis of Pu and minor actinides content in actual VVER-440 spent fuel stored in Slovakia. The next part presents the possibilities of reprocessing and Pu re-use in fast reactor under Slovak conditions. The fuel cycle consisting of the VVER-440 reactor, PUREX reprocessing plant and a sodium fast reactor was designed. The last section compares two parts of this fuel cycle: one is UOX cycle in VVER-440 reactor and the other is cycle in the fast reactor - SUPER PHENIX loaded with MOX fuel (Pu + Minor Actinides). The starting point is a single recycling of Pu from VVER-440 in the FR. The next step is multirecycling of Pu in the FR to obtain equilibrium cycle. This article is dealing with the solution of power production and fuel cycle indicators. All kinds of calculations were performed by computer code HELIOS 1.10. (authors)

Dismantling of the reactor block of the FRJ-1 research reactor (MERLIN); Abbau des Reaktorblocks des Forschungsreaktors FRJ-1 (MERLIN)

Energy Technology Data Exchange (ETDEWEB)

Stahn, B.; Matela, K.; Zehbe, C. [Forschungszentrum Juelich GmbH (Germany); Poeppinghaus, J. [Gesellschaft fuer Nuklear-Service mbH, Essen (Germany); Cremer, J. [Siempelkamp Nukleartechnik GmbH, Heidelberg (Germany)

2003-07-01

By the end of 1998 the complete secondary cooling system and the major part of the primary cooling system were dismantled. Furthermore, the experimental devices, including a rabbit system conceived as an in-core irradiation device, were disassembled and disposed of. In total, approx. 65 t of contaminated and/or activated material as well as approx. 70 t of clearance-measured material were disposed of within the framework of these activities. The dismantling of the coolant loops and experimental devices was followed in 2000 by the removal of the reactor tank internals and the subsequent draining of the reactor tank water. The reactor tank internals were essentially the core support plate, the core box, the flow channel and the neutron flux bridges (s. Fig. 2, detailed reactor core). All components consisted of aluminium, the connecting elements such as bolts and nuts, however, of stainless steel. Due to the high activation of the core internals, disassembly had to be remotely controlled under water. All removal work was carried out from a tank intermediate floor (s. Fig. 2). These activities, which served for preparing the dismantling of the reactor block, were completed in summer 2001. The waste parts arising were transferred to the Service Department for Decontamination of the Research Centre. This included approx. 2.5 t of waste parts with a total activity of approx. 8 x 10{sup 11} Bq. (orig.)

The Advanced High-Temperature Reactor (AHTR) for Producing Hydrogen to Manufacture Liquid Fuels

International Nuclear Information System (INIS)

Forsberg, C.W.; Peterson, P.F.; Ott, L.

2004-01-01

Conventional world oil production is expected to peak within a decade. Shortfalls in production of liquid fuels (gasoline, diesel, and jet fuel) from conventional oil sources are expected to be offset by increased production of fuels from heavy oils and tar sands that are primarily located in the Western Hemisphere (Canada, Venezuela, the United States, and Mexico). Simultaneously, there is a renewed interest in liquid fuels from biomass, such as alcohol; but, biomass production requires fertilizer. Massive quantities of hydrogen (H2) are required (1) to convert heavy oils and tar sands to liquid fuels and (2) to produce fertilizer for production of biomass that can be converted to liquid fuels. If these liquid fuels are to be used while simultaneously minimizing greenhouse emissions, nonfossil methods for the production of H2 are required. Nuclear energy can be used to produce H2. The most efficient methods to produce H2 from nuclear energy involve thermochemical cycles in which high-temperature heat (700 to 850 C) and water are converted to H2 and oxygen. The peak nuclear reactor fuel and coolant temperatures must be significantly higher than the chemical process temperatures to transport heat from the reactor core to an intermediate heat transfer loop and from the intermediate heat transfer loop to the chemical plant. The reactor temperatures required for H2 production are at the limits of practical engineering materials. A new high-temperature reactor concept is being developed for H2 and electricity production: the Advanced High-Temperature Reactor (AHTR). The fuel is a graphite-matrix, coated-particle fuel, the same type that is used in modular high-temperature gas-cooled reactors (MHTGRs). The coolant is a clean molten fluoride salt with a boiling point near 1400 C. The use of a liquid coolant, rather than helium, reduces peak reactor fuel and coolant temperatures 100 to 200 C relative to those of a MHTGR. Liquids are better heat transfer fluids than gases

Nuclear heat source component design considerations for HTGR process heat reactor plant concept

International Nuclear Information System (INIS)

McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

1982-05-01

The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported [via an intermediate heat exchanger (IHX)] to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

Transporting spent fuel and reactor waste in Sweden experience from 5 years of operation

International Nuclear Information System (INIS)

Dybeck, P.; Gustafsson, B.

1990-01-01

This paper reports that since the Final Repository for Reactor Waste, SFR, was taken into operation in 1988, the SKB sea transportation system is operating at full capacity by transporting spent fuel and now also reactor waste from the 12 Swedish reactors to CLAB and SFR. Transports from the National Research Center, Studsvik to the repository has recently also been integrated in the system. CLAB, the central intermediate storage for spent fuel, has been in operation since 1985. The SKB Sea Transportation System consists today of the purpose built ship M/s Sigyn, 10 transport casks for spent fuel, 2 casks for spent core components, 27 IP-2 shielded steel containers for reactor waste and 5 terminal vehicles. During an average year about 250 tonnes of spent fuel and 3 -- 4000 m 3 of reactor waste are transported to CLAB and SFR respectively, corresponding to around 30 sea voyages

Intermediate probabilistic safety assessment approach for safety critical digital systems

International Nuclear Information System (INIS)

Taeyong, Sung; Hyun Gook, Kang

2001-01-01

Even though the conventional probabilistic safety assessment methods are immature for applying to microprocessor-based digital systems, practical needs force to apply it. In the Korea, UCN 5 and 6 units are being constructed and Korean Next Generation Reactor is being designed using the digital instrumentation and control equipment for the safety related functions. Korean regulatory body requires probabilistic safety assessment. This paper analyzes the difficulties on the assessment of digital systems and suggests an intermediate framework for evaluating their safety using fault tree models. The framework deals with several important characteristics of digital systems including software modules and fault-tolerant features. We expect that the analysis result will provide valuable design feedback. (authors)

Intermediate neutron spectrum problems and the intermediate neutron spectrum experiment

International Nuclear Information System (INIS)

Jaegers, P.J.; Sanchez, R.G.

1996-01-01

Criticality benchmark data for intermediate energy spectrum systems does not exist. These systems are dominated by scattering and fission events induced by neutrons with energies between 1 eV and 1 MeV. Nuclear data uncertainties have been reported for such systems which can not be resolved without benchmark critical experiments. Intermediate energy spectrum systems have been proposed for the geological disposition of surplus fissile materials. Without the proper benchmarking of the nuclear data in the intermediate energy spectrum, adequate criticality safety margins can not be guaranteed. The Zeus critical experiment now under construction will provide this necessary benchmark data

Activity report of Reactor Physics Division - 1989

International Nuclear Information System (INIS)

1990-01-01

The highlights of the various studies carried out during the year 1989 in Reactor Physics Division are presented in this report in the form of summaries. The topics are organised under the following subjects: (1) nuclear data evaluation, processing and validation, (2) core physics and analysis, (3) reacto r kinetics and safety analysis, (4) noise analysis, and radiation transport and shielding. It is observed that with the restart and operation of FBTR at low power for some time, some of the low power physics experiments were completed and plans and procedures for the remaining physics experiments at intermediate and high power (upto 10 MWt) have been prepared. The lists of publications by the members of Division and the Reactor Physics Seminars held during the year 19 89, are included at the end of the report. (author). refs., figs., tabs

Low and intermediate level radioactive waste in Mexico

International Nuclear Information System (INIS)

Paredes, L.C.; Ortiz, J.R.; Sanchez, S.

2002-01-01

Currently, it is necessary to establish, in a few years, a definitive repository for low and intermediate level radioactive waste in order to satisfy the necessities of Mexico for the next 50 years. Consequently, it is required to estimate the volumes of the radioactive waste generated annually, the stored volumes to-date and their projection to medium-term. On this subject, the annual average production of low and intermediate level radioactive waste from the electricity production by means of nuclear power reactors is 250 m 3 /y which consist of humid and dry solid waste from the 2 units of the Laguna Verde Nuclear Power plant having a re-use efficiency of effluents of 95%. On the other hand, the applications in medicine, industry and research generate 20 m 3 /y of solid waste, 280 m 3 /y of liquid waste and approximately 10 m 3 /y from 300 spent sealed radioactive sources. The estimation of the total volume of these waste to the year 2035 is 17500 m 3 corresponding to the 46% of the volume generated by the operation and maintenance of the 2 units of the Laguna Verde Nuclear Power plant, 34% to the decommissioning of these 2 units at the end of their useful life and 20% to the waste generated by applications in medicine, industry and research. (author)

Assessment of very high temperature reactors in process applications

International Nuclear Information System (INIS)

Jones, J.E. Jr.; Spiewak, I.; Gambill, W.R.

1976-01-01

In April 1974, the United States Energy Research and Development Administration (ERDA) authorized General Atomic Company, General Electric Company, and Westinghouse Astronuclear Laboratory to assess the available technology for producing process heat utilizing a very high temperature nuclear reactor (VHTR). The VHTR is defined as a gas-cooled graphite-moderated reactor. Oak Ridge National Laboratory has been given a lead role in evaluating the VHTR reactor studies and potential applications of the VHTR. Process temperatures up to the 760 to 871 0 C range appear to be achievable with near-term technology. The major development considerations are high temperature materials, the safety questions (especially regarding the need for an intermediate heat exchanger) and the process heat exchanger. The potential advantages of the VHTR over competing fossil energy sources are conservation of fossil fuels and reduced atmospheric impacts. Costs are developed for nuclear process heat supplied from a 3000-MW(th) VHTR. The range of cost in process applications is competitive with current fossil fuel alternatives

Heat exchanger for transfering heat produced in a high temperature reactor to an intermediate circuit gas

International Nuclear Information System (INIS)

Barchewitz, E.; Baumgaertner, H.

1985-01-01

The invention is concerned with improving the arrangement of a heat exchanger designed to transfer heat from the coolant gas circuit of a high temperature reactor to a gas which is to be used for a process heat plant. In the plant the material stresses are to be kept low at high differential pressures and temperatures. According to the invention the tube bundles designed as boxes are fixed within the heat exchanger closure by means of supply pipes having got loops. For conducting the hot gas the heat exchanger has got a central pipe leading out of the reactor vessel through the pod closure and having got only one point of fixation, lying in this closure. Additional advantageous designs are mentioned. (orig./PW)

Design and operation of a filter reactor for continuous production of a selected pharmaceutical intermediate

DEFF Research Database (Denmark)

Christensen, Kim Müller; Pedersen, Michael Jønch; Dam-Johansen, Kim

2012-01-01

in tetrahydrofuran solvent. The use of the filter reactor design was explored by examining the transferability of a synthesis step in a present full-scale semi-batch pharmaceutical production into continuous processing. The main advantages of the new continuous minireactor system, compared to the conventional semi...

Present status of high-temperature engineering test reactor (HTTR) program

International Nuclear Information System (INIS)

Tanaka, Toshiyuki; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru; Tobioka, Toshiaki

1994-01-01

The 30MWt HTTR is a high-temperature gas-cooled reactor (HTGR), with a maximum helium coolant temperature of 950degC at the reactor outlet. The construction of the HTTR started in March 1991, with first criticality to be followed in 1998 after commissioning testing. At present the HTTR reactor building (underground part) and its containment vessel have been almost completed and its main components, such as a reactor pressure vessel (RPV), an intermediate heat exchanger, hot gas pipings and graphite core structures, are now manufacturing at their factories at the target of their installation starting in 1994. The project is intended to establish and upgrade the technology basis necessary for HTGR developments. Japan Atomic Energy Research Institute (JAERI) also plans to conduct material and fuel irradiation tests as an innovative basic research after attaining rated power and coolant temperature. Innovative basic researches are now in great request. The paper describes major features of HTTR, present status of its construction and research and test using HTTR. (author)

Present status of High-Temperature engineering Test Reactor (HTTR) program

International Nuclear Information System (INIS)

Tanaka, Toshiyuki; Baba, Osamu; Shiozawa, Shusaku; Okubo, Minoru; Tobioka, Toshiaki

1993-01-01

The 30MWt HTTR is a high-temperature gas-cooled reactor (HTGR), with a maximum helium coolant temperature of 950 deg C at the reactor outlet. The construction of the HTTR started in March 1991, with first criticality to be followed in 1998 after commissioning testing. At present the HTTR reactor building (underground part) and its containment vessel have been almost completed and its main components, such as a reactor pressure vessel (RPV), an intermediate heat exchanger, hot gas pipings and graphite core structures, are now manufacturing at their factories at the target of their installation starting in 1994. The project is intended to establish and upgrade the technology basis necessary for HTGR developments. Japan Atomic Energy Research Institute (JAERI) also plans to conduct material and fuel irradiation tests as an innovative basic research after attaining rated power and coolant temperature. Innovative basic researches are now in great request. The paper describes major features of HTTR, present status of its construction and research and test plan using HTTR. (author)

Characterization of intermediate products of solar photocatalytic degradation of ranitidine at pilot-scale.

Science.gov (United States)

Radjenović, Jelena; Sirtori, Carla; Petrović, Mira; Barceló, Damià; Malato, Sixto

2010-04-01

In the present study the mechanisms of solar photodegradation of H(2)-receptor antagonist ranitidine (RNTD) were studied in a well-defined system of a pilot plant scale Compound Parabolic Collector (CPC) reactor. Two types of heterogeneous photocatalytic experiments were performed: catalysed by titanium-dioxide (TiO(2)) semiconductor and by Fenton reagent (Fe(2+)/H(2)O(2)), each one with distilled water and synthetic wastewater effluent matrix. Complete disappearance of the parent compounds and discreet mineralization were attained in all experiments. Furthermore, kinetic parameters, main intermediate products, release of heteroatoms and formation of carboxylic acids are discussed. The main intermediate products of photocatalytic degradation of RNTD have been structurally elucidated by tandem mass spectrometry (MS(2)) experiments performed at quadrupole-time of flight (QqToF) mass analyzer coupled to ultra-performance liquid chromatograph (UPLC). RNTD displayed high reactivity towards OH radicals, although a product of conduction band electrons reduction was also present in the experiment with TiO(2). In the absence of standards, quantification of intermediates was not possible and only qualitative profiles of their evolution could be determined. The proposed TiO(2) and photo-Fenton degradation routes of RNTD are reported for the first time. (c) 2010 Elsevier Ltd. All rights reserved.

Anaerobic stabilization and conversion of transformed intermediates of antibiotic pharmaceutical effluent in a fluidized bed reactor.

Science.gov (United States)

Tamijevendane, S; Saravanane, R; Rajesh, R; Sivacoumar, R

2011-07-01

The formulation and implementation of regulatory standards for the ultimate disposal and reuse of transformed products of antibiotic drugs and solvents have been a pending issue in the waste management of pharmaceutical industries especially in the developing countries like India. A case study has been identified and the current issues in one of the major pharmaceutical industry (manufacturing cephalosporin drugs) located in Chennai, India, has been discussed for the possible implementation of anaerobically transformed intermediates of antibiotic pharmaceutical waste sludge. The objective of the study was to determine the effect of bioaugmentation on the convertibility of anaerobically transformed intermediates of antibiotic pharmaceutical waste sludge into residuals and biocompost. Cephalosporin is a common name refers to cephradine (C16H19N3O4S) and cephalexin (C16H17N3O4S.H2O). Based on the critical examination of results, the industry is looking for the alternatives of either direct disposal of 7-amino-3-deacetoxycephalosporanic acid (7-ADCA) and phenyl acetic acid or for further degradation and disposal, which will essentially require additional cost and maintenance. The present regulatory standard implemented in India does not envisage such disposal alternatives and hence this would invite suggestions and recommendations of the expertise for the possible implementation on the pending issue in the antibiotic based pharmaceutical industries. The presence of cephalosporin increases total strength (Chemical Oxygen Demand) of the effluent and indirectly increases the cost of the treatment. Hence the biotransformation of cephalosporin either alone or in combination with other energetic compounds, offers the potential for an economical and environment friendly disposal alternative for the anaerobically transformed intermediates of antibiotic pharmaceutical waste sludge.

Study on conceptual design system of tritium production fusion reactor

International Nuclear Information System (INIS)

He Kaihui

2004-11-01

Conceptual design of an advanced tritium production reactor based on spherical torus, which is intermediate application of fusion energy, was presented. Different from traditional tokamak tritium production reactor design, advanced plasma physics performance and compact structural characteristics of ST were used to minimize tritium leakage and to maximize tritium breeding ratio with arrangement of tritium production blankets as possible as it can within vacuum vessel in order to produce 1 kg excess tritium except self-sufficient plasma core, corresponding plant availability 40% or more. Based on 2D neutronics calculation, preliminary conceptual design of ST-TPR was presented. Besides systematical analyses; design risk, uncertainty and backup are introduced generally for the backgrounds of next detailed conceptual design. (author)

Hydrogen production by water-splitting using heat supplied by a high-temperature reactor

International Nuclear Information System (INIS)

Courvoisier, P.; Rastoin, J.; Titiliette, Z.

1976-01-01

Some aspects of the use of heat of nuclear origin for the production of hydrogen by water-splitting are considered. General notions pertaining to the yield of chemical cycles are discussed and the heat balance corresponding to two specific processes is evaluated. The possibilities of high temperature reactors, with respect to the coolant temperature levels, are examined from the standpoint of core design and technology of some components. Furthermore, subject to a judicious selection of their characteristics, these reactors can lead to excellent use of nuclear fuel. The coupling of the nuclear reactor with the chemical plant by means of a secondary helium circuit gives rise to the design of an intermediate heat exchanger, which is an important component of the overall installation. (orig.) [de

The SILENE reactor: an instrument suitable for studying the impact of intermediate and high radiation doses

International Nuclear Information System (INIS)

Verray, B.; Leo, Y.; Fouillaud, P.

2002-01-01

Designed in 1974 to study the phenomenology and consequences of a criticality accident, the SILENE experimental reactor, an intense source of mixed neutron and gamma radiation, is also suited to radiobiological studies. (author)

High and very high temperature reactor research for multipurpose energy applications (RAPHAEL)

International Nuclear Information System (INIS)

Hittner, Dominique; Bogusch, Edgar; Fuetterer, Michal; De Groot, Sander; Ruer, Jacques

2010-01-01

The sections of the contribution are as follows: Achievements and future R and D needs in baseline technologies (Fuel and fuel cycle; Materials and components - reactor vessel, intermediate heat exchanger, graphite internals; Safety; Computer code qualification; Waste management; Education and training; International dimension of the European R and D); Towards application; and The path forward (P.A.)

The generation of denatured reactor plutonium by different options of the fuel cycle

Energy Technology Data Exchange (ETDEWEB)

Broeders, C.H.M.; Kessler, G. [Inst. for Neutron Physics and Reactor Technology, Research Center Karlsruhe (Germany)

2006-11-15

Denatured (proliferation resistant) reactor plutonium can be generated in a number of different fuel cycle options. First denatured reactor plutonium can be obtained if, instead of low enriched U-235 PWR fuel, re-enriched U-235/U-236 from reprocessed uranium is used (fuel type A). Also the envisaged existing 2,500 t of reactor plutonium (being generated world wide up to the year 2010), mostly stored in intermediate fuel storage facilities at present, could be converted during a transition phase into denatured reactor plutonium by the options fuel type B and D. Denatured reactor plutonium could have the same safeguards standard as present low enriched (<20% U-235) LWR fuel. It could be incinerated by recycling once or twice in PWRs and subsequently by multi-recycling in FRs (CAPRA type or IFRs). Once denatured, such reactor plutonium could remain denatured during multiple recycling. In a PWR, e.g., denatured reactor plutonium could be destroyed at a rate of about 250 kg/GWey. While denatured reactor plutonium could be recycled and incinerated under relieved IAEA safeguards, neptunium would still have to be monitored by the IAEA in future for all cases in which considerable amounts of neptunium are produced. (orig.)

Fuel handling grapple for nuclear reactor plants

International Nuclear Information System (INIS)

Rousar, D.L.

1992-01-01

This patent describes a fuel handling system for nuclear reactor plants. It comprises: a reactor vessel having an openable top and removable cover and containing therein, submerged in water substantially filling the reactor vessel, a fuel core including a multiplicity of fuel bundles formed of groups of sealed tube elements enclosing fissionable fuel assembled into units, the fuel handling system consisting essentially of the combination of: a fuel bundle handling platform movable over the open top of the reactor vessel; a fuel bundle handling mast extendable downward from the platform with a lower end projecting into the open top reactor vessel to the fuel core submerged in water; a grapple head mounted on the lower end of the mast provided with grapple means comprising complementary hooks which pivot inward toward each other to securely grasp a bail handle of a nuclear reactor fuel bundle and pivot backward away from each other to release a bail handle; the grapple means having a hollow cylindrical support shaft fixed within the grapple head with hollow cylindrical sleeves rotatably mounted and fixed in longitudinal axial position on the support shaft and each sleeve having complementary hooks secured thereto whereby each hook pivots with the rotation of the sleeve secured thereto; and the hollow cylindrical support shaft being provided with complementary orifices on opposite sides of its hollow cylindrical and intermediate to the sleeves mounted thereon whereby the orifices on both sides of the hollow cylindrical support shaft are vertically aligned providing a direct in-line optical viewing path downward there-through and a remote operator positioned above the grapple means can observe from overhead the area immediately below the grapple hooks

GENIUS & the Swedish Fast Reactor programme

International Nuclear Information System (INIS)

Wallenius, Janne

2012-01-01

Concluding remarks: Sweden’s growing fast reactor programme focuses on LFR technology, but we also participate in ASTRID. • An innovative facility for UN fabrication, an LBE thermal hydraulics loop and a lead corrosion facility are operational. • A plutonium fuel fabrication lab is is under installation (this week!) • The government is assessing the construction of ELECTRA-FCC, a centre for Gen IV-system R&D, at a tentative cost of ~ 140±20 M€. • Location: Oskarshamn (adjacent to intermediate repository) • Date of criticality: 2023 (best case) • Swedish participation in IAEA TWG-FR should intensify

European supercritical water cooled reactor

International Nuclear Information System (INIS)

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

2011-01-01

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

An optimized power conversion system concept of the integral, inherently-safe light water reactor

International Nuclear Information System (INIS)

Memmott, Matthew J.; Wilding, Paul R.; Petrovic, Bojan

2017-01-01

Highlights: • Three power conversion systems (PCS) for the I"2S-LWR are presented. • An optimization analyses was performed to evaluate these PCS alternatives. • The ideal PCS consists of 5 turbines, and obtains an overall efficiency of 35.7%. - Abstract: The integral, inherently safe light water reactor (I"2S-LWR) has been developed to significantly enhance passive safety capabilities while maintaining cost competitiveness relative to the current light water reactor (LWR) fleet. The compact heat exchangers of the I"2S-LWR preclude boiling of the secondary fluid, which decreases the probability of heat exchanger failure, but this requires the addition of a flash drum, which negatively affects the overall plant thermodynamic efficiency. A state of the art Rankine cycle is proposed for the I"2S-LWR to increase the thermodynamic efficiency by utilizing a flash drum with optimized operational parameters. In presenting this option for power conversion in the I"2S-LWR power plant, the key metric used in rating the performance is the overall net thermodynamic efficiency of the cycle. In evaluating the flash-Rankine cycle, three basic industrial concepts are evaluated, one without an intermediate pressure turbine, one with an intermediate turbine and one reheat stream, and one with an intermediate turbine and two reheat streams. For each configuration, a single-path multi-variable optimization is undertaken to maximize the thermal efficiency. The third configuration with an intermediate turbine and 2 reheat streams is the most effective concept, with an optimized efficiency of 35.7%.

Cooling of nuclear power stations with high temperature reactors and helium turbine cycles

International Nuclear Information System (INIS)

Foerster, S.; Hewing, G.

1977-01-01

On nuclear power stations with high temperature reactors and helium turbine cycles (HTR-single circuits) the residual heat from the energy conversion process in the primary and intermediate coolers is removed from cycled gas, helium. Water, which is circulated for safety reasons through a closed circuit, is used for cooling. The primary and intermediate coolers as well as other cooling equipment of the power plant are installed within the reactor building. The heat from the helium turbine cycle is removed to the environment most effectively by natural draught cooling towers. In this way a net plant efficiency of about 40% is attainable. The low quantities of residual heat thereby produced and the high (in comparison with power stations with steam turbine cycles) cooling agent pressure and cooling water reheat pressure in the circulating coolers enable an economically favourable design of the overall 'cold end' to be expected. In the so-called unit range it is possible to make do with one or two cooling towers. Known techniques and existing operating experience can be used for these dry cooling towers. After-heat removal reactor shutdown is effected by a separate, redundant cooling system with forced air dry coolers. The heat from the cooling process at such locations in the power station is removed to the environment either by a forced air dry cooling installation or by a wet cooling system. (orig.) [de

Seawater desalination plant using nuclear heating reactor coupled with MED process

Institute of Scientific and Technical Information of China (English)

无

2000-01-01

A small size plant for seawater desalination using nuclear heating reactor coupled with MED process was developed by the Institute of Nuclear Energy Technology, Tsinghua University, China. This seawater desalination plant was designed to supply potable water demand to some coastal location or island where both fresh water and energy source are severely lacking. It is also recommended as a demonstration and training facility for seawater desalination using nuclear energy. The design of small size of seawater desalination plant couples two proven technologies: Nuclear Heating Reactor (NHR) and Multi-Effect Destination (MED) process. The NHR design possesses intrinsic and passive safety features, which was demonstrated by the experiences of the project NHR-5. The intermediate circuit and steam circuit were designed as the safety barriers between the NHR reactor and MED desalination system. Within 10～200 MWt of the power range of the heating reactor, the desalination plant could provide 8000 to 150,000 m3/d of high quality potable water. The design concept and parameters, safety features and coupling scheme are presented.

Seawater desalination plant using nuclear heating reactor coupled with MED process

International Nuclear Information System (INIS)

Wu Shaorong; Dong Duo; Zhang Dafang; Wang Xiuzhen

2000-01-01

A small size plant for seawater desalination using nuclear heating reactor coupled with MED process was developed by the Institute of Nuclear Energy Technology, Tsinghua University, China. this seawater desalination plant was designed to supply potable water demand to some coastal location or island where both fresh water and energy source are severely lacking. It is also recommended as a demonstration and training facility for seawater desalination using nuclear energy. The design of small size of seawater desalination plant couples two proven technologies: Nuclear Heating Reactor (NHR) and Multi-Effect Destination (MED) process. The NHR design possesses intrinsic and passive safety features, which was demonstrated by the experiences of the project NHR-5. the intermediate circuit and steam circuit were designed as the safety barriers between the NHR reactor and MED desalination system. Within 10-200 MWt of the power range of the heating reactor, the desalination plant could provide 8000 to 150,000 m 3 /d of high quality potable water. The design concept and parameters, safety features and coupling scheme are presented

What is past is prologue: future directions in tokamak power reactor design research

International Nuclear Information System (INIS)

Conn, R.W.

1976-01-01

Conceptual tokamak power reactor designs over the last five years have provided us with many fundamental insights regarding tokamaks as fusion reactors. This first generation of studies has helped lay the groundwork upon which to build improvements in reactor design and begin a process of optimization. After reviewing the first generation of studies and the primary conclusions they produced, we discuss four current designs that are representative of present trends in this area of research. In particular, we discuss the trends towards reduced reactor size and higher neutron wall loadings. Moving in this direction requires new approaches to many subsystem designs. We describe new approaches and future directions in first wall and blanket designs that can achieve reliable operation and reasonable lifetime, the use of cryogenic but normal aluminum magnets for the pulsed coils in a tokamak, blanket designs that allow elimination of the intermediate loop, and low activity shields and toroidal field magnets. We close with a discussion of the future role of conceptual reactor design research and the need for close interaction with ongoing experiments in fusion technology

Intermediality and media change

OpenAIRE

2012-01-01

This book is about intermediality as an approach to analysing and understanding media change. Intermediality and Media Change is critical of technological determinism that characterises 'new media discourse' about the ongoing digitalization, framed as a revolution and creating sharp contrasts between old and new media. Intermediality instead emphasises paying attention to continuities between media of all types and privileges a comparative perspective on technological changes in media over ti...

Performance analysis of a repository for low and intermediate level reactor waste

International Nuclear Information System (INIS)

Vieno, T.; Nordman, H.; Vuori, S.; Peltonen, E.

1987-01-01

In Finland, utilities producing nuclear energy are responsible for the management of the radioactive waste, including final disposal. As regards low and intermediate level waste, the approach has been adopted to employ the power plant sites for locations of repositories. The repositories will be excavated at the depth of about 50 to 125 m in the bedrock of the two Finnish nuclear power plant sites, Loviisa and Olkiluoto. The performance analysis presented in this paper has been carried out for the Preliminary Safety Analysis Report (PSAR) of the Olkiluoto repository. A flexible model has been developed to estimate the release of radionuclides from waste packages and their subsequent transport through the engineered barriers in the repository. Gradual degradation of the engineered barriers is accounted for by altering parameters at fixed time points. Safety margins of the disposal concept have been evaluated by including disturbed evolution scenarios in the analysis. 13 references, 10 figures, 2 tables

Conceptual design of tritium production fusion reactor based on spherical torus

International Nuclear Information System (INIS)

He Kaihui; Huang Jinhua

2003-01-01

Conceptual design of an advanced tritium production fusion reactor based on spherical torus, which is intermediate application of fusion energy, was presented in this paper. Differing from the traditional tokamak tritium production reactor design, advanced plasma physics performance and compact structural characteristics of ST were used to minimize tritium leakage and maximize tritium breeding ratio with arrangement of tritium production blankets within vacuum vessel as possible in order to produce 1 kg excess tritium except need of self-sufficient plasma core with 40% or more corresponding plant availability. Based on 2D neutronics calculation, preliminary conceptual design of ST-TPR was presented, providing the backgrounds and reference for next detailed conceptual design

Inherent reactor power controller for a metal-fueled ALMR

International Nuclear Information System (INIS)

Wood, R.T.; Wilson, T.L. Jr.

1990-01-01

Inherent power control for metal-fueled ALMR designs involves using reactivity thermal feedback effects to control reactor power. This paper describes how, using classical control design techniques, a control system for normal load following maneuvers was deigned for a pool-type ALMR. This design provides active control of power removal in the balance of plant, direct control of selected primary and intermediate loop temperatures, and passive control of reactor power. The inherent stability of the strong, fast reactivity feedback effects bring heat production in the core into balance with the heat removal system temperatures, which are controlled to meet power demand. A simulation of the control system successfully responded to a 10% step change in power demand by changing power at an acceptable rate without causing large temperature fluctuations or exceeding thermal limits

A preliminary design study of a pool-type FBR 'ARES' eliminating intermediate heat transport systems

International Nuclear Information System (INIS)

Ueda, N.; Nishi, Y.; Kinoshita, I.; Yoshida, K.

