Irradiation and testing of compact ignition tokamak toroidal field coil insulation materials

International Nuclear Information System (INIS)

Kanemoto, G.K.; Sherick, M.J.; Sparks, D.C.

1990-05-01

This report documents the results of an irradiation and testing program performed on behalf of Martin Marietta Energy Systems, Inc. in support of the Compact Ignition Tokamak Research and Development program. The purpose of the irradiation and testing program was to determine the effects of neutron and gamma irradiation on the mechanical and electrical properties of candidate toroidal field coil insulation materials. Insulation samples were irradiated in the Advanced Test Reactor (ATR) in a large I-hole. The insulation samples were irradiated within a lead shield to reduce exposure to gamma radiation to better approximate the desired ration of neutron to gamma exposure. Two different exposure levels were specified for the insulation samples. To accomplish this, the samples were encapsulated in two separate aluminum capsules; the capsules positioned at the ATR core mid-plane and at the top of the fueled region to take advantage of the axial cosine distribution of the neutron and gamma flux; and by varying the length of irradiation time of the two capsules. Disassembly of the irradiated capsules and testing of the insulation samples were performed at the Test Reactor Area (TRA) Hot Cell Facilities. Testing of the samples included shear compression static, shear compression fatigue, flexure static, and electrical resistance measurements

Conceptual design of a fast-ignition laser fusion reactor FALCON-D

International Nuclear Information System (INIS)

Goto, T.; Ogawa, Y.; Okano, K.; Hiwatari, R.; Asaoka, Y.; Someya, Y.; Sunahara, A.; Johzaki, T.

2008-10-01

A new conceptual design of the laser fusion power plant FALCON-D (Fast ignition Advanced Laser fusion reactor CONcept with a Dry wall chamber) has been proposed. The fast ignition method can achieve the sufficient fusion gain for a commercial operation (∼100) with about 10 times smaller fusion yield than the conventional central ignition method. FALCON-D makes full use of this property and aims at designing with a compact dry wall chamber (5 - 6 m radius). 1-D/2-D hydrodynamic simulations showed the possibility of the sufficient gain achievement with a 40 MJ target yield. The design feasibility of the compact dry wall chamber and solid breeder blanket system was shown through the thermomechanical analysis of the dry wall and neutronics analysis of the blanket system. A moderate electric output (∼400 MWe) can be achieved with a high repetition (30 Hz) laser. This dry wall concept not only reduces some difficulties accompanied with a liquid wall but also enables a simple cask maintenance method for the replacement of the blanket system, which can shorten the maintenance time. The basic idea of the maintenance method for the final optics system has also been proposed. Some critical R and D issues required for this design are also discussed. (author)

A cryogenic system for TIBER II [Tokamak Ignition/Burn Experimental Reactor

International Nuclear Information System (INIS)

Slack, D.S.; Kerns, J.A.

1987-01-01

Phase II of the Tokamak Ignition/Burn Experimental Reactor (TIBER II) study describes one option for a small, economical, next-generation tokamak [1,2]. Because of its small size, minimum shielding is used between the plasma and the toroidal-field (TF) coils. Consequently, a large cryogenic system (approximately 70 kW at 4.5 K) capable of delivering forced-flow helium is required. This paper describes a cryogenic system that meets this requirement and includes TIBER-II requirements. 3 refs

Approach to ignition of tokamak reactors

International Nuclear Information System (INIS)

Sigmar, D.J.

1981-02-01

Recent transport modeling results for JET, INTOR, and ETF are reviewed and analyzed with respect to existing uncertainties in the underlying physics, the self-consistency of the very large numerical codes, and the margin for ignition. The codes show ignition to occur in ETF/INTOR-sized machines if empirical scaling can be extrapolated to ion temperatures (and beta values) much higher than those presently achieved, if there is no significant impurity accumulation over the first 7 s, and if the known ideal and resistive MHD instabilities remain controllable for the evolving plasma profiles during ignition startup

Ignition characteristics of coal blends in an entrained flow furnace

Energy Technology Data Exchange (ETDEWEB)

J. Faundez; B. Arias; F. Rubiera; A. Arenillas; X. Garcia; A.L. Gordon; J.J. Pis [Universidad de Concepcion, Concepcion (Chile)

2007-09-15

Ignition tests were carried out on blends of three coals of different rank - subbituminous, high volatile and low volatile bituminous - in two entrained flow reactors. The ignition temperatures were determined from the gas evolution profiles (CO, CO{sub 2}, NO, O{sub 2}), while the mechanism of ignition was elucidated from these profiles and corroborated by high-speed video recording. Under the experimental conditions of high carbon loading, clear interactive effects were observed for all the blends. Ignition of the lower rank coals (subbituminous, high volatile bituminous) enhanced the ignition of the higher rank coal (low volatile bituminous) in the blends. The ignition temperatures of the blends of the low rank coals (subbituminous-high volatile bituminous) were additive. However, for the rest of the blends the ignition temperatures were always closer to the lower rank coal in the blend. 21 refs., 8 figs.

Conceptual design of a fast-ignition laser fusion reactor based on a dry wall chamber

International Nuclear Information System (INIS)

Ogawa, Y; Goto, T; Okano, K; Asaoka, Y; Hiwatari, R; Someya, Y

2008-01-01

The fast ignition is quite attractive for a compact laser fusion reactor, because a sufficiently high pellet gain is available with a small input energy. We designed an inertial fusion reactor based on Fast-ignition Advanced Laser fusion reactor CONcept, called FALCON-D, where a dry wall is employed for a chamber wall. A simple point model shows that the pellet gain G∼100 is available with laser energies of 350kJ for implosion, 50kJ for heating. This results in the fusion yield of 40 MJ in one shot. By increasing the repetition rate up to 30 Hz, the fusion power of 1.2 GWth becomes available. Plant system analysis shows the net electric power to be about 0.4 GWe In the fast ignition it is available to employ a low aspect ratio pellet, which is favorable for the stability during the implosion phase. Here the pellet aspect ratio is reduced to be 2 ∼ 4, and the optimization of the pulse shape for the implosion laser are carried out by using the 1-D hydrodynamic simulation code ILESTA-1D. A ferritic steel with a tungsten armour is employed for the chamber wall. The feasibility of this dry wall concept is studied from various engineering aspects such as surface melting, physical and chemical sputtering, blistering and exfoliation by helium retention, and thermo-mechanical fatigue, and it is found that blistering and exfoliation due to the helium retention and fatigue failure due to cyclic thermal load are major concerns. The cost analysis shows that the construction cost is moderate but the cost of electricity is slightly expensive

Conceptual design of a fast-ignition laser fusion reactor based on a dry wall chamber

Energy Technology Data Exchange (ETDEWEB)

Ogawa, Y [High Temperature Plasma Center, University of Tokyo, Chiba (Japan); Goto, T; Okano, K [Graduate School of Frontier Sciences, University of Tokyo, Chiba (Japan); Asaoka, Y; Hiwatari, R [Central Research Institute for Electric Power Industry, Komae, Tokyo (Japan); Someya, Y [Graduate School of Engineering, Musashi Institute of Technology, Tokyo (Japan)], E-mail: ogawa@ppl.k.u-tokyo.ac.jp

2008-05-15

The fast ignition is quite attractive for a compact laser fusion reactor, because a sufficiently high pellet gain is available with a small input energy. We designed an inertial fusion reactor based on Fast-ignition Advanced Laser fusion reactor CONcept, called FALCON-D, where a dry wall is employed for a chamber wall. A simple point model shows that the pellet gain G{approx}100 is available with laser energies of 350kJ for implosion, 50kJ for heating. This results in the fusion yield of 40 MJ in one shot. By increasing the repetition rate up to 30 Hz, the fusion power of 1.2 GWth becomes available. Plant system analysis shows the net electric power to be about 0.4 GWe In the fast ignition it is available to employ a low aspect ratio pellet, which is favorable for the stability during the implosion phase. Here the pellet aspect ratio is reduced to be 2 {approx} 4, and the optimization of the pulse shape for the implosion laser are carried out by using the 1-D hydrodynamic simulation code ILESTA-1D. A ferritic steel with a tungsten armour is employed for the chamber wall. The feasibility of this dry wall concept is studied from various engineering aspects such as surface melting, physical and chemical sputtering, blistering and exfoliation by helium retention, and thermo-mechanical fatigue, and it is found that blistering and exfoliation due to the helium retention and fatigue failure due to cyclic thermal load are major concerns. The cost analysis shows that the construction cost is moderate but the cost of electricity is slightly expensive.

Conceptual design of a fast-ignition laser fusion reactor based on a dry wall chamber

Science.gov (United States)

Ogawa, Y.; Goto, T.; Okano, K.; Asaoka, Y.; Hiwatari, R.; Someya, Y.

2008-05-01

The fast ignition is quite attractive for a compact laser fusion reactor, because a sufficiently high pellet gain is available with a small input energy. We designed an inertial fusion reactor based on Fast-ignition Advanced Laser fusion reactor CONcept, called FALCON-D, where a dry wall is employed for a chamber wall. A simple point model shows that the pellet gain G~100 is available with laser energies of 350kJ for implosion, 50kJ for heating. This results in the fusion yield of 40 MJ in one shot. By increasing the repetition rate up to 30 Hz, the fusion power of 1.2 GWth becomes available. Plant system analysis shows the net electric power to be about 0.4 GWe In the fast ignition it is available to employ a low aspect ratio pellet, which is favorable for the stability during the implosion phase. Here the pellet aspect ratio is reduced to be 2 ~ 4, and the optimization of the pulse shape for the implosion laser are carried out by using the 1-D hydrodynamic simulation code ILESTA-1D. A ferritic steel with a tungsten armour is employed for the chamber wall. The feasibility of this dry wall concept is studied from various engineering aspects such as surface melting, physical and chemical sputtering, blistering and exfoliation by helium retention, and thermo-mechanical fatigue, and it is found that blistering and exfoliation due to the helium retention and fatigue failure due to cyclic thermal load are major concerns. The cost analysis shows that the construction cost is moderate but the cost of electricity is slightly expensive.

Neutral beam system for an ignition tokamak

International Nuclear Information System (INIS)

Fasolo, J.; Fuja, R.; Jung, J.; Moenich, J.; Norem, J.; Praeg, W.; Stevens, H.

1978-01-01

We have attempted to make detailed designs of several neutral beam systems which would be applicable to a large machine, e.g. an ITR (Ignition Test Reactor), EPR (Experimental Power Reactor), or reactor. Detailed studies of beam transport to the reactor and neutron transport from the reactor have been made. We have also considered constraints imposed by the neutron radiation environment in the injectors, and the resulting shielding, radiation-damage, and maintenance problems. The effects of neutron heat loads on cryopanels and ZrAl getter panels have been considered. Design studies of power supplies, vacuum systems, bending magnets, and injector layouts are in progress and will be discussed

Tokamak and RFP ignition requirements

International Nuclear Information System (INIS)

Werley, K.A.

1991-01-01

A plasma model is applied to calculate numerically transport- confinement (nτ E ) requirements and steady-state operation tokamak. The CIT tokamak and RFP ignition conditions are examined. Physics differences between RFP and tokamaks, and their consequences for a DT ignition machine, are discussed. The ignition RFP, compared to a tokamak, has many physics advantages, including ohmic heating to ignition (no need for auxiliary heating systems), higher beta, low ignition current, less sensitivity of ignition requirements to impurity effects, no hard disruptions (associated with beta or density limits), and successful operation with high radiation fractions (f RAD ∼ 0.95). These physics advantages, coupled with important engineering advantages associated with lower external magnetic fields, larger aspect ratios, and smaller plasma cross sections translate into significant cost reductions for both ignition and power reactor. The primary drawback of the RFP is the uncertainty that the present confinement scaling will extrapolate to reactor regimes. The 4-MA ZTH was expected to extend the nτ E transport scaling data three order of magnitude above ZT-40M results, and if the present scaling held, to achieve a DT-equivalent scientific energy breakeven, Q=1. A basecase RFP ignition point is identified with a plasma current of 8.1 MA and no auxiliary heating. 16 refs., 4 figs., 1 tab

Experimental results pertaining to the performance of thermal igniters

International Nuclear Information System (INIS)

Carmel, M.K.

1989-10-01

This report summarizes the results of various experimental programs regarding the performance of thermal igniters for the deliberate ignition of hydrogen in light water reactors. Experiments involving both premixed combustion and combustion with continuous hydrogen injection are reviewed. Combustion characteristics examined include flammability limits of hydrogen:air and hydrogen:air:steam mixtures, combustion pressure rises, combustion completeness, flame speeds, and heat transfer aspects. Comparisons of igniter type and igniter reliability under simulated reactor accident conditions are included. The results of the research programs provide a broad data base covering nearly all aspects of hydrogen combustion related to the performance of deliberate ignition systems

RF-driven tokamak reactor with sub-ignited, thermally stable operation

International Nuclear Information System (INIS)

Harten, L.P.; Bers, A.; Fuchs, V.; Shoucri, M.M.

1981-02-01

A Radio-Frequency Driven Tokamak Reactor (RFDTR) can use RF-power, programmed by a delayed temperature measurement, to thermally stabilize a power equilibrium below ignition, and to drive a steady state current. We propose the parameters for such a device generating approx. = 1600 MW thermal power and operating with Q approx. = 40 (= power out/power in). A one temperature zero-dimensional model allows simple analytical formulation of the problem. The relevance of injected impurities for locating the equilibrium is discussed. We present the results of a one-dimensional (radial) code which includes the deposition of the supplementary power, and compare with our zero-dimensional model

TIBER II/ETR [Engineering Test Reactor] nuclear shielding and optional tritium breeding system: An overview

International Nuclear Information System (INIS)

Lee, J.D.; Sawan, M.

1987-01-01

TIBER II, the Tokamak Ignition/Burn Experimental Reactor II, is a design concept developed as the US candidate for an International Engineering Test Reactor (ETR). An important objective of this design is to minimize cost by minimizing major radius while providing a wall loading greater than 1.0 MW/m2 and a total fluence greater than 3.0 MWY/m2 needed for blanket module testing. The shielding required for the superconducting TF coils is an important element in setting TIBER II's 3.0m major radius. 6 refs., 1 fig., 1 tab

Report on ignitability testing of ''no-flow'' push bit

International Nuclear Information System (INIS)

Witwer, K.S.

1997-01-01

Testing was done to determine if an ignition occurs during a sixty foot drop of a Universal Sampler onto a push-mode bit in a flammable gas environment. Ten drops each of the sampler using both a push-mode and rotary mode insert onto a push-mode bit were completed. No ignition occurred during any of the drops

Reversed field pinch ignition requirements

International Nuclear Information System (INIS)

Werley, K.A.

1991-01-01

Plasma models are described and used to calculated numerically the transport confinement (nτ E ) requirements and steady state operation points for both the reversed field pinch (RFP) and the tokamak. The models are used to examine the CIT tokamak ignition conditions and the RFP experimental and ignition conditions. Physics differences between RFPs and tokamaks and their consequences for a D-T ignition machine are discussed. Compared with a tokamak, the ignition RFP has many physics advantages, including Ohmic heating to ignition (no need for auxiliary heating systems), higher beta, lower ignition current, less sensitivity of ignition requirements to impurity effects, no hard disruptions (associated with beta or density limits) and successful operation with high radiation fractions (f RAD ∼ 0.95). These physics advantages, coupled with important engineering advantages associated with lower external magnetic field, larger aspect ratios and smaller plasma cross-sections, translate to significant cost reductions for both ignition and reactor applications. The primary drawback of the RFP is the uncertainty that the present scaling will extrapolate to reactor regimes. Devices that are under construction should go a long way toward resolving this scaling uncertainty. The 4 MA ZTH is expected to extend the nτ E transport scaling data by three orders of magnitude above the results of ZT-40M, and, if the present scaling holds, ZTH is expected to achieve a D-T equivalent scientific energy breakeven, Q = 1. A base case RFP ignition point is identified with a plasma current of 8.1 MA and no auxiliary heating. (author). 19 refs, 11 figs, 3 tabs

Two-stage Lagrangian modeling of ignition processes in ignition quality tester and constant volume combustion chambers

KAUST Repository

Alfazazi, Adamu; Kuti, Olawole Abiola; Naser, Nimal; Chung, Suk-Ho; Sarathy, Mani

2016-01-01

The ignition characteristics of isooctane and n-heptane in an ignition quality tester (IQT) were simulated using a two-stage Lagrangian (TSL) model, which is a zero-dimensional (0-D) reactor network method. The TSL model was also used to simulate

Analysis of the NASA White Sands Test Facility (WSTF) Test System for Friction-Ignition of Metallic Materials

Science.gov (United States)

Shoffstall, Michael S.; Wilson, D. Bruce; Stoltzfus, Joel M.