2001-01-01

An innovative reactor concept 'ARES' (Advanced Reactor Eliminating Secondary system) is proposed to aim at reducing the construction cost of a liquid metal cooled fast breeder reactor (LMFBR). This concept is developed to show the ultimate cost down potential of LMFBR's at their commercial stage. The electrical output is 1500 MW, while the thermal output is 3900 MW. Main components of the primary cooling system are four electromagnetic pumps (EMP) and eight double-wall-tube steam generators (SG). Both of them are installed in a reactor vessel like pool type LMFBR's. An intermediate heat transport system which a previous LMFBR has it eliminated, main components of which are intermediate heat exchangers (IHX), secondary pumps and secondary piping. Further, a high reliable SG could decrease the occurrence of water leak accidents and reduce the related mitigation systems. In this study, structure concept, approach to embody a high reliable SG and accidents analyses are carried out. Flow path configuration is mainly discussed in investigation of the structure concept. In case of a water leak accident in a SG, the fault SG must be isolated to prevent a reaction production from flowing into the core. The measure to cut both inlet and outlet coolant flow paths by siphon-break mechanism is adopted to be consistent with the decay heat removal operation. The safety design approach of the double-wall-tube SG is investigated to limit the accident occurrence below 10 -7 (1/ry). A tube-to-tube weld is excluded from the reference design, because the welding process is too difficult and complicated to prevent adhesion of the double-wall-tube effectively. The reliability of the tube-to-tube-sheet was evaluated as 10 -10 (1/hr) for an inner tube and 10 -9 (1/hr) for an outer tube with reference to the failure experience of previous SG's. The failure must be detected within 60 to 120 minutes. Finally, a seamless U tube type of double-wall-tube SG is adopted. Transient events due to

Intermediate Fragment

DEFF Research Database (Denmark)

Kruse Aagaard, Anders

2015-01-01

This text and its connected exhibition are aiming to reflect both on the thoughts, the processes and the outcome of the design and production of the artefact ‘Intermediate Fragment’ and making as a contemporary architectural tool in general. Intermediate Fragment was made for the exhibition ‘Enga...... of realising an exhibition object was conceived, but expanded, refined and concretised through this process. The context of the work shown here is an interest in a tighter, deeper connection between experimentally obtained material knowledge and architectural design....

Governance-Default Risk Relationship and the Demand for Intermediated and Non-Intermediated Debt

Directory of Open Access Journals (Sweden)

Husam Aldamen

2012-09-01

Full Text Available This paper explores the impact of corporate governance on the demand for intermediated debt (asset finance, bank debt, non-bank private debt and non-intermediated debt (public debt in the Australian debt market. Relative to other countries the Australian debt market is characterised by higher proportions of intermediated or private debt with a lower inherent level of information asymmetry in that private lenders have greater access to financial information (Gray, Koh & Tong 2009. Our firm level, cross-sectional evidence suggests that higher corporate governance impacts demand for debt via the mitigation of default risk. However, this relationship is not uniform across all debt types. Intermediated debt such as bank and asset finance debt are more responsive to changes in governance-default risk relationship than non-bank and non-intermediated debt. The implication is that a firm’s demand for different debt types will reflect its governance-default risk profile.

Applications of the Dow TRIGA research reactor

International Nuclear Information System (INIS)

Kocher, C.W.; Quinn, T.J.; Krueger, D.A.

1982-01-01

The Dow TRIGA Mark I reactor is a one-hundred kilowatt nuclear reactor installed by General Atomics using the Torrey Pines reactor console, seventy-five used stainless-steel clad fuel elements and one new aluminium clad fuel element. The reactor is equipped with a forty-position rotating Lazy Susan in the reflector, a pneumatic transfer system with its terminal in the F-ring of the core, and a central thimble which can be used for irradiation of samples in the center of the core or which can be emptied of the shielding water to produce a beam of neutrons and gamma rays in the area atop the pool. Samples can also be irradiated in or near the core. There is no provision for pulsing this TRIGA reactor. The neutron activation analysis program uses the Dow TRIGA reactor as a source of thermal neutrons and a Kaman A711 generator as a source of 14-MeV neutrons. The associated counting equipment includes one Gel(Li) detector and two Nal(Tl) detectors, each using a 100-position sample changer and all interfaced to a Tracor-Northern TN-11 data acquisition and computing system, one Ge(Li) detector and its TN-11 system for the pneumatic transfer system and the beam tube experiments, and two NaKTl)detectors with a TN-4000 system used with the Kaman neutron generator. The activation analysis program gets samples from all parts of the manufacturing and research efforts at Dow: raw materials, intermediates, products, effluents, research samples, samples from customers who use Dow products, and environmental samples. This presentation is devoted to the progress made in the past year on the pneumatic transfer system and the renewed work on prompt gamma-ray spectroscopy including the extensive process of method validation

Overview of gas cooled reactors' applications with CATHARE

International Nuclear Information System (INIS)

Genevieve Geffraye; Fabrice Bentivoglio; Anne Messie; Alain Ruby; Manuel Saez; Nicolas Tauveron; Ola Widlund

2005-01-01

Full text of publication follows: For about four years, CEA has launched feasibility studies of future nuclear advanced systems in a consistent series of Gas Cooled Reactors (GCR) ranging from thermal reactors, as the Very High Temperature Reactor (VHTR) for the mid term, to fast reactors (GFR) for the long term. Thermal hydraulic performances are a key issue for the core design, the evaluation of the thermal stresses on the structures and the decay heat removal systems. This analysis requires a 1D code able to simulate the whole reactor, including the core, the vessel, the piping and the components (turbine, compressors, heat exchangers). CATHARE is the reference code developed and extensively validated in collaboration between CEA, EDF, IRSN and FRAMATOME-ANP for the French Pressurized Water Reactors. CATHARE has the capabilities to model a Gas Cooled Reactor using standard 0D and 1D modules with some adaptations to treat the specificities of the GCR designs. In this paper, the different adaptations are presented and discussed. The direct coupling of a Gas Cooled Reactor with a closed gas-turbine cycle leads to a specific dynamic plant behaviour and a specific turbomachinery module has been developed. The thermal reactors' core consists of hexagonal graphite blocks with an annular-fueled region surrounded by reflectors and a special attention is paid on the thermal modeling of such a core leading to a quasi-2D thermal description. First designs of the VHTR are proposed and are based on an indirect cycle concept with a primary circuit, cooled by helium, and containing the core and a circulator. The core power is transmitted to the secondary circuit via an intermediate heat exchanger (IHX). The secondary circuit contains a turbine and a compressor coupled on a single shaft. It uses a mixture of helium and nitrogen, in order to benefit from both the favourable thermal properties of helium for the heat exchanger, and from existing experience of turbomachines using

Pressure vessel for nuclear reactors

International Nuclear Information System (INIS)

Schulten, R.; Kugeler, K.; Kugeler, M.; Petersen, K.; Decken, C.B. von der.

1983-01-01

This construction of a container, which is pressure-relieved by axial-central tensioning cables or reinforcing cables distributed over the circumference, makes a reduction of the wall thickness for the floor and roof, which was previously 2.5 metres by about 40% possible, and thus reduce manufacturing and cost problems. This is achieved by an appreciable increase of the prestressing exerted by the tensioning cables as this is taken up, not by the elasticity of the roof and floor, but instead by an intermediate part of pressure-resisting material. Such a container consists of a vertical cylindrical jacket of, for example, 20 metres diameter and 18 metres height, of a roof and floor of, for example, 1.50 metres thickness each and the intermediate part, which keeps the spacing of floor and roof as a central piece. This intermediate part which is taken through seals through the container can be imagined as a double tube of outside tube diameter of, for example, 4 metres and inside tube diameter of 2 metres with both tubes having thick walls. 4 tensioning cables displaced vertically by 900 run in the cylindrical annulus between the outer and inner tubes which are brought to the required pretension, e.g. 80,000 tonnes by nuts situated on the outside. The inner tube projects through the floor and roof. Its openings act as manholes and for the introduction of pipelines. These can, for example, carry a cooling medium for a reactor core via further ducts into the inside of the container. Container wall, floor and roof and the intermediate part in the form of a double tube are made up of cast steel segments or sectors in several layers. (RW)

Effect of an Intermediate Heat-treatment on a Change of the Corrosion Resistance and Hardness of a HANA-4 Outer Strip

International Nuclear Information System (INIS)

Lee, Myung Ho; Jung, Yang Il; Park, Sang Yoon; Choi, Byoung Kwon; Park, Jeong Yong; Jeong, Yong Hwan; Eom, Kyong Bo; Park, Nam Gyu; Lim, Yoon Soo

2008-01-01

KAERI (Korea Atomic Energy Research Institute) in collaboration with KNF (Korea Nuclear Fuel) undertook some researches on the applicability of HANA-4 and HANA-6 alloys for the spacer grid for a PWR (Pressurized Water Reactor) nuclear fuel. As a part of the research, KAERI studied the effect of the final heat-treatment on the mechanical and corrosion properties of a HANA-4 inner strip. The strip was manufactured with a sheet which had been intermediately heat-treated at about 580 .deg. C for 2.5-4 hours after each cold rolling before being processed into the final strip product. It was mentioned that the process with the intermediate heat treatment needed reviewing to establish an improved manufacturing process for the cold rolling. So, this work tried to check the effect of an intermediate heat-treatment on the properties of a HANA-4 strip using a specimen that was taken from a second hot rolled material before a cold-rolling. The manufacturing processes, with three different kinds of annealings, were introduced to investigate the applicable intermediate heat-treatment process. After all the cold-rolling processes, the Vickers hardness was measured for the final annealed specimens and 60 days of corrosion tests were carried out to check on the effect of the intermediate heat-treatment. Finally, an appropriate intermediate heat-treatment was proposed to improve the manufacturability of the HANA-4 strip

The bottom-supported fast reactor - system simplifications and enhanced safety

International Nuclear Information System (INIS)

Petrozelli, J.; Golan, S.; Kawamura, Yutaka; Kumaoka, Yoshio; Nakagawa, Hiroshi

1992-01-01

The 600-MW(electric) bottom-supported fast reactor (BSFR) incorporates the following key features: (1) modular upper internal structure (UIS); (2) electromagnetic pumps (EMPs); (3) low-sodium-void-worth metal-fuel core; and (4) bottom supported reactor vessel (BSRV), which is entirely supported by the basement, except for the control rods, control rod drives (CRDs), UIS, and the stationary plug; by comparison, a top-supported reactor vessel (TSRV) is completely supported by the operating floor. The diameter of the reactor vessel (RV) is 12.8 m (42 ft), and the height (distance from the basemat to the operating floor) is 19.8 m (65 ft). The RV is supported by a single support cylinder anchored to the basemat. The core has 210 driver assemblies and 192 radial blanket assemblies in an annular configuration. The primary heat transport system components consist of four intermediate heat exchangers (IHXs), four EMPs, and four primary reactor auxillary cooling systems. All these components are supported by the BSRV and hang from their tops. Six modular, vertically movable UIS mechanisms clear the UIS from the space over the core during refueling. The top closure is designed to operate at the reactor outlet temperature and is free to expand and contract. Small bellows between the top closure and each UIS model accommodate differential movements and comprise a portion of the cover gas boundary. A 1200-MW(electric) plant with two 600-MW(electric) (twin) nuclear steam supply systems is being studied

ORTURB, HTGR Steam Turbine Dynamic for FSV Reactor

International Nuclear Information System (INIS)

Conklin, J.C.

2001-01-01

1 - Description of program or function: ORTURB was written specifically to calculate the dynamic behavior of the Fort St. Vrain (FSV) High- Temperature Gas-Cooled Reactor (HTGR) steam turbines. The program is divided into three main parts: the driver subroutine; turbine subroutines to calculate the pressure-flow balance of the high-, intermediate-, and low-pressure turbines; and feedwater heater subroutines. 2 - Method of solution: The program uses a relationship derived for ideal gas flow in an iterative fashion that minimizes computational time to determine the pressure and flow in the FSV steam turbines as a function of plant transient operating conditions. An important computer modeling characteristic, unique to FSV, is that the high-pressure turbine exhaust steam is used to drive the reactor core coolant circulators prior to entering the reheater. A feedwater heater dynamic simulation model utilizing seven state variables for each of the five heaters is included in the ORTURB computer simulation of the regenerative Rankine cycle steam turbines. The seven temperature differential equations are solved at each time- step using a matrix exponential method. 3 - Restrictions on the complexity of the problem: The turbine shaft is assumed to rotate at a constant (rated) speed of 3600 rpm. Energy and mass storage of steam in the high-, intermediate-, and low-pressure turbines is assumed to be negligible. These limitations exclude the use of ORTURB during a turbine transient such as startup from zero power or very low turbine flows

Feasibility study of self sustaining capability on water cooled thorium reactors for different power reactors

International Nuclear Information System (INIS)

Permana, S.; Takaki, N.; Sekimoto, H.

2007-01-01

fertile and fissile nuclides with the contribution from intermediate nuclides such as 2 34U and 2 33Pa. The conversion ratio is evaluated by considering the conversion capability of the reactor to convert the fertile material into fissile material. The fissile accumulation capability for different conditions is investigated for estimating the fissile production capability during operation. The result shows the negative reactivity coefficient, and its feasibility of breeding for different MFR and burnup. The very tight lattice pin of MFR ≤ 0.3 is preferable for obtaining breeding condition for relatively higher burnup. The breeding capability of the reactor increases with increasing power output and decreasing power density. In relation to the self sustaining system, the large power output is easier to reach than the small power output

European development of ferritic-martensitic steels for fast reactor wrapper applications

International Nuclear Information System (INIS)

Bagley, K.; Little, E.A.; Levy, V.; Alamo, A.

1987-01-01

9-12%Cr ferritic-martensitic stainless steels are under development in Europe for fast reactor sub-assembly wrapper applications. Within this class of alloys, attention is focussed on three key specifications, viz. FV448 and DIN 1.4914 (both 10-12%CrMoVNb steels) and EM10 (an 8-10%Cr-0.15%C steel), which can be optimized to give acceptably low ductile-brittle transition characteristics. The results of studies on these steels, and earlier choices, covering heat treatment and compositional optimization, evolution of wrapper fabrication routes, pre and post-irradiation mechanical property and fracture toughness behaviour, microstructural stability, void swelling and in-reactor creep characteristics are reviewed. The retention of high void swelling to displacement doses in excess of 100 dpa in reactor irradiations reaffirms the selection of 9-12%Cr steels for on-going wrapper development. Moreover, irradiation-induced changes in mechanical properties (e.g. in-reactor creep and impact behaviour), measured to intermediate doses, do not give cause for concern; however, additional data to higher doses and at the lower irradiation temperatures of 370 0 -400 0 C are needed in order to fully endorse these alloys for high burnup applications in advanced reactor systems

FIST/6IB1, BWR/6 System Responses to Intermediate Break in Recirculation Suction Line LINE

International Nuclear Information System (INIS)

1993-01-01

1 - Description of test facility: BWR/6-218 standard plant. A full size bundle with electrically heated rods is used to simulate the reactor core. A scaling ratio of 1/624 is applied in the design of the system components. Key features of the FIST facility include: (1) Full height test vessel and internals; (2) correctly scaled fluid volume distribution; (3) simulation of ECCS, S/RV, and ADS; (4) level trip capability; (5) heated feedwater supply system, which provides the capability for steady state operation. 2 - Description of test: Test 6IB1 investigates system responses to an intermediate break in the recirculation suction line. BWR system licensing evaluations for various size recirculation break LOCA's indicates that a break size of about 0.2 sq.ft., without LPCS operation, is the highest PCT case for the intermediate break LOCA. Test 6IB1 simulates this event

Physical particularities of nuclear reactors using heavy moderators of neutrons

Energy Technology Data Exchange (ETDEWEB)

Kulikov, G. G., E-mail: ggkulikov@mephi.ru; Shmelev, A. N. [National Research Nuclear University MEPhI (Moscow Engineering Physics Institute) (Russian Federation)

2016-12-15

In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using {sup 233}U as a fissile nuclide and {sup 232}Th and {sup 231}Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

Physical particularities of nuclear reactors using heavy moderators of neutrons

International Nuclear Information System (INIS)

Kulikov, G. G.; Shmelev, A. N.

2016-01-01

In nuclear reactors, thermal neutron spectra are formed using moderators with small atomic weights. For fast reactors, inserting such moderators in the core may create problems since they efficiently decelerate the neutrons. In order to form an intermediate neutron spectrum, it is preferable to employ neutron moderators with sufficiently large atomic weights, using "2"3"3U as a fissile nuclide and "2"3"2Th and "2"3"1Pa as fertile ones. The aim of the work is to investigate the properties of heavy neutron moderators and to assess their advantages. The analysis employs the JENDL-4.0 nuclear data library and the SCALE program package for simulating the variation of fuel composition caused by irradiation in the reactor. The following main results are obtained. By using heavy moderators with small neutron moderation steps, one is able to (1) increase the rate of resonance capture, so that the amount of fertile material in the fuel may be reduced while maintaining the breeding factor of the core; (2) use the vacant space for improving the fuel-element properties by adding inert, strong, and thermally conductive materials and by implementing dispersive fuel elements in which the fissile material is self-replenished and neutron multiplication remains stable during the process of fuel burnup; and (3) employ mixtures of different fertile materials with resonance capture cross sections in order to increase the resonance-lattice density and the probability of resonance neutron capture leading to formation of fissile material. The general conclusion is that, by forming an intermediate neutron spectrum with heavy neutron moderators, one can use the fuel more efficiently and improve nuclear safety.

Comparative Study of the Reactor Burner Efficiency for Transmutation of Minor Actinides

Energy Technology Data Exchange (ETDEWEB)

Gulevich, A.; Zemskov, E. [Institute of Physics and Power Engineering, Bondarenko sq. 1, Obninsk, Kaluga region, 249020 (Russian Federation); Degtyarev, A.; Kalugin, A.; Ponomarev, L. [Russian Research Center ' Kurchatov Institute' , Kurchatov sq. 1, Moscow, 123182 (Russian Federation); Konev, V.; Seliverstov, V. [Institute of Theoretical and Experimental Physics, ul. B. Cheremushinskaya 25, Moscow, 117259 (Russian Federation)

2009-06-15

Transmutation of minor actinides (MA) in the closed nuclear fuel cycle (NFC) is a one of the most important problem for future nuclear energetic. There are several approaches for MA transmutation but there are no common criteria for the comparison of their efficiency. In paper [1] we turned out the attention to the importance of taking into account the duration of the closed NFC in addition to a usual criterion of the neutron economy. In accordance with these criteria the transmutation efficiency are compared of two fast reactors (sodium and lead cooled) and three types of ADS-burners: LBE-cooled reactors (fast neutron spectrum), molten-salt reactor (intermediate spectrum) and heavy water reactor (thermal spectrum). It is shown that the time of transmutation of loaded MA in the closed nuclear fuel cycle is more than 50 years. References: A. Gulevich, A. Kalugin, L. Ponomarev, V. Seliverstov, M. Seregin, 'Comparative Study of ADS for Minor Actinides Transmutation', Progress in Nuclear Energy, 50, March-August, p. 358, 2008. (authors)

Personal neutron dosimetry at a research reactor facility

International Nuclear Information System (INIS)

Kamenopoulou, V.; Carinou, E.; Stamatelatos, I.E.

2001-01-01

Individual neutron monitoring presents several difficulties due to the differences in energy response of the dosemeters. In the present study, an individual dosemeter (TLD) calibration approach is attempted for the personnel of a research reactor facility. The neutron energy response function of the dosemeter was derived using the MCNP code. The results were verified by measurements to three different neutron spectra and were found to be in good agreement. Three different calibration curves were defined for thermal, intermediate and fast neutrons. At the different working positions around the reactor, neutron spectra were defined using the Monte Carlo technique and ambient dose rate measurements were performed. An estimation of the neutrons energy is provided by the ratio of the different TLD pellets of each dosemeter in combination with the information concerning the worker's position; then the dose equivalent is deduced according to the appropriate calibration curve. (author)

Economic Contributions of Radioisotope Production Reactor in Korea

International Nuclear Information System (INIS)

Nam, Ji Hee; Kim, Seung Su; Moon, Kee Whan

2010-01-01

Radioisotopes (RIs) have been used extensively in the fields of industrial, the agricultural, and the medical applications. Especially the deficiency of radioisotopes such as Mo-99 and I-131 in the medical applications recently is becoming the main issue in our society. Radioisotope with the characteristics of public goods in some aspects is mainly playing as the intermediate inputs or goods in the process of the industrial production, with being expected to produce the economic benefits by creating the new demand in the market or enlarging the value added for the related goods and services. In this study, the contribution effects for Korean economy by the construction and operation of the reactor for radioisotope production would be evaluated the effects produced by the activities such as a RI supplies into domestic industry, the RI exports, the neutron transmutation doping services called NTD, and the exports of RI production reactors

MHD intermediate shock discontinuities: Pt. 1

International Nuclear Information System (INIS)

Kennel, C.F.; Blandford, R.D.; Coppi, P.

1989-01-01

Recent numerical investigations have focused attention once more on the role of intermediate shocks in MHD. Four types of intermediate shock are identified using a graphical representation of the MHD Rankine-Hugoniot conditions. This same representation can be used to exhibit the close relationship of intermediate shocks to switch-on shocks and rotational discontinuities. The conditions under which intermediate discontinuities can be found are elucidated. The variations in velocity, pressure, entropy and magnetic-field jumps with upstream parameters in intermediate shocks are exhibited graphically. The evolutionary arguments traditionally advanced against intermediate shocks may fail because the equations of classical MHD are not strictly hyperbolic. (author)

Cooling system of the core of a nuclear reactor while it is being stopped or normally operating

International Nuclear Information System (INIS)

Tilliette, Z.

1986-01-01

The present invention proposes a cooling system with intermediate gas flow which ensures the reactor core cooling when the primary pumps are stopped either directly by means of main heat-exchange circuits ensuring normally the reactor operation, or by means of separated loops, these ones being able so to operate in an autonomous way for they produce their own electricity needs and also an excedent which is added to the power plant production. The cooling circuit and the heat exchanger are described in detail [fr

Analysis for mechanical consequences of a core disruptive accident in Prototype Fast Breeder Reactor

International Nuclear Information System (INIS)

Chellapandi, P.; Velusamy, K.; Chetal, S.C.; Bhoje, S.B.; Lal, H.; Sethi, V.S.

2003-01-01

The mechanical consequences of a core disruptive accident (CDA) in a fast breeder reactor are described. The consequences are development of deformations and strains in the vessels, intermediate heat exchangers (IHX) and decay heat exchangers (DHX), impact of sodium slug on the bottom surface of the top shield, sodium release to reactor containment building through top shield penetrations, sodium fire and consequent temperature and pressure rise in reactor containment building (RCB). These are quantified for 500 MWe Prototype Fast Breeder Reactor (PFBR) for a CDA with 100 MJ work potential. The results are validated by conducting a series of experiments on 1/30 and 1/13 scaled down models with increasing complexities. Mechanical energy release due to nuclear excursion is simulated by chemical explosion of specially developed low density explosive charge. Based on these studies, structural integrity of primary containment, IHX and DHX is demonstrated. The sodium release to RCB is 350 kg which causes pressure rise of 12 kPa in RCB. (author)

Treatment and conditioning of metallic intermediate level waste

International Nuclear Information System (INIS)

Lidar, Per; Larsson, Arne; Huutoniemi, Tommi; Blank, Eva; Elfwing, Mattias

2014-01-01

In 2011 SKB started an R and D program for evaluating different disposal concepts for LL-LILW. The purpose was to develop alternative repository concepts and conditioning methods for LL-LILW and to evaluate and compare them from a range of parameters. The goal is to present a comparison between identified repository concepts by 2013. The material should be of such a quality that SKB can make decisions of which concepts that are to be further investigated in a safety analysis. As a part of the R and D program for the LL-LILW disposal facility, Studsvik was assigned to investigate whether melting of metallic LL-LILW is technically feasible and, if so, what the requirements are to build and operate such a facility. Specific concern was given to the following metallic components: - Core components and reactor internals from both boiling water reactors (BWRs) and pressurized water reactors (PWRs). - Reactor pressure vessels from PWRs. The paper presents a feasibility study of a melting facility for core components and reactor internals. An overview is given of how such a facility for treatment of intermediate level waste might be designed, constructed and operated and highlights both the possibilities and challenges. A cost estimate and a risk analysis are presented in order to make a conclusion of the technical feasibility of such a facility. Based on the Studsvik authors' experience in operating a low level waste melting facility, their conclusion is presented in the paper, considering cost of constructing and operating such a facility, in conjunction with the radio-logical risks associated with operation and the benefits to disposal and long term safety. Studsvik also investigated alternative techniques for embedding of metallic ILW components. Embedding of radioactive metallic ILW components protects the component from corrosion and leakage of radionuclides from repository to biosphere can thereby be both delayed and decreased. Conditioning by embedding has

Design of performance and analysis of dynamic and transient thermal behaviors on the intermediate heat exchanger for HTGR

International Nuclear Information System (INIS)

Mori, Michitsugu; Mizuno, Minoru; Itoh, Mitsuyoshi; Urabe, Shigemi

1985-01-01

The intermediate heat exchanger (IHX) is designed as the high temperature heat exchanger for HTGR (High Temperature Gas-cooled Reactor), which transmits the primary coolant helium's heat raised up to about 950 0 C in the reactor core to the secondary helium or the nuclear heat utilization. Having to meet, in addition, the requirement of the primary coolant pressure boundary as the Class-1 component, it must be secured integrity throughout the service life. This paper will show (1) the design of the thermal performance; (2) the results of the dynamic analyses of the 1.5 MWt-IHX with its comparison to the experimental data; (3) the analytical predictions of the dynamic thermal behaviors under start-up and of the transient thermal behaviors during the accident on the 25 MWt-IHX. (author)

Immobilisation of Higher Activity Wastes from Nuclear Reactor Production of 99Mo

Directory of Open Access Journals (Sweden)

Martin W. A. Stewart

2013-01-01

Full Text Available A variety of intermediate- and low-level liquid and solid wastes are produced from reactor production of 99Mo using UAl alloy or UO2 targets and in principle can be collectively or individually converted into waste forms. At ANSTO, we have legacy acidic uranyl-nitrate-rich intermediate level waste (ILW from the latter, and an alkaline liquid ILW, a U-rich filter cake, plus a shorter lived liquid stream that rapidly decays to low-level waste (LLW standards, from the former. The options considered consist of cementitious products, glasses, glass-ceramics, or ceramics produced by vitrification or hot isostatic pressing for intermediate-level wastes. This paper discusses the progress in waste form development and processing to treat ANSTO’s ILW streams arising from 99Mo. The various waste forms and the reason for the process option chosen will be reviewed. We also address the concerns over adapting our chosen process for use in a hot-cell environment.

Management of historical waste from research reactors: the Dutch experience

International Nuclear Information System (INIS)

Van Heek, Aliki; Metz, Bert; Janssen, Bas; Groothuis, Ron

2013-01-01

Most radioactive waste emerges as well-defined waste streams from operating power reactors. The management of this is an on-going practice, based on comprehensive (IAEA) guidelines. A special waste category however consists of the historical waste from research reactors, mostly originating from various experiments in the early years of the nuclear era. Removal of the waste from the research site, often required by law, raises challenges: the waste packages must fulfill the acceptance criteria from the receiving storage site as well as the criteria for nuclear transports. Often the aged waste containers do not fulfill today's requirements anymore, and their contents are not well documented. Therefore removal of historical waste requires advanced characterization, sorting, sustainable repackaging and sometimes conditioning of the waste. This paper describes the Dutch experience of a historical waste removal campaign from the Petten High Flux research reactor. The reactor is still in operation, but Dutch legislation asks for central storage of all radioactive waste at the COVRA site in Vlissingen since the availability of the high- and intermediate-level waste storage facility HABOG in 2004. In order to comply with COVRA's acceptance criteria, the complex and mixed inventory of intermediate and low level waste must be characterized and conditioned, identifying the relevant nuclides and their activities. Sorting and segregation of the waste in a Hot Cell offers the possibility to reduce the environmental footprint of the historical waste, by repackaging it into different classes of intermediate and low level waste. In this way, most of the waste volume can be separated into lower level categories not needing to be stored in the HABOG, but in the less demanding LOG facility for low-level waste instead. The characterization and sorting is done on the basis of a combination of gamma scanning with high energy resolution of the closed waste canister and low

Centrifugal pump assembly for use in nuclear reactor plants or the like

International Nuclear Information System (INIS)

Honold, E.; Ruepp, M.

1975-01-01

A description is given of a vertical centrifugal pump assembly for use as a primary recirculating pump in a nuclear reactor plant which has an uppermost shaft which is driven by a motor, a lowermost shaft which drives the impeller of the pump, and an intermediate shaft which is movable axially between the uppermost and lowermost shafts and carries a housing for radial and/or thrust bearings. A rigid coupling between the lowermost shaft and the intermediate shaft is disengaged when the intermediate shaft is lifted to thereby afford access to certain parts of the lowermost shaft. The uppermost shaft can drive the intermediate shaft through the medium of a flexible coupling having coupling elements mounted on the lower end of the uppermost shaft and the upper end of the intermediate shaft, and a distancing sleeve whose internal threads are in permanent mesh with external threads of the two coupling elements. A shroud which surrounds the flexible coupling has a maximum-diameter tubular section rigid with the housing for the bearings, a minimum-diameter tubular section rigid with but detachable from the casing of the motor, and a median tubular section which penetrates into the maximum-diameter section when the intermediate shaft is lifted. When the minimum-diameter section is detached from the casing of the motor, it can be slid, together with the intermediate section, into the maximum-diameter section to thereby afford access to the flexible coupling. The rigid coupling has two coaxial coupling elements which are mounted at the upper end of the pump shaft and the lower end of the intermediate shaft and may have teeth which mate when the intermediate shaft assumes its lower end position in which it can drive the pump

The gamma two-step cascade method at Dalat Nuclear Research Reactor

International Nuclear Information System (INIS)

Vuong Huu Tan; Pham Dinh Khang; Nguyen Nhi Dien; Nguyen Xuan Hai; Tran Tuan Anh; Ho Huu Thang; Pham Ngoc Son; Mangengo Lumengano

2014-01-01

The event-event coincidence spectroscopy system was successfully established and operated on thermal neutron beam of channel No. 3 at Dalat Nuclear Research Reactor (DNRR) with resolving time value of about 10 ns. The studies on level density, gamma strength function and decay scheme of intermediate-mass and heavy nuclei have been performed on this system. The achieved results are opening a new research of nuclear structure based on (n, 2γ) reaction. (author)

Kinetic modeling of the photocatalytic degradation of clofibric acid in a slurry reactor.

Science.gov (United States)

Manassero, Agustina; Satuf, María Lucila; Alfano, Orlando Mario

2015-01-01

A kinetic study of the photocatalytic degradation of the pharmaceutical clofibric acid is presented. Experiments were carried out under UV radiation employing titanium dioxide in water suspension. The main reaction intermediates were identified and quantified. Intrinsic expressions to represent the kinetics of clofibric acid and the main intermediates were derived. The modeling of the radiation field in the reactor was carried out by Monte Carlo simulation. Experimental runs were performed by varying the catalyst concentration and the incident radiation. Kinetic parameters were estimated from the experiments by applying a non-linear regression procedure. Good agreement was obtained between model predictions and experimental data, with an error of 5.9 % in the estimations of the primary pollutant concentration.

Nuclear Reactor/Hydrogen Process Interface Including the HyPEP Model

International Nuclear Information System (INIS)

Steven R. Sherman

2007-01-01

The Nuclear Reactor/Hydrogen Plant interface is the intermediate heat transport loop that will connect a very high temperature gas-cooled nuclear reactor (VHTR) to a thermochemical, high-temperature electrolysis, or hybrid hydrogen production plant. A prototype plant called the Next Generation Nuclear Plant (NGNP) is planned for construction and operation at the Idaho National Laboratory in the 2018-2021 timeframe, and will involve a VHTR, a high-temperature interface, and a hydrogen production plant. The interface is responsible for transporting high-temperature thermal energy from the nuclear reactor to the hydrogen production plant while protecting the nuclear plant from operational disturbances at the hydrogen plant. Development of the interface is occurring under the DOE Nuclear Hydrogen Initiative (NHI) and involves the study, design, and development of high-temperature heat exchangers, heat transport systems, materials, safety, and integrated system models. Research and development work on the system interface began in 2004 and is expected to continue at least until the start of construction of an engineering-scale demonstration plant

Structural integrity assessment of intermediate heat exchanger in the HTTR. Based on results of rise-to-power test

Energy Technology Data Exchange (ETDEWEB)

Takeda, Takeshi; Tachibana, Yukio; Nakagawa, Shigeaki [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment

2002-12-01

A helium/helium intermediate heat exchanger (IHX) in the high temperature engineering test reactor (HTTR) is an essential component for demonstration of future nuclear process heat utilization of high temperature gas-cooled reactor (HTGR). The IHX with a heat capacity of 10 MW has 96 helically-coiled heat transfer tubes. Structural design for the IHX had been conducted through elastic-creep analysis of superalloy Hastelloy XR components such as heat transfer tubes and center pipe. In the HTTR rise-to-power test, it was clarified that temperature of the coolant in the IHX at the reactor scrams changes more rapidly than expected in the design. Effects of the IHX coolant temperature change, at anticipated reactor scram from the full power of 30 MW at high temperature test operation, on structural integrity of the heat transfer tubes and the lower reducer of the center pipe were investigated analytically based on the coolant temperature data obtained from the rise-to-power test. As results of the assessment, it was confirmed that cumulative principal creep strain, cumulative creep and fatigue damage factor of the IHX components during 10{sup 5} h of the HTTR lifetime should be below the allowable limits, which are established in the high-temperature structural design code for the HTGR Class 1 components. (author)

Effect of an intermediate tungsten layer on thermal properties of TiC coatings ion plated onto molybdenum

International Nuclear Information System (INIS)

Fukutomi, M.; Fujitsuka, M.; Shikama, T.; Okada, M.

1985-01-01

Among the various low-Z coating-substrate systems proposed for fusion reactor first-wall applications, molybdenum coated with titanium carbide is considered very promising since it has a good capability of receiving heat from the plasma. The thermal stabilities of TiC layers ion plated onto the molybdenum substrate are discussed with particular reference to the interfacial reaction between the TiC coating and molybdenum. The deposition of an intermediate tungsten layer was found to be very effective in suppressing the formation of reaction layers, resulting in a marked improvement in thermal stabilities of TiC--Mo systems. Thermal shock test using a pulsed electron beam showed that the TiC coatings remained adherent to the molybdenum substrates during energy depositions high enough to melt the substrates within the area of beam deposition. The melt area of the TiC coatings apparently decreased when a tungsten intermediate layer was applied

Information acquisition and financial intermediation

OpenAIRE

Boyarchenko, Nina

2012-01-01

This paper considers the problem of information acquisition in an intermediated market, where the specialists have access to superior technology for acquiring information. These informational advantages of specialists relative to households lead to disagreement between the two groups, changing the shape of the intermediation-constrained region of the economy and increasing the frequency of periods when the intermediation constraint binds. Acquiring the additional information is, however, cost...

First conceptual design of the experimental multi-purpose high temperature reactor

Energy Technology Data Exchange (ETDEWEB)

Tsunoda, T [Fuji Electric Co. Ltd., Tokyo (Japan)

1976-02-01

A part of the multi-purpose high temperature reactor (VHTR) was designed by the First Atomic Power Industry Group (FAPIG). Both Fuji Electric Co., Ltd. and Kawasaki Heavy Industries, Ltd. of the FAPIG group took charge of the design of main parts of the reactor Kobe Steel, Ltd., Ebara Manufacturing Co., Ltd., Shimizu Construction Co., Ltd. and the Nuclear Fuel Corp. have associated with this group. The reactor system includes a nuclear reactor and two cooling loops provided through intermediate heat exchangers in order to utilize the heat of helium gas delivered from the reactor outlet at 1,000 deg C. One is a reformer loop to produce the reducing gas for steel manufacture. The other is a testing loop for a reducing gas heater and a gas turbine. These loops transfer heat of about 25 MW at 930 deg C at rated capacity. The reformer can supply the reducing gas equivalent to the production of 100 tons per day sponge iron. A housing of the reactor is composed of a primary steel container, internal concrete and a secondary container made of reinforced concrete. The construction is based on the following principles. (1) For the very high temperature portion at 1,000 deg C, a non-metallic material such as graphite should be used. (2) The metallic construction shall be cooled with return gas below 400 deg C. (3) The steel pressure vessel shall be employed. (4) The design shall be based on the existing gas furnace.

A vacuum disengager for tritium removal from HYLIFE-II Reactor Flibe

International Nuclear Information System (INIS)

Dolan, T.J.; Longhurst, G.R.; Garcia-Otero, E.

1992-01-01

We have designed a vacuum disengager system to remove tritium from the Flibe (Li 2 BeF 4 ) molten salt coolant of the HYLIFE-II fusion reactor. There is a two-stage vacuum disengager in each of three intermediate heat exchanger (IHX) loops. Each stage consists of a vacuum chamber 4 m in diameter and 7 m tall. As 0.2 mm diameter molten salt droplets fall vertically downward into the vacuum, most of the tritium diffuses out of the droplets and is pumped away. A fraction Φ ∼10 -5 of the 8.6 MCi/day tritium source (from breeding in the Flibe and from unburned fuel) remains in the Flibe as it leaves the vacuum disengagers, and about 21% of that permeates into the intermediate coolant loop, so about 20 Ci/day leak into the steam system. With Flibe primary coolant and a vacuum disengager, it appears that an intermediate coolant loop is not needed to prevent tritium from leaking into the steam system. An experiment is needed to demonstrate Flibe vacuum disengager operation

Next generation reactor development activity at Hitachi, Ltd

International Nuclear Information System (INIS)

Yamashita, Junichi

2005-01-01

Developments of innovative nuclear systems in Japan have been highly requested to cope with uncertain future nuclear power generation and fuel cycle situation. Next generation reactor system shall be surely deployed earlier to be capable to provide with several options such as plutonium multi-recycle, intermediate storage of spent fuels, simplified reprocessing of spent fuels and separated storage of 'Pu+FP' and 'U', spent fuels storage after Pu LWR recycle and their combinations, while future reactor system will be targeted at ideal fuel recycle system of higher breeding gain and transmutation of radioactive wastes. Modified designs of the ABWR at large size and medium and small size have been investigated as well as a BWR based RMWR and a supercritical-pressure LWR to ensure safety and improve economics. Advanced fuel cycle technologies of a combination of fluoride volatility process and PUREX process with high decontamination (FLUOREX process) and a modified fluoride volatility process with low decontamination have been developed. (T. Tanaka)

Investigation of vessel exterior air cooling for a HLMC reactor

International Nuclear Information System (INIS)

Sienicki, J. J.; Spencer, B. W.

2000-01-01

The Secure Transportable Autonomous Reactor (STAR) concept under development at Argonne National Laboratory provides a small (300 MWt) reactor module for steam supply that incorporates design features to attain proliferation resistance, heightened passive safety, and improved cost competitiveness through extreme simplification. Examples are the achievement of 100%+ natural circulation heat removal from the low power density/low pressure drop ultra-long lifetime core and utilization of lead-bismuth eutectic (LBE) coolant enabling elimination of main coolant pumps as well as the need for an intermediate heat transport circuit. It is required to provide a passive means of removing decay heat and effecting reactor cooldown in the event that the normal steam generator heat sink, including its normal shutdown heat removal mode, is postulated to be unavailable. In the present approach, denoted as the Reactor Exterior Cooling System (RECS), passive decay heat removal is provided by cooling the outside of the containment/guard vessel with air. RECS is similar to the Reactor Vessel Auxiliary Cooling System (RVACS) incorporated into the PRISM design. However, to enhance the heat removal, RECS incorporates fins on the containment vessel exterior to enhance heat transfer to air as well as removable steel venetian conductors that provide a conduction heat transfer path across the reactor vessel-containment vessel gap to enhance heat transfer between the vessels. The objective of the present work is to investigate the effectiveness of air cooling in removing heat from the vessel and limiting the coolant temperature increase following a sudden complete loss of the steam generator heat sink

Advanced Reactor PSA Methodologies for System Reliability Analysis and Source Term Assessment

Energy Technology Data Exchange (ETDEWEB)

Grabaskas, D.; Brunett, A.; Passerini, S.; Grelle, A.; Bucknor, M.

2017-06-26

Beginning in 2015, a project was initiated to update and modernize the probabilistic safety assessment (PSA) of the GE-Hitachi PRISM sodium fast reactor. This project is a collaboration between GE-Hitachi and Argonne National Laboratory (Argonne), and funded in part by the U.S. Department of Energy. Specifically, the role of Argonne is to assess the reliability of passive safety systems, complete a mechanistic source term calculation, and provide component reliability estimates. The assessment of passive system reliability focused on the performance of the Reactor Vessel Auxiliary Cooling System (RVACS) and the inherent reactivity feedback mechanisms of the metal fuel core. The mechanistic source term assessment attempted to provide a sequence specific source term evaluation to quantify offsite consequences. Lastly, the reliability assessment focused on components specific to the sodium fast reactor, including electromagnetic pumps, intermediate heat exchangers, the steam generator, and sodium valves and piping.

Mechanism of propylene oxidation over bismuth molybdate revealed by transient response method--1. An intermediate in the complete oxidation of C/sub 3/H/sub 6/

Energy Technology Data Exchange (ETDEWEB)

Kobayashi, M.; Futaya, R.

1979-01-01

An intermediate in the complete oxidation of C/sub 3/H/sub 6/ and its nature were deduced from transient response studies of a flow reactor system with a 1:1 bismuth-molybdenum catalyst at temperatures below ordinary operating temperatures. The formation of a stable surface intermediate with a non-allylic structure resembling that of propylene on every site of the catalyst surface occurred at 310/sup 0/C in the presence of propylene/oxygen or acrolein/oxygen; under oxidative conditions at 310/sup 0/C about eight times more intermediate formed directly from propylene than from adsorbed acrolein. The intermediate reacted with weakly adsorbed oxygen from the gas phase, but not with lattice oxygen, to decompose into CO/sub 2/; in temperature programed desorption studies (2.3/sup 0/C/min) it desorbed as CO/sub 2/ into a stream of helium above 220/sup 0/C (peak at 420/sup 0/C), apparently by reaction with oxygen diffused from the bulk to the surface.

Degradation of Acid Orange 7 Dye in Two Hybrid Plasma Discharge Reactors

Science.gov (United States)

Shen, Yongjun; Lei, Lecheng; Zhang, Xingwang; Ding, Jiandong

2014-11-01

To get an optimized pulsed electrical plasma discharge reactor and to increase the energy utilization efficiency in the removal of pollutants, two hybrid plasma discharge reactors were designed and optimized. The reactors were compared via the discharge characteristics, energy transfer efficiency, the yields of the active species and the energy utilization in dye wastewater degradation. The results showed that under the same AC input power, the characteristics of the discharge waveform of the point-to-plate reactor were better. Under the same AC input power, the two reactors both had almost the same peak voltage of 22 kV. The peak current of the point-to-plate reactor was 146 A, while that of the wire-to-cylinder reactor was only 48.8 A. The peak powers of the point-to-plate reactor and the wire-to-cylinder reactor were 1.38 MW and 1.01 MW, respectively. The energy per pulse of the point-to-plate reactor was 0.2221 J, which was about 29.4% higher than that of the wire-to-cylinder reactor (0.1716 J). To remove 50% Acid Orange 7 (AO7), the energy utilizations of the point-to-plate reactor and the wire-to-cylinder reactor were 1.02 × 10-9 mol/L and 0.61 × 10-9 mol/L, respectively. In the point-to-plate reactor, the concentration of hydrogen peroxide in pure water was 3.6 mmol/L after 40 min of discharge, which was higher than that of the wire-to-cylinder reactor (2.5 mmol/L). The concentration of liquid phase ozone in the point-to-plate reactor (5.7 × 10-2 mmol/L) was about 26.7% higher than that in the wire-to-cylinder reactor (4.5 × 10-2 mmol/L). The analysis results of the variance showed that the type of reactor and reaction time had significant impacts on the yields of the hydrogen peroxide and ozone. The main degradation intermediates of AO7 identified by gas chromatography and mass spectrometry (GCMS) were acetic acid, maleic anhydride, p-benzoquinone, phenol, benzoic acid, phthalic anhydride, coumarin and 2-naphthol. Proposed degradation pathways were

Degradation of Acid Orange 7 Dye in Two Hybrid Plasma Discharge Reactors

International Nuclear Information System (INIS)

Shen Yongjun; Ding Jiandong; Lei Lecheng; Zhang Xingwang

2014-01-01

To get an optimized pulsed electrical plasma discharge reactor and to increase the energy utilization efficiency in the removal of pollutants, two hybrid plasma discharge reactors were designed and optimized. The reactors were compared via the discharge characteristics, energy transfer efficiency, the yields of the active species and the energy utilization in dye wastewater degradation. The results showed that under the same AC input power, the characteristics of the discharge waveform of the point-to-plate reactor were better. Under the same AC input power, the two reactors both had almost the same peak voltage of 22 kV. The peak current of the point-to-plate reactor was 146 A, while that of the wire-to-cylinder reactor was only 48.8 A. The peak powers of the point-to-plate reactor and the wire-to-cylinder reactor were 1.38 MW and 1.01 MW, respectively. The energy per pulse of the point-to-plate reactor was 0.2221 J, which was about 29.4% higher than that of the wire-to-cylinder reactor (0.1716 J). To remove 50% Acid Orange 7 (AO7), the energy utilizations of the point-to-plate reactor and the wire-to-cylinder reactor were 1.02 × 10 −9 mol/L and 0.61 × 10 −9 mol/L, respectively. In the point-to-plate reactor, the concentration of hydrogen peroxide in pure water was 3.6 mmol/L after 40 min of discharge, which was higher than that of the wire-to-cylinder reactor (2.5 mmol/L). The concentration of liquid phase ozone in the point-to-plate reactor (5.7 × 10 −2 mmol/L) was about 26.7% higher than that in the wire-to-cylinder reactor (4.5 × 10 −2 mmol/L). The analysis results of the variance showed that the type of reactor and reaction time had significant impacts on the yields of the hydrogen peroxide and ozone. The main degradation intermediates of AO7 identified by gas chromatography and mass spectrometry (GCMS) were acetic acid, maleic anhydride, p-benzoquinone, phenol, benzoic acid, phthalic anhydride, coumarin and 2-naphthol. Proposed degradation

In situ mid-infrared analyses of reactive gas-phase intermediates in TEOS/Ozone SAPCVD

International Nuclear Information System (INIS)

Whidden, Thomas K.; Doiron, Sarah

1998-01-01

In this report, we present in situ characterizations of chemical vapour deposition (CVD) reactors used in silicon dioxide thin film depositions. The characterizations are based on Fourier transform infrared spectroscopy. The infrared absorption data are interpreted within the context of process and thin film properties and the bearing of the spectroscopic data upon the chemical mechanisms extant in the deposition reaction. The relevance of the interpretations to real-time process control is discussed. The process under study in this work is TEOS/ozone-based deposition of silicon dioxide thin films at subatmospheric pressures. This process exhibits many desirable properties but has fundamental problems that may be solvable by reaction control based on in situ analyses and the real-time manipulation of reagent concentrations and process conditions. Herein we discuss our preliminary data on characterizations of TEOS/ozone chemistries in commercial reactor configurations. Reaction products and reactive intermediate species are detected and identified. Quantitative in situ measurements of the reagent materials are demonstrated. Preliminary correlations of these data with process and thin film properties are discussed

Public acceptance of fusion energy and scientific feasibility of a fusion reactor. DREAM (DRastically EAsy Maintenance) tokamak

International Nuclear Information System (INIS)

Nishio, Satoshi

1998-01-01

If the major part of the electric power demand will be supplied by tokamak fusion power plants, a suitable tokamak reactor must be an ultimate goal, i.e., the reactor must be excellent both in terms of construction cost and safety aspects including operation availability (maintainability and reliability). In attaining this goal, an approach focusing on both safety and availability (including reliability and maintainability) issues is the most promising strategy. The tokamak reactor concept with a very high aspect ratio configuration and SiC/SiC composite structural materials is compatible with this approach, which is called the DREAM (DRastically EAsy Maintenance) approach. The SiC/SiC composite is a low activation material and an insulation material, and the high aspect ratio configuration leads to good accessibility for the maintenance of machines. As an intermediate steps between an experimental reactor such as ITER and the ultimate goal, the development of prototype reactor which demonstrates electric power generation and an initial-phase commercial reactor which demonstrates for COE (cost of electricity) competitiveness has been investigated. Especially for the prototype reactor, material and technological immaturity must be considered. (J.P.N.)

Ignition of DME and DME/CH4 at High Pressure: Flow Reactor Experiments and Kinetic Modeling

DEFF Research Database (Denmark)

Hashemi, Hamid; Christensen, Jakob Munkholt; Glarborg, Peter

The pyrolysis and oxidation of dimethyl ether (DME) and its mixtures with methane were investigated at high pressures (50 and 100 bar) and intermediate temperatures (450―900 K) in a laminar flow reactor. DME pyrolysis started at 825 K (at 50 bar). The onset of DME reaction was detected at 525―550 K...

Method for operating a nuclear reactor with scrammable part length rod

International Nuclear Information System (INIS)

Bevilacqua, F.

1979-01-01

A new part length rod is provided which may be used to control xenon induced power oscillations but also to contribute to shutdown reactivity when a rapid shutdown of the reactor is required. The part length rod consists of a control rod with three regions. The lower control region is a longer weaker active portion separated from an upper stronger shorter poison section by an intermediate section which is a relative non-absorber of neutrons. The combination of the longer weaker control section with the upper high worth poison section permits the part length rod to be scrammed into the core. When a reactor shutdown is required but also permits the control rod to be used as a tool to control power distribution in both the axial and radial directions during normal operation

Discourses and Models of Intermediality

OpenAIRE

Schröter, Jens

2011-01-01

In his article "Discourses and Models of Intermediality" Jens Schröter discusses the question as to what relations do different discourses pose between different "media." Schröter identifies four models of discourse: 1) synthetic intermediality: a "fusion" of different media to super-media, a model with roots in the Wagnerian concept of Gesamtkunstwerk with political connotations, 2) formal (or transmedial) intermediality: a concept based on formal structures not "specific" to one medium but ...

Time-resolved resonance Raman spectroscopy of intermediates of bacteriorhodopsin: The bK(590) intermediate.

Science.gov (United States)

Terner, J; Hsieh, C L; Burns, A R; El-Sayed, M A

1979-07-01

We have combined microbeam and flow techniques with computer subtraction methods to obtain the resonance Raman spectrum of the short lived batho-intermediate (bK(590)) of bacteriorhodopsin. Comparison of the spectra obtained in (1)H(2)O and (2)H(2)O, as well as the fact that the bK(590) intermediate shows large optical red shifts, suggests that the Schiff base linkage of this intermediate is protonated. The fingerprint region of the spectrum of bK(590), sensitive to the isomeric configuration of the retinal chromophore, does not resemble the corresponding region of the parent bR(570) form. The resonance Raman spectrum of bK(590) as well as the spectra of all of the other main intermediates in the photoreaction cycle of bacteriorhodopsin are discussed and compared with resonance Raman spectra of published model compounds.

Characterization of surface dielectric barrier discharge influenced by intermediate frequency for ozone production

Science.gov (United States)

Abdelaziz, Ayman A.; Ishijima, Tatsuo; Seto, Takafumi; Osawa, Naoki; Wedaa, Hassan; Otani, Yoshio

2016-06-01

The aim of this study is to investigate the effect of the intermediate frequency (1-10 kHz) of the sinusoidal driving voltage on the characteristics of a developed surface dielectric barrier discharge (SDBD)-based reactor having spikes on its discharge electrode. Moreover, its influence on the production of ozone and nitrogen oxide byproducts is evaluated. The results show that SDBD is operated in the filamentary mode at all the frequencies. Nevertheless, the pulses of the discharge current at high frequencies are much denser and have higher amplitudes than those at low frequencies. The analysis of the power consumed in the reactor shows that a small portion of the input power is dissipated in the dielectric material of SDBD source, whereas the major part of the power is consumed in the plasma discharge. The results of the ozone production show that higher frequencies have a slightly adverse effect on the ozone production at relatively high energy density values, where the ozone concentration is slightly decreased when the frequency is increased at the same energy density. The temperature of the discharge channels and gas is not a crucial factor for the decomposition of ozone in this reactor, while the results of the measurements of nitrogen oxides characteristics indicate that the formation of NO and NO2 has a significant adverse effect on the production efficiency of ozone due to their oxidation to another nitrogen oxides and their catalytic effect.

Characterization of surface dielectric barrier discharge influenced by intermediate frequency for ozone production

International Nuclear Information System (INIS)

Abdelaziz, Ayman A; Ishijima, Tatsuo; Seto, Takafumi; Otani, Yoshio; Osawa, Naoki; Wedaa, Hassan

2016-01-01

The aim of this study is to investigate the effect of the intermediate frequency (1–10 kHz) of the sinusoidal driving voltage on the characteristics of a developed surface dielectric barrier discharge (SDBD)-based reactor having spikes on its discharge electrode. Moreover, its influence on the production of ozone and nitrogen oxide byproducts is evaluated. The results show that SDBD is operated in the filamentary mode at all the frequencies. Nevertheless, the pulses of the discharge current at high frequencies are much denser and have higher amplitudes than those at low frequencies. The analysis of the power consumed in the reactor shows that a small portion of the input power is dissipated in the dielectric material of SDBD source, whereas the major part of the power is consumed in the plasma discharge. The results of the ozone production show that higher frequencies have a slightly adverse effect on the ozone production at relatively high energy density values, where the ozone concentration is slightly decreased when the frequency is increased at the same energy density. The temperature of the discharge channels and gas is not a crucial factor for the decomposition of ozone in this reactor, while the results of the measurements of nitrogen oxides characteristics indicate that the formation of NO and NO 2 has a significant adverse effect on the production efficiency of ozone due to their oxidation to another nitrogen oxides and their catalytic effect. (paper)

Role of Intermediate Filaments in Vesicular Traffic

Directory of Open Access Journals (Sweden)

Azzurra Margiotta

2016-04-01

Full Text Available Intermediate filaments are an important component of the cellular cytoskeleton. The first established role attributed to intermediate filaments was the mechanical support to cells. However, it is now clear that intermediate filaments have many different roles affecting a variety of other biological functions, such as the organization of microtubules and microfilaments, the regulation of nuclear structure and activity, the control of cell cycle and the regulation of signal transduction pathways. Furthermore, a number of intermediate filament proteins have been involved in the acquisition of tumorigenic properties. Over the last years, a strong involvement of intermediate filament proteins in the regulation of several aspects of intracellular trafficking has strongly emerged. Here, we review the functions of intermediate filaments proteins focusing mainly on the recent knowledge gained from the discovery that intermediate filaments associate with key proteins of the vesicular membrane transport machinery. In particular, we analyze the current understanding of the contribution of intermediate filaments to the endocytic pathway.

Reactor theory and power reactors. 1. Calculational methods for reactors. 2. Reactor kinetics

International Nuclear Information System (INIS)

Henry, A.F.

1980-01-01

Various methods for calculation of neutron flux in power reactors are discussed. Some mathematical models used to describe transients in nuclear reactors and techniques for the reactor kinetics' relevant equations solution are also presented

Intermediality: Bridge to Critical Media Literacy.

Science.gov (United States)

Pailliotet, Ann Watts; Semali, Ladislaus; Rodenberg, Rita K.; Giles, Jackie K.; Macaul, Sherry L.

2000-01-01

Defines "intermediality" as the ability to critically read and write with and across varied symbol systems. Relates it to critical media literacy. Offers rationales for teaching critical media literacy in general, and intermedial instruction in particular. Identifies seven guiding intermedial elements: theory, texts, processes, contexts,…

Non-catalytic production of fatty acid ethyl esters from soybean oil with supercritical ethanol in a two-step process using a microtube reactor

International Nuclear Information System (INIS)

Silva, Camila da; Lima, Ana Paula de; Castilhos, Fernanda de; Cardozo Filho, Lucio; Oliveira, J. Vladimir

2011-01-01

This work reports the production of fatty acid ethyl esters (FAEE) from the transesterification of soybean oil in supercritical ethanol in a continuous catalyst-free process using different reactor configurations. Experiments were performed in a microtube reactor with experimental simulation of two reactors operated in series and a reactor with recycle, both configurations at a constant temperature of 573 K, pressure of 20 MPa and oil to ethanol mass ratio of 1:1. Results show that the configurations studied with intermediate separation of glycerol afford higher conversions of vegetable oil to its fatty acid ethyl ester derivatives when compared to the one-step reaction, with relatively low decomposition of fatty acids (<3.0 wt%).

Reactor core in FBR type reactor

International Nuclear Information System (INIS)

Masumi, Ryoji; Kawashima, Katsuyuki; Kurihara, Kunitoshi.

1989-01-01

In a reactor core in FBR type reactors, a portion of homogenous fuels constituting the homogenous reactor core is replaced with multi-region fuels in which the enrichment degree of fissile materials is lower nearer to the axial center. This enables to condition the composition such that a reactor core having neutron flux distribution either of a homogenous reactor core or a heterogenous reactor core has substantially identical reactivity. Accordingly, in the transfer from the homogenous reactor core to the axially heterogenous reactor core, the average reactivity in the reactor core is substantially equal in each of the cycles. Further, by replacing a portion of the homogenous fuels with a multi-region fuels, thereby increasing the heat generation near the axial center, it is possiable to reduce the linear power output in the regions above and below thereof and, in addition, to improve the thermal margin in the reactor core. (T.M.)

Full-fluence tests of experimental thermosetting fuel rods for the high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Bullock, R.E.

1981-01-01

The irradiation performance of injected thermosetting fuel rods is compared to that of standard pitch-temperature gas-cooled reactor requirements. The primary objective of the experiments reported here was to obtain additional irradiation data at higher fluences for resin-based rods with intermediate binder char contents within the 15 to 30 wt% ''window of acceptability'' that had been previously established. 12 refs

Spanish experience in managing low and intermediate activity radioactive wastes

International Nuclear Information System (INIS)

Granero, J.J.

1986-01-01

The Spanish experience in management of low and intermediated level radioactive wastes is presented. The radioactive wastes stored come from research reactors, nuclear power plants, nuclear fuel cycle, scientific research, radiodiagnostic and medical applications. The commonest method is incorporation in cement inside special drums, even though some facilities use processes based on urea formal dehyde and on asphalt. Transport of the wastes is carried out by private undertakings and the Nuclear Energy Board. The sites used for storing are temporary in nature. The wastes produced by nuclear power plants are stored on site, with those processed by the Nuclear Energy Board are taken to a province of Cordoba. The National Company ENRESA for managing of all kinds of wastes was created. The Spanish legislation on this subject and the research being carried out by Spain itself and in cooperation with other States, are described. (Author) [pt

Electron-atom scattering at intermediate energies

International Nuclear Information System (INIS)

Kingston, A.E.; Walters, H.R.J.

1982-01-01

The problems of intermediate energy scattering are approached from the low and high energy ends. At low intermediate energies difficulties associated with the use of pseudostates and correlation terms are discussed, special consideration being given to nonphysical pseudoresonances. Perturbation methods appropriate to high intermediate energies are described and attempts to extend these high energy approximations down to low intermediate energies are studied. It is shown how the importance of electron exchange effects develops with decreasing energy. The problem of assessing the 'effective completeness' of pseudostate sets at intermediate energies is mentioned and an instructive analysis of a 2p pseudostate approximation to elastic e - -H scattering is given. It is suggested that at low energies the Pauli Exclusion Principle can act to hide short range defects in pseudostate approximations. (author)

On Brazil's participation in the International Project on Innovative Nuclear Reactors and Fuels Cycles (INPRO)

International Nuclear Information System (INIS)

Goncalves Filho, Orlando Joao Agostinho

2007-01-01

In response to a resolution of its 44th General Conference (GC(44)/RES/21) held in September 2000, the International Atomic Energy Agency launched in May 2001 the International Project on Innovative Nuclear Reactors and Fuels Cycles (INPRO) with the objective of supporting the safe, sustainable, economic and proliferation-resistant use of nuclear technology to meet the global energy needs of the 21st century. Brazil joined the project from its beginnings and in 2005 submitted a proposal for the screening assessment using INPRO methodology of two small-size light-water reactors as potential components of an innovative nuclear reactor system (INS) completed with a conventional open nuclear fuel cycle. The INS reactor components currently being assessed are the International Reactor Innovative and Secure (IRIS) that is being developed by an international consortium made of 21 organizations from 10 countries (Brazil included) led by the Westinghouse Company, and the Fixed Bed Nuclear Reactor (FBNR) that is being developed at the Federal University of Rio Grande do Sul. This paper gives an overview of Brazil's participation in INPRO, highlighting the objective, scope and intermediate results of the assessment study being performed, and the possibilities for participation in one or two collaborative research projects under INPRO Phase 2 Action Plan for 2008-2009. (author)

Unusual occurrences in fast breeder test reactor

International Nuclear Information System (INIS)

Kapoor, R.P.; Srinivasan, G.; Ellappan, T.R.; Ramalingam, P.V.; Vasudevan, A.T.; Iyer, M.A.K.; Lee, S.M.; Bhoje, S.B.

2000-01-01

Fast Breeder Test Reactor (FBTR) is a 40 MWt/13.2 MWe sodium cooled mixed carbide fuelled reactor. Its main aim is to generate experience in the design, construction and operation of fast reactors including sodium systems and to serve as an irradiation facility for the development of fuel and structural materials for future fast reactors. It achieved first criticality in Oct 85 with Mark I core (70% PuC - 30% UC). Steam generator was put in service in Jan 93 and power was raised to 10.5 MWt in Dec 93. Turbine generator was synchronised to the grid in Jul 97. The indigenously developed mixed carbide fuel has achieved a burnup of 44,000 MW-d/t max at a linear heat rating of 320 W/cm max without any fuel clad failure. The commissioning and operation of sodium systems and components have been smooth and performance of major components, viz., sodium pumps, intermediate heat exchangers and once through sodium heated steam generators (SG) have been excellent. There have been three minor incidents of Na/NaK leaks during the past 14 years, which are described in the paper. There have been no incident of a tube leak in SG. However, three incidents of water leaks from water / steam headers have been detailed. The plant has encountered some unusual occurrences, which were critically analysed and remedial measures, in terms of system and procedural modifications, incorporated to prevent recurrence. This paper describes unusual occurrences of fuel handling incident of May 1987, main boiler feed pump seizure in Apr 1992, reactivity transients in Nov 1994 and Apr 1995, and malfunctioning of the core cover plate mechanism in Jul 1995. These incidents have resulted in long plant shutdowns. During the course of investigation, various theoretical and experimental studies were carried out for better understanding of the phenomena and several inspection techniques and tools were developed resulting in enriching the technology of sodium cooled reactors. FBTR has 36 neutronic and process

The intermediate state in Patd

African Journals Online (AJOL)

) Jesus had assumed. (concerning the 'intermediate state') as existing, anything which does not exist. Three basic things about the intermediate state emerge from the parable: (a) Jesus recognizes that at the moment of death, in ipso articulo.