2000-01-01

Friction is a known ignition source for metals in oxygen-enriched atmospheres. The test system developed by the NASA White Sands Test Facility in response to ASTM G-94 has been used successfully to determine the relative ignition from friction of numerous metallic materials and metallic materials pairs. These results have been ranked in terms of a pressure-velocity product (PV) as measured under the prescribed test conditions. A high value of 4.1(exp 8) watts per square meter for Inconel MA 754 is used to imply resistance to friction ignition, whereas a low value of 1.04(exp 8) watts per square meter for stainless steel 304 is taken as indicating material susceptible to friction ignition. No attempt has been made to relate PV values to other material properties. This work reports the analysis of the WSTF friction-ignition test system for producing fundamental properties of metallic materials relating to ignition through friction. Three materials, aluminum, titanium, and nickel were tested in the WSTF frictional ignition instrument system under atmospheres of oxygen or nitrogen. Test conditions were modified to reach a steady state of operation, that is applied, the force was reduced and the rotational speed was reduced. Additional temperature measurements were made on the stator sample. The aluminum immediately galled on contact (reproducible) and the test was stopped. Titanium immediately ignited as a result of non-uniform contact of the stator and rotor. This was reproducible. A portion of the stator sampled burned, but the test continued. Temperature measurements on the stator were used to validate the mathematical model used for estimating the interface (stator/rotor) temperature. These interface temperature measurements and the associate thermal flux into the stator were used to distinguish material-phase transitions, chemical reaction, and mechanical work. The mechanical work was used to analyze surface asperities in the materials and to estimate a

Electrically heated 3D-macro cellular SiC structures for ignition and combustion application

International Nuclear Information System (INIS)

Falgenhauer, Ralf; Rambacher, Patrick; Schlier, Lorenz; Volkert, Jochen; Travitzky, Nahum; Greil, Peter; Weclas, Miroslaw

2017-01-01

Highlights: • 3D-printed macro cellular SiC structure. • Directly integrated electrically heated ignition element used in combustion reactor. • Experimental investigation of the ignition process. - Abstract: The paper describes different aspects of porous combustion reactor operation especially at cold start conditions. Under cold start conditions it is necessary to increase the internal energy of the combustion reactor, to accumulate enough energy inside its solid phase and to reach at least the ignition temperature on the reactors inner surface. The most practicable method to preheat a cold porous reactor is to use its surface as a flame holder and to apply free flame combustion as a heat source for the preheating process. This paper presents a new electrically heated ignition element, which gets integrated in a three dimensional macro-cellular SiSiC reactor structure. For the development of the ignition element it was assumed, that the element is made of the same material as the combustion reactor itself and is fully integrated within the three-dimensional macro-cellular structure of the combustion reactor. Additive manufacturing like three-dimensional (3D) printing permits the production of regular SiSiC structures with constant strut thickness and a defined current flow path. To get a controlled temperature distribution on the ignition element it is necessary to control the current density distribution in the three-dimensional macro-cellular reactor structure. The ignition element used is designed to be an electrical resistance in an electric current system, converting flowing current into heat with the goal to get the highest temperature in the ignition region (glow plug). First experiments show that the ignition element integrated in a combustion reactor exhibits high dynamics and can be heated to the temperatures much above 1000 °C in a very short time (approx. 800 ms) for current of I = 150 A.

Ignition properties of nuclear grade activated carbons

International Nuclear Information System (INIS)

Freeman, W.P.; Hunt, J.R.; Kovach, J.L.

1983-01-01

The ignition property of new activated carbons used in air cleaning systems of nuclear facilities has been evaluated in the past, however very little information has been generated on the behavior of aged, weathered carbons which have been exposed to normal nuclear facility environment. Additionally the standard procedure for evaluation of ignition temperature of carbon is performed under very different conditions than those used in the design of nuclear air cleaning systems. Data were generated evaluating the ageing of activated carbons and comparing their CH 3 131 I removal histories to their ignition temperatures. A series of tests were performed on samples from one nuclear power reactor versus use time, a second series evaluated samples from several plants showing the variability of atmospheric effects. The ignition temperatures were evaluated simulating the conditions existing in nuclear air cleaning systems, such as velocity, bed depth, etc., to eliminate potential confusion resulting from artifically set current standard conditions

Activation analysis of the compact ignition tokamak

International Nuclear Information System (INIS)

Selcow, E.C.

1986-01-01

The US fusion program has completed the conceptual design of a compact tokamak device that achieves ignition. The high neutron wall loadings associated with this compact deuterium-tritium-burning device indicate that radiation-related issues may be significant considerations in the overall system design. Sufficient shielding will be requied for the radiation protection of both reactor components and occupational personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure to permit personnel access into the test cell after shutdown and limit the total activation of the test cell components. This paper describes the conceptual design of the igloo shield system and discusses the major neutronic concerns related to the design of the Compact Ignition Tokamak

Evaluation of the ignition behaviour of coals and blends

Energy Technology Data Exchange (ETDEWEB)

J. Faundez; F. Rubiera; X. Garcia; A. Arenillas; A.L. Gordon; J.J. Pis [CSIC, Instituto Nacional del Carbon, Oviedo (Spain). Department of Energy and Environment

2003-07-01

An experimental study about ignition of coals and blends was carried out by using an entrained flow reactor (EFR) with continuous feed. Seven coals of varying rank, from subbituminous to semianthracite, were tested and evolving gases (O{sub 2}, CO, CO{sub 2}, NO) were measured. The ignition temperature was evaluated from the evolution profiles of these gases, and correlated inversely to the reactivity of coals, as reflected by increasing values of ignition temperatures in the sequence subbituminous, high volatile bituminous, low volatile bituminous and semianthracite coals. Mechanism of ignition varied from an heterogeneous mechanism (for subbituminous, low volatile bituminous and semianthracite coals) to an homogeneous mechanism (for high volatile bituminous coal). Experiments with coal blends showed that if a low volatile bituminous coal is blended with a high volatile bituminous coal, the latter determines the value of the ignition temperature and ignition mechanism of the blend, when its percentage in the blend is 50% or higher. For blends of subbituminous and high volatile bituminous coals, the ignition mechanism of the blend is determined by the ignition mechanism of the coal with a higher content in the blend. 12 refs., 9 figs., 1 tab.

Laser fusion reactor design in a fast ignition with a dry wall chamber

International Nuclear Information System (INIS)

Ogawa, Yichi; Goto, Takuya; Ninomiya, Daisuke; Hiwatari, Ryoji; Asaoka, Yoshiyuki; Okano, Kunihiko

2007-01-01

One of the critical issues in laser fusion reactor design is high pulse heat load on the first wall by the X-rays and the fast/debris ions from fusion burn. There are mainly two concepts for the first wall of laser fusion reactor, a dry wall and a liquid metal wall. We should notice that the fast ignition method can achieve sufficiently high pellet gain with smaller (about 1/10 of the conventional central ignition method) input energy. To take advantage of this property, the design of a laser fusion reactor with a small size dry wall chamber may become possible. Since a small fusion pulse leads to a small electric power, high repetition of laser irradiation is required to keep sufficient electric power. Then we tried to design a laser fusion reactor with a dry wall chamber and a high repetition laser. This is a new challenging path to realize a laser fusion plant. Based on the point model of the core plasma, we have estimated that fusion energy in one pulse can be reduced to be 40 MJ with a pellet gain around G>100. To evaluate the validity of this simple estimation and to optimize the pellet design and the pulse shaping for the fast ignition scenario, we have introduced 1-D hydrodynamic simulation code ILESTA-1D and carried out implosion simulations. Since the code is one-dimensional, the detailed physics process of fast heating cannot be reproduced. Thus the fast heating is reflected in the code as the additional artificial heating source in the energy equation. It is modeled as a homogeneous heating of electrons in core region at the time just before when the maximum compression is achieved. At present we obtained the pellet gain G∝100 with the same input energy as the above estimation by a simple point model (350kJ for implosion, 50kJ for heating and assuming 20% coupling of heating laser). A dry wall is exposed to several threats due to the cyclic load by the high energy X-ray and charged particles: surface melting, physical and chemical sputtering

Rotating shield ceiling for the compact ignition tokamak test cell

International Nuclear Information System (INIS)

Commander, J.C.

1986-01-01

For the next phase of the United States fusion program, a compact, high-field, toroidal ignition machine with liquid nitrogen cooled copper coils, designated the Compact Ignition Tokamak (CIT), is proposed. The CIT machine will be housed in a test cell with design features developed during preconceptual design. Configured as a right cylinder, the selected test cell design features: a test cell and basement with thick concrete shielding walls, and floor; leak tight tritium seals; and operational characteristics well suited to the circular CIT machine configuration and radially oriented ancillary equipment and systems

Tests of an experimental slash ignition unit

Science.gov (United States)

James L. Murphy; Harry E. Schimke

1965-01-01

A prototype ignition package containing an incendiary powder and designed for slash and brush burning jobs showed some promise, but the unit tested was not superior to such conventional devices as fusees, diesel backpack type flamethrowers, Very pistols, and drip torches.

The Tokamak Fusion Test Reactor D-T modifications and operations

International Nuclear Information System (INIS)

1992-01-01

This Environmental Assessment (EA) was prepared in accordance with the National Environmental Policy Act (NEPA) of 1969, as amended, in support of the Department of Energy's proposal for the Tokamak Fusion Test Reactor (TFTR) D-T program. The objective of the proposed D-T program is to take the initial step in studying the effects of alpha particle heating and transport in a magnetic fusion device. These studies would enable the successful completion of the original TFTR program objectives, and would support the research and development needs of the Burning Plasma Experiment, BPX (formerly the Compact Ignition Tokamak (CIT)) and International Thermonuclear Experimental Reactor (ITER) in the areas of alpha particle physics, tritium retention, alpha particle diagnostic development, and tritium handling

Elements of a method to scale ignition reactor Tokamak

International Nuclear Information System (INIS)

Cotsaftis, M.

1984-08-01

Due to unavoidable uncertainties from present scaling laws when projected to thermonuclear regime, a method is proposed to minimize these uncertainties in order to figure out the main parameters of ignited tokamak. The method mainly consists in searching, if any, a domain in adapted parameters space which allows Ignition, but is the least sensitive to possible change in scaling laws. In other words, Ignition domain is researched which is the intersection of all possible Ignition domains corresponding to all possible scaling laws produced by all possible transports

Physics-based modeling of live wildland fuel ignition experiments in the Forced Ignition and Flame Spread Test apparatus

Science.gov (United States)

C. Anand; B. Shotorban; S. Mahalingam; S. McAllister; D. R. Weise

2017-01-01

A computational study was performed to improve our understanding of the ignition of live fuel in the forced ignition and flame spread test apparatus, a setup where the impact of the heating mode is investigated by subjecting the fuel to forced convection and radiation. An improvement was first made in the physics-based model WFDS where the fuel is treated as fixed...

High-Gain Shock Ignition on the National Ignition Facility

Science.gov (United States)

Perkins, L. J.; Lafortune, K.; Bailey, D.; Lambert, M.; MacKinnon, A.; Blackfield, D.; Comley, A.; Schurtz, G.; Ribeyre, X.; Lebel, E.; Casner, A.; Craxton, R. S.; Betti, R.; McKenty, P.; Anderson, K.; Theobald, W.; Schmitt, A.; Atzeni, S.; Schiavi, A.

2010-11-01

Shock ignition offers the possibility for a near-term test of high-gain ICF on the NIF at less than 1MJ drive energy and with day-1 laser hardware. We will summarize the status of target performance simulations, delineate the critical issues and describe the R&D program to be performed in order to test the potential of a shock-ignited target on NIF. In shock ignition, compressed fuel is separately ignited by a late-time laser-driven shock and, because capsule implosion velocities are significantly lower than those required for conventional hotpot ignition, simulations indicate that fusion energy gains of 60 may be achievable at laser energies around 0.5MJ. Like fast ignition, shock ignition offers high gain but requires only a single laser with less demanding timing and focusing requirements. Conventional symmetry and stability constraints apply, thus a key immediate step towards attempting shock ignition on NIF is to demonstrate adequacy of low-mode uniformity and shock symmetry under polar drive

Space Launch System Scale Model Acoustic Test Ignition Overpressure Testing

Science.gov (United States)

Nance, Donald; Liever, Peter; Nielsen, Tanner

2015-01-01

The overpressure phenomenon is a transient fluid dynamic event occurring during rocket propulsion system ignition. This phenomenon results from fluid compression of the accelerating plume gas, subsequent rarefaction, and subsequent propagation from the exhaust trench and duct holes. The high-amplitude unsteady fluid-dynamic perturbations can adversely affect the vehicle and surrounding structure. Commonly known as ignition overpressure (IOP), this is an important design-to environment for the Space Launch System (SLS) that NASA is currently developing. Subscale testing is useful in validating and verifying the IOP environment. This was one of the objectives of the Scale Model Acoustic Test, conducted at Marshall Space Flight Center. The test data quantifies the effectiveness of the SLS IOP suppression system and improves the analytical models used to predict the SLS IOP environments. The reduction and analysis of the data gathered during the SMAT IOP test series requires identification and characterization of multiple dynamic events and scaling of the event waveforms to provide the most accurate comparisons to determine the effectiveness of the IOP suppression systems. The identification and characterization of the overpressure events, the waveform scaling, the computation of the IOP suppression system knockdown factors, and preliminary comparisons to the analytical models are discussed.

Space Launch System Scale Model Acoustic Test Ignition Overpressure Testing

Science.gov (United States)

Nance, Donald K.; Liever, Peter A.

2015-01-01

The overpressure phenomenon is a transient fluid dynamic event occurring during rocket propulsion system ignition. This phenomenon results from fluid compression of the accelerating plume gas, subsequent rarefaction, and subsequent propagation from the exhaust trench and duct holes. The high-amplitude unsteady fluid-dynamic perturbations can adversely affect the vehicle and surrounding structure. Commonly known as ignition overpressure (IOP), this is an important design-to environment for the Space Launch System (SLS) that NASA is currently developing. Subscale testing is useful in validating and verifying the IOP environment. This was one of the objectives of the Scale Model Acoustic Test (SMAT), conducted at Marshall Space Flight Center (MSFC). The test data quantifies the effectiveness of the SLS IOP suppression system and improves the analytical models used to predict the SLS IOP environments. The reduction and analysis of the data gathered during the SMAT IOP test series requires identification and characterization of multiple dynamic events and scaling of the event waveforms to provide the most accurate comparisons to determine the effectiveness of the IOP suppression systems. The identification and characterization of the overpressure events, the waveform scaling, the computation of the IOP suppression system knockdown factors, and preliminary comparisons to the analytical models are discussed.

Effect of oxy-fuel combustion with steam addition on coal ignition and burnout in an entrained flow reactor

International Nuclear Information System (INIS)

Riaza, J.; Alvarez, L.; Gil, M.V.; Pevida, C.; Pis, J.J.; Rubiera, F.

2011-01-01

The ignition temperature and burnout of a semi-anthracite and a high-volatile bituminous coal were studied under oxy-fuel combustion conditions in an entrained flow reactor (EFR). The results obtained under oxy-fuel atmospheres (21%O 2 -79%CO 2 , 30%O 2 -70% O 2 and 35%O 2 -65%CO 2 ) were compared with those attained in air. The replacement of CO 2 by 5, 10 and 20% of steam in the oxy-fuel combustion atmospheres was also evaluated in order to study the wet recirculation of flue gas. For the 21%O 2 -79%CO 2 atmosphere, the results indicated that the ignition temperature was higher and the coal burnout was lower than in air. However, when the O 2 concentration was increased to 30 and 35% in the oxy-fuel combustion atmosphere, the ignition temperature was lower and coal burnout was improved in comparison with air conditions. On the other hand, an increase in ignition temperature and a worsening of the coal burnout was observed when steam was added to the oxy-fuel combustion atmospheres though no relevant differences between the different steam concentrations were detected. -- Highlights: → The ignition temperature and the burnout of two thermal coals under oxy-fuel combustion conditions were determined. → The effect of the wet recirculation of flue gas on combustion behaviour was evaluated. → Addition of steam caused a worsening of the ignition temperature and coal burnout.

Investigation and analysis of hydrogen ignition and explosion events in foreign nuclear power plants

Energy Technology Data Exchange (ETDEWEB)

Okuda, Yasunori [Institute of Nuclear Safety System, Inc., Mihama, Fukui (Japan)

2002-09-01

Reports about hydrogen ignition and explosion events in foreign nuclear power plants from 1980 to 2001 were investigated, and 31 events were identified. Analysis showed that they were categorized in (1) outer leakage ignition events and (2) inner accumulation ignition events. The dominant event for PWR (pressurized water reactor) was outer leakage ignition in the main generator, and in BWR (boiling water reactor) it was inner accumulation ignition in the off-gas system. The outer leakage ignition was a result of work process failure with the ignition source, operator error, or main generator hydrogen leakage. The inner accumulation ignition events were caused by equipment failure or insufficient monitoring. With careful preventive measures, the factors leading to these events could be eliminated. (author)

Two-stage Lagrangian modeling of ignition processes in ignition quality tester and constant volume combustion chambers

KAUST Repository

Alfazazi, Adamu

2016-08-10

The ignition characteristics of isooctane and n-heptane in an ignition quality tester (IQT) were simulated using a two-stage Lagrangian (TSL) model, which is a zero-dimensional (0-D) reactor network method. The TSL model was also used to simulate the ignition delay of n-dodecane and n-heptane in a constant volume combustion chamber (CVCC), which is archived in the engine combustion network (ECN) library (http://www.ca.sandia.gov/ecn). A detailed chemical kinetic model for gasoline surrogates from the Lawrence Livermore National Laboratory (LLNL) was utilized for the simulation of n-heptane and isooctane. Additional simulations were performed using an optimized gasoline surrogate mechanism from RWTH Aachen University. Validations of the simulated data were also performed with experimental results from an IQT at KAUST. For simulation of n-dodecane in the CVCC, two n-dodecane kinetic models from the literature were utilized. The primary aim of this study is to test the ability of TSL to replicate ignition timings in the IQT and the CVCC. The agreement between the model and the experiment is acceptable except for isooctane in the IQT and n-heptane and n-dodecane in the CVCC. The ability of the simulations to replicate observable trends in ignition delay times with regard to changes in ambient temperature and pressure allows the model to provide insights into the reactions contributing towards ignition. Thus, the TSL model was further employed to investigate the physical and chemical processes responsible for controlling the overall ignition under various conditions. The effects of exothermicity, ambient pressure, and ambient oxygen concentration on first stage ignition were also studied. Increasing ambient pressure and oxygen concentration was found to shorten the overall ignition delay time, but does not affect the timing of the first stage ignition. Additionally, the temperature at the end of the first stage ignition was found to increase at higher ambient pressure

The Ignition Physics Study Group

International Nuclear Information System (INIS)

Sheffield, J.

1987-01-01

In the US magnetic fusion program there have been relatively few standing committees of experts, with the mandate to review a particular sub-area on a continuing basis. Generally, ad hoc committees of experts have been assembled to advise on a particular issue. There has been a lack of broad, systematic and continuing review and analysis, combining the wisdom of experts in the field, in support of decision making. The Ignition Physics Study Group (IPSG) provides one forum for the systematic discussion of fusion science, complementing the other exchanges of information, and providing a most important continuity in this critical area. In a similar manner to the European program, this continuity of discussion and the focus provided by a national effort, Compact Ignition Tokamak (CIT), and international effort, Engineering Test Reactor (ETR), are helping to lower those barriers which previously were an impediment to rational debate

Use of a fluidized bed combustor and thermogravimetric analyzer for the study of coal ignition temperature

International Nuclear Information System (INIS)

Ávila, Ivonete; Crnkovic, Paula M.; Luna, Carlos M.R.; Milioli, Fernando E.

2017-01-01

Highlights: • Coal ignition tests were conducted in a fluidized bed and thermogravimetric conditions. • The use of two different ignition criteria showed a similar coal ignition temperature. • Coal ignition temperature was obtained by the changes of gas concentrations in FBC. • Ignition temperatures were associated with the activation energy of coal combustion. - Abstract: Ignition experiments with two bituminous coals were carried out in an atmospheric bubbling fluidized bed combustor (FBC) and a thermogravimetric analyzer (TGA). In the FBC tests, the rapid increase in O_2, CO_2, and SO_2 concentrations is an indication of the coal ignition. In the TGA technique, the ignition temperature was determined by the evaluation of the TGA curves in both combustion and pyrolysis processes. Model-Free Kinetics was applied and the coal ignition temperatures were associated with changes in the activation energy values during the combustion process. The results show the coal with the lowest activation energy also showed the lowest ignition temperature, highest values of volatile content and a higher heating value. The application of two different ignition criteria (TGA and FBC) resulted in similar ignition temperatures. The FBC curves indicated the high volatile coal ignites in the freeboard, i.e. during the feeding in the reactor, whereas the low volatile coal ignites in the bed. Finally, the physicochemical characteristics of the investigated coal types were correlated with their reactivities for the prediction of the ignition temperatures behaviors under different operating conditions as those in FBC.

Analysis of Hydrogen Control Strategy Using Igniter during Severe Accident

International Nuclear Information System (INIS)

Lee, Sung Bok; Kim, Hyeong Taek; Lee, Keo Hyoung

2008-01-01

The Severe Accident Management Guidelines (SAMGs) for the operating pressurized water reactor (PWR) have been completed within 2006. Among the SAMG strategies, mitigation-07 is the most important strategy for managing a severe accident of a PWR in order to reduce containment hydrogen. The fastest way to reduce the containment hydrogen concentration is to intentionally ignite the hydrogen. For this strategy, igniters exist in Optimized Power Reactor 1000 (OPR 1000) to burn hydrogen for a severe accident. For using the igniters during a severe accident, the adverse effects such as the explosion of the hydrogen mixture should be considered for containment integrity. However, an applicable discrimination method to activate the igniters does not exist, so that the hydrogen control strategy using the igniters cannot be chosen during a severe accident. Thus, this study focused on suggesting an applicable discrimination method to carry out the strategy of using the igniters. In this study, the specific plant used for this analysis is Ulchin Unit 5 and 6, OPR 1000 plant, in Korea

Near-term tokamak-reactor designs with high-performance resistive magnets

International Nuclear Information System (INIS)

Cohn, D.R.; Bromberg, L.; Williams, J.E.C.; Becker, H.; Leclaire, R.; Yang, T.

1981-10-01

Advanced Fusion Test Reactors (AFTR) designs have been developed using BITTER type magnets which are capable of steady state operation. The goals of compact AFTR designs (with major radii R approx. 2.5 - 4 m), include DT ignition with large physics margins; high duty cycle, long pulse operation; and DD-DT operation with low tritium concentration. Larger AFTR designs (R approx. 5 m), have the additional goal of early demonstration of self sufficiency in tritium production. The AFTR devices could also serve as prototypes for commercial reactors. Compact ignition test reactors have also been designed (R approx. 1 - 2 m). These designs use BITTER magnets that are inertially cooled starting at liquid nitrogen temperature. A detailed engineering design was developed for ZEPHYR

Development And Testing Of Biogas-Petrol Blend As An Alternative Fuel For Spark Ignition Engine

Directory of Open Access Journals (Sweden)

Awogbemi

2015-08-01

Full Text Available Abstract This research is on the development and testing of a biogas-petrol blend to run a spark ignition engine. A2080 ratio biogaspetrol blend was developed as an alternative fuel for spark ignition engine test bed. Petrol and biogas-petrol blend were comparatively tested on the test bed to determine the effectiveness of the fuels. The results of the tests showed that biogas petrol blend generated higher torque brake power indicated power brake thermal efficiency and brake mean effective pressure but lower fuel consumption and exhaust temperature than petrol. The research concluded that a spark ignition engine powered by biogas-petrol blend was found to be economical consumed less fuel and contributes to sanitation and production of fertilizer.

The ignition of magnesium and magnox alloys. A review of data

International Nuclear Information System (INIS)

Pearce, R.J.

1987-07-01

Available data on the ignition temperatures of magnesium and the Magnox alloys of interest in CEGB Magnox reactor operations have been reviewed. Two basic ignition modes, instantaneous where a continuing and high rate of temperature rise is applied to the metal and delayed where the metal is held at a predetermined ambient temperature until ignition occurs, are identified. (author)

Fusion power demonstration - a baseline for the mirror engineering test reactor

International Nuclear Information System (INIS)

Henning, C.D.; Logan, B.G.; Neef, W.S.

1983-01-01

Developing a definition of an engineering test reactor (ETR) is a current goal of the Office of Fusion Energy (OFE). As a baseline for the mirror ETR, the Fusion Power Demonstration (FPD) concept has been pursued at Lawrence Livermore National Laboratory (LLNL) in cooperation with Grumman Aerospace, TRW, and the Idaho National Engineering Laboratory. Envisioned as an intermediate step to fusion power applications, the FPD would achieve DT ignition in the central cell, after which blankets and power conversion would be added to produce net power. To achieve ignition, a minimum central cell length of 67.5 m is needed to supply the ion and alpha particles radial drift pumping losses in the transition region. The resulting fusion power is 360 MW. Low electron-cyclotron heating power of 12 MW, ion-cyclotron heating of 2.5 MW, and a sloshing ion beam power of 1.0 MW result in a net plasma Q of 22. A primary technological challenge is the 24-T, 45-cm bore choke coil, comprising a copper hybrid insert within a 15 to 18 T superconducting coil

Insulation irradiation test programme for the Compact Ignition Tokamak

International Nuclear Information System (INIS)

McManamy, T.J.; Kanemoto, G.; Snook, P.

1991-01-01

In a programme to evaluate the effects of radiation exposure on the electrical insulation for the toroidal field coils of the Compact Ignition Tokamak, three types of boron-free insulation were irradiated at room temperature in the Advanced Technology Reactor (ATR) and tested at the Idaho National Engineering Laboratory. The materials were Spaulrad-S, Shikishima PG5-1 and Shikishima PG3-1. The first two use a bismaleimide resin and the third an aromatic amine hardened epoxy. Spaulrad-S is a two-dimensional (2-D) weave of S-glass, while the others are 3-D weaves of T-glass. Flexure and shear/compression samples were irradiated to ≅ 5 x 10 9 and 3 x 10 10 rad with 35-40% of the total dose from neutrons. The shear/compression samples were tested in pairs by applying an average compression of 345 MPa and then a shear load. After static tests were completed, fatigue testing was performed by cycling the shear load for up to 30000 cycles with a constant compression. The static shear strength of the samples that did not fail was then determined. Generally, shear strengths of the order of 120 MPa were measured. The behaviour of the flexure and shear/compression samples was significantly different; large reductions in the flexure strength were observed, while the shear strength stayed the same or increased slightly. The 3-D weave material demonstrated higher strength and significantly less radiation damage than the 2-D material in flexure but performed almost identically when tested with combined shear and compression. The epoxy system was much more sensitive to fatigue damage than the bismaleimide materials. No swelling was measured; however, the epoxy samples did twist slightly. (author)

Preliminary results of thermal igniter experiments in H2-air-steam environments

International Nuclear Information System (INIS)

Lowry, W.

1981-01-01

Thermal igniters (glow plugs), proposed by the Tennessee Valley Authority for intentional ignition of hydrogen in nuclear reactor containment, have been tested for functionability in mixtures of air, hydrogen, and steam. Test environments included 6% to 16% hydrogen concentrations in air, and 8%, 10%, and 12% hydrogen in mixtures with 30% and 40% steam fractions. All were conducted in a 10.6 ft 3 insulated pressure vessel. For all of these tests the glow plug successfully initiated combustion. Dry air/hydrogen tests exhibited a distinct tendency for complete combustion at hydrogen concentrations between 8% and 9%. Steam suppressed both peak pressures and completeness of combustion. No combustion could be initiated at or above a 50% steam fraction. Circulation of the mixture with a fan increased the completeness of combustion. The glow plug showed no evidence of performance degradation throughout the program

Deliberate ignition of hydrogen-air-steam mixtures in condensing steam environments

International Nuclear Information System (INIS)

Blanchat, T.K.; Stamps, D.W.

1997-05-01

Large scale experiments were performed to determine the effectiveness of thermal glow plug igniters to burn hydrogen in a condensing steam environment due to the presence of water sprays. The experiments were designed to determine if a detonation or accelerated flame could occur in a hydrogen-air-steam mixture which was initially nonflammable due to steam dilution but was rendered flammable by rapid steam condensation due to water sprays. Eleven Hydrogen Igniter Tests were conducted in the test vessel. The vessel was instrumented with pressure transducers, thermocouple rakes, gas grab sample bottles, hydrogen microsensors, and cameras. The vessel contained two prototypic engineered systems: (1) a deliberate hydrogen ignition system and (2) a water spray system. Experiments were conducted under conditions scaled to be nearly prototypic of those expected in Advanced Light Water Reactors (such as the Combustion Engineering (CE) System 80+), with prototypic spray drop diameter, spray mass flux, steam condensation rates, hydrogen injection flow rates, and using the actual proposed plant igniters. The lack of any significant pressure increase during the majority of the burn and condensation events signified that localized, benign hydrogen deflagration(s) occurred with no significant pressure load on the containment vessel. Igniter location did not appear to be a factor in the open geometry. Initially stratified tests with a stoichiometric mixture in the top showed that the water spray effectively mixes the initially stratified atmosphere prior to the deflagration event. All tests demonstrated that thermal glow plugs ignite hydrogen-air-steam mixtures under conditions with water sprays near the flammability limits previously determined for hydrogen-air-steam mixtures under quiescent conditions. This report describes these experiments, gives experimental results, and provides interpretation of the results. 12 refs., 127 figs., 16 tabs

OECD/NEA Sandia Fuel Project phase I: Benchmark of the ignition testing

Energy Technology Data Exchange (ETDEWEB)

Adorni, Martina, E-mail: martina_adorni@hotmail.it [UNIPI (Italy); Herranz, Luis E. [CIEMAT (Spain); Hollands, Thorsten [GRS (Germany); Ahn, Kwang-II [KAERI (Korea, Republic of); Bals, Christine [GRS (Germany); D' Auria, Francesco [UNIPI (Italy); Horvath, Gabor L. [NUBIKI (Hungary); Jaeckel, Bernd S. [PSI (Switzerland); Kim, Han-Chul; Lee, Jung-Jae [KINS (Korea, Republic of); Ogino, Masao [JNES (Japan); Techy, Zsolt [NUBIKI (Hungary); Velazquez-Lozad, Alexander; Zigh, Abdelghani [USNRC (United States); Rehacek, Radomir [OECD/NEA (France)

2016-10-15

Highlights: • A unique PWR spent fuel pool experimental project is analytically investigated. • Predictability of fuel clad ignition in case of a complete loss of coolant in SFPs is assessed. • Computer codes reasonably estimate peak cladding temperature and time of ignition. - Abstract: The OECD/NEA Sandia Fuel Project provided unique thermal-hydraulic experimental data associated with Spent Fuel Pool (SFP) complete drain down. The study conducted at Sandia National Laboratories (SNL) was successfully completed (July 2009 to February 2013). The accident conditions of interest for the SFP were simulated in a full scale prototypic fashion (electrically heated, prototypic assemblies in a prototypic SFP rack) so that the experimental results closely represent actual fuel assembly responses. A major impetus for this work was to facilitate severe accident code validation and to reduce modeling uncertainties within the codes. Phase I focused on axial heating and burn propagation in a single PWR 17 × 17 assembly (i.e. “hot neighbors” configuration). Phase II addressed axial and radial heating and zirconium fire propagation including effects of fuel rod ballooning in a 1 × 4 assembly configuration (i.e. single, hot center assembly and four, “cooler neighbors”). This paper summarizes the comparative analysis regarding the final destructive ignition test of the phase I of the project. The objective of the benchmark is to evaluate and compare the predictive capabilities of computer codes concerning the ignition testing of PWR fuel assemblies. Nine institutions from eight different countries were involved in the benchmark calculations. The time to ignition and the maximum temperature are adequately captured by the calculations. It is believed that the benchmark constitutes an enlargement of the validation range for the codes to the conditions tested, thus enhancing the code applicability to other fuel assembly designs and configurations. The comparison of

TESTING OF GAS REACTOR MATERIALS AND FUEL IN THE ADVANCED TEST REACTOR

International Nuclear Information System (INIS)

Grover, S.B.