Higher order antibunching in intermediate states

International Nuclear Information System (INIS)

Verma, Amit; Sharma, Navneet K.; Pathak, Anirban

2008-01-01

Since the introduction of binomial state as an intermediate state, different intermediate states have been proposed. Different nonclassical effects have also been reported in these intermediate states. But till now higher order antibunching is predicted in only one type of intermediate state, which is known as shadowed negative binomial state. Recently we have shown that the higher order antibunching is not a rare phenomenon [P. Gupta, P. Pandey, A. Pathak, J. Phys. B 39 (2006) 1137]. To establish our earlier claim further, here we have shown that the higher order antibunching can be seen in different intermediate states, such as binomial state, reciprocal binomial state, hypergeometric state, generalized binomial state, negative binomial state and photon added coherent state. We have studied the possibility of observing the higher order subpoissonian photon statistics in different limits of intermediate states. The effects of different control parameters on the depth of non classicality have also been studied in this connection and it has been shown that the depth of nonclassicality can be tuned by controlling various physical parameters

Comparison of In-Vessel Shielding Design Concepts between Sodium-cooled Fast Burner Reactor and the Sodium-cooled Fast Breeder Reactor

International Nuclear Information System (INIS)

Yun, Sunghwan; Kim, Sang Ji

2015-01-01

In this study, quantities of in-vessel shields were derived and compared each other based on the replaceable shield assembly concept for both of the breeder and burner SFRs. Korean Prototype Gen-IV Sodium-cooled Fast Reactor (PGSFR) like SFR was used as the reference reactor and calculation method reported in the reference was used for shielding analysis. In this paper, characteristics of in-vessel shielding design were studied for the burner SFR and breeder SFR based on the replaceable shield assembly concept. An in-vessel shield to prevent secondary sodium activation (SSA) in the intermediate heat exchangers (IHXs) is one of the most important structures for the pool type Sodium-cooled Fast Reactor (SFR). In our previous work, two in-vessel shielding design concepts were compared each other for the burner SFR. However, a number of SFRs have been designed and operated with the breeder concept, in which axial and radial blankets were loaded for fuel breeding, during the past several decades. Since axial and radial blanket plays a role of neutron shield, comparison of required in-vessel shield amount between the breeder and burner SFRs may be an interesting work for SFR designer. Due to the blanket, the breeder SFR showed better performance in axial neutron shielding. Hence, 10.1 m diameter reactor vessel satisfied the design limit of SSA at the IHXs. In case of the burner SFR, due to more significant axial fast neutron leakage, 10.6 m diameter reactor vessel was required to satisfy the design limit of SSA at the IHXs. Although more efficient axial shied such as a mixture of ZrH 2 and B 4 C can improve shielding performance of the burner SFR, additional fabrication difficulty may mitigate the advantage of improved shielding performance. Therefore, it can be concluded that the breeder SFR has better characteristic in invessel shielding design to prevent SSA at the IHXs than the burner SFR in the pool-type reactor

Preparations for the shipment of RA-3 reactor irradiated fuel

International Nuclear Information System (INIS)

Goldschmidt, Adrian; Novara, Oscar; Lafuente, Jose

2002-01-01

During the last quarter of 2000, in the Radioactive Waste Management Area of the Argentine National Commission of Atomic Energy (CNEA), located at Ezeiza Atomic Center (CAE), activities associated to the shipment of 207 MTR spent fuels containing high enrichment uranium were carried out within the Foreign Research Reactor/Domestic Research Reactor Receipt Program launched by the US Department of Energy (DOE). The MTR spent fuel shipped to Savannah River Site (SRS) was fabricated in Argentina with 90% enriched uranium of US origin and it was utilized in the operation of the research and radioisotope production reactor RA-3 from 1968 until 1987. After a cooling period at the reactor, the spent fuel was transferred to the Central Storage Facility (CSF) located in the waste management area of CAE for interim storage. The spent fuel (SF) inventory consisted of 166 standard assemblies (SA) and 41 control assemblies (CA). Basically, the activities performed were the fuel conditioning operations inside the storage facility (remote transference of the assemblies to the operation pool, fuel cropping, fuel re-identification, loading in transport baskets, etc.) conducted by CNEA. The loading of the filled baskets in the transport casks (NAC-LWT) by means of intermediate transfer systems and loaded casks final preparations were conducted by NAC personnel (DOE's contractor) with the support of CNEA personnel. (author)

Prestudy of final disposal of long-lived low and intermediate level waste

International Nuclear Information System (INIS)

Wiborgh, M.

1995-01-01

The repository for long-lived low and intermediate level waste, SFL 3-5, is foreseen to be located adjacent to the deep repository for spent encapsulated fuel, SFL 2. The SFL 3-5 repository comprises of three repository parts which will be used for the different categories of waste. In this report the work performed within a pre-study of the SFL 3-5 repository concept is summarised. The aim was to make a first preliminary and simplified assessment of the near-field as a barrier to radionuclide dispersion. A major task has been to compile information on the waste foreseen to be disposed of in SFL 3-5. The waste comprises of; low and intermediate level waste from Studsvik, operational waste from the central interim storage for spent fuel, CLAB, and the encapsulation plant, decommissioning waste from these facilities, and core components and internal parts from the reactors. The total waste volume has been estimated to about 25000 m 3 . The total activity content at repository closure is estimated to be about 1 ·10 17 Bq in SFL 3-5. At repository closure the short-lived radionuclides, for example Co-60 and Fe-55, have decayed considerably and the activity is dominated by nickel isotopes in the metallic waste from the reactors, to be disposed of in SFL 5. However, other radionuclides may be more or equally important from a safety point of view, e.g cesium-isotopes and actinides which are found in largest amounts in the SFL 3 waste. A first evaluation of the long term performance or the SFL 3-5 repository has been made. A systematic methodology for scenario formulation was tested. The near-field release of contaminants was calculated for a selected number of radionuclides and chemo-toxic elements. The radionuclide release calculations revealed that Cs-137 and Ni-63 would dominate the annual release from all repository parts during the first 1000 years after repository closure and that Ni-59 would dominate at longer times

Prestudy of final disposal of long-lived low and intermediate level waste

Energy Technology Data Exchange (ETDEWEB)

Wiborgh, M [ed.; Kemakta Konsult AB., Stockholm (Sweden)

1995-01-01

The repository for long-lived low and intermediate level waste, SFL 3-5, is foreseen to be located adjacent to the deep repository for spent encapsulated fuel, SFL 2. The SFL 3-5 repository comprises of three repository parts which will be used for the different categories of waste. In this report the work performed within a pre-study of the SFL 3-5 repository concept is summarised. The aim was to make a first preliminary and simplified assessment of the near-field as a barrier to radionuclide dispersion. A major task has been to compile information on the waste foreseen to be disposed of in SFL 3-5. The waste comprises of; low and intermediate level waste from Studsvik, operational waste from the central interim storage for spent fuel, CLAB, and the encapsulation plant, decommissioning waste from these facilities, and core components and internal parts from the reactors. The total waste volume has been estimated to about 25000 m{sup 3}. The total activity content at repository closure is estimated to be about 1 {center_dot}10{sup 17} Bq in SFL 3-5. At repository closure the short-lived radionuclides, for example Co-60 and Fe-55, have decayed considerably and the activity is dominated by nickel isotopes in the metallic waste from the reactors, to be disposed of in SFL 5. However, other radionuclides may be more or equally important from a safety point of view, e.g cesium-isotopes and actinides which are found in largest amounts in the SFL 3 waste. A first evaluation of the long term performance or the SFL 3-5 repository has been made. A systematic methodology for scenario formulation was tested. The near-field release of contaminants was calculated for a selected number of radionuclides and chemo-toxic elements. The radionuclide release calculations revealed that Cs-137 and Ni-63 would dominate the annual release from all repository parts during the first 1000 years after repository closure and that Ni-59 would dominate at longer times.

Reactor core and initially loaded reactor core of nuclear reactor

International Nuclear Information System (INIS)

Koyama, Jun-ichi; Aoyama, Motoo.

1989-01-01

In BWR type reactors, improvement for the reactor shutdown margin is an important characteristic condition togehter with power distribution flattening . However, in the reactor core at high burnup degree, the reactor shutdown margin is different depending on the radial position of the reactor core. That is , the reactor shutdown margin is smaller in the outer peripheral region than in the central region of the reactor core. In view of the above, the reactor core is divided radially into a central region and as outer region. The amount of fissionable material of first fuel assemblies newly loaded in the outer region is made less than the amount of the fissionable material of second fuel assemblies newly loaded in the central region, to thereby improve the reactor shutdown margin in the outer region. Further, the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower portion of the first fuel assemblies is made smaller than the ratio between the amount of the fissionable material in the upper region and that of the fissionable material in the lower region of the second fuel assemblies, to thereby obtain a sufficient thermal margin in the central region. (K.M.)

UF6 breeder reactor power plants for electric power generation

International Nuclear Information System (INIS)

Rust, J.H.; Clement, J.D.; Hohl, F.

1976-01-01

The reactor concept analyzed is a 233 UF 6 core surrounded by a molten salt (Li 7 F, BeF 2 , ThF 4 ) blanket. Nuclear survey calculations were carried out for both spherical and cylindrical geometries. A maximum breeding ratio of 1.22 was found. Thermodynamic cycle calculations were performed for a variety of Rankine cycles. Optimization of a Rankine cycle for a gas core breeder reactor employing an intermediate heat exchanger gave a maximum efficiency of 37 percent. A conceptual design is presented along with a system layout for a 1000 MW stationary power plant. The advantages of the GCBR are as follows: (1) high efficiency, (2) simplified on-line reprocessing, (3) inherent safety considerations, (4) high breeding ratio, (5) possibility of burning all or most of the long-lived nuclear waste actinides, and (6) possibility of extrapolating the technology to higher temperatures and MHD direct conversion

Calculational framework for safety analyses of non-reactor nuclear facilities

International Nuclear Information System (INIS)

Coleman, J.R.

1994-01-01

A calculational framework for the consequences analysis of non-reactor nuclear facilities is presented. The analysis framework starts with accident scenarios which are developed through a traditional hazard analysis and continues with a probabilistic framework for the consequences analysis. The framework encourages the use of response continua derived from engineering judgment and traditional deterministic engineering analyses. The general approach consists of dividing the overall problem into a series of interrelated analysis cells and then devising Markov chain like probability transition matrices for each of the cells. An advantage of this division of the problem is that intermediate output (as probability state vectors) are generated at each calculational interface. The series of analyses when combined yield risk analysis output. The analysis approach is illustrated through application to two non-reactor nuclear analyses: the Ulysses Space Mission, and a hydrogen burn in the Hanford waste storage tanks

Experimental fuel channel for samples irradiation at the RB reactor

International Nuclear Information System (INIS)

Pesic, M.; Markovic, H.; Sokcic-Kostic, M.; Miric, I.; Prokic, M.; Strugar, P.

1984-12-01

An 80% enriched UO 2 fuel channel at the RB nuclear reactor in the 'Boris Kidric' Institute of Nuclear Sciences is modified for samples irradiation by fast neutrons. Maximum sample diameter is 25 mm and length up to 1000 mm. Characteristics of neutron and gamma radiation fields of this new experimental channel are investigated. In the centre of the channel, the main contribution to the total neutron absorbed dose, i.e. 0.29 Gy/Wh of reactor operation, is due to the fast neutron spectrum component. Only 0.05 Gy and 0.07 Gy in the total neutron absorbed dose are due to intermediate and thermal neutrons, respectively. At the same time the gamma absorbed dose is 0.35 Gy. The developed experimental fuel channel, EFC, has wide possibilities for utilization, from fast neutron spectrum studies, electronic component irradiations, dosemeters testing, up to cross-section measurements. (author)

Analysis of a bio-electrochemical reactor containing carbon fiber textiles for the anaerobic digestion of tomato plant residues.

Science.gov (United States)

Hirano, Shin-Ichi; Matsumoto, Norio

2018-02-01

A bio-electrochemical system packed with supporting material can promote anaerobic digestion for several types of organic waste. To expand the target organic matters of a BES, tomato plant residues (TPRs), generated year-round as agricultural and cellulosic waste, were treated using three methanogenic reactors: a continuous stirred tank reactor (CSTR), a carbon fiber textile (CFT) reactor, and a bio-electrochemical reactor (BER) including CFT with electrochemical regulation (BER + CFT). CFT had positive effects on methane fermentation and methanogen abundance. The microbial population stimulated by electrochemical regulation, including hydrogenotrophic methanogens, cellulose-degrading bacteria, and acetate-degrading bacteria, suppressed acetate accumulation, as evidenced by the low acetate concentration in the suspended fraction in the BER + CFT. These results indicated that the microbial community in the BER + CFT facilitated the efficient decomposition of TPR and its intermediates such as acetate to methane. Copyright © 2017 Elsevier Ltd. All rights reserved.

A nuclear desalination complex with a VK-300 boiling type reactor facility

International Nuclear Information System (INIS)

Kuznetzov, Y.N.; Mishanina, Y.A.; Romenkov, A.A.

2004-01-01

RDIPE has developed a detailed design of an enhanced safety nuclear steam supply system (NSSS) with a VK-300 boiling water reactor for combined heat and power generation. The thermal power of the reactor is 750 MW. The maximum electrical power in the condensation mode is 250 MWe. The maximum heat generation capacity of 400 Gcal/h is reached at 150 MWe. This report describes, in brief, the basic technical concepts for the VK-300 NSSS and the power unit, with an emphasis on enhanced safety and good economic performance. With relatively small power, good technical and economic performance of the VK-300 reactor that is a base for the desalination complex is attained through: reduced capital costs of the nuclear plant construction thanks to technical approaches ensuring maximum simplicity of the reactor design and the NSSS layout; a single-circuit power unit configuration (reactor-turbine) excluding expensive equipment with a lot of metal, less pipelines and valves; reduced construction costs of the basic buildings thanks to reduced construction volumes due to rational arrangement concepts; higher reliability of equipment and reduced maintenance and repair costs; longer reactor design service life of up to 60 years; selection of the best reactor and desalination equipment interface pattern. The report considers the potential application of the VK-300 reactor as a source of energy for distillation desalination units. The heat from the reactor is transferred to the desalination unit via an intermediate circuit. Comparison is made between variants of the reactor integration with desalination units of the following types: multi-stage flash (MSF technology); multi-effect distillation horizontal-tube film units of the DOU GTPA type (MED technology). The NDC capacity with the VK-300 reactor, in terms of distillate, will be more than 200,000 m 3 /day, with the simultaneous output of electric power from the turbine generator buses of around 150 MWe. The variants of the

Licence applications for low and intermediate level waste predisposal facilities: A manual for operators

International Nuclear Information System (INIS)

2009-07-01

This publication covers all predisposal waste management facilities and practices for receipt, pretreatment (sorting, segregation, characterization), treatment, conditioning, internal relocation and storage of low and intermediate level radioactive waste, including disused sealed radioactive sources. The publication contains an Annex presenting the example of a safety assessment for a small radioactive waste storage facility. Facilities dealing with both short lived and long lived low and intermediate level waste generated from nuclear applications and from operation of small nuclear research reactors are included in the scope. Processing and storage facilities for high activity disused sealed sources and sealed sources containing long lived radionuclides are also covered. The publication does not cover facilities processing or storing radioactive waste from nuclear power plants or any other industrial scale nuclear fuel cycle facilities. Disposal facilities are excluded from the scope of this publication. Authorization process can be implemented in several stages, which may start at the site planning and the feasibility study stage and will continue through preliminary design, final design, commissioning, operation and decommissioning stages. This publication covers primarily the authorization needed to take the facility into operation

Advanced Reactor Passive System Reliability Demonstration Analysis for an External Event

Directory of Open Access Journals (Sweden)

Matthew Bucknor

2017-03-01

Full Text Available Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general for the postulated transient event.

Advanced reactor passive system reliability demonstration analysis for an external event

Energy Technology Data Exchange (ETDEWEB)

Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; Grelle, Austin [Argonne National Laboratory, Argonne (United States)

2017-03-15

Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event.

Advanced reactor passive system reliability demonstration analysis for an external event

International Nuclear Information System (INIS)

Bucknor, Matthew; Grabaskas, David; Brunett, Acacia J.; Grelle, Austin

2017-01-01

Many advanced reactor designs rely on passive systems to fulfill safety functions during accident sequences. These systems depend heavily on boundary conditions to induce a motive force, meaning the system can fail to operate as intended because of deviations in boundary conditions, rather than as the result of physical failures. Furthermore, passive systems may operate in intermediate or degraded modes. These factors make passive system operation difficult to characterize within a traditional probabilistic framework that only recognizes discrete operating modes and does not allow for the explicit consideration of time-dependent boundary conditions. Argonne National Laboratory has been examining various methodologies for assessing passive system reliability within a probabilistic risk assessment for a station blackout event at an advanced small modular reactor. This paper provides an overview of a passive system reliability demonstration analysis for an external event. Considering an earthquake with the possibility of site flooding, the analysis focuses on the behavior of the passive Reactor Cavity Cooling System following potential physical damage and system flooding. The assessment approach seeks to combine mechanistic and simulation-based methods to leverage the benefits of the simulation-based approach without the need to substantially deviate from conventional probabilistic risk assessment techniques. Although this study is presented as only an example analysis, the results appear to demonstrate a high level of reliability of the Reactor Cavity Cooling System (and the reactor system in general) for the postulated transient event

[Therapy of intermediate uveitis].

Science.gov (United States)

Doycheva, D; Deuter, C; Zierhut, M

2014-12-01

Intermediate uveitis is a form of intraocular inflammation in which the vitreous body is the major site of inflammation. Intermediate uveitis is primarily treated medicinally and systemic corticosteroids are the mainstay of therapy. When recurrence of uveitis or side effects occur during corticosteroid therapy an immunosuppressive treatment is required. Cyclosporine A is the only immunosuppressive agent that is approved for therapy of uveitis in Germany; however, other immunosuppressive drugs have also been shown to be effective and well-tolerated in patients with intermediate uveitis. In severe therapy-refractory cases when conventional immunosuppressive therapy has failed, biologics can be used. In patients with unilateral uveitis or when the systemic therapy is contraindicated because of side effects, an intravitreal steroid treatment can be carried out. In certain cases a vitrectomy may be used.

Thermal hydrodynamic modeling and simulation of hot-gas duct for next-generation nuclear reactor

Energy Technology Data Exchange (ETDEWEB)

Lee, Injun [School of Mechanical Engineering, Yeungnam University, Gyeongsan 712-749 (Korea, Republic of); Hong, Sungdeok; Kim, Chansoo [Korea Atomic Energy Research Institute, Daejeon 305-353 (Korea, Republic of); Bai, Cheolho; Hong, Sungyull [School of Mechanical Engineering, Yeungnam University, Gyeongsan 712-749 (Korea, Republic of); Shim, Jaesool, E-mail: jshim@ynu.ac.kr [School of Mechanical Engineering, Yeungnam University, Gyeongsan 712-749 (Korea, Republic of)

2016-12-15

Highlights: • Thermal hydrodynamic nonlinear model is presented to examine a hot gas duct (HGD) used in a fourth-generation nuclear power reactor. • Experiments and simulation were compared to validate the nonlinear porous model. • Natural convection and radiation are considered to study the effect on the surface temperature of the HGD. • Local Nusselt number is obtained for the optimum design of a possible next-generation HGD. - Abstract: A very high-temperature gas-cooled reactor (VHTR) is a fourth-generation nuclear power reactor that requires an intermediate loop that consists of a hot-gas duct (HGD), an intermediate heat exchanger (IHX), and a process heat exchanger for massive hydrogen production. In this study, a mathematical model and simulation were developed for the HGD in a small-scale nitrogen gas loop that was designed and manufactured by the Korea Atomic Energy Research Institute. These were used to investigate the effect of various important factors on the surface of the HGD. In the modeling, a porous model was considered for a Kaowool insulator inside the HGD. The natural convection and radiation are included in the model. For validation, the modeled external surface temperatures are compared with experimental results obtained while changing the inlet temperatures of the nitrogen working fluid. The simulation results show very good agreement with the experiments. The external surface temperatures of the HGD are obtained with respect to the porosity of insulator, emissivity of radiation, and pressure of the working fluid. The local Nusselt number is also obtained for the optimum design of a possible next-generation HGD.

Fuel element for a nuclear reactor

International Nuclear Information System (INIS)

Linning, D.L.

1977-01-01

An improvement of the fuel element for a fast nuclear reactor described in patent 15 89 010 is proposed which should avoid possible damage due to swelling of the fuel. While the fuel element according to patent 15 89 010 is made in the form of a tube, here a further metal jacket is inserted in the centre of the fuel rod and the intermediate layer (ceramic uranium compound) is provided on both sides, so that the nuclear fuel is situated in the centre of the annular construction. Ceramic uranium or plutonium compounds (preferably carbide) form the fuel zone in the form of circular pellets, which are surrounded by annular gaps, so that gaseous fission products can escape. (UWI) [de

Store and process for intermediate or final storage of used fuel elements from a nuclear reactor

International Nuclear Information System (INIS)

Kumpf, H.

1986-01-01

The fuel elements are enclosed in boxes at the nuclear reactor and transported in these to the incoming station. Transport is a by truck, which makes it possible for the transport container to move in a vertical position, where the upper side is on the top side of the truck. The fuel elements in their boxes are handed over to a magazine there, which can be reached by a loading machine serving the storage room. (orig./HP) [de

Canadian experiences in characterizing two low-level and intermediate-level radioactive waste management sites

International Nuclear Information System (INIS)

Heystee, R.J.; Rao, P.K.M.

1984-02-01

Low-level waste (LLW) and intermediate-level reactor waste (ILW) arise in Canada from the operation of nuclear power reactors for the generation of electricity and from the operation of reactors for nuclear research and development as well as for the production of separated radioisotopes. The majority of this waste is currently being safely managed at two sites in the Province of Ontario: (1) Chalk River Nuclear Laboratories, and (2) Ontario Hydro's Bruce Nuclear Power Development Radioactive Waste Operations Site 2. Although these storage facilities can safely manage the waste for a long period of time, there are advantages in disposal of the LLW and ILW. The design of the disposal facilities and the assessment of long-term performance will require that the hydrologic and geologic data be gathered for a potential disposal site. Past site characterization programs at the two aforementioned waste storage sites have produced information which will be useful to future disposal studies in similar geologic materials. The assessment of long-term performance will require that predictions be made regarding the potential subsurface migration of radionuclides. However there still remain many uncertainties regarding the chemical and physical processes which affect radionuclide mobility and concentrations, in particular hydrodynamic dispersion, geochemical reactions, and transport through fractured media. These uncertainties have to be borne in mind when conducting the performance assessments and adequate conservatism must be included to account for the uncertainties. (author)

Nuclear reactor physics course for reactor operators

International Nuclear Information System (INIS)

Baeten, P.

2006-01-01

The education and training of nuclear reactor operators is important to guarantee the safe operation of present and future nuclear reactors. Therefore, a course on basic 'Nuclear reactor physics' in the initial and continuous training of reactor operators has proven to be indispensable. In most countries, such training also results from the direct request from the safety authorities to assure the high level of competence of the staff in nuclear reactors. The aim of the basic course on 'Nuclear Reactor Physics for reactor operators' is to provide the reactor operators with a basic understanding of the main concepts relevant to nuclear reactors. Seen the education level of the participants, mathematical derivations are simplified and reduced to a minimum, but not completely eliminated

Reactor core of FBR type reactor

International Nuclear Information System (INIS)

Hayashi, Hideyuki; Ichimiya, Masakazu.

1994-01-01

A reactor core is a homogeneous reactor core divided into two regions of an inner reactor core region at the center and an outer reactor core region surrounding the outside of the inner reactor core region. In this case, the inner reactor core region has a lower plutonium enrichment degree and less amount of neutron leakage in the radial direction, and the outer reactor core region has higher plutonium enrichment degree and greater amount of neutron leakage in the radial direction. Moderator materials containing hydrogen are added only to the inner reactor core fuels in the inner reactor core region. Pins loaded with the fuels with addition of the moderator materials are inserted at a ratio of from 3 to 10% of the total number of the fuel pins. The moderator materials containing hydrogen comprise zirconium hydride, titanium hydride, or calcium hydride. With such a constitution, fluctuation of the power distribution in the radial direction along with burning is suppressed. In addition, an absolute value of the Doppler coefficient can be increased, and a temperature coefficient of coolants can be reduced. (I.N.)

Design Configurations and Coupling High Temperature Gas-Cooled Reactor and Hydrogen Plant

International Nuclear Information System (INIS)

Chang H. Oh; Eung Soo Kim; Steven Sherman

2008-01-01

The US Department of Energy is investigating the use of high-temperature nuclear reactors to produce hydrogen using either thermochemical cycles or high-temperature electrolysis. Although the hydrogen production processes are in an early stage of development, coupling either of these processes to the high-temperature reactor requires both efficient heat transfer and adequate separation of the facilities to assure that off-normal events in the production facility do not impact the nuclear power plant. An intermediate heat transport loop will be required to separate the operations and safety functions of the nuclear and hydrogen plants. A next generation high-temperature reactor could be envisioned as a single-purpose facility that produces hydrogen or a dual-purpose facility that produces hydrogen and electricity. Early plants, such as the proposed Next Generation Nuclear Plant (NGNP), may be dual-purpose facilities that demonstrate both hydrogen and efficient electrical generation. Later plants could be single-purpose facilities. At this stage of development, both single- and dual-purpose facilities need to be understood

Heavy section steel technology program technical report No. 38. Fracture toughness characterization of HSST intermediate pressure vessel material

International Nuclear Information System (INIS)

Mager, T.R.; Yanichko, S.E.; Singer, L.R.

1974-12-01

The primary objective of the Heavy Section Steel Technology (HSST) Program is to develop pertinent fracture technology to demonstrate the structural reliability of present and contemplated water-cooled nuclear reactor pressure vessels. In order to demonstrate the ability to predict failure of large, heavy-walled pressure vessels under service type loading conditions, the fracture toughness properties of the vessel's materials must be characterized. The sampling procedure and test results are presented for vessel material supplied by the Oak Ridge National Laboratory that were used to characterize the fracture toughness of the HSST Intermediate Test Vessels. The metallurgical condition and heat treatment of the test material was representative of the vessel simulated service test condition. Test specimen locations and orientations were selected by the Oak Ridge National Laboratory and are representative of flaw orientations incorporated in the test vessels. The fracture toughness is documented for the materials from each of the eight HSST Intermediate Pressure Vessels tested to date. 7 references. (U.S.)

Possibility of the use of intermediate carbidsiliconoxide nanolayers on polydiamond substrates for gallium nitride layers epitaxy

Energy Technology Data Exchange (ETDEWEB)

Averichkin, P. A., E-mail: P-Yugov@mail.ru; Donskov, A. A. [State Research and Design Institute of Rare-Metal Industry Giredmet AO (Russian Federation); Dukhnovsky, M. P. [R & D Enterprise Istok (Russian Federation); Knyazev, S. N. [State Research and Design Institute of Rare-Metal Industry Giredmet AO (Russian Federation); Kozlova, Yu. P. [Russian Academy of Sciences, Institute for Nuclear Research (Russian Federation); Yugova, T. G.; Belogorokhov, I. A. [State Research and Design Institute of Rare-Metal Industry Giredmet AO (Russian Federation)

2016-04-15

The results of using carbidsiliconoxide (a-C:SiO1{sub .5}) films with a thickness of 30–60 nm, produced by the pyrolysis annealing of oligomethylsilseskvioksana (CH{sub 3}–SiO{sub 1.5}){sub n} with cyclolinear (staircased) molecular structure, as intermediate films in the hydride vapor phase epitaxy of gallium nitride on polycrystalline CVD-diamond substrates are presented. In the pyrolysis annealing of (CH{sub 3}–SiO{sub 1.5}){sub n} films in an atmosphere of nitrogen at a temperature of 1060°C, methyl radicals are carbonized to yield carbon atoms chemically bound to silicon. In turn, these atoms form a SiC monolayer on the surface of a-C:SiO{sub 1.5} films via covalent bonding with silicon. It is shown that GaN islands grow on such an intermediate layer on CVD-polydiamond substrates in the process of hydride vapor phase epitaxy in a vertical reactor from the GaCl–NH{sub 3}–N{sub 2} gas mixture.

Pre evaluation for heat balance of prototype sodium cooled fast reactor

International Nuclear Information System (INIS)

Han, Ji Woong; Kim, De Hee; Yoon, Jung; Kim, Eui Kwang; Lee, Tae Ho

2012-01-01

Under the long term advanced SFR R and D plan, the design of prototype reactor has been carried out toward the construction of the prototype SFR plant by 2028. The R and D efforts in fluid system design will be focused on developing a prototype design of primary heat transport system(PHTS), intermediate heat transport system (IHTS), decay heat removal system(DHRS), steam generation system(SGS), and related auxiliary system design for a prototype reactor as shown in Fig. 1. In order to make progress system design, top tier requirements for prototype reactor related to design parameters of NSSS and BOP should be decided at first. The top tier requirement includes general design basis, capacity and characteristics of reactor, various requirements related to safety, performance, securities, economics, site, and etc.. Extensive discussion has been done within Korea Atomic Energy Research Institute(KAERI) for the decision of top tier requirements of the prototype reactor. The core outlet temperature, which should be described as top tier requirements, is one of the critical parameter for system design. The higher core exit temperature could contribute to increase the plant efficiency. However, it could also contribute to decrease the design margin for structure and safety. Therefore various operating strategies based on different core outlet temperatures should be examined and evaluated. For the prototype reactor two core outlet temperatures are taken into accounted. The lower temperature is for the operation condition and the higher temperature is for the system design and licensing process of the prototype reactor. In order to evaluate the operability of prototype reactor designed based on higher temperature, the heat balance calculations have been performed at different core outlet temperature conditions. The electrical power of prototype reactor was assumed to be 100MWe and reference operating conditions were decided based on existing available data. The

Vitrification of reactor wastes

International Nuclear Information System (INIS)

Jouan, A.

1993-01-01

The vitrification of low and intermediate level wastes from the NPP operation has been studied in the frame of a Franco-Czech agreement. The laboratory experiments concentrated on a search for a suitable borosilicate glass matrix which could incorporate relatively high quantities of boron and sodium, main components of liquid wastes from the WWER reactor type NPPs. A relatively wide area of waste compositions has been studied and properties of glasses suitable for the technology and waste disposal were measured. Great attention has been paid to the chemical stability (leachability), other properties like thermal dependence of viscosity and electrical conductivity of melts, and the microstructure of the final solidification product have also been evaluated. The feasibility of the vitrification process has been proved during pilot plant tests which were accomplished at the French establishment in Marcoule. The results of tests were promising. (authors). 4 tabs., 7 figs

Vitrification of reactor wastes

Energy Technology Data Exchange (ETDEWEB)

Jouan, A [CEA Centre d` Etudes de la Vallee du Rhone, 30 - Marcoule (France). Dept. des Procedes de Retraitement; Sussmilch, J [Nuclear Research Institut, Rez (Czech Republic)

1994-12-31

The vitrification of low and intermediate level wastes from the NPP operation has been studied in the frame of a Franco-Czech agreement. The laboratory experiments concentrated on a search for a suitable borosilicate glass matrix which could incorporate relatively high quantities of boron and sodium, main components of liquid wastes from the WWER reactor type NPPs. A relatively wide area of waste compositions has been studied and properties of glasses suitable for the technology and waste disposal were measured. Great attention has been paid to the chemical stability (leachability), other properties like thermal dependence of viscosity and electrical conductivity of melts, and the microstructure of the final solidification product have also been evaluated. The feasibility of the vitrification process has been proved during pilot plant tests which were accomplished at the French establishment in Marcoule. The results of tests were promising. (authors). 4 tabs., 7 figs.

Reprocessing technology for present water reactor fuels

International Nuclear Information System (INIS)

McMurray, P.R.

1977-01-01

The basic Purex solvent extraction technology developed and applied in the U.S. in the 1950's provides a well-demonstrated and efficient process for recovering uranium and plutonium for fuel recycle and separating the wastes for further treatment and packaging. The technologies for confinement of radioactive effluents have been developed but have had limited utilization in the processing of commercial light water reactor fuels. Technologies for solidification and packaging of radioactive wastes have not yet been demonstrated but significant experience has been gained in laboratory and engineering scale experiments with simulated commercial reprocessing wastes and intermediate level wastes. Commercial scale experience with combined operations of all the required processes and equipment are needed to demonstrate reliable reprocessing centers

Proceedings of the 4. CSNI workshop on the chemistry of iodine in reactor safety

Energy Technology Data Exchange (ETDEWEB)

Guentay, S [ed.; Paul Scherrer Inst. (PSI), Villigen (Switzerland)

1996-12-01

The 4. OECD workshop on the chemistry of iodine in reactor safety was held in Wuerenlingen, Switzerland from June 10th to 12th, 1996. It was organised in collaboration with the Laboratory for Safety and Accident Research of the Paul Scherrer Institute. About seventy experts from fourteen OECD member countries attended the meeting, as well as experts from Latvia and the Commission of the European Communities. Thirty-four papers were presented in five sessions on various aspects of national and international programmes, integral and intermediate-scale experiments, experimental homogeneous phase chemistry, surface processes, thermodynamic and kinetic studies and safety applications. Throughout the meeting, emphasis was placed on detailed and open discussions. The purpose of the workshop was to exchange information on the iodine chemistry and other important fission products relevant to reactor safety, to discuss the status of the open issues identified during the previous workshop held in 1991, to define reactor safety issues and to discuss developments and future plans. (author) figs., tabs., refs.

Proceedings of the 4. CSNI workshop on the chemistry of iodine in reactor safety

International Nuclear Information System (INIS)

Guentay, S.