2004-01-01

The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

Testing of Gas Reactor Materials and Fuel in the Advanced Test Reactor

International Nuclear Information System (INIS)

S. Blaine Grover

2004-01-01

The Advanced Test Reactor (ATR) has long been involved in testing gas reactor materials, and has developed facilities well suited for providing the right conditions and environment for gas reactor tests. This paper discusses the different types of irradiation hardware that have been utilized in past ATR irradiation tests of gas reactor materials. The new Gas Test Loop facility currently being developed for the ATR is discussed and the different approaches being considered in the design of the facility. The different options for an irradiation experiment such as active versus passive temperature control, neutron spectrum tailoring, and different types of lead experiment sweep gas monitors are also discussed. The paper is then concluded with examples of different past and present gas reactor material and fuel irradiations

Experimental study of hydrogen jet ignition and jet extinguishment

International Nuclear Information System (INIS)

Wierman, R.W.

1979-04-01

Two phases are described of an experimental study that investigated: (1) the ignition characteristics of hydrogen--sodium jets, (2) the formation of hydrogen in sodium--humid air atmospheres, and (3) the extinguishment characteristics of burning hydrogen--sodium jets. Test conditions were similar to those postulated for highly-improbable breeder reactor core melt-through accidents and included: jet temperature, jet velocity, jet hydrogen concentration, jet sodium concentration, atmospheric oxygen concentration, and atmospheric water vapor concentration

Reactor coolant pump for a nuclear reactor

International Nuclear Information System (INIS)

Burkhardt, W.; Richter, G.

1976-01-01

An improvement is proposed concerning the easier disengagement of the coupling at the reactor coolant pump for a nuclear reactor transporting a pressurized coolant. According to the invention the disengaging coupling consists of two parts separated by screws. At least one of the screws contains a propellent charge ananged within a bore and provided with a speed-dependent ignition device in such a way that by separation of the screws at overspeeds the coupling is disengaged. The sub-claims are concerned with the kind of ignition ot the propellent charge. (UWI) [de

Studies of conceptual spheromak fusion reactors

International Nuclear Information System (INIS)

Katsurai, M.; Yamada, M.

1982-01-01

Preliminary design studies are carried out for a spheromak fusion reactor. Simplified circuit theory is applied to obtain the characteristic relations among various parameters of the spheromak configuration for an aspect ratio of A >or approx. 1.6. These relations are used to calculate the parameters for the conceptual designs of three types of fusion reactor: (1) the DT reactor with two-component-type operation, (2) the ignited DT reactor, and (3) the ignited catalysed-type DD reactor. With a total wall loading of approx. 4 MW.m -2 , it is found that edge magnetic fields of only approx. 4 T (DT) and approx. 9 T (Cat. DD) are required for ignited reactors of 1 m plasma (minor) radius with output powers in the gigawatt range. An assessment of various schemes of generation, compression and translation of spheromak plasmas is presented. (author)

Physics analysis of the TIBER-II engineering test reactor

International Nuclear Information System (INIS)

Uckan, N.A.; Houlberg, W.A.; Attenberger, S.E.; Dory, R.A.; Spong, D.A.; Tolliver, J.S.; Sheffield, J.

1987-11-01

Confinement capability, burn characteristics, heating and fueling requirements, and fast alpha particle effects are assessed for the TIBER-II engineering test reactor (ETR/ITER). Confinement predictions for a wide variety of empirical scaling laws show that ignition in TIBER-II (or similar ETR-like devices) is marginal at 10 MA, whereas the design goal to achieve noninductively driven, steady-state burn with Q > 5 can easily be attained. Operation at the higher plasma currents being discussed for ITER or the attainment of higher density limits and/or favorable H-mode scalings improves the ignition capability. Pellet penetration calculations indicate that density profile control with pellets may not be feasible even for pellet velocities up to about 50 km/s, however, density peaking could result from inward pinch effects, as frequently inferred from experiments. The fast alpha contribution to pressure is substantial (10 to 30%) at TIBER (or any ETR/ITER) burn temperatures (8 to 20 keV). A relatively low level of fast alpha radial diffusion or a modest level of thermal alpha buildup significantly influences the ignition and steady-state burn capability. The fast alpha population can also modify the background plasma ballooning mode stability boundaries, lowering the beta limit β/sub crit/ - in particular, operation at the high electron temperatures needed for efficient current drive can exacerbate this effect. The use of high-energy neutral beams offers the promise of two important improvements in projected performance: an effective method for noninductive current drive and a means for controlling the current density profile deep within the plasma, as required for stable operation at high beta levels. 14 refs., 10 figs., 1 tab

Physics analysis of the TIBER-II engineering test reactor

International Nuclear Information System (INIS)

Uckan, N.A.; Houlberg, W.A.; Attenberger, S.E.; Dory, R.A.; Spong, D.A.; Tolliver, J.S.; Sheffield, J.

1987-01-01

Confinement capability, burn characteristics, heating and fueling requirements, and fast-alpha particle effects are assessed for the TIBER-II engineering test reactor (ETR/ITER). Confinement predictions for a wide variety of empirical scaling laws show that ignition on TIBER-II (or similar ETR-like devices) is marginal at 10 MA, whereas the design goal to achieve noninductively driven, steady-state burn with Q > 5 can easily be attained. Operation at the higher plasma currents being discussed for ITER or the attainment of higher density limits and/or favorable H-mode scalings improves the ignition capability. Pellet penetration calculations indicate that density profile control with pellets may not be feasible even for pellet velocities up to bout 50 km/s; however, density peaking could result from inward pinch effects, as frequently inferred from experiments. The fast alpha contribution to pressure is substantial (10-30%) at TIBER (or any ETR/ITER) burn temperatures (8-20 keV). A relatively low level of fast alpha radial diffusion or a modest level of thermal alpha buildup significantly influences the ignition and steady-state burn capability. The fast alpha population can also modify the background plasma ballooning mode stability boundaries, lowering the beta limit β/sub crit/ - in particular, operation at the high electron temperatures needed for efficient current drive can exacerbate this effect. The use of high-energy neutral beams offers the promise of two important improvements in projected performance: an effective method for noninductive current drive and a means for controlling the current density profile deep within the plasma, as required for stable operation at high beta levels

Conceptual design of the fast ignition laser fusion power plant (KOYO-Fast). 6. Design of chamber and reactor system

International Nuclear Information System (INIS)

Kozaki, Yasuji; Norimatsu, Takayoshi; Furukawa, Hiroyuki; Hayashi, Takumi; Souman, Yoshihito; Nishikawa, Masabumi; Tomabechi, Ken

2007-01-01

A conceptual design of the reactor chamber system with LiPb liquid wall based on the fast ignition cone target design and the related reactor systems with exhaust system, laser beam shutter, blanket and cooling system are summarized. The multi overflow fall method was investigated as the structure of chamber and repeating 4 Hz pulse potential. The ablation depth of LiPb liquid wall was estimated and the conditions of repeat of operation were evaluated. The basic design of chamber, selection and conditions of liquid wall chamber, recycle type multi overflow fall (MOF) wall, LiPb two layers blanket structure, basic specification of reactor system, laser beam line shutter, design of chamber exhaust system, cooling system, tritium recovery system, power plant total design and arrangement of chamber and laser beam, and issues are stated. (S.Y.)

Wildfire ignition resistant home design(WIRHD) program: Full-scale testing and demonstration final report.

Energy Technology Data Exchange (ETDEWEB)

Quarles, Stephen, L.; Sindelar, Melissa

2011-12-13

The primary goal of the Wildfire ignition resistant home design(WIRHD) program was to develop a home evaluation tool that could assess the ignition potential of a structure subjected to wildfire exposures. This report describes the tests that were conducted, summarizes the results, and discusses the implications of these results with regard to the vulnerabilities to homes and buildings.

Influence of fast alpha diffusion and thermal alpha buildup on tokamak reactor performance

International Nuclear Information System (INIS)

Uckan, N.A.; Tolliver, J.S.; Houlberg, W.A.; Attenberger, S.E.

1988-01-01

The effect of fast alpha diffusion and thermal alpha accumulation on the confinement capability of a candidate Engineering Test Reactor plasma (Tokamak Ignition/Burn Experimental Reactor) in achieving ignition and steady-state driven operation has been assessed using both global and 1-1/2-dimensional transport models. Estimates are made of the threshold for radial diffusion of fast alphas and thermal alpha buildup. It is shown that a relatively low level of radial transport, when combined with large gradients in the fast alpha density, leads to a significant radial flow with a deleterious effect on plasma performance. Similarly, modest levels of thermal alpha concentration significantly influence the ignition and steady-state burn capability

Confinement requirements for OHMIC-compressive ignition of a Spheromak plasma

International Nuclear Information System (INIS)

Olson, R.; Gilligan, J.; Miley, G.

1980-01-01

The Moving Plasmoid Reactor (MPR) is an attractive alternative magnetic fusion scheme in which Spheromak plasmoids are envisioned to be formed, compressed, burned, and expanded as the plasmoids translate through a series of linear reactor modules. Although auxiliary heating of the plasmoids may be possible, the MPR scenario would be especially interesting if ohmic decay and compression along were sufficient to heat the plasmoids to an ignition temperature. In the present work, we will study the transport conditions under which a Spheromak plasmoid could be expected to reach ignition via a combination of ohmic and compression heating

Confinement requirements for ohmic-compressive ignition of a Spheromak plasma

International Nuclear Information System (INIS)

Olson, R.E.; Miley, G.H.

1981-01-01

The Moving Plasmoid Reactor (MPR) is an attractive alternative magnetic fusion scheme in which Spheromak plasmoids are envisioned to be formed, compressed, burned, and expanded as the plasmoids translate through a series of linear reactor modules. Although auxiliary heating of the plasmoids may be possible, the MPR scenario would be especially interesting if ohmic decay and compression alone is sufficient to heat the plasmoids to an ignition temperature. In the present work, we examine the transport conditions under which a Spheromak plasmoid can be expected to reach ignition via a combination of ohmic and compression heating

Recent progress in ignition fusion research on the National Ignition Facility

International Nuclear Information System (INIS)

Leeper, Ramon J.

2011-01-01

This paper will review the ignition fusion research program that is currently being carried out on the National Ignition Facility (NIF) located at Lawrence Livermore National Laboratory. This work is being conducted under the auspices of the National Ignition Campaign (NIC) that is a broad collaboration of national laboratories and universities that together have developed a detailed research plan whose goal is ignition in the laboratory. The paper will begin with a description of the NIF facility and associated experimental facilities. The paper will then focus on the ignition target and hohlraum designs that will be tested in the first ignition attempts on NIF. The next topic to be introduced will be a description of the diagnostic suite that has been developed for the initial ignition experiments on NIF. The paper will then describe the experimental results that were obtained in experiments conducted during the fall of 2009 on NIF. Finally, the paper will end with a description of the detailed experimental plans that have been developed for the first ignition campaign that will begin later this year. (author)

New facilities in Japan materials testing reactor for irradiation test of fusion reactor components

International Nuclear Information System (INIS)

Kawamura, H.; Sagawa, H.; Ishitsuka, E.; Sakamoto, N.; Niiho, T.

1996-01-01

The testing and evaluation of fusion reactor components, i.e. blanket, plasma facing components (divertor, etc.) and vacuum vessel with neutron irradiation is required for the design of fusion reactor components. Therefore, four new test facilities were developed in the Japan Materials Testing Reactor: an in-pile functional testing facility, a neutron multiplication test facility, an electron beam facility, and a re-weldability facility. The paper describes these facilities

A mechanistic approach to safe igniter implementation for hydrogen mitigation

International Nuclear Information System (INIS)

Breitung, W.; Dorofeev, S.B.; Travis, J.R.

1997-01-01

A new methodology for safe igniter implementation in a full-scale 3-d containment is described. The method consists of the following steps: determination of bounding H 2 /steam sources; high-resolution analysis of the 3-d transport and mixing processes; evaluation of the detonation potential at the time of ignition; optimization of the igniter system such that only early ignition and nonenergetic combustion occurs; and modelling of the continuous deflagration processes during H 2 -release. The method was implemented into the GASFLOW code. The principle and the feasibility is demonstrated for a single room geometry. A full-scale 3-d reactor case is analyzed without and with deliberate ignition, assuming a severe dry H 2 release sequence (1200 kg). In the unmitigated case significant DDT potential in the whole containment develops, including the possibility of global detonations. The analysis with igniters in different positions predicted deflagration or detonation in the break compartment, depending on the igniter location. Igniter positions were found which lead to early ignition, effective H 2 -removal, and negligible pressure loads. The approach can be used to determine number, position and frequency of a safe igniter system for a given large dry containment. (author)

Ignitability of hydrogen/oxygen/diluent mixtures in the presence of hot surfaces

International Nuclear Information System (INIS)

Kumar, R.K.; Koroll, G.W.

1995-01-01

In the licensing process for CANDU nuclear power stations it is necessary to demonstrate tolerance to a wide range of low-probability accidents. These include loss of moderator accidents that may lead to the formation of flammable mixtures of deuterium, oxygen, helium, and steam in the reactor calandria vessel. Uncovered adjuster or control rods are considered as possible sources of ignition when a flammable mixture is present. A knowledge of the minimum hot-surface temperature required for ignition is important in assessing the reactor safety. These hot surface temperatures were measured using electrically heated adjuster rod simulators in a large spherical vessel (2.3-m internal diameter). Whereas the effects of geometry on ignition temperature were studied in the large-scale apparatus, some of the effects, such as those produced by a strong radiation field, were studied using a small-scale apparatus. Investigations carried our over a range of hydrogen and diluent concentrations indicated that, although the ignition temperatures were fairly insensitive to the hydrogen concentration, they were strongly affected by the presence of steam The addition of 30% steam to a dry combustible mixture increased the minimum surface temperature required for ignition by approximates 100 degrees C of the diluents investigated, steam had the most effect on ignition. The effect of initial temperature of the mixture on the ignition temperature was small, whereas the effect of initial pressure was significant. The effect of substituting deuterium for hydrogen on ignition temperature was small. The effect of a high-intensity gamma-radiation field on the minimum hot-surface temperature required for ignition was investigated using a 2-dm 3 ignition vessel placed in a linear accelerator. Radiation had no measurable effect on ignition temperature

Ignition experiment - alternatives

International Nuclear Information System (INIS)

Knobloch, A.F.