1996-12-01

The 4. OECD workshop on the chemistry of iodine in reactor safety was held in Wuerenlingen, Switzerland from June 10th to 12th, 1996. It was organised in collaboration with the Laboratory for Safety and Accident Research of the Paul Scherrer Institute. About seventy experts from fourteen OECD member countries attended the meeting, as well as experts from Latvia and the Commission of the European Communities. Thirty-four papers were presented in five sessions on various aspects of national and international programmes, integral and intermediate-scale experiments, experimental homogeneous phase chemistry, surface processes, thermodynamic and kinetic studies and safety applications. Throughout the meeting, emphasis was placed on detailed and open discussions. The purpose of the workshop was to exchange information on the iodine chemistry and other important fission products relevant to reactor safety, to discuss the status of the open issues identified during the previous workshop held in 1991, to define reactor safety issues and to discuss developments and future plans. (author) figs., tabs., refs

Nuclear reactors

International Nuclear Information System (INIS)

Barre, Bertrand

2015-10-01

After some remarks on the nuclear fuel, on the chain reaction control, on fuel loading and unloading, this article proposes descriptions of the design, principles and operations of different types of nuclear reactors as well as comments on their presence and use in different countries: pressurized water reactors (design of the primary and secondary circuits, volume and chemistry control, backup injection circuits), boiling water reactors, heavy water reactors, graphite and boiling water reactors, graphite-gas reactors, fast breeder reactors, and fourth generation reactors (definition, fast breeding). For these last ones, six concepts are presented: sodium-cooled fast reactor, lead-cooled fast reactor, gas-cooled fast reactor, high temperature gas-cooled reactor, supercritical water-cooled reactor, and molten salt reactor

Advanced In-Service Inspection Approaches Applied to the Phenix Fast Breeder Reactor

International Nuclear Information System (INIS)

Guidez, J.; Martin, L.; Dupraz, R.

2006-01-01

The safety upgrading of the Phenix plant undertaken between 1994 and 1997 involved a vast inspection programme of the reactor, the external storage drum and the secondary sodium circuits in order to meet the requirements of the defence-in-depth safety approach. The three lines of defence were analysed for every safety related component: demonstration of the quality of design and construction, appropriate in-service inspection and controlling the consequences of an accident. The in-service reactor block inspection programme consisted in controlling the core support structures and the high-temperature elements. Despite the fact that limited consideration had been given to inspection constraints during the design stage of the reactor in the 1960's, as compared to more recent reactor projects such as the European Fast Reactor (EFR), all the core support line elements were able to be inspected. The three following main operations are described: Ultrasonic inspection of the upper hangers of the main vessel, using small transducers able to withstand temperatures of 130 deg. C, Inspection of the conical shell supporting the core dia-grid. A specific ultrasonic method and a special implementation technique were used to control the under sodium structure welds, located up to several meters away from the scan surface. Remote inspection of the hot pool structures, particularly the core cover plug after partial sodium drainage of the reactor vessel. Other inspections are also summarized: control of secondary sodium circuit piping, intermediate heat exchangers, primary sodium pumps, steam generator units and external storage drum. The pool type reactor concept, developed in France since the 1960's, presents several favourable safety and operational features. The feedback from the Phenix plant also shows real potential for in-service inspection. The design of future generation IV sodium fast reactors will benefit from the experience acquired from the Phenix plant. (authors)

Aspects on thorium utilization in heavy water reactors

International Nuclear Information System (INIS)

1978-11-01

Some of the main problems of the Th - PHWR cycles are analyzed. With respect to the burnup limitations introduced by SSET cycle conditions and the burnup sensitivities of this reference cycle, estimates are presented using an integrated neutron-heavy element balance method. A PHWR of 1 GW(e) very similar to the CANDU current design was selected. In the case of 0.5% uranium losses, 11000 MWD/tHE and 13000 MWD/tHE were considered for U-235, respectively, for the Pu initialization of the cycle, the corresponding inventory being 4 t U-235/GW(e) and 5 t Pu (with 72% fissile content) per GW(e) for one year delay time between reactor out to reactor in, 66% capacity factor, 27 MW (fission)/tHE medium specific power. The following aspects are also analyzed: Safety problems associated with low delayed neutron fraction values; High and intermediate burnup fuel elements conceptual problems; Specific problems of thorium reprocessing; Specific problems for radioactive wastes and thorium storage; U-232 content evaluations and related fuel fabrication problems

Effect of Intermediate Hosts on Emerging Zoonoses.

Science.gov (United States)

Cui, Jing-An; Chen, Fangyuan; Fan, Shengjie

2017-08-01

Most emerging zoonotic pathogens originate from animals. They can directly infect humans through natural reservoirs or indirectly through intermediate hosts. As a bridge, an intermediate host plays different roles in the transmission of zoonotic pathogens. In this study, we present three types of pathogen transmission to evaluate the effect of intermediate hosts on emerging zoonotic diseases in human epidemics. These types are identified as follows: TYPE 1, pathogen transmission without an intermediate host for comparison; TYPE 2, pathogen transmission with an intermediate host as an amplifier; and TYPE 3, pathogen transmission with an intermediate host as a vessel for genetic variation. In addition, we established three mathematical models to elucidate the mechanisms underlying zoonotic disease transmission according to these three types. Stability analysis indicated that the existence of intermediate hosts increased the difficulty of controlling zoonotic diseases because of more difficult conditions to satisfy for the disease to die out. The human epidemic would die out under the following conditions: TYPE 1: [Formula: see text] and [Formula: see text]; TYPE 2: [Formula: see text], [Formula: see text], and [Formula: see text]; and TYPE 3: [Formula: see text], [Formula: see text], [Formula: see text], and [Formula: see text] Simulation with similar parameters demonstrated that intermediate hosts could change the peak time and number of infected humans during a human epidemic; intermediate hosts also exerted different effects on controlling the prevalence of a human epidemic with natural reservoirs in different periods, which is important in addressing problems in public health. Monitoring and controlling the number of natural reservoirs and intermediate hosts at the right time would successfully manage and prevent the prevalence of emerging zoonoses in humans.

Core Design and Deployment Strategy of Heavy Water Cooled Sustainable Thorium Reactor

Directory of Open Access Journals (Sweden)

Naoyuki Takaki

2012-08-01

Full Text Available Our previous studies on water cooled thorium breeder reactor based on matured pressurized water reactor (PWR plant technology concluded that reduced moderated core by arranging fuel pins in a triangular tight lattice array and using heavy water as coolant is appropriate for achieving better breeding performance and higher burn-up simultaneously [1–6]. One optimum core that produces 3.5 GW thermal energy using Th-²³³U oxide fuel shows a breeding ratio of 1.07 and averaged burn-up of about 80 GWd/t with long cycle length of 1300 days. The moderator to fuel volume ratio is 0.6 and required enrichment of ²³³U for the fresh fuel is about 7%. The coolant reactivity coefficient is negative during all cycles despite it being a large scale breeder reactor. In order to introduce this sustainable thorium reactor, three-step deployment scenario, with intermediate transition phase between current light water reactor (LWR phase and future sustainer phase, is proposed. Both in transition phase and sustainer phase, almost the same core design can be applicable only by changing fissile materials mixed with thorium from plutonium to ²³³U with slight modification in the fuel assembly design. Assuming total capacity of 60 GWe in current LWR phase and reprocessing capacity of 800 ton/y with further extensions to 1600 ton/y, all LWRs will be replaced by heavy water cooled thorium reactors within about one century then thorium reactors will be kept operational owing to its potential to sustain fissile fuels while reprocessing all spent fuels until exhaustion of massive thorium resource.

The final report of ''on-the-job training'' on the CANDU reactor

International Nuclear Information System (INIS)

Kim, D.H.; Koh, B.J.

1983-01-01

This is the final Report for the technical ''on-the-job traning'' for the Wolsung CANDU nuclear power plant which is the first Pressurized Heavy Water Reactor setting up in Korea. The technical ''on-the-job traning'' was established to increase the capability for the nuclear safety evaluation in order to contribute the future safe operation of the CANDU nuclear power plant. The training has been excuted through three level courses as elementary, intermediate and ''on-the-job training'' at Wolsung power plant. The elementary course was introduction to the CANDU basics and fundamentals. The intermediate course was the more advanced course, and the detailed concepts and engineering explanations of the CANDU system had been instructed. The third course was the ''on-the-job training'' at the Wolsung plant site, which was the most emphasized course during the project. (Author)

Intermediate structure and threshold phenomena

International Nuclear Information System (INIS)

Hategan, Cornel

2004-01-01

The Intermediate Structure, evidenced through microstructures of the neutron strength function, is reflected in open reaction channels as fluctuations in excitation function of nuclear threshold effects. The intermediate state supporting both neutron strength function and nuclear threshold effect is a micro-giant neutron threshold state. (author)

Biodegradation of Methylene Blue Dye by Sequential Treatment Using Anaerobic Hybrid Reactor and Submerged Aerobic Fixed Film Bioreactor

Science.gov (United States)

Farooqi, Izharul H.; Basheer, Farrukh; Tiwari, Pradeepika

2017-12-01

Laboratory scale experiments were carried out to access the feasibility of sequential anaerobic/aerobic biological treatment for the biodegradation of Methylene Blue (MB) dye. Anaerobic studies were performed using anaerobic hybrid reactor (consisting of UASB and Anaerobic filter) whereas submerged aerobic fixed film reactor was used as aerobic reactor. Degradation of MB dye was attempted using neutralized acetic acid (1000 mg/L) as co-substrate. MB dye concentration was stepwise increased from 10 to 70 mg/L after reaching steady state in each dye concentration. Such a gradual increase in the dye concentration helps in the proper acclimatization of the sludge to dyes thereby avoiding the possible inhibitory effects to biological activities at high dye concentrations. The overall treatment efficiency of MB through sequential anaerobic-aerobic reactor operation was 90% at maximum attempted dye concentration of 70 mg/L. The effluent from anaerobic reactor was analysed for intermediate biodegradation products through HPLC. It was observed that catechol, quinone, amino pyrine, 1,4 diamino benzene were present. However they were absent in final effluent.

Fuel requirements for isotope production and reasearch reactors: Possible alternative ways of meeting non-proliferation objectives

International Nuclear Information System (INIS)

There is a continuing need for access to medium-to-high flux research reactors of intermediate power level (5-50 MW) for the production of industrial and medical radioisotopes, for the provision of neutron beams and for materials research. The construction of further reactors of this type is likely. To obtain the required flux levels in adequate volumes and at the lowest capital cost, past practice has been to design a small-core reactor around a fuel element concept using fully enriched uranium, that is, uranium enriched to 80% U-235 or greater. In recent years, however, it has been recognised that the use of fully enriched uranium in research reactors could give rise to significant risks of nuclear weapons proliferation. Accordingly, there would be advantage if research reactors could be operated on low enriched fuel, that is, enrichment levels of 20% or less. It is the purpose of this paper to explore the implications for proliferation of the enrichment level of research reactor fuel and to draw attention to possible options for reducing proliferation concerns which warrant further study. It does not, however, consider research reactors using very low enriched or natural uranium fuel. The paper is offered to stimulate discussion of the issues and the views expressed do not necessarily represent any formal Australian position

Intermediate algebra & analytic geometry

CERN Document Server

Gondin, William R

1967-01-01

Intermediate Algebra & Analytic Geometry Made Simple focuses on the principles, processes, calculations, and methodologies involved in intermediate algebra and analytic geometry. The publication first offers information on linear equations in two unknowns and variables, functions, and graphs. Discussions focus on graphic interpretations, explicit and implicit functions, first quadrant graphs, variables and functions, determinate and indeterminate systems, independent and dependent equations, and defective and redundant systems. The text then examines quadratic equations in one variable, system

Reactor core for LMFBR type reactors

International Nuclear Information System (INIS)

Masumi, Ryoji; Azekura, Kazuo; Kurihara, Kunitoshi; Bando, Masaru; Watari, Yoshio.

1987-01-01

Purpose: To reduce the power distribution fluctuations and obtain flat and stable power distribution throughout the operation period in an LMFBR type reactor. Constitution: In the inner reactor core region and the outer reactor core region surrounding the same, the thickness of the inner region is made smaller than the axial height of the reactor core region and the radial width thereof is made smaller than that of the reactor core region and the volume thereof is made to 30 - 50 % for the reactor core region. Further, the amount of the fuel material per unit volume in the inner region is made to 70 - 90 % of that in the outer region. The difference in the neutron infinite multiplication factor between the inner region and the outer region is substantially constant irrespective of the burnup degree and the power distribution fluctuation can be reduced to about 2/3, by which the effect of thermal striping to the reactor core upper mechanisms can be moderated. Further, the maximum linear power during operation can be reduced by 3 %, by which the thermal margin in the reactor core is increased and the reactor core fuels can be saved by 3 %. (Kamimura, M.)

A study on conceptual design of tritium production fusion reactor based on spherical torus

International Nuclear Information System (INIS)

He Kaihui; Huang Jinhua

2003-01-01

Conceptual design of an advanced tritium production reactor based on spherical torus (ST), which is an intermediate application of fusion energy, is presented. Different from traditional Tokamak tritium production reactor design, advanced plasma physics performance and compact structural characteristics of ST are used to minimize tritium leakage and to maximize tritium breeding ratio with arrangement of tritium production blankets as possible as it can do within vacuum vessel in order to produce certain amount of excess tritium except self-sufficient plasma core, corresponding plant availability 40% or more. Based on 2D neutronics calculation, preliminary conceptual design of ST-TPR is presented. Based on systematical analysis, design risk, uncertainty and backup are introduced generally for the backgrounds of next detailed conceptual design. (authors)

Enhanced situation awareness and decision making for an intelligent reconfigurable reactor power controller

International Nuclear Information System (INIS)

Kenney, S.J.; Edwards, R.M.

1996-01-01

A Learning Automata based intelligent reconfigurable controller has been adapted for use as a reactor power controller to achieve improved reactor temperature performance. The intelligent reconfigurable controller is capable of enforcing either a classical or an optimal reactor power controller based on control performance feedback. Four control performance evaluation measures: dynamically estimated average quadratic temperature error, power, rod reactivity and rod reactivity rate were developed to provide feedback to the control decision component of the intelligent reconfigurable controller. Fuzzy Logic and Neural Network controllers have been studied for inclusion in the bank of controllers that form the intermediate level of an enhanced intelligent reconfigurable reactor power controller (IRRPC). The increased number of alternatives available to the supervisory level of the IRRPC requires enhanced situation awareness. Additional performance measures have been designed and a method for synthesizing them into a single indication of the overall performance of the currently enforced reactor power controller has been conceptualized. Modification of the reward/penalty scheme implemented in the existing IRRPC to increase the quality of the supervisory level decision process has been studied. The logogen model of human memory (Morton, 1969) and individual controller design information could be used to allocate reward to the most appropriate controller. Methods for allocating supervisory level attention were also studied with the goal of maximizing learning rate

Flow reactor studies of non-equilibrium plasma-assisted oxidation of n-alkanes.

Science.gov (United States)

Tsolas, Nicholas; Lee, Jong Guen; Yetter, Richard A

2015-08-13

The oxidation of n-alkanes (C1-C7) has been studied with and without the effects of a nanosecond, non-equilibrium plasma discharge at 1 atm pressure from 420 to 1250 K. Experiments have been performed under nearly isothermal conditions in a flow reactor, where reactive mixtures are diluted in Ar to minimize temperature changes from chemical reactions. Sample extraction performed at the exit of the reactor captures product and intermediate species and stores them in a multi-position valve for subsequent identification and quantification using gas chromatography. By fixing the flow rate in the reactor and varying the temperature, reactivity maps for the oxidation of fuels are achieved. Considering all the fuels studied, fuel consumption under the effects of the plasma is shown to have been enhanced significantly, particularly for the low-temperature regime (T<800 K). In fact, multiple transitions in the rates of fuel consumption are observed depending on fuel with the emergence of a negative-temperature-coefficient regime. For all fuels, the temperature for the transition into the high-temperature chemistry is lowered as a consequence of the plasma being able to increase the rate of fuel consumption. Using a phenomenological interpretation of the intermediate species formed, it can be shown that the active particles produced from the plasma enhance alkyl radical formation at all temperatures and enable low-temperature chain branching for fuels C3 and greater. The significance of this result demonstrates that the plasma provides an opportunity for low-temperature chain branching to occur at reduced pressures, which is typically observed at elevated pressures in thermal induced systems. © 2015 The Author(s) Published by the Royal Society. All rights reserved.

Continuous Hydrolysis and Liquid–Liquid Phase Separation of an Active Pharmaceutical Ingredient Intermediate Using a Miniscale Hydrophobic Membrane Separator

DEFF Research Database (Denmark)

Cervera Padrell, Albert Emili; Morthensen, Sofie Thage; Lewandowski, Daniel Jacob

2012-01-01

Continuous hydrolysis of an active pharmaceutical ingredient intermediate, and subsequent liquid–liquid (L-L) separation of the resulting organic and aqueous phases, have been achieved using a simple PTFE tube reactor connected to a miniscale hydrophobic membrane separator. An alkoxide product......, obtained in continuous mode by a Grignard reaction in THF, reacted with acidic water to produce partially miscible organic and aqueous phases containing Mg salts. Despite the partial THF–water miscibility, the two phases could be separated at total flow rates up to 40 mL/min at different flow ratios, using...

Different types of intermediate storages

International Nuclear Information System (INIS)

Spilker, H.

1982-01-01

Spent nuclear fuel elements require a period of ''cooling'' after withdrawal from a nuclear reactor. This is done in water storage pools located at the reactor facility. The minimum storage time prior to subsequent operations with the spent fuel is recognized as six to twelve months. After the initial storage period for cooling, the subsequent storage time in water pools could vary from a few years to several decades. Up to 1980, about 15.700 MTHM of spent LWR and HWR fuel have been discharged from nuclear stations. Since only a small fraction thereof has been already reprocessed. The spent fuel assemblies are stored at the reactor-side (AR) in reactor basins in normal or compact racks and away from reactor (AFR) in water pool storages awaiting either reprocessing or final disposal depending on the fuel cycle concept chosen by the individual countries. (orig.)

Regularities of intermediate adsorption complex relaxation

International Nuclear Information System (INIS)

Manukova, L.A.

1982-01-01

The experimental data, characterizing the regularities of intermediate adsorption complex relaxation in the polycrystalline Mo-N 2 system at 77 K are given. The method of molecular beam has been used in the investigation. The analytical expressions of change regularity in the relaxation process of full and specific rates - of transition from intermediate state into ''non-reversible'', of desorption into the gas phase and accumUlation of the particles in the intermediate state are obtained

FINANCIAL INTERMEDIATION, ENTREPRENEURSHIP AND ECONOMIC GROWTH

OpenAIRE

Wenli Cheng

2007-01-01

This paper presents a simple general equilibrium model of financial intermediation, entrepreneurship and economic growth. In this model, the role of financial intermediation is to pool savings and to lend the pooled funds to an entrepreneur, who in turn invests the funds in a new production technology. The adoption of the new production technology improves individual real income. Thus financial intermediation promotes economic growth through affecting individuals’ saving behaviour and enabl...

What is past is prologue: future directions in Tokamak Power Reactor Design Research

International Nuclear Information System (INIS)

Conn, R.W.

1976-01-01

After reviewing the first generation of studies and the primary conclusions they produced, four current designs are discussed that are representative of present trends in this area of research. In particular, the trends towards reduced reactor size and higher neutron wall loadings are discussed. Moving in this direction requires new approaches to many subsystem designs. New approaches and future directions in first wall and blanket designs that can achieve reliable operation and reasonable lifetime, the use of cryogenic but normal aluminum magnets for the pulsed coils in a tokamak, blanket designs that allow elimination of the intermediate loop, and low activity shields and toroidal field magnets are described. A discussion is given of the future role of conceptual reactor design research and the need for close interactions with ongoing experiments in fusion technology

Methods for monitoring the initial load to critical in the fast test reactor

International Nuclear Information System (INIS)

Johnson, D.L.

1975-08-01

Conventional symmetric fuel loadings for the initial loading to critical of the Fast Test Reactor (FTR) are predicted to be more time consuming than asymmetric or trisector loadings. Potentially significant time savings can be realized by the latter, since adequate intermediate assessments of neutron multiplication can be made periodically without control rod reconnection in all trisectors. Experimental simulation of both loading schemes was carried out in the Reverse Approach to Critical (RAC) experiments in the Fast Test Reactor-Engineering Mockup Critical facility. Analyses of these experiments indicated that conventional source multiplication methods can be applied for monitoring either a symmetric or asymmetric fuel loading scheme equally well provided that detection efficiency corrections are employed. Methods for refining predictions of reactivity and count rates for the stages in a load to critical were also investigated. (auth)

Some Intermediate-Level Violin Concertos.

Science.gov (United States)

Abramson, Michael

1997-01-01

Contends that many violin students attempt difficult concertos before they are technically or musically prepared. Identifies a variety of concertos at the intermediate and advanced intermediate-level for students to study and master before attempting the advanced works by Bach and Mozart. Includes concertos by Vivaldi, Leclair, Viotti, Haydn,…

The dynamic characteristics of HTGR (High Temperature Gas Cooled Reactor) system, (2)

International Nuclear Information System (INIS)

Kudo, Kazuhiko; Ohta, Masao; Kawasaki, Hidenori

1979-01-01

The dynamic characteristics of a HTGR plant, which has two cooling loops, was investigated. The analytical model consists of the core with fuel sleeves, coolant channels and blocks, the upper and lower reflectors, the high and low temperature plenums, two double wall pipings, two intermediate heat exchangers and the secondary system. The key plant parameters for calculation were as follows: the core outlet gas temperature 1000 deg C, the reactor thermal output 50 MW, the flow rate of primary coolant gas 7.96 kg/sec-loop and the pressure of primary coolant gas 40 kg/cm 2 at the rated operating condition. The calculating parameters were fixed as follows: the time interval for core characteristic analysis 0.1 sec, the time interval for thermal characteristic analysis 5.0 sec, the number of division of fuel channels 130, and the number of division of an intermediate heat exchanger 200. The assumptions for making the model were evaluated especially for the power distribution in the core and the heat transmission coefficients in the core, the double wall piping and the intermediate heat exchangers. Concerning the analytical results, the self-control to the outer disturbance of reactivity and the plant dynamic behavior due to the change of flow rate of primary and secondary coolants, and the change of gas temperature of secondary coolant at the inlet of intermediate heat exchangers, are presented. (Nakai, Y.)

Welding. Performance Objectives. Intermediate Course.

Science.gov (United States)

Vincent, Kenneth

Several intermediate performance objectives and corresponding criterion measures are listed for each of nine terminal objectives for an intermediate welding course. The materials were developed for a 36-week (3 hours daily) course designed to prepare the student for employment in the field of welding. Electric welding and specialized (TIG & MIG)…

Review of intermediate and final product characterization on coated particles preparation

International Nuclear Information System (INIS)

Sukarsono; Kristanti Nurwidyaningrum

2015-01-01

Review of the intermediate and final product characterization on preparation of coated particles was done. Product characterization included a tool to measure the character of raw materials, intermediate product and the final product of the process, which affects the success of getting the high temperature reactor fuel are eligible. Equipment's for the characterization of such materials were pH meter, viscometer, microbalance, turbidity meter, tab density measurement, true density measurement and auto pycnometer. Being for the measurement of particles there are two types destructive testing and non destructive. Destructive testing was done by polished the particles then cross sectional imaging of particle observed using an optical microscope. In this way contains errors due to polishing treatment that could not be right on the equator section so it needs correction. Destructive testing also create waste that must be processed from the remnants of the polishing process. By using non-destructive testing, waste was not formed but the imaging results are often unclear due to lack of contrast. Development of non-destructive test equipment has been made using radiographic method and automated microscopy. The overall activity is still much needed additional tools for measurement and for processing, so that the results obtained will not rejected as the specification of nuclear fuel. Similarly, in the case of a sampling test method and limits to a product accepted or rejected, it should be determined based on statistical methods. (author)

Magnetic switch for reactor control rod

International Nuclear Information System (INIS)

Germer, J.H.

1986-01-01

This patent describes a control rod system for a nuclear reactor utilizing an electromagnetic grapple mechanism for holding and releasing a control rod, the improvement comprising a magnetic reed switch assembly having a Curie-point magnetic shunt and responsive to reactor coolant temperature for short circuiting the electromagnetic grapple mechanism causing release of the control rod when the coolant temperature reaches the Curie-point of the magnetic shunt. The magnetic reed switch assembly includes a: a permanent magnet, a pair of magnetic pole pieces located at and in contact with opposite ends of the permanent magnet, the Curie-point magnetic shunt being positioned adjacent the permanent magnet and in contact with the pair of magnetic pole pieces, and a reed switch positioned intermediate the pole pieces and provided with a pair of ferromagnetic reeds, a nonmagnetic enclosure around the reeds, a first of the reeds being secured at one end to a first of the pair of pole pieces, a second of the reeds having one end extending into and secured to a hollow member positioned in and extending through a second of the pair of pole pieces, the one end of the second of the reeds secured to a condector adapted to be connected to the electromagnetic grapple mechanism

Nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

This patent describes an improved liquid metal nuclear reactor construction comprising: (a) a nuclear reactor core having a bottom platform support structure; (b) a reactor vessel for holding a large pool of low pressure liquid metal coolant and housing the core; (c) a containment structure surrounding the reactor vessel and having a sidewall spaced outwardly from the reactor vessel side wall and having a base mat spaced below the reactor vessel bottom end wall; (d) a central small diameter post anchored to the containment structure base mat and extending upwardly to the reactor vessel to axially fix the bottom end wall of the reactor vessel and provide a center column support for the lower end of the reactor core; (e) annular support structure disposed in the reactor vessel on the bottom end wall and extending about the lower end of the core; (f) structural support means disposed between the containment structure base mat and bottom end of the reactor vessel wall and cooperating for supporting the reactor vessel at its bottom end wall on the containment structure base mat to allow the reactor vessel to expand radially but substantially prevent any lateral motions that might be imposed by the occurrence of a seismic event; (g) a bed of insulating material disposed between the containment structure base mat and the bottom end wall of the reactor vessel and uniformly supporting the reactor vessel at its bottom end wall; freely expand radially from the central post as it heats up while providing continuous support thereof; (h) a deck supported upon the wall of the containment vessel above the top open end of the reactor vessel; and (i) extendible and retractable coupling means extending between the deck and the top open end of the reactor vessel and flexibly and sealably interconnecting the reactor vessel at its top end to the deck

A Takagi–Sugeno fuzzy power-distribution method for a prototypical advanced reactor considering pump degradation

Directory of Open Access Journals (Sweden)

Yue Yuan

2017-08-01

Full Text Available Advanced reactor designs often feature longer operating cycles between refueling and new concepts of operation beyond traditional baseload electricity production. Owing to this increased complexity, traditional proportional–integral control may not be sufficient across all potential operating regimes. The prototypical advanced reactor (PAR design features two independent reactor modules, each connected to a single dedicated steam generator that feeds a common balance of plant for electricity generation and process heat applications. In the current research, the PAR is expected to operate in a load-following manner to produce electricity to meet grid demand over a 24-hour period. Over the operational lifetime of the PAR system, primary and intermediate sodium pumps are expected to degrade in performance. The independent operation of the two reactor modules in the PAR may allow the system to continue operating under degraded pump performance by shifting the power production between reactor modules in order to meet overall load demands. This paper proposes a Takagi–Sugeno (T–S fuzzy logic-based power distribution system. Two T–S fuzzy power distribution controllers have been designed and tested. Simulation shows that the devised T–S fuzzy controllers provide improved performance over traditional controls during daily load-following operation under different levels of pump degradation.

A Takagi-Sugeno fuzzy power-distribution method for a prototypical advanced reactor considering pump degradation

Energy Technology Data Exchange (ETDEWEB)

Yuan, Yue [Institute of Nuclear and New Energy Technology, Tsinghua University, Collaborative Innovation Center of Advanced Nuclear Energy Technology, Key Laboratory of Advanced Reactor Engineering and Safety of Ministry of Education, Beijing (China); Coble, Jamie [Dept. of Nuclear Engineering, University of Tennessee, Knoxville (United States)

2017-08-15

Advanced reactor designs often feature longer operating cycles between refueling and new concepts of operation beyond traditional baseload electricity production. Owing to this increased complexity, traditional proportional–integral control may not be sufficient across all potential operating regimes. The prototypical advanced reactor (PAR) design features two independent reactor modules, each connected to a single dedicated steam generator that feeds a common balance of plant for electricity generation and process heat applications. In the current research, the PAR is expected to operate in a load-following manner to produce electricity to meet grid demand over a 24-hour period. Over the operational lifetime of the PAR system, primary and intermediate sodium pumps are expected to degrade in performance. The independent operation of the two reactor modules in the PAR may allow the system to continue operating under degraded pump performance by shifting the power production between reactor modules in order to meet overall load demands. This paper proposes a Takagi–Sugeno (T–S) fuzzy logic-based power distribution system. Two T–S fuzzy power distribution controllers have been designed and tested. Simulation shows that the devised T–S fuzzy controllers provide improved performance over traditional controls during daily load-following operation under different levels of pump degradation.

The pebble-bed high-temperature reactor as a source of nuclear process heat. Vol. 3

International Nuclear Information System (INIS)

Kugeler, K.; Schulten, R.; Kugeler, M.; Niessen, H.F.; Roeth-Kamat, M.; Hohn, H.; Woike, O.; Germer, J.H.

1974-08-01

The characteristic questions concerning a process heat reactor with high helium outlet temperatures are dealt with in this volume like e.g. fuel element design, corrosion, and fission product release. Furthermore, some possibilities of the technical realization of the hot-gas ducting and intermediate heat exchangers are described. Important parameters for the design of the reactor such as core power density, helium inlet and outlet temperatures, helium pressure and fuel cycle burn-up and conversion and the effect of these on the primary circuit are investigated. The important question regarding which reactor vessel is to be chosen for nuclear process heat plants is discussed with the aid of the integrated and non-integrated concepts using prestressed concrete, cast iron and cast steel. Thereafter, considerations on the safety of the nuclear plant are given. Finally, mention is made of the availability of the nuclear plant and of the status of development of the HTR technology. (orig.) [de

Analysis on approach of safeguards implementation at research reactor handling item count and bulk material

International Nuclear Information System (INIS)

Kim, Hyun Jo; Lee, Sung Ho; Lee, Byung Doo; Jung, Juang

2016-01-01

KiJang research reactor (KJRR) will be constructed to produce the radioisotope such as Mo-99 etc., provide the neutron transmutation doping (NTD) service of silicon, and develop the core technologies of research reactor. In this paper, the features of the process and nuclear material flow are reviewed and the material balance area (MBA) and key measurement point (KMP) are established based on the nuclear material flow. Also, this paper reviews the approach on safeguards and nuclear material accountancy at the facility level for Safeguards-by-Design at research reactor handling item count and bulk material. In this paper, MBA and KMPs are established through the analysis on facility features and major process at KJRR handling item count and bulk material. Also, this paper reviews the IAEA safeguards implementation and nuclear material accountancy at KJRR. It is necessary to discuss the safeguards approach on the fresh FM target assemblies and remaining uranium in the intermediate level liquid wastes

Analysis on approach of safeguards implementation at research reactor handling item count and bulk material

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyun Jo; Lee, Sung Ho; Lee, Byung Doo; Jung, Juang [KAERI, Daejeon (Korea, Republic of)

2016-05-15

KiJang research reactor (KJRR) will be constructed to produce the radioisotope such as Mo-99 etc., provide the neutron transmutation doping (NTD) service of silicon, and develop the core technologies of research reactor. In this paper, the features of the process and nuclear material flow are reviewed and the material balance area (MBA) and key measurement point (KMP) are established based on the nuclear material flow. Also, this paper reviews the approach on safeguards and nuclear material accountancy at the facility level for Safeguards-by-Design at research reactor handling item count and bulk material. In this paper, MBA and KMPs are established through the analysis on facility features and major process at KJRR handling item count and bulk material. Also, this paper reviews the IAEA safeguards implementation and nuclear material accountancy at KJRR. It is necessary to discuss the safeguards approach on the fresh FM target assemblies and remaining uranium in the intermediate level liquid wastes.

Money distribution with intermediation

OpenAIRE

Teles, Caio Augusto Colnago

2013-01-01

This pap er analyzes the distribution of money holdings in a commo dity money search-based mo del with intermediation. Intro ducing heterogeneity of costs to the Kiyotaki e Wright ( 1989 ) mo del, Cavalcanti e Puzzello ( 2010) gives rise to a non-degenerated distribution of money. We extend further this mo del intro ducing intermediation in the trading pro cess. We show that the distribution of money matters for savings decisions. This gives rises to a xed p oint problem for the ...