1979-10-01

This report comprises three short papers on cost estimates, integral burn time and alternative versions of Tokamak ignition experiments. These papers were discussed at the ZEPHYR workshop with participants from IPP Garching, MIT Cambridge and PPPL Princeton (Garching July 30 - August 2 1979) (Chapters A, B, C). It is shown, that starting from a practical parameter independent minimum integral burn time of Tokamak ignition experiments (some 10 3 s) by adding a shield for protection of the magnet insulation (permitted neutron dose 10 9 rad) an integral burn time of some 10 4 s can be achieved for only about 30% more outlay. For a substantially longer integral burn time the outlay approaches rather quickly that for a Tokamak reactor. Some examples for alternatives to ZEPHYR are being given, including some with low or no compression. In a further chapter D some early results of evaluating an ignition experiment on the basis of the energy confinement scaling put forward by Coppi and Mazzucato are presented. As opposed to the case of the Alcator scaling used in chapters A through C the minimum integral burn time of Tokamak ignition experiments here depends on the plasma current. Provided neutral injectors up to about 160 keV are available compression boosting is not required with this scaling. The results presented have been obtained neglecting the effects of the toroidal field ripple. (orig.) 891 HT/orig. 892 RKD [de

Conceptual design of a Bitter-magnet toroidal-field system for the ZEPHYR Ignition Test Reactor

International Nuclear Information System (INIS)

Williams, J.E.C.; Becker, H.D.; Bobrov, E.S.; Bromberg, L.; Cohn, D.R.; Davin, J.M.; Erez, E.

1981-05-01

The following problems are described and discussed: (1) parametric studies - these studies examine among other things the interdependence of throat stresses, plasma parameters (margins of ignition) and stored energy. The latter is a measure of cost and is minimized in the present design; (2) magnet configuration - the shape of the plates are considered in detail including standard turns, turns located at beam ports, diagnostic and closure flanges; (3) ripple computation - this section describes the codes by which ripple is computed; (4) field diffusion and nuclear heating - the effect of magnetic field diffusion on heating is considered along with neutron heating. Current, field and temperature profiles are computed; (5) finite element analysis - the two and three dimensional finite element codes are described and the results discussed in detail; (6) structures engineering - this considers the calculation of critical stresses due to toroidal and overturning forces and discusses the method of constraint of these forces. The Materials Testing Program is also discussed; (7) fabrication - the methods available for the manufacture of the constituent parts of the Bitter plates, the method of assembly and remote maintenance are summarized

Maintainability features of the compact ignition tokamak

International Nuclear Information System (INIS)

Spampinato, P.T.; Bushnell, C.W.

1986-01-01

The Compact Ignition Tokamak (CIT) is a deuterium-tritium (D-T) device envisaged to be the next experimental reactor in the US Fusion Program. The reactor will initially operate in a nonactivated hydrogen phase for approximately two years. This will permit verification of the integrity of the total system and allow hands-on repair to equipment which has experienced shakedown and early operation failures. Once D-T operations commence, reactor maintenance will require remote handling techniques. An evaluation has been completed to determine what maintenance operations must be performed on the CIT. A maintenance philosophy has been developed which is based upon the use of manipulator systems and robotics in the test cell. Replacement of life-limited equipment will be accomplished using a modular design approach for components, with simple remotely operable interfaces. Examples of operations to be done remotely include: (1) replacing of rf antennae and Faraday shields, (2) uncoupling diagnostic and fueling penetrations, (3) removing of all port covers, and (4) replacing first wall armor tiles, optical mirrors, and vacuum windows

National Ignition Facility TestController for automated and manual testing

Energy Technology Data Exchange (ETDEWEB)

Zielinski, Jason, E-mail: fishler2@llnl.gov [Lawrence Livermore National Laboratory, Livermore, CA 94551 (United States)

2012-12-15

The Controls and Information Systems (CIS) organization for the National Ignition Facility (NIF) has developed controls, configuration and analysis software applications that combine for several million lines of code. The team delivers updates throughout the year, from major releases containing hundreds of changes to patch releases containing a small number of focused updates. To ensure the quality of each delivery, manual and automated tests are performed using the NIF TestController test infrastructure. The TestController system provides test inventory management, test planning, automated and manual test execution, release testing summaries and results search, all through a web browser interface. As part of the three-stage software testing strategy, the NIF TestController system helps plan, evaluate and track the readiness of each release to the NIF production environment. After several years of use in testing NIF software applications, the TestController's manual testing features have been leveraged for verifying the installation and operation of NIF Target Diagnostic hardware. The TestController recorded its first test results in 2004. Today, the system has recorded the execution of more than 160,000 tests and continues to play a central role in ensuring that NIF hardware and software meet the requirements of a high reliability facility. This paper describes the TestController system and discusses its use in assuring the quality of software delivered to the NIF.

National Ignition Facility TestController for automated and manual testing

International Nuclear Information System (INIS)

Zielinski, Jason

2012-01-01

The Controls and Information Systems (CIS) organization for the National Ignition Facility (NIF) has developed controls, configuration and analysis software applications that combine for several million lines of code. The team delivers updates throughout the year, from major releases containing hundreds of changes to patch releases containing a small number of focused updates. To ensure the quality of each delivery, manual and automated tests are performed using the NIF TestController test infrastructure. The TestController system provides test inventory management, test planning, automated and manual test execution, release testing summaries and results search, all through a web browser interface. As part of the three-stage software testing strategy, the NIF TestController system helps plan, evaluate and track the readiness of each release to the NIF production environment. After several years of use in testing NIF software applications, the TestController's manual testing features have been leveraged for verifying the installation and operation of NIF Target Diagnostic hardware. The TestController recorded its first test results in 2004. Today, the system has recorded the execution of more than 160,000 tests and continues to play a central role in ensuring that NIF hardware and software meet the requirements of a high reliability facility. This paper describes the TestController system and discusses its use in assuring the quality of software delivered to the NIF.

Broad-Application Test Reactor

International Nuclear Information System (INIS)

Motloch, C.G.

1992-05-01

This report is about a new, safe, and operationally efficient DOE reactor of nuclear research and testing proposed for the early to mid- 21st Century. Dubbed the Broad-Application Test Reactor (BATR), the proposed facility incorporates a multiple-application, multiple-mission design to support DOE programs such as naval reactors and space power and propulsion, as well as research in medical, science, isotope, and electronics arenas. DOE research reactors are aging, and implementing major replacement projects requires long lead times. Primary design drivers include safety, low risk, minimum operation cost, mission flexibility, waste minimization, and long life. Scientists and engineers at the Idaho National Engineering Laboratory are evaluating possible fuel forms, structural materials, reactor geometries, coolants, and moderators

Influence of test configuration on the combustion characteristics of polymers as ignition sources

Science.gov (United States)

Julien, Howard L.

1993-01-01

The experimental evaluation of polymers as ignition sources for metals was accomplished at the NASA White Sands Test Facility (WSTF) using a standard promoted combustion test. These tests involve the transient burning of materials in high-pressure oxygen environments. They have provided data from which design decisions can be made; data include video recordings of ignition and non-ignition for specific combinations of metals and polymers. Other tests provide the measured compositions of combustion products for polymers at select burn times and an empirical basis for estimating burn rates. With the current test configuration, the detailed analysis of test results requires modeling a three-dimensional, transient convection process involving fluid motion, thermal conduction and convection, the diffusion of chemical species, and the erosion of sample surface. At the high pressure extremes, it even requires the analysis of turbulent, transient convection where the physics of the problem are not well known and the computation requirements are not practical at this time. An alternative test configuration that can be analyzed with a relatively-simple convection model was developed during the summer period. The principal change constitutes replacing a large-diameter polymer disk at the end of the metal test rod with coaxial polymer cylinders that have a diameter nearer to that of the metal rod. The experimental objective is to assess the importance of test geometries on the promotion of metal ignition by testing with different lengths of the polymer and, with an extended effort, to analyze the surface combustion in the redesigned promoted combustion tests through analytical modeling of the process. The analysis shall use the results of cone-calorimeter tests of the polymer material to model primary chemical reactions and, with proper design of the promoted combustion test, modeling of the convection process could be conveniently limited to a quasi-steady boundary layer

The insulation irradiation test program for the Compact Ignition Tokamak

International Nuclear Information System (INIS)

McManamy, T.J.; Kanemoto, G.; Snook, P.

1990-01-01

The electrical insulation for the toroidal field coils of the Compact Ignition Tokamak (CIT) is expected to be exposed to radiation doses on the order of 10 10 rad with ∼90% of the dose from neutrons. The coils are cooled to liquid nitrogen temperature and then heated during the pulse to a peak temperature >300 K. In a program to evaluate the effects of radiation exposure on the insulators, three types of boron-free insulation were irradiated at room temperature in the Advanced Technology Reactor (ATR) and tested at the Idaho National Engineering Laboratory. The materials were Spaulrad-S, Shikishima PG5-1, and Shikishima PG3-1. The first two use a bismaleimide resin and the third an aromatic amine hardened epoxy. Spaulrad-S is a two-dimensional (2-D) weave of S-glass, while the others are 3-D weaves of T-glass. Flexure and shear/compression samples were irradiated to approximately 5 x 10 9 rad and 3 x 10 10 rad with 35 to 40% of the total dose from neutrons. The shear/compression samples were tested in pairs by applying an average compression of 345 MPa and then a shear load. After static tests were completed, fatigue testing was done by cycling the shear load for up to 30,000 cycles with a constant compression. The static shear strength of the samples that did not fail was then determined. Generally, shear strengths on the order of 120 MPa were measured. The behavior of the flexure and shear/compression samples was significantly different; large reductions in the flexure strength were observed, while the shear strength stayed the same or increased slightly. The 3-D weave material demonstrated higher strength and significantly less radiation damage than the 2-D material in flexure but performed nearly identically when tested with combined shear and compression. The epoxy system was much more sensitive to fatigue damage than the bismaleimide materials. 9 refs., 5 figs

Status and update of the National Ignition Facility radiation effects testing program

International Nuclear Information System (INIS)

Davis, J F; Serduke, F J; Wuest, C R.

1998-01-01

We are progressing in our efforts to make the National Ignition Facility (NIF) available to the nation as a radiation effects simulator to support the Services needs for nuclear hardness and survivability testing and validation. Details of our program were summarized in a paper presented at the 1998 HEART Conference [1]. This paper describes recent activities and updates plans for NIF radiation effects testing. research. Radiation Effects Testing

Mirror Advanced Reactor Study (MARS) final report summary

International Nuclear Information System (INIS)

Henning, C.D.; Logan, B.G.; Carlson, G.A.

1983-01-01

The Mirror Advanced Reactor Study (MARS) has resulted in an overview of a first-generation tandem mirror reactor. The central cell fusion plasma is self-sustained by alpha heating (ignition), while electron-cyclotron resonance heating and negative ion beams maintain the electrostatic confining potentials in the end plugs. Plug injection power is reduced by the use of high-field choke coils and thermal barriers, concepts to be tested in the Tandem Mirror Experiment-Upgrade (TMX-U) and Mirror Fusion Test Facility (MFTF-B) at Lawrence Livermore National Laboratory

Ignition delay times of Gasoline Distillation Cuts measured with Ignition Quality Tester

KAUST Repository

Naser, Nimal

2017-04-21

Tailoring fuel properties to maximize the efficiency of internal combustion engines is a way towards achieving cleaner combustion systems. In this work, the ignition properties of various gasoline fuel distillation cuts are analyzed to better understand fuel properties of the full boiling range fuel. An advanced distillation column (ADC) provides a more realistic representation of volatility characteristics, which can be modeled using equilibrium thermodynamic methods. The temperature reported is that of the liquid, as opposed to the vapor temperature in conventional ASTM D86 distillation standard. Various FACE (fuels for advanced combustion engines) gasolines were distilled and various cuts were obtained. The separated fractions were then tested in an ignition quality tester (IQT) to see the effect of chemical composition of different fractions on their ignition delay time. Fuels with lower aromatic content showed decreasing ignition delay time with increasing boiling point (i.e., molecular weight). However, fuels with higher aromatic content showed an initial decrease in ignition delay time with increasing boiling point, followed by drastic increase in ignition delay time due to fractions containing aromatics. This study also provides an understanding on contribution of different fractions to the ignition delay time of the fuel, which provides insights into fuel stratification utilized in gasoline compression ignition (GCI) engines to tailor heat release rates.

Ground test facility for nuclear testing of space reactor subsystems

International Nuclear Information System (INIS)

Quapp, W.J.; Watts, K.D.

1985-01-01

Two major reactor facilities at the INEL have been identified as easily adaptable for supporting the nuclear testing of the SP-100 reactor subsystem. They are the Engineering Test Reactor (ETR) and the Loss of Fluid Test Reactor (LOFT). In addition, there are machine shops, analytical laboratories, hot cells, and the supporting services (fire protection, safety, security, medical, waste management, etc.) necessary to conducting a nuclear test program. This paper presents the conceptual approach for modifying these reactor facilities for the ground engineering test facility for the SP-100 nuclear subsystem. 4 figs

Simulator for materials testing reactors

International Nuclear Information System (INIS)

Takemoto, Noriyuki; Sugaya, Naoto; Ohtsuka, Kaoru; Hanakawa, Hiroki; Onuma, Yuichi; Hosokawa, Jinsaku; Hori, Naohiko; Kaminaga, Masanori; Tamura, Kazuo; Hotta, Kohji; Ishitsuka, Tatsuo

2013-06-01

A real-time simulator for both reactor and irradiation facilities of a materials testing reactor, “Simulator of Materials Testing Reactors”, was developed for understanding reactor behavior and operational training in order to utilize it for nuclear human resource development and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR (Japan Materials Testing Reactor), and it simulates operation, irradiation tests and various kinds of anticipated operational transients and accident conditions caused by the reactor and irradiation facilities. The development of the simulator was sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. This report summarizes the simulation components, hardware specification and operation procedure of the simulator. (author)

Improving the ignition quality of fuels

KAUST Repository

Sarathy, Mani

2017-06-08

Provided herein are compounds and methods of producing compounds for improving ignition quality and combustion efficiency of fuels, for example fossil fuels. In various aspects we generate highly oxygenated compounds from hydrocarbon feedstocks. The feedstock can be a branched alkane or n-alkane having a chain length greater than or equal to 6, a cycloalkane with a 5 or 6 membered ring structure, or a alkylated cycloalkane with 5 or more carbon atoms. The reactant can be fed in the gas- phase to a partial oxidation reactor (with or without a catalyst), and at a fixed temperature, mixture composition, and residence time. The reactant can be converted to a mixture of products including keto hydroperoxides, diketo hydroperoxides, keto dihydroperoxides, hydroperoxyl cyclic ethers, and alkenyl hydroperoxides. The compounds are inherently unstable and can quickly decompose to highly reactive radical species that can be used to improve the ignition quality of a fuel and advance ignition in an engine.

Improving the ignition quality of fuels

KAUST Repository

Sarathy, Mani; Wang, Zhandong; Shankar, Vijai Shankar Bhavani

2017-01-01

Provided herein are compounds and methods of producing compounds for improving ignition quality and combustion efficiency of fuels, for example fossil fuels. In various aspects we generate highly oxygenated compounds from hydrocarbon feedstocks. The feedstock can be a branched alkane or n-alkane having a chain length greater than or equal to 6, a cycloalkane with a 5 or 6 membered ring structure, or a alkylated cycloalkane with 5 or more carbon atoms. The reactant can be fed in the gas- phase to a partial oxidation reactor (with or without a catalyst), and at a fixed temperature, mixture composition, and residence time. The reactant can be converted to a mixture of products including keto hydroperoxides, diketo hydroperoxides, keto dihydroperoxides, hydroperoxyl cyclic ethers, and alkenyl hydroperoxides. The compounds are inherently unstable and can quickly decompose to highly reactive radical species that can be used to improve the ignition quality of a fuel and advance ignition in an engine.

An experimental and analytical investigation of glow plug performance in ignition and flame propagation through low concentrations of H2 in a steam/fog environment

International Nuclear Information System (INIS)

Davis, B.W.

1982-01-01

Thermal igniters proposed by the Tennessee Valley Authority for intentional ignition of hydrogen in nuclear reactor containments have been tested in mixtures of air, hydrogen, and steam. The igniters, conventional diesel engine glow plugs, were tested in a 10.6 ft 3 pressure vessel with dry hydrogen concentrations from 4% to 29%, and in steam fractions of up to 50%. Dry tests indicated complete combustion consistently occurred at H 2 fractions above 8% with no combustion for concentrations below 5%. Combustion tests in the presence of steam were conducted with hydrogen volume fractions of 8%, 10%, and 12%. Steam concentrations of up to 30% consistently resulted in ignition. Most of the 40% steam fraction tests indicated a pressure rise. Circulation of the mixture improved combustion in both the dry and the steam tests, most notably at low H 2 concentrations. An analysis of the high steam fraction test data yielded evidence of the presence of small, suspended, water droplets in the combustion mixture. The suppressive influence of condensation-generated fog on combustion is evaluated. Analysis of experimental results along with results derived from analytic models have provided consistent evidence of the strong influence of mass condensation rates and fog on experimentally observed ignition and flame propagation phenomena

Test reactor: basic to U.S. breeder reactor development

International Nuclear Information System (INIS)

Miller, B.J.; Harness, A.J.

1975-01-01

Long-range energy planning in the U. S. includes development of a national commercial breeder reactor program. U. S. development of the LMFBR is following a conservative sequence of extensive technology development through use of test reactors and demonstration plants prior to construction of commercial plants. Because materials and fuel technology development is considered the first vital step in this sequence, initial U. S. efforts have been directed to the design and construction of a unique test reactor. The Fast Flux Test Facility, FFTF, is a 400 MW(t) reactor with driver fuel locations, open test locations, and closed loops for higher risk experiments. The FFTF will provide a prototypic LMFBR core environment with sufficient instrumentation for detailed core environmental characterization and a testing capability substituted for breeder capability. The unique comprehensive fuel and materials testing capability of the FFTF will be key to achieving long-range objectives of increased power density, improved breeding gain and shorter doubling times. (auth)

Ion beam heating for fast ignition

International Nuclear Information System (INIS)

Gus'kov, S.Yu.; Limpouch, J.; Klimo, O.

2010-01-01

Complete text of publication follows. The characteristics features of the formation of the spatial distribution of the energy transferred to the plasma from a beam of ions with different initial energies, masses and charges under fast ignition conditions are determined. The motion of the Bragg peak is extended with respect to the spatial distribution of the temperature of the ion-beam-heated medium. The parameters of the ion beams are determined to initiate different regimes of fast ignition of thermonuclear fuel precompressed to a density of 300-500 g/cm 3 - the edge regime, in which the ignition region is formed at the outer boundary of the fuel, and the internal regime, in which the ignition region is formed in central parts of the fuel. The conclusion on the requirements for fast ignition by light and heavy ion beams is presented. It is shown that the edge heating with negative temperature gradient is described by a self-similar solution. Such a temperature distribution is the reason of the fact that the ignited beam energy at the edge heating is larger than the minimal ignition energy by factor 1.65. The temperature Bragg peak may be produced by ion beam heating in the reactor scale targets with pR-parameter larger than 3-4 g/cm 2 . In particular, for central ignition of the targets with pR-parameters in the range of 4-8 g/cm 2 the ion beam energy should be, respectively, from 5 to 7 times larger than the minimal ignition energy. The work by S.Ye. Gus'kov, D.V. Il'in, and V.E. Sherman was supported by the Ministry of Education and Science of the Russian Federation under the program 'Development of the Scientific Potential of High Education for 2009-2010' (project no. 2.1.1/1505) and the Russian Foundation for Basic Research (project no. 08-02-01394 a ). The work by J. Limpouch and O. Klimo was supported by the Czech Ministry of Education (project no. LC528, MSM6840770022).

Irradiation Facilities at the Advanced Test Reactor

International Nuclear Information System (INIS)

S. Blaine Grover

2005-01-01

The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC) (formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950s with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens

Rise-to-power test in High Temperature Engineering Test Reactor. Test progress and summary of test results up to 30 MW of reactor thermal power

International Nuclear Information System (INIS)

Nakagawa, Shigeaki; Fujimoto, Nozomu; Shimakawa, Satoshi

2002-08-01

The High Temperature Engineering Test Reactor (HTTR) is a graphite moderated and gas cooled reactor with the thermal power of 30 MW and the reactor outlet coolant temperature of 850degC/950degC. Rise-to-power test in the HTTR was performed from April 23rd to June 6th in 2000 as phase 1 test up to 10 MW in the rated operation mode, from January 29th to March 1st in 2001 as phase 2 test up to 20 MW in the rated operation mode and from April 14th to June 8th in 2001 as phase 3 test up to 20 MW in the high temperature test the mechanism of the reactor outlet coolant temperature becomes 850degC at 30 MW in the rated operation mode and 950degC in the high temperature test operation mode. Phase 4 rise-to-power test to achieve the thermal reactor power of 30 MW started on October 23rd in 2001. On December 7th in 2001 it was confirmed that the thermal reactor power and the reactor outlet coolant temperature reached to 30 MW and 850degC respectively in the single loaded operation mode in which only the primary pressurized water cooler is operating. Phase 4 test was performed until March 6th in 2002. JAERI (Japan Atomic Energy Research Institute) obtained the certificate of the pre-operation test from MEXT (Ministry of Education Culture Sports Science and Technology) after all the pre-operation tests by MEXT were passed successfully with the reactor transient test at an abnormal event as a final pre-operation test. From the test results of the rise-up-power test up to 30 MW in the rated operation mode, performance of the reactor and cooling system were confirmed, and it was also confirmed that an operation of reactor facility can be performed safely. Some problems to be solved were found through the tests. By solving them, the reactor operation with the reactor outlet coolant temperature of 950degC will be achievable. (author)

Liquid Oxygen Rotating Friction Ignition Testing of Aluminum and Titanium with Monel and Inconel for Rocket Engine Propulsion System Contamination Investigation

Science.gov (United States)

Peralta, S.; Rosales, Keisa R.; Stoltzfus, Joel M.

2009-01-01

Metallic contaminant was found in the liquid oxygen (LOX) pre-valve screen of the shuttle main engine propulsion system on two orbiter vehicles. To investigate the potential for an ignition, NASA Johnson Space Center White Sands Test Facility performed (modified) rotating friction ignition testing in LOX. This testing simulated a contaminant particle in the low-pressure oxygen turbo pump (LPOTP) and the high-pressure oxygen turbo pump (HPOTP) of the shuttle main propulsion system. Monel(R) K-500 and Inconel(R) 718 samples represented the LPOTP and HPOTP materials. Aluminum foil tape and titanium foil represented the contaminant particles. In both the Monel(R) and Inconel(R) material configurations, the aluminum foil tape samples did not ignite after 30 s of rubbing. In contrast, all of the titanium foil samples ignited regardless of the rubbing duration or material configuration. However, the titanium foil ignitions did not propagate to the Monel and Inconel materials.

Ignition and flame spread properties of wood, elaborated during a new test method based on convective heat flux

DEFF Research Database (Denmark)

Sørensen, Lars Schiøtt

Ignition and flame spread properties on selected types of wood and wall papers are elaborated. Tests are established in a new test setup in which the test specimen can be fixed in different angles due to a horizontal level. The heat exposing the test objects is arranged as a convective flux......, established from a Bunsen burners pilot flame. This principal is somewhat in contrast to the more typical radiation established fluxes. For instance, the ISO 9239 (DS 2000) test method is based on a gas fired radiant panel. And in the ISO 5657 standard, the ignition properties are investigated on test...

Irradiation facilitates at the advanced test reactor

International Nuclear Information System (INIS)

Grover, Blaine S.

2006-01-01

The Advanced Test Reactor (ATR) is the third generation and largest test reactor built in the Reactor Technology Complex (RTC - formerly known as the Test Reactor Area), located at the Idaho National Laboratory (INL), to study the effects of intense neutron and gamma radiation on reactor materials and fuels. The RTC was established in the early 1950's with the development of the Materials Testing Reactor (MTR), which operated until 1970. The second major reactor was the Engineering Test Reactor (ETR), which operated from 1957 to 1981, and finally the ATR, which began operation in 1967 and will continue operation well into the future. These reactors have produced a significant portion of the world's data on materials response to reactor environments. The wide range of experiment facilities in the ATR and the unique ability to vary the neutron flux in different areas of the core allow numerous experiment conditions to co-exist during the same reactor operating cycle. Simple experiments may involve a non-instrumented capsule containing test specimens with no real-time monitoring or control capabilities. More sophisticated testing facilities include inert gas temperature control systems and pressurized water loops that have continuous chemistry, pressure, temperature, and flow control as well as numerous test specimen monitoring capabilities. There are also apparatus that allow for the simulation of reactor transients on test specimens. The paper has the following contents: ATR description and capabilities; ATR operations, quality and safety requirements; Static capsule experiments; Lead experiments; Irradiation test vehicle; In-pile loop experiments; Gas test loop; Future testing; Support facilities at RTC; Conclusions. To summarize, the ATR has a long history in fuel and material irradiations, and will be fulfilling a critical role in the future fuel and material testing necessary to develop the next generation reactor systems and advanced fuel cycles. The

Correlations between power and test reactor data bases

International Nuclear Information System (INIS)

Guthrie, G.L.; Simonen, E.P.

1989-02-01

Differences between power reactor and test reactor data bases have been evaluated. Charpy shift data has been assembled from specimens irradiated in both high-flux test reactors and low-flux power reactors. Preliminary tests for the existence of a bias between test and power reactor data bases indicate a possible bias between the weld data bases. The bias is nonconservative for power predictive purposes, using test reactor data. The lesser shift for test reactor data compared to power reactor data is interpreted primarily in terms of greater point defect recombination for test reactor fluxes compared to power reactor fluxes. The possibility of greater thermal aging effects during lower damage rates is also discussed. 15 refs., 5 figs., 2 tabs

Ignition of deuterium based fuel cycles in a high beta system

International Nuclear Information System (INIS)

Hirano, K.

1987-01-01

A steady state self-consistent plasma modeling applied to a system having close to unity, such as FRC or like, is found to be quite effective in solving the problems independently of any anomalous process and proves the existence of ignited state of deuterium based fuel cycles. The temperature ranges that the plasma falls into ignited state are obtained as a function of relative feeding rates of tritium and 3 He to deuterium's. We find pure DD cycle will not ignite so that 3 He or/and tritium must be added as catalyzer to achieve ignition. Standing on the points to construct a cleaner system yielding smaller amount of 14 MeV neutrons and to burn the fuel in steady state for long periods of time, we have confirmed superiority of the complex composed of the master reactor of 3 He-Cat.D cycle (catalyzed DD cycle reinjecting only fusion produced 3 He) and the satellite reactor of 3 He enriched D 3 He cycle. In case storage of tritium for 3 He by β - decay is turned out not to be allowed environmentally, we may utilize conventional catalyzed DD cycle although 14 MeV neutron yields will be increased by 35 % over the complex. It is demonstrated that advanced fuel cycle reactors can be very simple in constructions and compact in size such that the field strength and the plasma volume of the order of JT-60's may be enough for 1000 MW power plant. (author)

TIBER II: an upgraded tokamak igntion/burn experimental reactor

International Nuclear Information System (INIS)

Henning, C.D.; Logan, B.G.; Perkins, L.J.

1986-01-01

We are disIgning a minimum-size Tokamak ignition/Burn Reactor (TIBER II). This design incorporates physics requirements, neutron wall loading and fluence parameters that will make it compatible with a nuclear testing mission. Reactor relevant physics will be tested by using current drive and steady-state operation. Although the design accommodates several current drive options, including neutral beams, the base case uses a combination of lower hybrid and electron-cyclotron radio frequency power. Minimum neutron shielding, compact structures, high magnet-current densities, and remotely maintainable vacuum seals, all contribute to the compact size

Building a resistance to ignition testing device for sunglasses and analysing data: a continuing study for sunglasses standards.

Science.gov (United States)

Magri, Renan; Masili, Mauro; Duarte, Fernanda Oliveira; Ventura, Liliane

2017-09-21

Sunglasses popularity skyrocketed since its advent. The ongoing trend led to the creation of standards to protect consumers from injuries and secondary hazards due to spectacles use. In Brazil, the corresponding standard is NBR ISO 12312-1:2015 and since there is no mandatory testing, evaluating sunglasses performance provides an insight into compliance with the standard. In a continuing revision of sunglasses standards requirements, resistance to ignition is one of the concerns, since sunglasses should be protected from burning into flames at a pre-determined temperature, which may protect user of getting their sunglasses into flames if some, cigarette sparks reaches the spectacles, as an example. This paper describes the building of a resistance to ignition system and the results of 410 samples that have been tested accordingly to ISO 12312-1. The procedure is in accordance with the resistance to ignition test. It consists of heating a steel rod to 650 °C and pressing it against the sample surface for 5 s, with a force equivalent to the rod weight. For carrying out the assessments, we have build resistance to ignition testing system and assured the testing requirements of the standard. The apparatus has an electrical furnace with a temperature acquisition circuit and electronic control that maintains the temperature of the steel rod at 650 °C. A linear actuator was designed for the project to drive the steel rod vertically and pressing it against the sunglasses samples. The control system is composed by a Freescale development board FRDM-KL25Z with an ARM Cortex-M0 embedded. We have also provided a LabView PC interface for acquiring, displaying, and storing data as well as added a physical control panel to the equipment for performing the evaluations. We assessed 410 sunglasses frames at the built apparatus, where the 410 lenses came out to be in accordance with the guidelines provided by the ignition to resistance test. Out of the 410 tested frames, 50

Startup testing of Romania dual-core test reactor

International Nuclear Information System (INIS)

Whittemore, W.L.

1980-01-01

Late in 1979 both the Annular Core Pulsed Reactor (ACPR) and the 14-MW steady-state reactor (SSR) were loaded to critical. The fuel loading in both was then carried to completion and low-power testing was conducted. Early in 1980 both reactors successfully underwent high-power testing. The ACPR was operated for several hours at 500 kW and underwent pulse tests culminating in pulses with reactivity insertions of $4.60, peak power levels of about 20,000 MW, energy releases of 100 MW-sec, and peak measured fuel temperatures of 830 deg. C. The SSR was operated in several modes, both with natural convection and forced cooling with one or more pumps. The reactor successfully completed a 120-hr full-power test. Subsequent fuel element inspections confirmed that the fuel has performed without fuel damage or distortion. (author)

Low power arcjet thruster pulse ignition

Science.gov (United States)

Sarmiento, Charles J.; Gruber, Robert P.

1987-01-01

An investigation of the pulse ignition characteristics of a 1 kW class arcjet using an inductive energy storage pulse generator with a pulse width modulated power converter identified several thruster and pulse generator parameters that influence breakdown voltage including pulse generator rate of voltage rise. This work was conducted with an arcjet tested on hydrogen-nitrogen gas mixtures to simulate fully decomposed hydrazine. Over all ranges of thruster and pulser parameters investigated, the mean breakdown voltages varied from 1.4 to 2.7 kV. Ignition tests at elevated thruster temperatures under certain conditions revealed occasional breakdowns to thruster voltages higher than the power converter output voltage. These post breakdown discharges sometimes failed to transition to the lower voltage arc discharge mode and the thruster would not ignite. Under the same conditions, a transition to the arc mode would occur for a subsequent pulse and the thruster would ignite. An automated 11 600 cycle starting and transition to steady state test demonstrated ignition on the first pulse and required application of a second pulse only two times to initiate breakdown.