Strategic review on management and disposal of low- and intermediate-level solid radwastes

International Nuclear Information System (INIS)

Li Xuequn

1993-01-01

An overview on the actual status of solid low- and intermediate-level wastes (L/ILW) management in China is described. Some of the main problems at present are analysed. The strategies on management and disposal of the wastes are discussed in light of systematology. A large amount of solid L/ILW and distilled residual solution to be solidified have been accumulated during the past 30 years development of nuclear industry in China. These wastes, containing fission products, activated products, and uranium and transuranium elements respectively, mainly come from nuclear reactors, spent fuel reprocessing plants, and nuclear fuel fabrication plants. In the century, solid L/ILW and solidified wastes are produced mainly by nuclear industry; but in the next century, solid wastes will be steadily produced mainly from nuclear power plants

Search for intermediate vector bosons

International Nuclear Information System (INIS)

Klajn, D.B.; Rubbia, K.; Meer, S.

1983-01-01

Problem of registration and search for intermediate vector bosons is discussed. According to weak-current theory there are three intermediate vector bosons with +1(W + )-1(W - ) and zero (Z 0 ) electric charges. It was suggested to conduct the investigation into particles in 1976 by cline, Rubbia and Makintair using proton-antiproton beams. Major difficulties of the experiment are related to the necessity of formation of sufficient amount of antiparticles and the method of antiproton beam ''cooling'' for the purpose of reduction of its random movements. The stochastic method was suggested by van der Meer in 1968 as one of possible cooling methods. Several large detectors were designed for searching intermediate vector bosons

Nuclear reactor seismic fuel assembly grid

International Nuclear Information System (INIS)

Anthony, A.J.

1977-01-01

The strength of a nuclear reactor fuel assembly is enhanced by increasing the crush strength of the zircaloy spacer grids which locate and support the fuel elements in the fuel assembly. Increased resistance to deformation as a result of laterally directed forces is achieved by increasing the section modulus of the perimeter strip through bending the upper and lower edges thereof inwardly. The perimeter strip is further rigidized by forming, in the central portion thereof, dimples which extend inwardly with respect to the fuel assembly. The integrity of the spacer grid may also be enhanced by providing back-up arches for some or all of the integral fuel element locating springs and the strength of the fuel assembly may be further enhanced by providing, intermediate its ends, a steel seismic grid. 13 claims, 6 figures

Validation of intermediate end points in cancer research.

Science.gov (United States)

Schatzkin, A; Freedman, L S; Schiffman, M H; Dawsey, S M

1990-11-21

Investigations using intermediate end points as cancer surrogates are quicker, smaller, and less expensive than studies that use malignancy as the end point. We present a strategy for determining whether a given biomarker is a valid intermediate end point between an exposure and incidence of cancer. Candidate intermediate end points may be selected from case series, ecologic studies, and animal experiments. Prospective cohort and sometimes case-control studies may be used to quantify the intermediate end point-cancer association. The most appropriate measure of this association is the attributable proportion. The intermediate end point is a valid cancer surrogate if the attributable proportion is close to 1.0, but not if it is close to 0. Usually, the attributable proportion is close to neither 1.0 nor 0; in this case, valid surrogacy requires that the intermediate end point mediate an established exposure-cancer relation. This would in turn imply that the exposure effect would vanish if adjusted for the intermediate end point. We discuss the relative advantages of intervention and observational studies for the validation of intermediate end points. This validation strategy also may be applied to intermediate end points for adverse reproductive outcomes and chronic diseases other than cancer.

Biocatalytic Synthesis of Chiral Pharmaceutical Intermediates

Directory of Open Access Journals (Sweden)

Ramesh N. Patel

2004-01-01

Full Text Available The production of single enantiomers of drug intermediates has become increasingly important in the pharmaceutical industry. Chiral intermediates and fine chemicals are in high demand from both the pharmaceutical and agrochemical industries for the preparation of bulk drug substances and agricultural products. The enormous potential of microorganisms and enzymes for the transformation of synthetic chemicals with high chemo-, regio- and enantioselectivities has been demonstrated. In this article, biocatalytic processes are described for the synthesis of chiral pharmaceutical intermediates.

Quantification of the Keto-Hydroperoxide (HOOCH2OCHO) and Other Elusive Intermediates during Low-Temperature Oxidation of Dimethyl Ether

KAUST Repository

Moshammer, Kai

2016-09-17

This work provides new temperature-dependent mole fractions of elusive intermediates relevant to the low-temperature oxidation of dimethyl ether (DME). It extends the previous study of Moshammer et al. [ J. Phys. Chem. A 2015, 119, 7361–7374] in which a combination of a jet-stirred reactor and molecular beam mass spectrometry with single-photon ionization via tunable synchrotron-generated vacuum-ultraviolet radiation was used to identify (but not quantify) several highly oxygenated species. Here, temperature-dependent concentration profiles of 17 components were determined in the range of 450–1000 K and compared to up-to-date kinetic modeling results. Special emphasis is paid toward the validation and application of a theoretical method for predicting photoionization cross sections that are hard to obtain experimentally but essential to turn mass spectral data into mole fraction profiles. The presented approach enabled the quantification of the hydroperoxymethyl formate (HOOCH2OCH2O), which is a key intermediate in the low-temperature oxidation of DME. The quantification of this keto-hydroperoxide together with the temperature-dependent concentration profiles of other intermediates including H2O2, HCOOH, CH3OCHO, and CH3OOH reveals new opportunities for the development of a next-generation DME combustion chemistry mechanism.

Nuclear reactor spacer assembly

International Nuclear Information System (INIS)

Anthony, A.J.; Groves, M.D.

1979-01-01

A fuel assembly for a nuclear reactor is disclosed wherein the fuel element receiving and supporting grid is comprised of a first metal, the guide tubes which pass through the grid assembly are comprised of a second metal and the grid is supported on the guide tubes by means of expanded sleeves located intermediate the grid and guide tubes. The fuel assembly is fabricated by inserting the sleeves, of initial outer diameter commensurate with the guide tube outer diameters, through the holes in the grid assembly provided for the guide tubes and thereafter expanding the sleeves radially outwardly along their entire length such that the guide tubes can subsequently be passed through the sleeves. The step of radial expansion, as a result of windows provided in the sleeves having dimensions commensurate with the geometry of the grid, mechanically captures the grid and simultaneously preloads the sleeve against the grid whereby relative motion between the grid and guide tube will be precluded

Prototype of thermal degradation for radioactive wastes of low and intermediate level

International Nuclear Information System (INIS)

Diaz A, L.V.; Pacheco S, J.O.; Pacheco P, M.; Monroy G, F.; Emeterio H, M.

2005-01-01

At the present time, the scientific, academic, industrial and technological activities, generate great quantity of radioactive wastes of low and intermediate level (DRNBI). For to assure an appropriate final disposal of these, it is intended their treatment and vitrification by means of thermal plasma. This alternative offers multiple advantages in an only process: elevated energy density (105W/cm 3 ), high enthalpy (1400 kJ/mol), elevated chemical reactivity, quick quenching (106K/s) and operation temperatures of 4000 to 15000K; this allows the treatment of a great diversity of waste. Those reactors are compact and they work to atmospheric pressure and reduced thermal inertia. This technology allows to degrade DRNBI and to contain them in a vitreous matrix by means of a system made up of a reactor, canyon of plasma, of monitoring, of washing of gases and of control. Besides the design and general characteristics of the Prototype of Thermal Degradation of DRNBI, they are reported in this work the advances achieved in the selection of the ceramic material for the vitrification. Their characterization was carried out by means of SEM and XRD. With the preliminary results it can discern that the material but appropriate to be used as vitreous matrix is a ceramic clay. With the development of the proposed technology and the material for the vitreous matrix, it will be to treat DRNBI. (Author)

The production of refined intermediate fuels with high temperature reactors

International Nuclear Information System (INIS)

Nowacki, P.J.

1977-01-01

Power plants can be divided into conventional steam plants, fueled with hard coal, lignite or liquid fuel, hydroelectric plants and nuclear plants, their chief use was or is the production of electric energy and - in certain cases only - of production of process heat, using steam or hot water for process heat in industry and district heating for residential and commercial purposes. The part played by electricity in the whole energy demand is of the order of 10% to 25% the total demand, the rest is necessary for supplying process heat below 200 0 C or above 200 0 C, up to some 1500 0 C. The present distribution of energy demands is covered chiefly by liquid fuel, coal and lignite, water energy and increasing steps by nuclear fuel. It is well known that the erection of nuclear energy plants is a necessity for today and for the future. There is another necessity, i.e. to utilize the primary energy resources in a complex way i.e. to supply electricity as energy vector and other fuels as process heat as new energy vectors. These manmade fuels - whether in a gaseous or liquid phase - contain hydrogen, and one can believe, the world is entering a new energy civilisation in utilizing hydrogen and its compounds as second energy vector. The author has taken up the task to investigate this new problem of process, heat in the form of hydrogen and its compounds, by evaluating their present and future production, based on the utilization of natural gas, oil coal, water and the nuclear heat of helium, available in a closed circuit as primary coolant in a High - Temeprature Helium cooled reactor, which is symbolized in the paper as HTR. The paper deals in more detail with the following application of Nuclear Heat: hydrogasification, direct reduction of ore, mainly iron ores, ammonia synthesis, methanol synthesis Hydrocracking, long distance transfer of process heat (chemical heat pipe), hydrogenation of coal, Fischer - Tropsch synthesis, oxosynthesis, coal gasification, coal

Simplifying biochemical models with intermediate species

DEFF Research Database (Denmark)

Feliu, Elisenda; Wiuf, Carsten

2013-01-01

techniques, we study systematically the effects of intermediate, or transient, species in biochemical systems and provide a simple, yet rigorous mathematical classification of all models obtained from a core model by including intermediates. Main examples include enzymatic and post-translational modification...... systems, where intermediates often are considered insignificant and neglected in a model, or they are not included because we are unaware of their existence. All possible models obtained from the core model are classified into a finite number of classes. Each class is defined by a mathematically simple...... canonical model that characterizes crucial dynamical properties, such as mono- and multistationarity and stability of steady states, of all models in the class. We show that if the core model does not have conservation laws, then the introduction of intermediates does not change the steady...

Specialists' meeting on heat exchanging components of gas-cooled reactors

Energy Technology Data Exchange (ETDEWEB)

NONE

1984-07-01

The objective of the Meeting sponsored by IAEA was to provide a forum for the exchange and discussion of technical information related to heat exchanging and heat conducting components for gas-cooled reactors. The technical part of the meeting covered eight subjects: Heat exchanging components for process heat applications, design and requirements, and research and development programs; Status of the design and construction of intermediate He/He exchangers; Design, construction and performance of steam generators; Metallic materials and design codes; Design and construction of valves and hot gas ducts; Description of component test facilities and test results; Manufacturing of heat exchanging components.

Specialists' meeting on heat exchanging components of gas-cooled reactors

International Nuclear Information System (INIS)

1984-01-01

The objective of the Meeting sponsored by IAEA was to provide a forum for the exchange and discussion of technical information related to heat exchanging and heat conducting components for gas-cooled reactors. The technical part of the meeting covered eight subjects: Heat exchanging components for process heat applications, design and requirements, and research and development programs; Status of the design and construction of intermediate He/He exchangers; Design, construction and performance of steam generators; Metallic materials and design codes; Design and construction of valves and hot gas ducts; Description of component test facilities and test results; Manufacturing of heat exchanging components

Balancing Near-Field Enhancement, Absorption, and Scattering for Effective Antenna-Reactor Plasmonic Photocatalysis.

Science.gov (United States)

Li, Kun; Hogan, Nathaniel J; Kale, Matthew J; Halas, Naomi J; Nordlander, Peter; Christopher, Phillip

2017-06-14

Efficient photocatalysis requires multifunctional materials that absorb photons and generate energetic charge carriers at catalytic active sites to facilitate a desired chemical reaction. Antenna-reactor complexes are an emerging multifunctional photocatalytic structure where the strong, localized near field of the plasmonic metal nanoparticle (e.g., Ag) is coupled to the catalytic properties of the nonplasmonic metal nanoparticle (e.g., Pt) to enable chemical transformations. With an eye toward sustainable solar driven photocatalysis, we investigate how the structure of antenna-reactor complexes governs their photocatalytic activity in the light-limited regime, where all photons need to be effectively utilized. By synthesizing core@shell/satellite (Ag@SiO 2 /Pt) antenna-reactor complexes with varying Ag nanoparticle diameters and performing photocatalytic CO oxidation, we observed plasmon-enhanced photocatalysis only for antenna-reactor complexes with antenna components of intermediate sizes (25 and 50 nm). Optimal photocatalytic performance was shown to be determined by a balance between maximized local field enhancements at the catalytically active Pt surface, minimized collective scattering of photons out of the catalyst bed by the complexes, and minimal light absorption in the Ag nanoparticle antenna. These results elucidate the critical aspects of local field enhancement, light scattering, and absorption in plasmonic photocatalyst design, especially under light-limited illumination conditions.

Nuclear heat source component design considerations for HTGR process heat reactor plant concept

International Nuclear Information System (INIS)

McDonald, C.F.; Kapich, D.; King, J.H.; Venkatesh, M.C.

1982-01-01

Using alternate energy sources abundant in the U.S.A. to help curb foreign oil imports is vitally important from both national security and economic standpoints. Perhaps the most forwardlooking opportunity to realize national energy goals involves the integrated use of two energy sources that have an established technology base in the U.S.A., namely nuclear energy and coal. The coupling of a high-temperature gas-cooled reactor (HTGR) and a chemical process facility has the potential for long-term synthetic fuel production (i.e., oil, gasoline, aviation fuel, hydrogen, etc.) using coal as the carbon source. Studies are in progress to exploit the high-temperature capability of an advanced HTGR variant for nuclear process heat. The process heat plant discussed in this paper has a 1170-MW(t) reactor as the heat source and the concept is based on indirect reforming, i.e., the high-temperature nuclear thermal energy is transported (via an intermediate heat exchanger (IHX)) to the externally located process plant by a secondary helium transport loop. Emphasis is placed on design considerations for the major nuclear heat source (NHS) components, and discussions are presented for the reactor core, prestressed concrete reactor vessel (PCRV), rotating machinery, and heat exchangers

Technical and economic studies of small reactors for supply of electricity and steam

International Nuclear Information System (INIS)

Spiewak, I.; Klepper, O.H.; Fuller, L.C.

1977-01-01

Several years ago conventional opinion held that nuclear power plants must be very large to be competitive with fossil fuels. This situation has changed markedly in most countries within recent years, as oil and gas supplies have become more scarce and costly. Studies have been carried out of several nuclear steam supply systems in the small and intermediate size range. Detail studies are reported of the Consolidated Nuclear Steam Generator (CNSG), a 313 MW(t) pressurized water reactor being developed by Babcock and Wilcox, as applied to industrial energy needs. Both conventional and barge-mounted nuclear steam supply systems are considered. Conceptual studies have been started of pressurized and boiling water reactors in the range of 1000 MW(t), which are envisioned for utility operation for supply of electric power and steam. Design studies of a 500 MW(t) high temperature reactor are also reported. The small reactors are expected to have higher unit costs than the large commercial plants, but to have compensating advantages in higher plant availability, shorter construction schedule and greater siting flexibility. Studies are also reported of power cycle parameters and cost allocations for extraction of steam from steam turbine plants. This steam could be used for industrial energy, district heating or desalination

Technical and economic studies of small reactors for supply of electricity and steam

International Nuclear Information System (INIS)

Spiewak, I.; Klepper, O.H.; Fuller, L.C.

1977-02-01

Several years ago conventional opinion held that nuclear power plants must be very large to be competitive with fossil fuels. This situation has changed markedly in most countries within recent years, as oil and gas supplies have become more scarce and costly. Studies have been carried out for several nuclear steam supply systems in the small and intermediate size range. Detail studies are reported of the Consolidated Nuclear Steam Generator (CNSG), a 365 MW(th) pressurized water reactor being developed by Babcock and Wilcox, as applied to industrial energy needs. Both conventional and barge-mounted nuclear steam supply systems are considered. Conceptual studies have been started of pressurized and boiling water reactors in the range of 1000 MW(th), which are envisioned for utility operation for supply of electric power and steam. Design studies of a 500 MW(th) high temperature reactor are also reported. The small reactors are expected to have higher unit costs than the large commercial plants, but to have compensating advantages in higher plant availability, shorter construction schedule, and greater siting flexibility. Studies are also reported of power cycle parameters and cost allocations for extraction of steam from steam turbine plants. This steam could be used for industrial energy, district heating, or desalination

Technical and economic studies of small reactors for supply of electricity and steam

International Nuclear Information System (INIS)

Spiewak, I.; Klepper, O.H.; Fuller, L.C.

1977-01-01

Several years ago conventional opinion held that nuclear power plants must be very large to be competitive with fossil fuels. This situation has changed markedly in most countries within recent years, as oil and gas supplies have become more scarce and costly. Studies have been carried out of several nuclear steam supply systems in the small and intermediate size range. Detail studies are reported of the Consolidated Nuclear Steam Generator (CNSG), a 313MW(th) pressurized water reactor being developed by Babcock and Wilcox, as applied to industrial energy needs. Both conventional and barge-mounted nuclear steam supply systems are considered. Conceptual studies have been started of pressurized and boiling water reactors in the range of 1000MW(th), which are envisioned for utility operation for supply of electric power and steam. Design studies of a 500MW(th) high temperature reactor are also reported. The small reactors are expected to have higher unit costs than the large commercial plants, but to have compensating advantages in higher plant availability, shorter construction schedule and greater siting flexibility. Studies are also reported of power cycle parameters and cost allocations for extraction of steam from steam turbine plants. This steam could be used for industrial energy, district heating or desalination. (author)

Mechanical Property and Its Comparison of Superalloys for High Temperature Gas Cooled Reactor

International Nuclear Information System (INIS)

Kim, Woo Gon; Kim, D. W.; Ryu, W. S.; Han, C. H.; Yoon, J. H.; Chang, J.

2005-01-01

Since structural materials for high temperature gas cooled reactor are used during long period in nuclear environment up to 1000 .deg. C, it is important to have good properties at elevated temperature such as mechanical properties (tensile, creep, fatigue, creep-fatigue), microstructural stability, interaction between metal and gas, friction and wear, hydrogen and tritium permeation, irradiation behavior, corrosion by impurity in He. Thus, in order to select excellent materials for the high temperature gas cooled reactor, it is necessary to understand the material properties and to gather the data for them. In this report, the items related to material properties which are needed for designing the high temperature gas cooled reactor were presented. Mechanical properties; tensile, creep, and fatigue etc. were investigated for Haynes 230, Hastelloy-X, In 617 and Alloy 800H, which can be used as the major structural components, such as intermediate heat exchanger (IHX), hot duct and piping and internals. Effect of He and irradiation on these structural materials was investigated. Also, mechanical properties; physical properties, tensile properties, creep and creep crack growth rate were compared for them, respectively. These results of this report can be used as important data to select superior materials for high temperature gas reactor

Reactions of stabilized Criegee Intermediates

Science.gov (United States)

Vereecken, Luc; Harder, Hartwig; Novelli, Anna

2014-05-01

Carbonyl oxides (Criegee intermediates) were proposed as key intermediates in the gas phase ozonolysis of alkenes in 1975 by Rudolf Criegee. Despite the importance of ozonolysis in atmospheric chemistry, direct observation of these intermediates remained elusive, with only indirect experimental evidence for their role in the oxidation of hydrocarbons, e.g. through scavenging experiments. Direct experimental observation of stabilized CI has only been achieved since 2008. Since then, a concerted effort using experimental and theoretical means is in motion to characterize the chemistry and kinetics of these reactive intermediates. We present the results of theoretical investigations of the chemistry of Criegee intermediates with a series of coreactants which may be of importance in the atmosphere, in experimental setups, or both. This includes the CI+CI cross-reaction, which proceeds with a rate coefficient near the collision limit and can be important in experimental conditions. The CI + alkene reactions show strong dependence of the rate coefficient depending on the coreactants, but is generally found to be rather slow. The CI + ozone reaction is sufficiently fast to occur both in experiment and the free troposphere, and acts as a sink for CI. The reaction of CI with hydroperoxides, ROOH, is complex, and leads both to the formation of oligomers, as to the formation of reactive etheroxides, with a moderately fast rate coefficient. The importance of these reactions is placed in the context of the reaction conditions in different atmospheric environments ranging from unpolluted to highly polluted.

Using activation method to measure neutron spectrum in an irradiation chamber of a research reactor

International Nuclear Information System (INIS)

Zhou Xuemei; Liu Guimin; Wang Xiaohe; Li Da; Meng Lingjie

2014-01-01

Neutron spectrum should be measured before test samples are irradiated. Neutron spectrum in an irradiation chamber of a research reactor was measured by using activation method when the reactor is in normal operation under 2 MW. Sixteen kinds of non-fission foils (19 reaction channels) were selected, of which 10 were sensitive to thermal and intermediate energy regions, while the others were of different threshold energy and sensitive to fast energy regions. By measuring the foil radioactivity, the neutron spectrum was unfolded with the iterative methods SAND-II and MSIT. Finally, shielding corrections of group cross-section and main factors affecting the calculation accuracy were studied and the uncertainty of solution was analyzed using the Monte Carlo method in the process of SAND-II. (authors)

The intermediate endpoint effect in logistic and probit regression

Science.gov (United States)

MacKinnon, DP; Lockwood, CM; Brown, CH; Wang, W; Hoffman, JM

2010-01-01

Background An intermediate endpoint is hypothesized to be in the middle of the causal sequence relating an independent variable to a dependent variable. The intermediate variable is also called a surrogate or mediating variable and the corresponding effect is called the mediated, surrogate endpoint, or intermediate endpoint effect. Clinical studies are often designed to change an intermediate or surrogate endpoint and through this intermediate change influence the ultimate endpoint. In many intermediate endpoint clinical studies the dependent variable is binary, and logistic or probit regression is used. Purpose The purpose of this study is to describe a limitation of a widely used approach to assessing intermediate endpoint effects and to propose an alternative method, based on products of coefficients, that yields more accurate results. Methods The intermediate endpoint model for a binary outcome is described for a true binary outcome and for a dichotomization of a latent continuous outcome. Plots of true values and a simulation study are used to evaluate the different methods. Results Distorted estimates of the intermediate endpoint effect and incorrect conclusions can result from the application of widely used methods to assess the intermediate endpoint effect. The same problem occurs for the proportion of an effect explained by an intermediate endpoint, which has been suggested as a useful measure for identifying intermediate endpoints. A solution to this problem is given based on the relationship between latent variable modeling and logistic or probit regression. Limitations More complicated intermediate variable models are not addressed in the study, although the methods described in the article can be extended to these more complicated models. Conclusions Researchers are encouraged to use an intermediate endpoint method based on the product of regression coefficients. A common method based on difference in coefficient methods can lead to distorted

Plant Control Concept for the Sodium Cooled Fast Reactor

International Nuclear Information System (INIS)

Kim, Eui Kwang; Kim, S. O.

2010-12-01

A power plant is designed for incorporation into a utility's grid system and follows the load demand through the steam generator, intermediate heat exchanger(IHX), from the nuclear core. During the load-following transients, various plant parameters must be controlled to protect the reactor core and other components in the plant. The purpose of this report is to review design considerations to establish SFR plant control and to design plant control concepts. The governing equations and solution procedure of the computer code to calculate plant temperature conditions during the part-load operation was reviewed and 4 types of plant operation concepts were designed, and the results of the calculations were compared

Generation IV reactors: reactor concepts

International Nuclear Information System (INIS)

Cardonnier, J.L.; Dumaz, P.; Antoni, O.; Arnoux, P.; Bergeron, A.; Renault, C.; Rimpault, G.; Delpech, M.; Garnier, J.C.; Anzieu, P.; Francois, G.; Lecomte, M.

2003-01-01

Liquid metal reactor concept looks promising because of its hard neutron spectrum. Sodium reactors benefit a large feedback experience in Japan and in France. Lead reactors have serious assets concerning safety but they require a great effort in technological research to overcome the corrosion issue and they lack a leader country to develop this innovative technology. In molten salt reactor concept, salt is both the nuclear fuel and the coolant fluid. The high exit temperature of the primary salt (700 Celsius degrees) allows a high energy efficiency (44%). Furthermore molten salts have interesting specificities concerning the transmutation of actinides: they are almost insensitive to irradiation damage, some salts can dissolve large quantities of actinides and they are compatible with most reprocessing processes based on pyro-chemistry. Supercritical water reactor concept is based on operating temperature and pressure conditions that infers water to be beyond its critical point. In this range water gets some useful characteristics: - boiling crisis is no more possible because liquid and vapour phase can not coexist, - a high heat transfer coefficient due to the low thermal conductivity of supercritical water, and - a high global energy efficiency due to the high temperature of water. Gas-cooled fast reactors combining hard neutron spectrum and closed fuel cycle open the way to a high valorization of natural uranium while minimizing ultimate radioactive wastes and proliferation risks. Very high temperature gas-cooled reactor concept is developed in the prospect of producing hydrogen from no-fossil fuels in large scale. This use implies a reactor producing helium over 1000 Celsius degrees. (A.C.)

Material development for gas-cooled high temperature reactors for the production of nuclear process heat

International Nuclear Information System (INIS)

Nickel, H.

1977-04-01

In the framework of the material development for gas-cooled high temperature reactors, considerable investigations of the materials for the reactor core and the primary cicuit are being conducted. Concerning the core components, the current state-of-the-art and the objectives of the development work on the spherical fuel elements, coated particles and structural graphite are discussed. As an example of the structural graphite, the non-replaceable reflector of the process heat reactor is discussed. The primary circuit will be constructed mainly from metallic materials, although some ceramics are also being considered. Components of interest are hot gas ducts, liners, methane reformer tubes and helium-helium intermediate heat exchangers. The gaseous impurities present in the helium coolant may cause oxidation and carburization of the nickel-base and iron-base alloys envisaged for use in these components, with a possible associated adverse effect on the mechanical properties such as creep and fatigue. Test capacity has therefore been installed to investigate materials behaviour in simulated reactor helium under both constant and alternating stress conditions. The first results on the creep behaviour of several alloys in impure helium are presented and discussed. (orig./GSC) [de

Experimental studies of tritium barrier concepts for fusion reactors

International Nuclear Information System (INIS)

Maroni, V.A.; Van Deventer, E.H.; Renner, T.A.; Pelto, R.H.; Wierdak, C.J.

1976-01-01

Ongoing experimental studies at ANL aimed at the development of methods to reduce tritium migration in fusion reactor systems currently include (1) work on the development of multilayered metal composites and impurity-coated refractory metals as barriers to tritium permeation in elevated temperature (greater than 300 0 C) structures and (2) investigations of the kinetics of tritium trapping reactions in inert gas purge streams under conditions that emulate fusion reactor environments. Significant results obtained thus far are (1) demonstration of greater than 50-fold reductions in the hydrogen permeability of stainless steel structures by using stainless steel-clad composites containing an intermediate layer of a selected copper alloy and (2) verification that surface-oxide coatings lead to greater than 100-fold reductions in the hydrogen permeability of vanadium, but that severe oxygen penetration and embrittlement of the vanadium occur at temperatures in the range from 300 to 800 0 C and under conditions of extremely low oxygen potential. Other considerations pertaining to the large-scale use of metal composites in fusion reactors are discussed, and progress in efforts to demonstrate the fabricability of metal composites is reviewed. Also presented are results of studies of the efficiencies of (1) CuO and CuO--MnO 2 beds in converting HT to HTO and (2) magnesium metal beds in converting HTO to HT

Gravity with Intermediate Goods Trade

Directory of Open Access Journals (Sweden)

Sujin Jang

2017-12-01

Full Text Available This paper derives the gravity equation with intermediate goods trade. We extend a standard monopolistic competition model to incorporate intermediate goods trade, and show that the gravity equation with intermediates trade is identical to the one without it except in that gross output should be used as the output measure instead of value added. We also show that the output elasticity of trade is significantly underestimated when value added is used as the output measure. This implies that with the conventional gravity equation, the contribution of output growth can be substantially underestimated and the role of trade costs reduction can be exaggerated in explaining trade expansion, as we demonstrate for the case of Korea's trade growth between 1995 and 2007.

A Review of Previous Research in Direct Energy Conversion Fission Reactors

International Nuclear Information System (INIS)

DUONG, HENRY; POLANSKY, GARY F.; SANDERS, THOMAS L.; SIEGEL, MALCOLM D.

1999-01-01

From the earliest days of power reactor development, direct energy conversion was an obvious choice to produce high efficiency electric power generation. Directly capturing the energy of the fission fragments produced during nuclear fission avoids the intermediate conversion to thermal energy and the efficiency limitations of classical thermodynamics. Efficiencies of more than 80% are possible, independent of operational temperature. Direct energy conversion fission reactors would possess a number of unique characteristics that would make them very attractive for commercial power generation. These reactors would be modular in design with integral power conversion and operate at low pressures and temperatures. They would operate at high efficiency and produce power well suited for long distance transmission. They would feature large safety margins and passively safe design. Ideally suited to production by advanced manufacturing techniques, direct energy conversion fission reactors could be produced more economically than conventional reactor designs. The history of direct energy conversion can be considered as dating back to 1913 when Moseleyl demonstrated that charged particle emission could be used to buildup a voltage. Soon after the successful operation of a nuclear reactor, E.P. Wigner suggested the use of fission fragments for direct energy conversion. Over a decade after Wigner's suggestion, the first theoretical treatment of the conversion of fission fragment kinetic energy into electrical potential appeared in the literature. Over the ten years that followed, a number of researchers investigated various aspects of fission fragment direct energy conversion. Experiments were performed that validated the basic physics of the concept, but a variety of technical challenges limited the efficiencies that were achieved. Most research in direct energy conversion ceased in the US by the late 1960s. Sporadic interest in the concept appears in the literature until this

Mobile communication and intermediality

DEFF Research Database (Denmark)

Helles, Rasmus

2013-01-01

communicative affordances of mobile devices in order to understand how people choose between them for different purposes. It is argued that mobile communication makes intermediality especially central, as the choice of medium is detached from the location of stationary media and begins to follow the user across......The article argues the importance of intermediality as a concept for research in mobile communication and media. The constant availability of several, partially overlapping channels for communication (texting, calls, email, Facebook, etc.) requires that we adopt an integrated view of the various...

Analysis of a sustainable gas cooled fast breeder reactor concept

International Nuclear Information System (INIS)

Kumar, Akansha; Chirayath, Sunil S.; Tsvetkov, Pavel V.

2014-01-01

Highlights: • A Thorium-GFBR breeder for actinide recycling ability, and thorium fuel feasibility. • A mixture of 232 Th and 233 U is used as fuel and LWR used fuel is used. • Detailed neutronics, fuel cycle, and thermal-hydraulics analysis has been presented. • Run this TGFBR for 20 years with breeding of 239 Pu and 233 U. • Neutronics analysis using MCNP and Brayton cycle for energy conversion are used. - Abstract: Analysis of a thorium fuelled gas cooled fast breeder reactor (TGFBR) concept has been done to demonstrate the self-sustainability, breeding capability, actinide recycling ability, and thorium fuel feasibility. Simultaneous use of 232 Th and used fuel from light water reactor in the core has been considered. Results obtained confirm the core neutron spectrum dominates in an intermediate energy range (peak at 100 keV) similar to that seen in a fast breeder reactor. The conceptual design achieves a breeding ratio of 1.034 and an average fuel burnup of 74.5 (GWd)/(MTHM) . TGFBR concept is to address the eventual shortage of 235 U and nuclear waste management issues. A mixture of thorium and uranium ( 232 Th + 233 U) is used as fuel and light water reactor used fuel is utilized as blanket, for the breeding of 239 Pu. Initial feed of 233 U has to be obtained from thorium based reactors; even though there are no thorium breeders to breed 233 U a theoretical evaluation has been used to derive the data for the source of 233 U. Reactor calculations have been performed with Monte Carlo radiation transport code, MCNP/MCNPX. It is determined that this reactor has to be fuelled once every 5 years assuming the design thermal power output as 445 MW. Detailed analysis of control rod worth has been performed and different reactivity coefficients have been evaluated as part of the safety analysis. The TGFBR concept demonstrates the sustainability of thorium, viability of 233 U as an alternate to 235 U and an alternate use for light water reactor used fuel as a

The CAREM reactor and present currents in reactor design

International Nuclear Information System (INIS)

Ordonez, J.P.

1990-01-01

INVAP has been working on the CAREM project since 1983. It concerns a very low power reactor for electrical energy generation. The design of the reactor and the basic criteria used were described in 1984. Since then, a series of designs have been presented for reactors which are similar to CAREM regarding the solutions presented to reduce the chance of major nuclear accidents. These designs have been grouped under different names: Advanced Reactors, Second Generation Reactors, Inherently Safe Reactors, or even, Revolutionary Reactors. Every reactor fabrication firm has, at least, one project which can be placed in this category. Presently, there are two main currents of Reactor Design; Evolutionary and Revolutionary. The present work discusses characteristics of these two types of reactors, some revolutionary designs and common criteria to both types. After, these criteria are compared with CAREM reactor design. (Author) [es

Very-high-temperature gas reactor environmental impacts assessment

International Nuclear Information System (INIS)

Baumann, C.D.; Barton, C.J.; Compere, E.L.; Row, T.H.

1977-08-01

The operation of a Very High Temperature Reactor (VHTR), a slightly modified General Atomic type High Temperature Gas-Cooled Reactor (HTGR) with 1600 F primary coolant, as a source of process heat for the 1400 0 F steam-methanation reformer step in a hydrogen producing plant (via hydrogasification of coal liquids) was examined. It was found that: (a) from the viewpoint of product contamination by fission and activation products, an Intermediate Heat Exchanger (IHX) is probably not necessary; and (b) long term steam corrosion of the core support posts may require increasing their diameter (a relatively minor design adjustment). However, the hydrogen contaminant in the primary coolant which permeates the reformer may reduce steam corrosion but may produce other problems which have not as yet been resolved. An IHX in parallel with both the reformer and steam generator would solve these problems, but probably at greater cost than that of increasing the size of the core support posts. It is recommended that this corrosion problem be examined in more detail, especially by investigating the performance of current fossil fuel heated reformers in industry. Detailed safety analysis of the VHTR would be required to establish definitely whether the IHX can be eliminated. Water and hydrogen ingress into the reactor system are potential problems which can be alleviated by an IHX. These problems will require analysis, research and development within the program required for development of the VHTR

Next Generation Nuclear Plant Intermediate Heat Exchanger Acquisition Strategy

Energy Technology Data Exchange (ETDEWEB)

Mizia, Ronald Eugene [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2008-04-01

DOE has selected the High Temperature Gas-cooled Reactor (HTGR) design for the Next Generation Nuclear Plant (NGNP) Project. The NGNP will demonstrate the use of nuclear power for electricity and hydrogen production. It will have an outlet gas temperature in the range of 900°C to 950°C and a plant design service life of 60 years. The reactor design will be a graphite moderated, helium cooled, prismatic or pebble-bed reactor, and use low-enriched uranium, TRISO-coated fuel. The plant size, reactor thermal power, and core configuration will ensure passive decay heat removal without fuel damage or radioactive material releases during accidents. The NGNP Materials Research and Development (R&D) Program is responsible for performing R&D on likely NGNP materials in support of the NGNP design, licensing, and construction activities. Selection of the technology and design configuration for the NGNP must consider both the cost and risk profiles to ensure that the demonstration plant establishes a sound foundation for future commercial deployments. The NGNP challenge is to achieve a significant advancement in nuclear technology while at the same time setting the stage for an economically viable deployment of the new technology in the commercial sector soon after 2020. The purpose of this report is to address the acquisition strategy for the NGNP Intermediate Heat Exchanger (IHX).This component will be operated in flowing, impure helium on the primary and secondary side at temperatures up to 950°C. There are major high temperature design, materials availability, and fabrication issues that need to be addressed. The prospective materials are Alloys 617, 230, 800H and X, with Alloy 617 being the leading candidate for the use at 950°C. The material delivery schedule for these materials does not pose a problem for a 2018 start up as the vendors can quote reasonable delivery times at the moment. The product forms and amount needed must be finalized as soon as possible. An

Three-dimensional fluid-structure interaction dynamics of a pool-reactor in-tank component

International Nuclear Information System (INIS)

Kulak, R.F.

1979-01-01

The safety evaluation of reactor-components often involves the analysis of various types of fluid/structural components interacting in three-dimensional space. For example, in the design of a pool-type reactor several vital in-tank components such as the primary pumps and the intermediate heat exchangers are contained within the primary tank. Typically, these components are suspended from the deck structure and largely submersed in the sodium pool. Because of this positioning these components are vulnerable to structural damage due to pressure wave propagation in the tank during a CDA. In order to assess the structural integrity of these components it is necessary to perform a dynamic analysis in three-dimensional space which accounts for the fluid-structure coupling. A model is developed which has many of the salient features of this fluid-structural component system

Swelling and swelling resistance possibilities of austenitic stainless steels in fusion reactors

International Nuclear Information System (INIS)

Maziasz, P.J.

1983-01-01

Fusion reactor helium generation rates in stainless steels are intermediate to those found in EBR-II and HFIR, and swelling in fusion reactors may differ from the fission swelling behavior. Advanced titanium-modified austenitic stainless steels exhibit much better void swelling resistance than AISI 316 under EBR-II (up to approx. 120 dpa) and HFIR (up to approx. 44 dpa) irradiations. The stability of fine titanium carbide (MC) precipitates plays an important role in void swelling resistance for the cold-worked titanium-modified steels irradiated in EBR-II. Futhermore, increased helium generation in these steels can (a) suppress void conversion, (b) suppress radiation-induced solute segregation (RIS), and (c) stabilize fine MC particles, if sufficient bubble nucleation occurs early in the irradation. The combined effects of helium-enhanced MC stability and helium-suppressed RIS suggest better void swelling resistance in these steels for fusion service than under EBR-II irradiation

Degradation pathway of malachite green in a novel dual-tank photoelectrochemical catalytic reactor

International Nuclear Information System (INIS)

Diao, Zenghui; Li, Mingyu; Zeng, Fanyin; Song, Lin; Qiu, Rongliang

2013-01-01

Highlights: • A novel dual-tank photoelectrochemical catalytic reactor was designed. • Malachite green degraded in bipolar double-effect mode. • Salt bridge replaced by a cation exchange membrane in the reactor. • Degradation pathways of malachite green in the cathode and anode tanks were similar. -- Abstract: A novel dual-tank photoelectrochemical catalytic reactor was designed to investigate the degradation pathway of malachite green. A thermally formed TiO 2 /Ti thin film electrode was used as photoanode, graphite was used as cathode, and a saturated calomel electrode was employed as the reference electrode in the reactor. In the reactor, the anode and cathode tanks were connected by a cation exchange membrane. Results showed that the decolorization ratio of malachite green in the anode and cathode was 98.5 and 96.5% after 120 min, respectively. Malachite green in the two anode and cathode tanks was oxidized, achieving the bipolar double effect. Malachite green in both the anode and cathode tanks exhibited similar catalytic degradation pathways. The double bond of the malachite green molecule was attacked by strong oxidative hydroxyl radicals, after which the organic compound was degraded by the two pathways into 4,4-bis(dimethylamino) benzophenone, 4-(dimethylamino) benzophenone, 4-(dimethylamino) phenol, and other intermediate products. Eventually, malachite green was degraded into oxalic acid as a small molecular organic acid, which was degraded by processes such as demethylation, deamination, nitration, substitution, addition, and other reactions

Classical model of intermediate statistics

International Nuclear Information System (INIS)

Kaniadakis, G.

1994-01-01

In this work we present a classical kinetic model of intermediate statistics. In the case of Brownian particles we show that the Fermi-Dirac (FD) and Bose-Einstein (BE) distributions can be obtained, just as the Maxwell-Boltzmann (MD) distribution, as steady states of a classical kinetic equation that intrinsically takes into account an exclusion-inclusion principle. In our model the intermediate statistics are obtained as steady states of a system of coupled nonlinear kinetic equations, where the coupling constants are the transmutational potentials η κκ' . We show that, besides the FD-BE intermediate statistics extensively studied from the quantum point of view, we can also study the MB-FD and MB-BE ones. Moreover, our model allows us to treat the three-state mixing FD-MB-BE intermediate statistics. For boson and fermion mixing in a D-dimensional space, we obtain a family of FD-BE intermediate statistics by varying the transmutational potential η BF . This family contains, as a particular case when η BF =0, the quantum statistics recently proposed by L. Wu, Z. Wu, and J. Sun [Phys. Lett. A 170, 280 (1992)]. When we consider the two-dimensional FD-BE statistics, we derive an analytic expression of the fraction of fermions. When the temperature T→∞, the system is composed by an equal number of bosons and fermions, regardless of the value of η BF . On the contrary, when T=0, η BF becomes important and, according to its value, the system can be completely bosonic or fermionic, or composed both by bosons and fermions

Determination of scaling factors to estimate the radionuclide inventory in waste with low and intermediate-level activity from the IEA-R1 reactor

International Nuclear Information System (INIS)

Taddei, Maria Helena Tirollo

2013-01-01

Regulations regarding transfer and final disposal of radioactive waste require that the inventory of radionuclides for each container enclosing such waste must be estimated and declared. The regulatory limits are established as a function of the annual radiation doses that members of the public could be exposed to from the radioactive waste repository, which mainly depend on the activity concentration of radionuclides, given in Bq/g, found in each waste container. Most of the radionuclides that emit gamma-rays can have their activity concentrations determined straightforwardly by measurements carried out externally to the containers. However, radionuclides that emit exclusively alpha or beta particles, as well as gamma-rays or X-rays with low energy and low absolute emission intensity, or whose activity is very low among the radioactive waste, are generically designated as Difficult to Measure Nuclides (DTMs). The activity concentrations of these DTMs are determined by means of complex radiochemical procedures that involve isolating the chemical species being studied from the interference in the waste matrix. Moreover, samples must be collected from each container in order to perform the analyses inherent to the radiochemical procedures, which exposes operators to high levels of radiation and is very costly because of the large number of radioactive waste containers that need to be characterized at a nuclear facility. An alternative methodology to approach this problem consists in obtaining empirical correlations between some radionuclides that can be measured directly – such as 60 Co and 137 Cs, therefore designated as Key Nuclides (KNs) – and the DTMs. This methodology, denominated Scaling Factor, was applied in the scope of the present work in order to obtain Scaling Factors or Correlation Functions for the most important radioactive wastes with low and intermediate-activity level from the IEA-R1 nuclear research reactor. (author)

Neutron behavior, reactor control, and reactor heat transfer. Volume four

International Nuclear Information System (INIS)

Anon.

1986-01-01

Volume four covers neutron behavior (neutron absorption, how big are nuclei, neutron slowing down, neutron losses, the self-sustaining reactor), reactor control (what is controlled in a reactor, controlling neutron population, is it easy to control a reactor, range of reactor control, what happens when the fuel burns up, controlling a PWR, controlling a BWR, inherent safety of reactors), and reactor heat transfer (heat generation in a nuclear reactor, how is heat removed from a reactor core, heat transfer rate, heat transfer properties of the reactor coolant)

European lead fast reactor (ELSY and LEADER projects)

International Nuclear Information System (INIS)

Alemberti, Alessandro; Carlsson, Johan; Malambu, Edouard; Orden, Alfredo; Cinotti, Luciano; Struwe, Dankward; Agostini, Pietro; Monti, Stefano

2010-01-01

The conceptual design of the European Lead Fast Reactor is being developed starting from September 2006, in the frame of the ELSY project. The ELSY reference design is a 600 MWe pool-type reactor cooled by pure lead. The ELSY project demonstrates the possibility of designing a competitive and safe fast critical reactor using simple engineered technical features, whilst fully complying with the Generation IV goal of sustainability and minor actinide (MA) burning capability. Sustainability was a leading criterion for option selection for core design, focusing on the demonstration of the potential to be self sustaining in plutonium and to burn its own generated MAs. To this end, different core configurations have been studied. Economics was a leading criterion for primary system design and plant layout. The use of a compact and simple primary circuit with the additional objective that all internal components be removable, are among the reactor features intended to assure competitive electric energy generation and long-term investment protection. Low capital cost and construction time are pursued through simplicity and compactness of the reactor building (reduced footprint and height). The reduced plant footprint is one of the benefits coming from the elimination of the Intermediate Cooling System, the low reactor building height is the result of the design approach which foresees the adoption of short-height components and two innovative DHR systems. Among the critical issues, the impact of the large mass of lead has been carefully analyzed; it has been demonstrated that the high density of lead can be mitigated by compact solutions and adoption of seismic isolators. Safety has been one of the major focuses all over the ELSY development. In addition to the inherent safety advantages of lead coolant (high boiling point and no exothermic reactions with air or water) a high safety grade of the overall system has been reached. In fact the overall primary system has been

Using Peephole Optimization on Intermediate Code

NARCIS (Netherlands)

Tanenbaum, A.S.; van Staveren, H.; Stevenson, J.W.

1982-01-01

Many portable compilers generate an intermediate code that is subsequently translated into the target machine's assembly language. In this paper a stack-machine-based intermediate code suitable for algebraic languages (e.g., PASCAL, C, FORTRAN) and most byte-addressed mini- and microcomputers is

Pair production of intermediate vector bosons

International Nuclear Information System (INIS)

Mikaelian, K.O.

1979-01-01

The production of intermediate vector boson pairs W + W - , Z 0 Z 0 , W +- Z 0 and W +- γ in pp and p anti p collisions is discussed. The motivation is to detect the self-interactions among the four intermediate vector bosons

Reactor science and technology: operation and control of reactors

International Nuclear Information System (INIS)

Qiu Junlong

1994-01-01

This article is a collection of short reports on reactor operation and research in China in 1991. The operation of and research activities linked with the Heavy Water Research Reactor, Swimming Pool Reactor and Miniature Neutron Source Reactor are briefly surveyed. A number of papers then follow on the developing strategies in Chinese fast breeder reactor technology including the conceptual design of an experimental fast reactor (FFR), theoretical studies of FFR thermo-hydraulics and a design for an immersed sodium flowmeter. Reactor physics studies cover a range of topics including several related to work on zero power reactors. The section on reactor safety analysis is concerned largely with the assessment of established, and the presentation of new, computer codes for use in PWR safety calculations. Experimental and theoretical studies of fuels and reactor materials for FBRs, PWRs, BWRs and fusion reactors are described. A final miscellaneous section covers Mo-Tc isotope production in the swimming pool reactor, convective heat transfer in tubes and diffusion of tritium through plastic/aluminium composite films and Li 2 SiO 3 . (UK)

Nuclear power reactors

International Nuclear Information System (INIS)

1982-11-01

After an introduction and general explanation of nuclear power the following reactor types are described: magnox thermal reactor; advanced gas-cooled reactor (AGR); pressurised water reactor (PWR); fast reactors (sodium cooled); boiling water reactor (BWR); CANDU thermal reactor; steam generating heavy water reactor (SGHWR); high temperature reactor (HTR); Leningrad (RMBK) type water-cooled graphite moderated reactor. (U.K.)

Liquid metal cooled fast breeder nuclear reactors

International Nuclear Information System (INIS)

Durston, J.G.

1976-01-01

It is stated that in a liquid metal cooled fast breeder reactor wherein the core, intermediate heat exchangers and liquid metal pumps are immersed in a pool of coolant such as Na, the intermediate heat exchangers are suspended from the roof, and ducting is provided in the form of a core tank or shroud interconnected with 'pods' housing the intermediate exchangers for directing coolant from the core over the heat exchanger tubes and thence back to the main pool of liquid metal. Seals are provided between the intermediate heat exchanger shells and the walls of their 'pods' to prevent liquid metal flow by-passing the heat exchanger tube bundles. As the heat exchangers must be withdrawable for servicing, and because linear differential thermal expansion of the heat exchanger and its 'pod' must be accommodated the seals hitherto have been of the sliding kind, generally known as 'piston ring type seals'. These present several disadvantages; for example sealing is not absolute, and the metal to metal seal gives rise to wear and fretting by rubbing and vibration. This could lead to seizure or jamming by the deposition of impurities in the coolant. Another difficulty arises in the need to accommodate lateral thermal expansion of the ducting, including the core tank and 'pods'. Hitherto some expansion has been allowed for by the use of expansible bellow pairs in the interconnections, or alternatively by allowing local deformations of the core tank 'pods'. Such bellows must be very flexible and hence constitute a weak section of the ducting, and local deformations give rise to high stress levels that could lead to premature failure. The arrangement described seeks to overcome these difficulties by use of a gas pocket trapping means to effect a seal against vertical liquid flow between the heat exchanger shell and the wall of the heat exchanger housing. Full details of the arrangement are described. (U.K.)

Research reactors

International Nuclear Information System (INIS)

Merchie, Francois

2015-10-01

This article proposes an overview of research reactors, i.e. nuclear reactors of less than 100 MW. Generally, these reactors are used as neutron generators for basic research in matter sciences and for technological research as a support to power reactors. The author proposes an overview of the general design of research reactors in terms of core size, of number of fissions, of neutron flow, of neutron space distribution. He outlines that this design is a compromise between a compact enough core, a sufficient experiment volume, and high enough power densities without affecting neutron performance or its experimental use. The author evokes the safety framework (same regulations as for power reactors, more constraining measures after Fukushima, international bodies). He presents the main characteristics and operation of the two families which represent almost all research reactors; firstly, heavy water reactors (photos, drawings and figures illustrate different examples); and secondly light water moderated and cooled reactors with a distinction between open core pool reactors like Melusine and Triton, pool reactors with containment, experimental fast breeder reactors (Rapsodie, the Russian BOR 60, the Chinese CEFR). The author describes the main uses of research reactors: basic research, applied and technological research, safety tests, production of radio-isotopes for medicine and industry, analysis of elements present under the form of traces at very low concentrations, non destructive testing, doping of silicon mono-crystalline ingots. The author then discusses the relationship between research reactors and non proliferation, and finally evokes perspectives (decrease of the number of research reactors in the world, the Jules Horowitz project)

Preliminary materials selection issues for the next generation nuclear plant reactor pressure vessel.

Energy Technology Data Exchange (ETDEWEB)

Natesan, K.; Majumdar, S.; Shankar, P. S.; Shah, V. N.; Nuclear Engineering Division

2007-03-21

In the coming decades, the United States and the entire world will need energy supplies to meet the growing demands due to population increase and increase in consumption due to global industrialization. One of the reactor system concepts, the Very High Temperature Reactor (VHTR), with helium as the coolant, has been identified as uniquely suited for producing hydrogen without consumption of fossil fuels or the emission of greenhouse gases [Generation IV 2002]. The U.S. Department of Energy (DOE) has selected this system for the Next Generation Nuclear Plant (NGNP) Project, to demonstrate emissions-free nuclear-assisted electricity and hydrogen production within the next 15 years. The NGNP reference concepts are helium-cooled, graphite-moderated, thermal neutron spectrum reactors with a design goal outlet helium temperature of {approx}1000 C [MacDonald et al. 2004]. The reactor core could be either a prismatic graphite block type core or a pebble bed core. The use of molten salt coolant, especially for the transfer of heat to hydrogen production, is also being considered. The NGNP is expected to produce both electricity and hydrogen. The process heat for hydrogen production will be transferred to the hydrogen plant through an intermediate heat exchanger (IHX). The basic technology for the NGNP has been established in the former high temperature gas reactor (HTGR) and demonstration plants (DRAGON, Peach Bottom, AVR, Fort St. Vrain, and THTR). In addition, the technologies for the NGNP are being advanced in the Gas Turbine-Modular Helium Reactor (GT-MHR) project, and the South African state utility ESKOM-sponsored project to develop the Pebble Bed Modular Reactor (PBMR). Furthermore, the Japanese HTTR and Chinese HTR-10 test reactors are demonstrating the feasibility of some of the planned components and materials. The proposed high operating temperatures in the VHTR place significant constraints on the choice of material selected for the reactor pressure vessel for

Roles of plasma neutron source reactor in development of fusion reactor engineering: Comparison with fission reactor engineering

International Nuclear Information System (INIS)

Hirayama, Shoichi; Kawabe, Takaya

1995-01-01

The history of development of fusion power reactor has come to a turning point, where the main research target is now shifting from the plasma heating and confinement physics toward the burning plasma physics and reactor engineering. Although the development of fusion reactor system is the first time for human beings, engineers have experience of development of fission power reactor. The common feature between them is that both are plants used for the generation of nuclear reactions for the production of energy, nucleon, and radiation on an industrial scale. By studying the history of the development of the fission reactor, one can find the existence of experimental neutron reactors including irradiation facilities for fission reactor materials. These research neutron reactors played very important roles in the development of fission power reactors. When one considers the strategy of development of fusion power reactors from the points of fusion reactor engineering, one finds that the fusion neutron source corresponds to the neutron reactor in fission reactor development. In this paper, the authors discuss the roles of the plasma-based neutron source reactors in the development of fusion reactor engineering, by comparing it with the neutron reactors in the history of fission power development, and make proposals for the strategy of the fusion reactor development. 21 refs., 6 figs

Language in use intermediate : classroom book

CERN Document Server

Doff, Adrian

1995-01-01

ach of the four levels comprises about 80 hours of class work, with additional time for the self-study work. The Teacher's Book contains all the pages from the Classroom Book, with interleaved teaching notes including optional activities to cater for different abilities. There is a video to accompany the Beginner, Pre-intermediate and Intermediate levels. Each video contains eight stimulating and entertaining short programmes, as well as a booklet of photocopiable activities. Free test material is available in booklet and web format for Beginner and Pre-intermediate levels. Visit www.cambridge.org/elt/liu or contact your local Cambridge University Press representative.

Language in use intermediate : teacher's book

CERN Document Server

Doff, Adrian

1998-01-01

Each of the four levels comprises about 80 hours of class work, with additional time for the self-study work. The Teacher's Book contains all the pages from the Classroom Book, with interleaved teaching notes including optional activities to cater for different abilities. There is a video to accompany the Beginner, Pre-intermediate and Intermediate levels. Each video contains eight stimulating and entertaining short programmes, as well as a booklet of photocopiable activities. Free test material is available in booklet and web format for Beginner and Pre-intermediate levels. Visit www.cambridge.org/elt/liu or contact your local Cambridge University Press representative.

Analysis of a Spanish energy scenario with Generation IV nuclear reactors

International Nuclear Information System (INIS)

Ochoa, Raquel; Jimenez, Gonzalo; Perez-Martin, Sara

2013-01-01

Highlights: • Spanish energy scenario for the hypothetical deployment of Gen-IV SFR reactors. • Availability of national resources is assessed, considering SFR’s breeding. • An assessment of the impact of transmuting MA on the final repository. • SERPENT code with own pre- and post-processing tools were employed. • The employed SFR core design is based on the specifications of the CP-ESFR. - Abstract: The advantages of fast-spectrum reactors consist not only of an efficient use of fuel through the breeding of fissile material and the use of natural or depleted uranium, but also of the potential reduction of the amount of actinides such as americium and neptunium contained in the irradiated fuel. The first aspect means a guaranteed future nuclear fuel supply. The second fact is key for high-level radioactive waste management, because these elements are the main responsible for the radioactivity of the irradiated fuel in the long term. The present study aims to analyze the hypothetical deployment of a Gen-IV Sodium Fast Reactor (SFR) fleet in Spain. A nuclear fleet of fast reactors would enable a fuel cycle strategy different than the open cycle, currently adopted by most of the countries with nuclear power. A transition from the current Gen-II to Gen-IV fleet is envisaged through an intermediate deployment of Gen-III reactors. Fuel reprocessing from the Gen-II and Gen-III Light Water Reactors (LWR) has been considered. In the so-called advanced fuel cycle, the reprocessed fuel used to produce energy will breed new fissile fuel and transmute minor actinides at the same time. A reference case scenario has been postulated and further sensitivity studies have been performed to analyze the impact of the different parameters on the required reactor fleet. The potential capability of Spain to supply the required fleet for the reference scenario using national resources has been verified. Finally, some consequences on irradiated final fuel inventory are assessed

Reactor physics aspects of CANDU reactors

International Nuclear Information System (INIS)

Critoph, E.

1980-01-01

These four lectures are being given at the Winter Course on Nuclear Physics at Trieste during 1978 February. They constitute part of the third week's lectures in Part II: Reactor Theory and Power Reactors. A physical description of CANDU reactors is given, followed by an overview of CANDU characteristics and some of the design options. Basic lattice physics is discussed in terms of zero energy lattice experiments, irradiation effects and analytical methods. Start-up and commissioning experiments in CANDU reactors are reviewed, and some of the more interesting aspects of operation discussed - fuel management, flux mapping and control of the power distribution. Finally, some of the characteristics of advanced fuel cycles that have been proposed for CANDU reactors are summarized. (author)

Interpretation and code generation based on intermediate languages

DEFF Research Database (Denmark)

Kornerup, Peter; Kristensen, Bent Bruun; Madsen, Ole Lehrmann

1980-01-01

The possibility of supporting high level languages through intermediate languages to be used for direct interpretation and as intermediate forms in compilers is investigated. An accomplished project in the construction of an interpreter and a code generator using one common intermediate form...

Japan: The Experimental Fast Reactor JOYO. Profile 12

International Nuclear Information System (INIS)

2017-01-01

The experimental fast reactor JOYO of the Japan Atomic Energy Agency (JAEA) is the first sodium-cooled fast reactor (SFR) in Japan. JOYO attained its initial criticality as a breeder core (MK-I core) in 1977. During the MK-I operation, which consisted of two 50 MWt and six 75 MWt duty cycles, the basic characteristics of plutonium (Pu) and uranium (U) mixed oxide (MOX) fuel core and sodium cooling system were investigated and the breeding performance was verified. In 1983, the reactor increased its thermal output up to 100 MWt in order to start the irradiation tests of fuels and materials to be used mainly for other SFRs. Thirty-five duty cycle operations and many irradiation tests were successfully carried out using the MK-II core by 2000. The core was then modified to the MK-III core in 2003. In order to obtain higher fast neutron flux, the core was modified from one region core to two region core with different Pu fissile contents. Accordingly, the reactor power increased up to 140 MWt together with a renewal of intermediate heat exchangers (IHXs) and dump heat exchangers (DHXs). The rated power operation of the MK-III core started in 2004. The MK-III core has been used for the irradiation tests of fuels and materials for future SFRs and other R&D fields like innovative nuclear energy systems and technologies as well. This powerful neutron irradiation flux has an advantage especially for high burn-up fuel irradiation and material irradiation with high neutron dose. This paper shows the outline of the irradiation irradiation irradiation irradiation irradiation capabilities and capacities to develop capabilities and capacities to develop capabilities and capacities to develop capabilities and capacities to develop innovative nuclear energy systems and technologies.

Research reactors

International Nuclear Information System (INIS)

Kowarski, L.

1955-01-01

It brings together the techniques data which are involved in the discussion about the utility for a research institute to acquire an atomic reactor for research purposes. This type of decision are often taken by non-specialist people who can need a brief presentation of a research reactor and its possibilities in term of research before asking advises to experts. In a first part, it draws up a list of the different research programs which can be studied by getting a research reactor. First of all is the reactor behaviour and kinetics studies (reproducibility factor, exploration of neutron density, effect of reactor structure, effect of material irradiation...). Physical studies includes study of the behaviour of the control system, studies of neutron resonance phenomena and study of the fission process for example. Chemical studies involves the study of manipulation and control of hot material, characterisation of nuclear species produced in the reactor and chemical effects of irradiation on chemical properties and reactions. Biology and medicine research involves studies of irradiation on man and animals, genetics research, food or medical tools sterilization and neutron beams effect on tumour for example. A large number of other subjects can be studied in a reactor research as reactor construction material research, fabrication of radioactive sources for radiographic techniques or applied research as in agriculture or electronic. The second part discussed the technological considerations when choosing the reactor type. The technological factors, which are considered for its choice, are the power of the reactor, the nature of the fuel which is used, the type of moderator (water, heavy water, graphite or BeO) and the reflector, the type of coolants, the protection shield and the control systems. In the third part, it described the characteristics (place of installation, type of combustible and comments) and performance (power, neutron flux ) of already existing

Strategic need for a multi-purpose thermal hydraulic loop for support of advanced reactor technologies

Energy Technology Data Exchange (ETDEWEB)

O' Brien, James E. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Sabharwall, Piyush [Idaho National Lab. (INL), Idaho Falls, ID (United States); Yoon, Su -Jong [Idaho National Lab. (INL), Idaho Falls, ID (United States); Housley, Gregory K. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2014-09-01

This report presents a conceptual design for a new high-temperature multi fluid, multi loop test facility for the INL to support thermal hydraulic, materials, and thermal energy storage research for nuclear and nuclear-hybrid applications. In its initial configuration, the facility will include a high-temperature helium loop, a liquid salt loop, and a hot water/steam loop. The three loops will be thermally coupled through an intermediate heat exchanger (IHX) and a secondary heat exchanger (SHX). Research topics to be addressed with this facility include the characterization and performance evaluation of candidate compact heat exchangers such as printed circuit heat exchangers (PCHEs) at prototypical operating conditions, flow and heat transfer issues related to core thermal hydraulics in advanced helium-cooled and salt-cooled reactors, and evaluation of corrosion behavior of new cladding materials and accident-tolerant fuels for LWRs at prototypical conditions. Based on its relevance to advanced reactor systems, the new facility has been named the Advanced Reactor Technology Integral System Test (ARTIST) facility. Research performed in this facility will advance the state of the art and technology readiness level of high temperature intermediate heat exchangers (IHXs) for nuclear applications while establishing the INL as a center of excellence for the development and certification of this technology. The thermal energy storage capability will support research and demonstration activities related to process heat delivery for a variety of hybrid energy systems and grid stabilization strategies. Experimental results obtained from this research will assist in development of reliable predictive models for thermal hydraulic design and safety codes over the range of expected advanced reactor operating conditions. Proposed/existing IHX heat transfer and friction correlations and criteria will be assessed with information on materials compatibility and instrumentation

Online failed fuel identification using delayed neutron detector signals in pool type reactors

International Nuclear Information System (INIS)

Upadhyay, Chandra Kant; Sivaramakrishna, M.; Nagaraj, C.P.; Madhusoodanan, K.

2011-01-01

In todays world, nuclear reactors are at the forefront of modern day innovation and reactor designs are increasingly incorporating cutting edge technology. It is of utmost importance to detect failure or defects in any part of a nuclear reactor for healthy operation of reactor as well as the safety aspects of the environment. Despite careful fabrication and manufacturing of fuel pins, there is a chance of clad failure. After fuel pin clad rupture takes place, it allows fission products to enter in to sodium pool. There are some potential consequences due to this such as Total Instantaneous Blockage (TIB) of coolant and primary component contamination. At present, the failed fuel detection techniques such as cover gas monitoring (alarming the operator), delayed neutron detection (DND-automatic trip) and standalone failed fuel localization module (FFLM) are exercised in various reactors. The first technique is a quantitative measurement of increase in the cover gas activity background whereas DND system causes automatic trip on detecting certain level of activity during clad wet rupture. FFLM is subsequently used to identify the failed fuel subassembly. The later although accurate, but mainly suffers from downtime and reduction in power during identification process. The proposed scheme, reported in this paper, reduces the operation of FFLM by predicting the faulty sector and therefore reducing reactor down time and thermal shocks. The neutron evolution pattern gets modulated because fission products are the delay neutron precursors. When they travel along with coolant to Intermediate heat Exchangers, experienced three effects i.e. delay; decay and dilution which make the neutron pulse frequency vary depending on the location of failed fuel sub assembly. This paper discusses the method that is followed to study the frequency domain properties, so that it is possible to detect exact fuel subassembly failure online, before the reactor automatically trips. (author)

Fundamental conceptual design of the experimental multi-purpose high-temperature gas-cooled reactor

International Nuclear Information System (INIS)

Shimokawa, Junichi; Yasuno, Takehiko; Yasukawa, Shigeru; Mitake, Susumu; Miyamoto, Yoshiaki

1975-06-01

The fundamental conceptual design of the experimental multi-purpose very high-temperature gas-cooled reactor (experimental VHTR of thermal output 50 MW with reactor outlet-gas temperature 1,000 0 C) has been carried out to provide the operation modes of the system consisting of the reactor and the heat-utilization system, including characteristics and performance of the components and safety of the plant system. For the heat-utilization system of the plant, heat distribution, temperature condition, cooling system constitution, and the containment facility are specified. For the operation of plant, testing capability of the reactor and controlability of the system are taken into consideration. Detail design is made of the fuel element, reactor core, reactivity control and pressure vessel, and also the heat exchanger, steam reformer, steam generator, helium circulator, helium-gas turbine, and helium-gas purification, fuel handling, and engineered safety systems. Emphasis is placed on providing the increase of the reactor outlet-gas temperature. Fuel element design is directed to the prismatic graphite blocks of hexagonal cross-section accommodating the hollow or tubular fuel pins sheathed in graphite sleeve. The reactor core is composed of 73 fuel columns in 7 stages, concerning the reference design MK-II. Orificing is made in the upper portion of core; one orifice for every 7 fuel columns. Average core power density is 2.5 watts/cm 3 . Fuel temperature is kept below 1,300 0 C in rated power. The main components, i.e. pressure vessel, reformer, gas turbine and intermediate heat exchanger are designed in detail; the IHX is of a double-shell and helically-wound tube coils, the reformer is of a byonet tube type, and the turbine-compressor unit is of an axial flow type (turbine in 6 stages and compressor in 16 stages). (auth.)