Study on the RF power necessary to ignite plasma for the ICP test facility at HUST

Energy Technology Data Exchange (ETDEWEB)

Yue, Haikun [School of Electronic Information and Communications, Huazhong University of Science and Technology, Wuhan (China); State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan (China); Li, Dong; Wang, Chenre; Li, Xiaofei; Chen, Dezhi; Liu, Kaifeng; Zhou, Chi; Pan, Ruimin [State Key Laboratory of Advanced Electromagnetic Engineering and Technology, Huazhong University of Science and Technology, Wuhan (China)

2015-10-15

An Radio-Frequency (RF) Inductively Coupled Plasma (ICP) ion source test facility has been successfully developed at Huazhong University of Science and Technology (HUST). As part of a study on hydrogen plasma, the influence of three main operation parameters on the RF power necessary to ignite plasma was investigated. At 6 Pa, the RF power necessary to ignite plasma influenced little by the filament heating current from 5 A to 9 A. The RF power necessary to ignite plasma increased rapidly with the operation pressure decreasing from 8 Pa to 4 Pa. The RF power necessary to ignite plasma decreased with the number of coil turns from 6 to 10. During the experiments, plasma was produced with the electron density of the order of 10{sup 16}m{sup -3} and the electron temperature of around 4 eV. (copyright 2015 WILEY-VCH Verlag GmbH and Co. KGaA, Weinheim)

FASTER Test Reactor Preconceptual Design Report

Energy Technology Data Exchange (ETDEWEB)

Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. J. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, S. [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-03-31

The FASTER test reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

IGNITION IMPROVEMENT OF LEAN NATURAL GAS MIXTURES

Energy Technology Data Exchange (ETDEWEB)

Jason M. Keith

2005-02-01

This report describes work performed during a thirty month project which involves the production of dimethyl ether (DME) on-site for use as an ignition-improving additive in a compression-ignition natural gas engine. A single cylinder spark ignition engine was converted to compression ignition operation. The engine was then fully instrumented with a cylinder pressure transducer, crank shaft position sensor, airflow meter, natural gas mass flow sensor, and an exhaust temperature sensor. Finally, the engine was interfaced with a control system for pilot injection of DME. The engine testing is currently in progress. In addition, a one-pass process to form DME from natural gas was simulated with chemical processing software. Natural gas is reformed to synthesis gas (a mixture of hydrogen and carbon monoxide), converted into methanol, and finally to DME in three steps. Of additional benefit to the internal combustion engine, the offgas from the pilot process can be mixed with the main natural gas charge and is expected to improve engine performance. Furthermore, a one-pass pilot facility was constructed to produce 3.7 liters/hour (0.98 gallons/hour) DME from methanol in order to characterize the effluent DME solution and determine suitability for engine use. Successful production of DME led to an economic estimate of completing a full natural gas-to-DME pilot process. Additional experimental work in constructing a synthesis gas to methanol reactor is in progress. The overall recommendation from this work is that natural gas to DME is not a suitable pathway to improved natural gas engine performance. The major reasons are difficulties in handling DME for pilot injection and the large capital costs associated with DME production from natural gas.

The combined use of test reactor experiments and power reactor tests for the development of PCI-resistant fuel

International Nuclear Information System (INIS)

Junkrans, S.; Vesterlund, G.; Vaernild, O.

1980-01-01

The theme of this paper is that for development of PCI-resistant fuel acceptable from the commercial and licensing aspects, extensive and time-consuming work is needed both in a test reactor and in power reactors. The test reactor is necessary for ramp testing to power levels not allowed in power reactors and with the aim of generating fuel failures. It is also used for other special irradiation experiments. The access to power reactors is necessary to generate information on performance in a real LWR core and to incubate at a reasonable cost the large amount of rods required for test reactor ramping. Selected results from the ASEA-ATOM work are used to support these conclusions. (author)

Ignition characteristics of 2-methyltetrahydrofuran: An experimental and kinetic study

KAUST Repository

Tripathi, Rupali; Lee, Changyoul; Fernandes, Ravi X.; Olivier, Herbert; Curran, Henry J.; Sarathy, Mani; Pitsch, Heinz

2016-01-01

The present paper elucidates oxidation behavior of 2-methyltetrahydrofuran (2-MTHF), a novel second-generation biofuel. New experimental data sets for 2-MTHF including ignition delay time measurements in two different combustion reactors, i.e. rapid

International Thermonuclear Experimental Reactor

International Nuclear Information System (INIS)

Blevins, J.D.; Stasko, R.R.

1989-09-01

An international design team comprised of members from Canada, Europe, Japan, the Soviet Union, and the United States of America, are designing an experimental fusion test reactor. The engineering and testing objectives of this International Thermonuclear Experimental Reactor (ITER) are to validate the design and to demonstrate controlled ignition, extended burn of a deuterium and tritium plasma, and achieve steady state using technology expected to be available by 1990. The concept maximizes flexibility while allowing for a variety of plasma configurations and operating scenarios. During physics phase operation, the machine produces a 22 MA plasma current. In the technology phase, the machine can be reconfigured with a thicker shield and a breeding blanket to operate with an 18 MA plasma current at a major radius of 5.5 meters. Canada's involvement in the areas of safety, facility design, reactor configuration and maintenance builds on our internationally recognized design and operational expertise in developing tritium processes and CANDU related technologies

Laser ignition of a multi-injector LOX/methane combustor

Science.gov (United States)

Börner, Michael; Manfletti, Chiara; Hardi, Justin; Suslov, Dmitry; Kroupa, Gerhard; Oschwald, Michael

2018-06-01

This paper reports the results of a test campaign of a laser-ignited combustion chamber with 15 shear coaxial injectors for the propellant combination LOX/methane. 259 ignition tests were performed for sea-level conditions. The igniter based on a monolithic ceramic laser system was directly attached to the combustion chamber and delivered 20 pulses with individual pulse energies of {33.2 ± 0.8 mJ } at 1064 nm wavelength and 2.3 ns FWHM pulse length. The applicability, reliability, and reusability of this ignition technology are demonstrated and the associated challenges during the start-up process induced by the oxygen two-phase flow are formulated. The ignition quality and pressure dynamics are evaluated using 14 dynamic pressure sensors distributed both azimuthally and axially along the combustion chamber wall. The influence of test sequencing on the ignition process is briefly discussed and the relevance of the injection timing of the propellants for the ignition process is described. The flame anchoring and stabilization process, as monitored using an optical probe system close to the injector faceplate connected to photomultiplier elements, is presented. For some of the ignition tests, non-uniform anchoring was detected with no influence onto the anchoring at steady-state conditions. The non-uniform anchoring can be explained by the inhomogeneous, transient injection of the two-phase flow of oxygen across the faceplate. This characteristic is verified by liquid nitrogen cold flow tests that were recorded by high-speed imaging. We conclude that by adapting the ignition sequence, laser ignition by optical breakdown of the propellants within the shear layer of a coaxial shear injector is a reliable ignition technology for LOX/methane combustors without significant over-pressure levels.

Laser ignition of a multi-injector LOX/methane combustor

Science.gov (United States)

Börner, Michael; Manfletti, Chiara; Hardi, Justin; Suslov, Dmitry; Kroupa, Gerhard; Oschwald, Michael

2018-02-01

This paper reports the results of a test campaign of a laser-ignited combustion chamber with 15 shear coaxial injectors for the propellant combination LOX/methane. 259 ignition tests were performed for sea-level conditions. The igniter based on a monolithic ceramic laser system was directly attached to the combustion chamber and delivered 20 pulses with individual pulse energies of {33.2 ± 0.8 mJ } at 1064 nm wavelength and 2.3 ns FWHM pulse length. The applicability, reliability, and reusability of this ignition technology are demonstrated and the associated challenges during the start-up process induced by the oxygen two-phase flow are formulated. The ignition quality and pressure dynamics are evaluated using 14 dynamic pressure sensors distributed both azimuthally and axially along the combustion chamber wall. The influence of test sequencing on the ignition process is briefly discussed and the relevance of the injection timing of the propellants for the ignition process is described. The flame anchoring and stabilization process, as monitored using an optical probe system close to the injector faceplate connected to photomultiplier elements, is presented. For some of the ignition tests, non-uniform anchoring was detected with no influence onto the anchoring at steady-state conditions. The non-uniform anchoring can be explained by the inhomogeneous, transient injection of the two-phase flow of oxygen across the faceplate. This characteristic is verified by liquid nitrogen cold flow tests that were recorded by high-speed imaging. We conclude that by adapting the ignition sequence, laser ignition by optical breakdown of the propellants within the shear layer of a coaxial shear injector is a reliable ignition technology for LOX/methane combustors without significant over-pressure levels.

Dynamic Response Testing in an Electrically Heated Reactor Test Facility

Science.gov (United States)

Bragg-Sitton, Shannon M.; Morton, T. J.

2006-01-01

Non-nuclear testing can be a valuable tool in development of a space nuclear power or propulsion system. In a non-nuclear test bed, electric heaters are used to simulate the heat from nuclear fuel. Standard testing allows one to fully assess thermal, heat transfer, and stress related attributes of a given system, but fails to demonstrate the dynamic response that would be present in an integrated, fueled reactor system. The integration of thermal hydraulic hardware tests with simulated neutronic response provides a bridge between electrically heated testing and full nuclear testing. By implementing a neutronic response model to simulate the dynamic response that would be expected in a fueled reactor system, one can better understand system integration issues, characterize integrated system response times and response characteristics, and assess potential design improvements at a relatively small fiscal investment. Initial system dynamic response testing was demonstrated on the integrated SAFE-100a heat pipe cooled, electrically heated reactor and heat exchanger hardware, utilizing a one-group solution to the point kinetics equations to simulate the expected neutronic response of the system (Bragg-Sitton, 2005). The current paper applies the same testing methodology to a direct drive gas cooled reactor system, demonstrating the applicability of the testing methodology to any reactor type and demonstrating the variation in system response characteristics in different reactor concepts. In each testing application, core power transients were controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. Although both system designs utilize a fast spectrum reactor, the method of cooling the reactor differs significantly, leading to a variable system response that can be demonstrated and assessed in a non-nuclear test facility.

CORONA DISCHARGE IGNITION FOR ADVANCED STATIONARY NATURAL GAS ENGINES

Energy Technology Data Exchange (ETDEWEB)

Dr. Paul D. Ronney

2003-09-12

An ignition source was constructed that is capable of producing a pulsed corona discharge for the purpose of igniting mixtures in a test chamber. This corona generator is adaptable for use as the ignition source for one cylinder on a test engine. The first tests were performed in a cylindrical shaped chamber to study the characteristics of the corona and analyze various electrode geometries. Next a test chamber was constructed that closely represented the dimensions of the combustion chamber of the test engine at USC. Combustion tests were performed in this chamber and various electrode diameters and geometries were tested. The data acquisition and control system hardware for the USC engine lab was updated with new equipment. New software was also developed to perform the engine control and data acquisition functions. Work is underway to design a corona electrode that will fit in the new test engine and be capable igniting the mixture in one cylinder at first and eventually in all four cylinders. A test engine was purchased for the project that has two spark plug ports per cylinder. With this configuration it will be possible to switch between corona ignition and conventional spark plug ignition without making any mechanical modifications.

Fusion ignition via a magnetically-assisted fast ignition approach

OpenAIRE

Wang, W. -M.; Gibbon, P.; Sheng, Z. -M.; Li, Y. T.; Zhang, J.

2016-01-01

Significant progress has been made towards laser-driven fusion ignition via different schemes, including direct and indirect central ignition, fast ignition, shock ignition, and impact ignition schemes. However, to reach ignition conditions, there are still various technical and physical challenges to be solved for all these schemes. Here, our multi-dimensional integrated simulation shows that the fast-ignition conditions could be achieved when two 2.8 petawatt heating laser pulses counter-pr...

Minimization of the external heating power by long fusion power rise-up time for self-ignition access in the helical reactor FFHR2m

International Nuclear Information System (INIS)

Mitarai, O.; Sagara, A.; Chikaraishi, H.; Imagawa, S.; Shishkin, A.A.; Motojima, O.

2006-10-01

Minimization of the external heating power to access self-ignition is advantageous to increase the reactor design flexibility and to reduce the capital and operating costs of the plasma heating device in a helical reactor. In this work we have discovered that a larger density limit leads to a smaller value of the required confinement enhancement factor, lower density limit margin reduces the external heating power, and over 300 s of the fusion power rise-up time makes it possible to reach a minimized heating power. While the fusion power rise-up time in a tokamak is limited by the OH transformer flux or the current drive capability, any fusion power rise-up time can be employed in a helical reactor for reducing the thermal stresses of the blanket and shields, because the confinement field is generated by the external helical coils. (author)

Remote ignitability analysis of high-level radioactive waste

International Nuclear Information System (INIS)

Lundholm, C.W.; Morgan, J.M.; Shurtliff, R.M.; Trejo, L.E.

1992-09-01

The Idaho Chemical Processing Plant (ICPP), was used to reprocess nuclear fuel from government owned reactors to recover the unused uranium-235. These processes generated highly radioactive liquid wastes which are stored in large underground tanks prior to being calcined into a granular solid. The Resource Conservation and Recovery Act (RCRA) and state/federal clean air statutes require waste characterization of these high level radioactive wastes for regulatory permitting and waste treatment purposes. The determination of the characteristic of ignitability is part of the required analyses prior to calcination and waste treatment. To perform this analysis in a radiologically safe manner, a remoted instrument was needed. The remote ignitability Method and Instrument will meet the 60 deg. C. requirement as prescribed for the ignitability in method 1020 of SW-846. The method for remote use will be equivalent to method 1020 of SW-846

Progress in catalytic ignition fabrication and modeling : fabrication part 2.

Science.gov (United States)

2012-06-01

The ignition temperature and heat generation from oxidation of methane on a platinum catalyst were : determined experimentally. A 127 micron diameter platinum coiled wire was placed crosswise in a : quartz tube of a plug flow reactor. A source meter ...

Effectiveness of thermal ignition devices in lean hydrogen-air-steam mixtures

International Nuclear Information System (INIS)

Tamm, H.; McFarlane, R.; Liu, D.D.S.

1985-03-01

Deliberate ignition of hydrogen at low concentrations in reactor containment systems is one method of controlling hydrogen during degraded core accidents. Since many postulated accident conditions have substantial amounts of steam present, experiments have been performed to determine the hydrogen-air-steam concentration regimes in which ignitors would be effective. In these experiments, both a GM AC 7G thermal flow plug and a Tayco Model 3442 ignitor have been used. These ignitors have been installed in PWR containments with ice condensers and in BWR Mark III containments. This report presents the results of these ignitor effectiveness experiments, and gives the ignition limits and the effect of steam on the ignitor surface temperatures required for ignition

Research reactors and materials testing

International Nuclear Information System (INIS)

Vidal, H.

1986-01-01

Research reactors can be classified in three main groups according to the moderator which is used. Their technical characteristics are given and the three most recent research and materials testing reactors are described: OSIRIS, ORPHEE and the high-flux reactor of Grenoble. The utilization of research reactors is reviewed in four fields of activity: training, fundamental or applied research and production (eg. radioisotopes) [fr

Ignition and fusion burn in fast ignition scheme

International Nuclear Information System (INIS)

Takabe, Hideaki

1998-01-01

The target physics of fast ignition is briefly reviewed by focusing on the ignition and fusion burn in the off-center ignition scheme. By the use of a two dimensional hydrodynamic code with an alpha heating process, the ignition condition is studied. It is shown that the ignition condition of the off-center ignition scheme coincides with that of the the central isochoric model. After the ignition, a nuclear burning wave is seen to burn the cold main fuel with a velocity of 2 - 3 x 10 8 cm/s. The spark energy required for the off-center ignition is 2 - 3 kJ or 10 - 15 kJ for the core density of 400 g/cm 3 or 200 g/cm 3 , respectively. It is demonstrated that a core gain of more than 2,000 is possible for a core energy of 100 kJ with a hot spark energy of 13 kJ. The requirement for the ignition region's heating time is also discussed by modeling a heating source in the 2-D code. (author)

Technical report on operating experience with boiling water reactor offgas systems

International Nuclear Information System (INIS)

Lo, R.; Barrett, L.; Grimes, B.; Eisenhut, D.