SOLUCIÓN ANALÍTICA PARA OBTENER EL VOLUMEN ÓPTIMO DE UNA SERIE DE REACTORES DE AGITACIÓN CONTINUA DONDE SE EFECTÚA UNA REACCIÓN DE PRIMER ORDEN

Directory of Open Access Journals (Sweden)

Ignacio Elizalde

2013-01-01

Full Text Available An analytical procedure for determining the optimum size of CSTR in series operating under isothermal and isobaric conditions sustaining first order reaction at constant density has been developed. The procedure requires the concentration of reactant at the entrance of the first reactor and at the outlet of the last reactor; it is also required the continuity of reaction rate as function of conversion, due to the later changes from one reactor to another. The optimization method involves the calculation of intermediate concentrations instead of their estimation, as it is done by graphical solution reported previously. Also, the procedure reported in this contribution is valid for any reactor number. Under these circumstances the method predicts that all reactors must have the same size in order to minimize the total volume of the system.

AERE contracts with DOE on the treatment and disposal of Intermediate Level Wastes

International Nuclear Information System (INIS)

Partridge, B.A.

1985-07-01

Individual summaries are provided for each contract report, under the titles: comparative evaluation of α and βγ irradiated medium level waste forms; modelling and characterisation of intermediate level waste forms based on polymers; optimisation of processing parameters for polymer and bitumen modified cements; α damage in non-reference matrix materials; leaching mechanisms and modelling; inorganic ion exchange treatment of medium active effluents; electrical processes for the treatment of medium active liquid waste; fast reactor fuel element cladding; dissolver residues; effects of radiation on the properties of cemented MTR waste forms; equilibrium leach testing of cemented MTR waste forms; radiolytic oxidation of radionuclides; immobilisation of liquid organic wastes; quality control, non-conformances and corrective action; application of gel processes in the treatment of actinide-containing liquid wastes; the role of colloids in the release of radionuclides from nuclear waste. (author)

Comparison between TRU burning reactors and commercial fast reactor

International Nuclear Information System (INIS)

Fujimura, Koji; Sanda, Toshio; Ogawa, Takashi

2001-03-01

Research and development for stabilizing or shortening the radioactive wastes including in spent nuclear fuel are widely conducted in view point of reducing the environmental impact. Especially it is effective way to irradiate and transmute long-lived TRU by fast reactors. Two types of loading way were previously proposed. The former is loading relatively small amount of TRU in all commercial fast reactors and the latter is loading large amount of TRU in a few TRU burning reactors. This study has been intended to contribute to the feasibility studies on commercialized fast reactor cycle system. The transmutation and nuclear characteristics of TRU burning reactors were evaluated and compared with those of conventional transmutation system using commercial type fast reactor based upon the investigation of technical information about TRU burning reactors. Major results are summarized as follows. (1) Investigation of technical information about TRU burning reactors. Based on published reports and papers, technical information about TRU burning reactor concepts transmutation system using convectional commercial type fast reactors were investigated. Transmutation and nuclear characteristics or R and D issue were investigated based on these results. Homogeneously loading of about 5 wt% MAs on core fuels in the conventional commercial type fast reactor may not cause significant impact on the nuclear core characteristics. Transmutation of MAs being produced in about five fast reactors generating the same output is feasible. The helium cooled MA burning fast reactor core concept propose by JAERI attains criticality using particle type nitride fuels which contain more than 60 wt% MA. This reactor could transmute MAs being produced in more than ten 1000 MWe-LWRs. Ultra-long life core concepts attaining more than 30 years operation without refueling by utilizing MA's nuclear characteristics as burnable absorber and fertile nuclides were proposed. Those were pointed out that

Nuclear reactor types

International Nuclear Information System (INIS)

Jones, P.M.S.

1987-01-01

The characteristics of different reactor types designed to exploit controlled fission reactions are explained. Reactors vary from low power research devices to high power devices especially designed to produce heat, either for direct use or to produce steam to drive turbines to generate electricity or propel ships. A general outline of basic reactors (thermal and fast) is given and then the different designs considered. The first are gas cooled, including the Magnox reactors (a list of UK Magnox stations and reactor performance is given), advanced gas cooled reactors (a list of UK AGRs is given) and the high temperature reactor. Light water cooled reactors (pressurized water [PWR] and boiling water [BWR] reactors) are considered next. Heavy water reactors are explained and listed. The pressurized heavy water reactors (including CANDU type reactors), boiling light water, steam generating heavy water reactors and gas cooled heavy water reactors all come into this category. Fast reactors (liquid metal fast breeder reactors and gas cooled fast reactors) and then water-cooled graphite-moderated reactors (RBMK) (the type at Chernobyl-4) are discussed. (U.K.)

Radioisotope tracer study in a pilot-scale trickle bed reactor

Energy Technology Data Exchange (ETDEWEB)

Pant, H.J.; Pendharkar, A.S. [Bhabha Atomic Research Centre, Isotope Applications Div, Bombay (India); Prasad, J.S.; Maiti, R.N.; Chawla, R.; Lahri, R.N.; Ram Babu, D. [Engineers India Limited, Gurgaon, Haryana (India); Berne, Ph. [CEA Grenoble, DTEN/SAT, 38 (France)

2001-07-01

Trickle bed reactor (TBR) is a reactor in which a liquid and a gas phase flow concurrently downwards through a fixed bed of catalyst particles while the reaction takes place. The trickle bed reactors find a number of applications in petroleum refining, chemical. petro-chemical and bio-chemical processes due to their suitability for hydro-processing operations.The knowledge of radial distribution, mean residence time (MRT). liquid holdup and degree of axial mixing is a basic requirement to evaluate the reactor performance of a TBR, its optimal size, the physical and chemical interactions and the pumping requirements. Measurement and analysis of residence time distribution (RTD) of liquid phase facilitate the determination of these parameters. This paper describes the measurement of RTD of liquid phase in a pilot-scale trickle bed reactor operating at high pressure. Kerosene and nitrogen were used as liquid and gas phase, respectively. Porous alumina catalyst particles were used as packing. Bromine-82 as para-di-bromo benzene was used as a tracer to measure RTD of organic liquid phase. The RTD of kerosene was measured as a function of liquid and gas flow rates and pressure. The axial dispersion model with exchange between active and dead zones was proposed and used to simulate the measured RTD data and model parameters i.e mean residence time ({tau}), Peclet Number (Pe), dynamic fraction of the liquid ({phi}) and the number of transfer units (N) were obtained. The results of the simulations indicated intermediate degree of axial mixing of liquid phase. The RTD predicted by the axial dispersion model with exchange between active and dead zones fits well to the experimentally measured RTD data and thus is a suitable model to describe the dynamics of the liquid phase in TBRs filled with porous catalyst particles. (authors)

Reactor containment and reactor safety in the United States

International Nuclear Information System (INIS)

Kouts, H.

1986-01-01

The reactor safety systems of two reactors are studied aiming at the reactor containment integrity. The first is a BWR type reactor and is called Peachbottom 2, and the second is a PWR type reactor, and is called surry. (E.G.) [pt

Core disruptive accident analysis in prototype fast breeder reactor

International Nuclear Information System (INIS)

Chellapandi, P.; Velusamy, K.; Kannan, S.E.; Singh, Om Pal; Chetal, S.C.; Bhoje, S.B.

2002-01-01

Liquid metal cooled fast breeder reactors, in particular, pool type have many inherent and engineered safety features and hence a core disruptive accident (CDA) involving melt down of the whole core is a very low probable event ( -6 /ry). The important mechanical consequences such as straining of the main vessel including top shield, structural integrity of safety grade decay heat exchangers (DHX) and intermediate heat exchangers (IHX) sodium release to reactor containment building (RCB) through the penetrations in the top shield, sodium fire and consequent temperature and pressure rise in RCB are theoretically analysed using computer codes. Through the analyses with these codes, it is demonstrated that an energetic CDA capability to the maximum 100 MJ mechanical energy in PFBR can be well contained in the primary containment. The sodium release to RCB is 350 kg and pressure rise in RCB is ∼10 kPa. In order to raise the confidence on the theoretical predictions, very systematic experimental program has been carried out. Totally 67 tests were conducted. This experimental study indicated that the primary containment is integral. The main vessel can withstand the energy release of ∼1200 MJ. The structural integrity of IHX and DHX is assured up to 200 MJ. The transient force transmitted to reactor vault is negligible. The average water leak measured under simulated tests for 122 MJ work potential is about 1.8 kg and the maximum leak is 2.41 kg. Extrapolation of the measured maximum leak based on simulation principles yields ∼ 233 kg of sodium leak in the reactor. Based on the above-mentioned theoretical and experimental investigations, the design pressure of 20 kPa is used for PFBR

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2002-01-01

SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2001-04-01

The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised.

Reactor Physics

International Nuclear Information System (INIS)

Ait Abderrahim, A.

2001-01-01

The Reactor Physics and MYRRHA Department of SCK-CEN offers expertise in various areas of reactor physics, in particular in neutronics calculations, reactor dosimetry, reactor operation, reactor safety and control and non-destructive analysis of reactor fuel. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2000 are summarised

Process Intensification of Alkynol Semihydrogenation in a Tube Reactor Coated with a Pd/ZnO Catalyst

Directory of Open Access Journals (Sweden)

Nikolay Cherkasov

2017-11-01

Full Text Available Semihydrogenation of 2-methyl-3-butyn-2-ol (MBY was studied in a 5 m tube reactor wall-coated with a 5 wt% Pd/ZnO catalyst. The system allowed for the excellent selectivity towards the intermediate alkene of 97.8 ± 0.2% at an ambient H2 pressure and a MBY conversion below 90%. The maximum alkene yield reached 94.6% under solvent-free conditions and 96.0% in a 30 vol % MBY aqueous solution. The reactor stability was studied for 80 h on stream with a deactivation rate of only 0.07% per hour. Such a low deactivation rate provides a continuous operation of one month with only a two-fold decrease in catalyst activity and a metal leaching below 1 parts per billion (ppb. The excellent turn-over numbers (TON of above 105 illustrates a very efficient utilisation of the noble metal inside catalyst-coated tube reactors. When compared to batch operation at 70 °C, the reaction rate in flow reactor can be increased by eight times at a higher reaction temperature, keeping the same product decomposition of about 1% in both cases.

Reactor

International Nuclear Information System (INIS)

Toyama, Masahiro; Kasai, Shigeo.

1978-01-01

Purpose: To provide a lmfbr type reactor wherein effusion of coolants through a loop contact portion is reduced even when fuel assemblies float up, and misloading of reactor core constituting elements is prevented thereby improving the reactor safety. Constitution: The reactor core constituents are secured in the reactor by utilizing the differential pressure between the high-pressure cooling chamber and low-pressure cooling chamber. A resistance port is formed at the upper part of a connecting pipe, and which is connect the low-pressure cooling chamber and the lower surface of the reactor core constituent. This resistance part is formed such that the internal sectional area of the connecting pipe is made larger stepwise toward the upper part, and the cylinder is formed larger so that it profiles the inner surface of the connecting pipe. (Aizawa, K.)

Search for intermediate vector bosons

International Nuclear Information System (INIS)

Cline, D.B.; Rubbia, C.; van der Meer, S.

1982-01-01

Over the past 15 years a new class of unified theories has been developed to describe the forces acting between elementary particles. The most successful of the new theories establishes a link between electromagnetism and the weak force. A crucial prediction of this unified electroweak theory is the existence of three massive particles called intermediate vector bosons. If these intermediate vector bosons exist and if they have properties attributed to them by electroweak theory, they should soon be detected, as the world's first particle accelerator with enough energy to create such particles has recently been completed at the European Organization for Nuclear Research (CERN) in Geneva. The accelerator has been converted to a colliding beam machine in which protons and antiprotons collide head on. According to electroweak theory, intermediate vector bosons can be created in proton-antiproton collisions. (SC)

The different facilities of the reactor Phenix for radio isotope production and fission product burner

International Nuclear Information System (INIS)

Coulon, P.; Clerc, R.; Tommasi, J.

1993-01-01

During the last few years different tests have been made to optimize the blanket of the reactor. Year after year the breeding ratio has lost a part of interest regarding the production and availability of plutonium in the world. A characteristic of a fast reactor is to have important neutron leaks from the core. The spectrum of those neutrons is intermediate, the idea was to find a moderator compatible with sodium and stable in temperature. After different tests we kept as a moderator the calcium hydride and as a samply support, a cluster which is separated from the carrier. At the end we present the model used for thermalized calculations. The scheme is then applied to a heavy nuclide transmutation example (Np237 Pu238) and to fission product transmutation (Tc99). (authors). 9 figs

Korean working towards low and intermediate level waste volume reduction

International Nuclear Information System (INIS)

Myung-Jae Song; Jong-Kil Park

2001-01-01

The safe management of radioactive waste is a national task required for sustainable generation of nuclear power and for energy self-reliance. This paper describes the results, efforts, and prospects for the safe management of radioactive wastes having been performed by the Nuclear Environment Technology Institute (NETEC) of the Korean Electric Power Corporation (KEPCO). Firstly, KEPCO's efforts and results for waste volume reduction are summarized to show how the number of waste drums generated per reactor-year could be reduced by about 60% during the last 10 years. Secondly, a new treatment technology, a technology for low and intermediate level waste (LILW) vitrification, was introduced to prospect how the technology reduces the waste volume and increases the inherent safety for LILW disposal. It is expected that the vitrification technology will contribute not only to reduce LILW volume to around 1/14 ∼ 1/32 but also to change the 'Not In My Back Yard' (NIMBY) syndrome to the 'Please In My Front Yard' (PIMFY) attitude of local communities/residents for LILW disposal. (author)

1980 Annual status report reactor safety

International Nuclear Information System (INIS)

1981-01-01

The JRC reactor safety programme involves theoretical and experimental activities to analyse accidents and their consequences for LWRs and LMFBRs. The first project deals with the improvement and the application of methodologies for risk and reliability assessment. This activity involves the identification and modelling of accident sequences and events and the analysis of fault trees. In this project, the implementation of a centralized data bank system (European Reliability Data System) is foreseen, which should provide the information needed for risk assessment studies. In project 2 a major effort on LWRs is centered on the study of the loss-of-coolant accident following large, intermediate or small breaks of the primary circuit. These accidents are simulated out of pile in the LOBI facility. In project 3 a contribution is made to solve material problems and to provide data and calculation methods for end of life predictions of reactor components. It involves a contribution to the programme for the inspection of steel components (PISC) as well as the study of fracture and creep fatigue properties of stainless steel. In the project 4 and 5 a deterministic approach is adopted to solve some problems of large hypothetical accidents in an LMFBR. The calculation tools developed concern sodium thermohydraulics in fuel element bundles, fuel coolant interaction, whole core accident analysis, containment loading and response and post accident heat removal

39 CFR 3001.39 - Intermediate decisions.

Science.gov (United States)

2010-07-01

... 39 Postal Service 1 2010-07-01 2010-07-01 false Intermediate decisions. 3001.39 Section 3001.39 Postal Service POSTAL REGULATORY COMMISSION PERSONNEL RULES OF PRACTICE AND PROCEDURE Rules of General Applicability § 3001.39 Intermediate decisions. (a) Initial decision by presiding officer. In any proceedings in...

42 CFR 54.12 - Treatment of intermediate organizations.

Science.gov (United States)

2010-10-01

... 42 Public Health 1 2010-10-01 2010-10-01 false Treatment of intermediate organizations. 54.12... intermediate organizations. If a nongovernmental organization (referred to here as an “intermediate organization”), acting under a contract or other agreement with the Federal Government or a State or local...

Formation, structure, and stability of MHD intermediate shocks

International Nuclear Information System (INIS)

Wu, C.C.

1990-01-01

Contrary to the usual belief that MHD intermediate shocks are extraneous, the author has recently shown by numerical solutions of dissipative MHD equations that intermediate shocks are admissible and can be formed through nonlinear wave steepening from continuous waves. In this paper, the formation, structure and stability of intermediate shocks in dissipative MHD are considered in detail. The differences between the conventional theory and his are pointed out and clarified. He shows that all four types of intermediate shocks can be formed from smooth waves. He also shows that there are free parameters in the structure of the intermediate shocks, and that these parameters are related to the shock stability. In addition, he shows that a rotational discontinuity can not exist with finite width, indicate how this is related to the existence of time-dependent intermediate shocks, and show why the conventional theory is not a good approximation to dissipative MHD solutions whenever there is rotation in magnetic field

Reactor Physics

Energy Technology Data Exchange (ETDEWEB)

Ait Abderrahim, A

2002-04-01

SCK-CEN's Reactor Physics and MYRRHA Department offers expertise in various areas of reactor physics, in particular in neutron and gamma calculations, reactor dosimetry, reactor operation and control, reactor code benchmarking and reactor safety calculations. This expertise is applied in the Department's own research projects in the VENUS critical facility, in the BR1 reactor and in the MYRRHA project (this project aims at designing a prototype Accelerator Driven System). Available expertise is also used in programmes external to the Department such as the reactor pressure steel vessel programme, the BR2 materials testing reactor dosimetry, and the preparation and interpretation of irradiation experiments by means of neutron and gamma calculations. The activities of the Fuzzy Logic and Intelligent Technologies in Nuclear Science programme cover several domains outside the department. Progress and achievements in these topical areas in 2001 are summarised.

Westinghouse Small Modular Reactor nuclear steam supply system design

Energy Technology Data Exchange (ETDEWEB)

Memmott, M. J.; Harkness, A. W.; Van Wyk, J. [Westinghouse Electric Company LLC, 600 Cranberry Woods Drive, Cranberry Twp. PA 16066 (United States)

2012-07-01

The Westinghouse Small Modular Reactor (SMR) is an 800 MWt (>225 MWe) integral pressurized water reactor (iPWR), in which all of the components typically associated with the nuclear steam supply system (NSSS) of a nuclear power plant are incorporated within a single reactor pressure vessel. This paper is the first in a series of four papers which describe the design and functionality of the Westinghouse SMR. Also described in this series are the key drivers influencing the design of the Westinghouse SMR and the unique passive safety features of the Westinghouse SMR. Several critical motivators contributed to the development and integration of the Westinghouse SMR design. These design driving motivators dictated the final configuration of the Westinghouse SMR to varying degrees, depending on the specific features under consideration. These design drivers include safety, economics, AP1000{sup R} reactor expertise and experience, research and development requirements, functionality of systems and components, size of the systems and vessels, simplicity of design, and licensing requirements. The Westinghouse SMR NSSS consists of an integral reactor vessel within a compact containment vessel. The core is located in the bottom of the reactor vessel and is composed of 89 modified Westinghouse 17x17 Robust Fuel Assemblies (RFA). These modified fuel assemblies have an active core length of only 2.4 m (8 ft) long, and the entirety of the core is encompassed by a radial reflector. The Westinghouse SMR core operates on a 24 month fuel cycle. The reactor vessel is approximately 24.4 m (80 ft) long and 3.7 m (12 ft) in diameter in order to facilitate standard rail shipping to the site. The reactor vessel houses hot and cold leg channels to facilitate coolant flow, control rod drive mechanisms (CRDM), instrumentation and cabling, an intermediate flange to separate flow and instrumentation and facilitate simpler refueling, a pressurizer, a straight tube, recirculating steam

Research reactor standards and their impact on the TRIGA reactor community

International Nuclear Information System (INIS)

Richards, W.J.

1980-01-01

The American Nuclear Society has established a standards committee devoted to writing standards for research reactors. This committee was formed in 1971 and has since that time written over 15 standards that cover all aspects of research reactor operation. The committee has representation from virtually every group concerned with research reactors and their operation. This organization includes University reactors, National laboratory reactors, Nuclear Regulatory commission, Department of Energy and private nuclear companies and insurers. Since its beginning the committee has developed standards in the following areas: Standard for the development of technical specifications for research reactors; Quality control for plate-type uranium-aluminium fuel elements; Records and reports for research reactors; Selection and training of personnel for research reactors; Review of experiments for research reactors; Research reactor site evaluation; Quality assurance program requirements for research reactors; Decommissioning of research reactors; Radiological control at research reactor facilities; Design objectives for and monitoring of systems controlling research reactor effluents; Physical security for research reactor facilities; Criteria for the reactor safety systems of research reactors; Emergency planning for research reactors; Fire protection program requirements for research reactors; Standard for administrative controls for research reactors. Besides writing the above standards, the committee is very active in using communications with the nuclear regulatory commission on proposed rules or positions which will affect the research reactor community

Promising design options for the encapsulated nuclear heat source reactor

International Nuclear Information System (INIS)

Conway, L.; Carelli, M.D.; Dzodzo, M.; Hossain, Q.; Brown, N.W.; Wade, D.C.; Sienick, J.J.; Greenspan, E.; Kastenberg, W.E.; Saphier, D.

2001-01-01

Promising design options for the Encapsulated Nuclear Heat Source (ENHS) liquid-metal cooled fast reactor were identified during the first year of the DOE NERI program sponsored feasibility study. Many opportunities for incorporation of innovations in design and fabrication were identified. Three of the innovations are hereby described: a novel IHX (intermediate heat exchanger) made of a relatively small number of rectangular channels, an ENHS module design featuring 100% natural circulation, and a novel conceptual design of core support and fuelling. As a result of the first year study the ENHS concept appears more practical and more promising than perceived at the outset of this study. (authors)

Promising design options for the encapsulated nuclear heat source reactor

Energy Technology Data Exchange (ETDEWEB)

Conway, L.; Carelli, M.D.; Dzodzo, M. [Westinghouse Science and Technology, Pittsburgh, PA (United States); Hossain, Q.; Brown, N.W. [Lawrence Livermore National Lab., CA (United States); Wade, D.C.; Sienick, J.J. [Argonne National Lab., IL (United States); Greenspan, E.; Kastenberg, W.E.; Saphier, D. [University of California Dept of Nuclear Engineering, Berkeley, CA (United States)

2001-07-01

Promising design options for the Encapsulated Nuclear Heat Source (ENHS) liquid-metal cooled fast reactor were identified during the first year of the DOE NERI program sponsored feasibility study. Many opportunities for incorporation of innovations in design and fabrication were identified. Three of the innovations are hereby described: a novel IHX (intermediate heat exchanger) made of a relatively small number of rectangular channels, an ENHS module design featuring 100% natural circulation, and a novel conceptual design of core support and fuelling. As a result of the first year study the ENHS concept appears more practical and more promising than perceived at the outset of this study. (authors)

Reactor operation

CERN Document Server

Shaw, J

2013-01-01

Reactor Operation covers the theoretical aspects and design information of nuclear reactors. This book is composed of nine chapters that also consider their control, calibration, and experimentation.The opening chapters present the general problems of reactor operation and the principles of reactor control and operation. The succeeding chapters deal with the instrumentation, start-up, pre-commissioning, and physical experiments of nuclear reactors. The remaining chapters are devoted to the control rod calibrations and temperature coefficient measurements in the reactor. These chapters also exp

Improved nuclear reactor construction with bottom supported reactor vessel

International Nuclear Information System (INIS)

Sharbaugh, J.E.

1987-01-01

An improved liquid metal nuclear reactor construction has a reactor core and a generally cylindrical reactor vessel for holding liquid metal coolant and housing the core within the pool. A generally cylindrical concrete containment structure surrounds the reactor vessel and a central support pedestal is anchored to the containment structure base mat and supports the bottom wall of the reactor vessel and the reactor core. The periphery of the reactor vessel bore is supported by an annular structure which allows thermal expansion but not seismic motion of the vessel, and a bed of thermally insulating material uniformly supports the vessel base whilst allowing expansion thereof. A guard ring prevents lateral seismic motion of the upper end of the reactor vessel. The periphery of the core is supported by an annular structure supported by the vessel base and keyed to the vessel wall so as to be able to expand but not undergo seismic motion. A deck is supported on the containment structure above the reactor vessel open top by annular bellows, the deck carrying the reactor control rods such that heating of the reactor vessel results in upward expansion against the control rods. (author)

Reactor as furnace and reactor as lamp

International Nuclear Information System (INIS)

Goldanskii, V.I.

1992-01-01

There are presented general characteristics of the following ways of transforming of nuclear energy released in reactors into chemical : ordinary way (i.e. trough the heat, mechanical energy and electricity); chemonuclear synthesis ; use of high-temperature fuel elements (reactor as furnace); use of the mixed nγ-radiation of reactors; use of the radiation loops; radiation - photochemical synthesis (reactor as lamp). Advantage and disadvantages of all above variants are compared. The yield of the primary product of fixation of nitrogen (nitric oxide NO) in reactor with the high-temperature (above ca. 1900degC) fuel elements (reactor-furnace) can exceed W ∼ 200 kg per gram of burned uranium. For the latter variant (reactor-lamp) the yield of chemical products can reach W ∼ 60 kg. per gram of uranium. Such values of W are close to or even strongly exceed the yields of chemical products for other abovementioned variants and - what is particularly important - are not connected to the necessity of archscrupulous removal of radioactive contamination of products. (author)

Development of Reactor Console Simulator for PUSPATI TRIGA Reactor

International Nuclear Information System (INIS)

Mohd Idris Taib; Izhar Abu Hussin; Mohd Khairulezwan Abdul Manan; Nufarhana Ayuni Joha; Mohd Sabri Minhat

2012-01-01

The Reactor Console Simulator will be an interactive tool for operator training and teaching of PUSPATI TRIGA Reactor. Behaviour and characteristic for reactor console and reactor itself can be evaluated and understand. This Simulator will be used as complement for actual present reactor console. Implementation of man-machine interface is using computer screens, keyboard and mouse. Multiple screens are used to match the physical of present reactor console. LabVIEW software are using for user interface and mathematical calculation. Polynomial equation based on control rods calibration data as well as operation parameters record was used to calculate the estimated reactor console parameters. (author)

Design and thermal dynamic analyses on the intermediate heat exchanger for HTGR

International Nuclear Information System (INIS)

Mori, M.; Mizuno, M.; Ito, M.; Urabe, S.

1986-01-01

The intermediate heat exchanger (IHX), one of the most important components in the high temperature gas cooled reactor (HTGR), is a high performance helium/helium (He/He) heat exchanger operated at a very high temperature above 900 0 C to transmit the nuclear heat from the reactor core to the nuclear heat utilization systems such as the chemical plant. Having to meet, in addition, the requirement of the pressure boundary as the Class-1 it demands the accurate estimation of thermal performance and analytical prediction of thermal behaviors to secure its integrity throughout the service life. In the present works, the newly-developed analytical codes carry out designing thermal performance and analyzing dynamic thermal behaviors of the IHX. These codes have been developed on a great deal of data and studies related to the research and development on the 1.5 MWt- and the 25 MWt-IHXs. This paper shows the design on the IHX, the results of the dynamic analyses on the 1.5 MWt-IHX with the comparison to the experimental data and the analytical predictions of the dynamic thermal behaviors on the 25 MWt-IHX. The results calculated are in fairly good agreement with the experimental data on the 1.5 MWt-IHX, the fact that has verified the analytical codes to be reasonable and much useful for the thermal design of the IHX. These presented results and data are available for the design of the IHX of HTGR

Degradation pathway of malachite green in a novel dual-tank photoelectrochemical catalytic reactor.

Science.gov (United States)

Diao, Zenghui; Li, Mingyu; Zeng, Fanyin; Song, Lin; Qiu, Rongliang

2013-09-15

A novel dual-tank photoelectrochemical catalytic reactor was designed to investigate the degradation pathway of malachite green. A thermally formed TiO₂/Ti thin film electrode was used as photoanode, graphite was used as cathode, and a saturated calomel electrode was employed as the reference electrode in the reactor. In the reactor, the anode and cathode tanks were connected by a cation exchange membrane. Results showed that the decolorization ratio of malachite green in the anode and cathode was 98.5 and 96.5% after 120 min, respectively. Malachite green in the two anode and cathode tanks was oxidized, achieving the bipolar double effect. Malachite green in both the anode and cathode tanks exhibited similar catalytic degradation pathways. The double bond of the malachite green molecule was attacked by strong oxidative hydroxyl radicals, after which the organic compound was degraded by the two pathways into 4,4-bis(dimethylamino) benzophenone, 4-(dimethylamino) benzophenone, 4-(dimethylamino) phenol, and other intermediate products. Eventually, malachite green was degraded into oxalic acid as a small molecular organic acid, which was degraded by processes such as demethylation, deamination, nitration, substitution, addition, and other reactions. Copyright © 2013 Elsevier B.V. All rights reserved.

Partially folded intermediates during trypsinogen denaturation

Directory of Open Access Journals (Sweden)

Martins N.F.

1999-01-01

Full Text Available The equilibrium unfolding of bovine trypsinogen was studied by circular dichroism, differential spectra and size exclusion HPLC. The change in free energy of denaturation was = 6.99 ± 1.40 kcal/mol for guanidine hydrochloride and = 6.37 ± 0.57 kcal/mol for urea. Satisfactory fits of equilibrium unfolding transitions required a three-state model involving an intermediate in addition to the native and unfolded forms. Size exclusion HPLC allowed the detection of an intermediate population of trypsinogen whose Stokes radii varied from 24.1 ± 0.4 Å to 26.0 ± 0.3 Å for 1.5 M and 2.5 M guanidine hydrochloride, respectively. During urea denaturation, the range of Stokes radii varied from 23.9 ± 0.3 Å to 25.7 ± 0.6 Å for 4.0 M and 6.0 M urea, respectively. Maximal intrinsic fluorescence was observed at about 3.8 M urea with 8-aniline-1-naphthalene sulfonate (ANS binding. These experimental data indicate that the unfolding of bovine trypsinogen is not a simple transition and suggest that the equilibrium intermediate population comprises one intermediate that may be characterized as a molten globule. To obtain further insight by studying intermediates representing different stages of unfolding, we hope to gain a better understanding of the complex interrelations between protein conformation and energetics.

Bacterial intermediate filaments

DEFF Research Database (Denmark)

Charbon, Godefroid; Cabeen, M.; Jacobs-Wagner, C.

2009-01-01

Crescentin, which is the founding member of a rapidly growing family of bacterial cytoskeletal proteins, was previously proposed to resemble eukaryotic intermediate filament (IF) proteins based on structural prediction and in vitro polymerization properties. Here, we demonstrate that crescentin...

Reactor safeguards

CERN Document Server

Russell, Charles R

1962-01-01

Reactor Safeguards provides information for all who are interested in the subject of reactor safeguards. Much of the material is descriptive although some sections are written for the engineer or physicist directly concerned with hazards analysis or site selection problems. The book opens with an introductory chapter on radiation hazards, the construction of nuclear reactors, safety issues, and the operation of nuclear reactors. This is followed by separate chapters that discuss radioactive materials, reactor kinetics, control and safety systems, containment, safety features for water reactor

Pressurized Water Reactors (PWR) and Boiling Water Reactors (BWR) are compared