1978-03-01

Over 100 reactor years of Boiling Water Reactor (BWR) operating experience have been accumulated since the first commercial operation of BWRs. A number of incidents have occurred involving the ''offgas'' of these Boiling Water Reactors. This report describes the generation and processing of ''offgas'' in Boiling Water Reactors, the safety considerations regarding systems processing the ''offgas'', operating experience involving ignitions or explosions of ''offgas'' and possible measures to reduce the likelihood of future ignitions or explosions and to mitigate the consequences of such incidents should they occur

Studies into laser ignition of confined pyrotechnics

Energy Technology Data Exchange (ETDEWEB)

Ahmad, S.R.; Russell, D.A. [Centre for Applied Laser Spectroscopy, DASSR, Defence Academy, Cranfield University, Shrivenham, Swindon (United Kingdom)

2008-10-15

Ignition tests were carried out on three different pyrotechnics using laser energy from the multimode output from an Ar-Ion laser (av) at 500 nm and a near-IR diode laser pigtailed to a fibre optic cable and operating at 808 nm. The pyrotechnics investigated were: G20 black powder, SR44 and SR371C. The confined ignition tests were conducted in a specially designed ignition chamber. Pyrotechnics were ignited by a free space beam entering the chamber through an industrial sapphire window in the case of the Ar-ion laser. For the NIR diode laser, fibre was ducted through a block into direct contact with the pyrotechnic. The Ar-Ion laser was chosen as this was found to ignite all three pyrotechnics in the unconfined condition. It also allowed for a direct comparison of confined/unconfined results to be made. The threshold laser flux densities to initiate reproducible ignitions at this wavelength were found to be between {proportional_to}12.7 and {proportional_to}0.16 kW cm{sup -2}. Plotted on the ignition maps are the laser flux densities versus the start of ignition times for the three confined pyrotechnics. It was found from these maps that the times for confined ignition were substantially lower than those obtained for unconfined ignition under similar experimental conditions. For the NIR diode laser flux densities varied between {proportional_to}6.8 and {proportional_to}0.2 kW cm{sup -2}. The minimum ignition times for the NIR diode laser for SR371C ({proportional_to}11.2 ms) and G20 ({proportional_to}17.1 ms) were faster than those achieved by the use of the Ar-ion laser. However, the minimum ignition time was shorter ({proportional_to}11.7 ms) with the Ar-ion laser for SR44. (Abstract Copyright [2008], Wiley Periodicals, Inc.)

Simulation of hydrogen and hydrogen-assisted propane ignition in Pt catalyzed microchannel

Energy Technology Data Exchange (ETDEWEB)

Seshadri, Vikram; Kaisare, Niket S. [Department of Chemical Engineering, Indian Institute of Technology - Madras, Chennai 600 036 (India)

2010-11-15

This paper deals with self-ignition of catalytic microburners from ambient cold-start conditions. First, reaction kinetics for hydrogen combustion is validated with experimental results from the literature, followed by validation of a simplified pseudo-2D microburner model. The model is then used to study the self-ignition behavior of lean hydrogen/air mixtures in a Platinum-catalyzed microburner. Hydrogen combustion on Pt is a very fast reaction. During cold start ignition, hydrogen conversion reaches 100% within the first few seconds and the reactor dynamics are governed by the ''thermal inertia'' of the microburner wall structure. The self-ignition property of hydrogen can be used to provide the energy required for propane ignition. Two different modes of hydrogen-assisted propane ignition are considered: co-feed mode, where the microburner inlet consists of premixed hydrogen/propane/air mixtures; and sequential feed mode, where the inlet feed is switched from hydrogen/air to propane/air mixtures after the microburner reaches propane ignition temperature. We show that hydrogen-assisted ignition is equivalent to selectively preheating the inlet section of the microburner. The time to reach steady state is lower at higher equivalence ratio, lower wall thermal conductivity, and higher inlet velocity for both the ignition modes. The ignition times and propane emissions are compared. Although the sequential feed mode requires slightly higher amount of hydrogen, the propane emissions are at least an order of magnitude lower than the other ignition modes. (author)

The SPHINX reactor for engineering tests

International Nuclear Information System (INIS)

Adamov, E.O.; Artamkin, K.N.; Bovin, A.P.; Bulkin, Y.M.; Kartashev, E.F.; Korneev, A.A.; Stenbok, I.A.; Terekhov, A.S.; Khmel'Shehikov, V.V.; Cherkashov, Y.M.

1990-01-01

A research reactor known as SPHINX is under development in the USSR. The reactor will be used mainly to carry out tests on mock-up power reactor fuel assemblies under close-to-normal parameters in experimental loop channels installed in the core and reflector of the reactor, as well as to test samples of structural materials in ampoule and loop channels. The SPHINX reactor is a channel-type reactor with light-water coolant and moderator. Maximum achievable neutron flux density in the experimental channels (cell composition 50% Fe, 50% H 2 O) is 1.1 X 10 15 neutrons/cm 2 · s for fast neutrons (E > 0.1 MeV) and 1.7 X 10 15 for thermal neutrons at a reactor power of 200 MW. The design concepts used represent a further development of the technical features which have met with approval in the MR and MIR channel-type engineering test reactors currently in use in the USSR. The 'in-pond channel' construction makes the facility flexible and eases the carrying out of experimental work while keeping discharges of radioactivity into the environment to a low level. The reactor and all associated buildings and constructions conform to modern radiation safety and environmental protection requirements

TREAT [Transient Reactor Test Facility] reactor control rod scram system simulations and testing

International Nuclear Information System (INIS)

Solbrig, C.W.; Stevens, W.W.

1990-01-01

Air cylinders moving heavy components (100 to 300 lbs) at high speeds (above 300 in/sec) present a formidable end-cushion-shock problem. With no speed control, the moving components can reach over 600 in/sec if the air cylinder has a 5 ft stroke. This paper presents an overview of a successful upgrade modification to an existing reactor control rod drive design using a computer model to simulate the modified system performance for system design analysis. This design uses a high speed air cylinder to rapidly insert control rods (278 lb moved 5 ft in less than 300 msec) to scram an air-cooled test reactor. Included is information about the computer models developed to simulate high-speed air cylinder operation and a unique new speed control and end cushion design. A patent application is pending with the US Patent ampersand Trade Mark Office for this system (DOE case number S-68,622). The evolution of the design, from computer simulations thru operational testing in a test stand (simulating in-reactor operating conditions) to installation and use in the reactor, is also described. 6 figs

Neutronic design of a traveling wave reactor core

International Nuclear Information System (INIS)

Lopez S, R. C.; Francois L, J. L.

2010-10-01

The traveling wave reactor is an innovative kind of fast breeder reactor, capable of operate for decades without refueling and whose operation requires only a small amount of enriched fuel for the ignition. Also, one of its advantages is its versatility; it can be designed as small modules of about 100 M We or large scale units of 1000 M We. In this paper the behaviour of the traveling wave reactor core is studied in order to determine whether the traveling breeding/burning wave moves (as theoretically predicted) or not. To achieve this, we consider a two pieces cylinder, the first one, the ignition zone, containing highly enriched fuel and the second, the breeding zone, which is the larger, containing natural or depleted uranium or thorium. We consider that both zones are homogeneous mixtures of fuel, sodium as coolant and iron as structural material. We also include a reflector material outside the cylinder to reduce the neutron leakages. Simulations were run with MCNPX version 2.6 code. We observed that the wave does move as time passes as predicted by theory, and reactor remains supercritical in the time we have simulated (3000 days). Also, we found that thorium does not perform as well as uranium for breeding in this type of reactor. Further test with different reflectors are planned for both U-Pu and Th-U fuel cycles. (Author)

Test reactor risk assessment methodology

International Nuclear Information System (INIS)

Jennings, R.H.; Rawlins, J.K.; Stewart, M.E.

1976-04-01

A methodology has been developed for the identification of accident initiating events and the fault modeling of systems, including common mode identification, as these methods are applied in overall test reactor risk assessment. The methods are exemplified by a determination of risks to a loss of primary coolant flow in the Engineering Test Reactor

Test reactors in the world

International Nuclear Information System (INIS)

Corella, M.R.; Gomez Alonso, M.

1983-01-01

INFCE work on research reactor core conversion from HEU to LEU, attracted a raising interest on this type of nuclear reactors. In this context, the present work shows a compilation of worldwide research and test nuclear reactors, now in operation, under construction, or planned, as well as decommissioned reactors (tables A to F). Brief descriptions of these reactors are included in tables G to L. In table M a summary view of reactors with power level between 10 and 30 MWt is shown. Attention is focused on that power range, as it has been considered in very preliminar studies for a new research reactor. Almost all data have been obtained from current available bibliography. (author)

Conceptual studies of plasma engineering test facility

International Nuclear Information System (INIS)

Hiraoka, Toru; Tazima, Teruhiko; Sugihara, Masayoshi; Kasai, Masao; Shinya, Kichiro

1979-04-01

Conceptual studies have been made of a Plasma Engineering Test Facility, which is to be constructed following JT-60 prior to the experimental power reactor. The physical aim of this machine is to examine self-ignition conditions. This machine possesses all essential technologies for reactor plasma, i.e. superconducting magnet, remote maintenance, shielding, blanket test modules, tritium handling. Emphasis in the conceptual studies was on structural consistency of the machine and whether the machine would be constructed practically. (author)

Compact Ignition Tokamak conventional facilities optimization

International Nuclear Information System (INIS)

Commander, J.C.; Spang, N.W.

1987-01-01

A high-field ignition machine with liquid-nitrogen-cooled copper coils, designated the Compact Ignition Tokamak (CIT), is proposed for the next phase of the United States magnetically confined fusion program. A team of national laboratory, university, and industrial participants completed the conceptual design for the CIT machine, support systems and conventional facilities. Following conceptual design, optimization studies were conducted with the goal of improving machine performance, support systems design, and conventional facilities configuration. This paper deals primarily with the conceptual design configuration of the CIT conventional facilities, the changes that evolved during optimization studies, and the revised changes resulting from functional and operational requirements (F and ORs). The CIT conventional facilities conceptual design is based on two premises: (1) satisfaction of the F and ORs developed in the CIT building and utilities requirements document, and (2) the assumption that the CIT project will be sited at the Princeton Plasma Physics Laboratory (PPPL) in order that maximum utilization can be made of existing Tokamak Fusion Test Reactor (TFTR) buildings and utilities. The optimization studies required reevaluation of the F and ORs and a second look at TFTR buildings and utilities. Some of the high-cost-impact optimization studies are discussed, including the evaluation criteria for a change from the conceptual design baseline configuration. The revised conventional facilities configuration are described and the estimated cost impact is summarized

The belt-screw-pinch reactor and other high-beta systems

International Nuclear Information System (INIS)

Bustraan, M.; Klippel, H.Th.; Veringa, H.J.; Verschuur, K.A.

1981-01-01

In a screw-pinch reactor the expenditure for plasma implosion and compression can be reduced and the reacting volume and burn time can be enlarged. This is possible by pinch ignition of only a few percent of the fuel. Fusion energy then ignites injected fuel pellets and expands the plasma. The magnitude of the pulsed magnetic fields is such as to make the application of superconducting coils feasible. An economical reactor model is described. A comparison is made with tokamak and reversed field pinch reactor designs. (author)

Prototype moving-ring reactor

International Nuclear Information System (INIS)

Smith, A.C. Jr.; Ashworth, C.P.; Abreu, K.E.

1982-01-01

We have completed a design of the Prototype Moving-Ring Reactor. The fusion fuel is confined in current-carrying rings of magnetically-field-reversed plasma (Compact Toroids). The plasma rings, formed by a coaxial plasma gun, undergo adiabatic magnetic compression to ignition temperature while they are being injected into the reactor's burner section. The cylindrical burner chamber is divided into three burn stations. Separator coils and a slight axial guide field gradient are used to shuttle the ignited toroids rapidly from one burn station to the next, pausing for 1/3 of the total burn time at each station. D-T- 3 He ice pellets refuel the rings at a rate which maintains constant radiated power

Probability of ignition - a better approach than ignition margin

International Nuclear Information System (INIS)

Ho, S.K.; Perkins, L.J.

1989-01-01

The use of a figure of merit - the probability of ignition - is proposed for the characterization of the ignition performance of projected ignition tokamaks. Monte Carlo and analytic models have been developed to compute the uncertainty distribution function for ignition of a given tokamak design, in terms of the uncertainties inherent in the tokamak physics database. A sample analysis with this method indicates that the risks of not achieving ignition may be unacceptably high unless the accepted margins for ignition are increased. (author). Letter-to-the-editor. 12 refs, 2 figs, 2 tabs

Advanced test reactor testing experience-past, present and future

International Nuclear Information System (INIS)

Marshall, Frances M.

2006-01-01

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is one of the world's premier test reactors for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The physical configuration of the ATR, a 4-leaf clover shape, allows the reactor to be operated at different power levels in the corner 'lobes' to allow for different testing conditions for multiple simultaneous experiments. The combination of high flux (maximum thermal neutron fluxes of 1E15 neutrons per square centimeter per second and maximum fast [E>1.0 MeV] neutron fluxes of 5E14 neutrons per square centimeter per second) and large test volumes (up to 122 cm long and 12.7 cm diameter) provide unique testing opportunities. The current experiments in the ATR are for a variety of test sponsors - US government, foreign governments, private researchers, and commercial companies needing neutron irradiation services. There are three basic types of test configurations in the ATR. The simplest configuration is the sealed static capsule, which places the capsule in direct contact with the primary coolant. The next level of experiment complexity is an instrumented lead experiment, which allows for active control of experiment conditions during the irradiation. The most complex experiment is the pressurized water loop, in which the test sample can be subjected to the exact environment of a pressurized water reactor. For future research, some ATR modifications and enhancements are currently planned. This paper provides more details on some of the ATR capabilities, key design features, experiments, and future plans

A Comparative Study of Cycle Variability of Laser Plug Ignition vs Classical Spark Plug Ignition in Combustion Engines

Science.gov (United States)

Done, Bogdan

2017-10-01

Over the past 30 years numerous studies and laboratory experiments have researched the use of laser energy to ignite gas and fuel-air mixtures. The actual implementation of this laser application has still to be fully achieved in a commercial automotive application. Laser Plug Ignition as a replacement for Spark Plug Ignition in the internal combustion engines of automotive vehicles, offers several potential benefits such as extending lean burn capability, reducing the cyclic variability between combustion cycles and decreasing the total amount of ignition costs, and implicitly weight and energy requirements. The paper presents preliminary results of cycle variability study carried on a SI Engine equipped with laser Plug Ignition system. Versus classic ignition system, the use of the laser Plug Ignition system assures the reduction of the combustion process variability, reflected in the lower values of the coefficient of variability evaluated for indicated mean effective pressure, maximum pressure, maximum pressure angle and maximum pressure rise rate. The laser plug ignition system was mounted on an experimental spark ignition engine and tested at the regime of 90% load and 2800 rev/min, at dosage of λ=1.1. Compared to conventional spark plug, laser ignition assures the efficiency at lean dosage.

Effect of flow velocity and temperature on ignition characteristics in laser ignition of natural gas and air mixtures

Science.gov (United States)

Griffiths, J.; Riley, M. J. W.; Borman, A.; Dowding, C.; Kirk, A.; Bickerton, R.

2015-03-01

Laser induced spark ignition offers the potential for greater reliability and consistency in ignition of lean air/fuel mixtures. This increased reliability is essential for the application of gas turbines as primary or secondary reserve energy sources in smart grid systems, enabling the integration of renewable energy sources whose output is prone to fluctuation over time. This work details a study into the effect of flow velocity and temperature on minimum ignition energies in laser-induced spark ignition in an atmospheric combustion test rig, representative of a sub 15 MW industrial gas turbine (Siemens Industrial Turbomachinery Ltd., Lincoln, UK). Determination of minimum ignition energies required for a range of temperatures and flow velocities is essential for establishing an operating window in which laser-induced spark ignition can operate under realistic, engine-like start conditions. Ignition of a natural gas and air mixture at atmospheric pressure was conducted using a laser ignition system utilizing a Q-switched Nd:YAG laser source operating at 532 nm wavelength and 4 ns pulse length. Analysis of the influence of flow velocity and temperature on ignition characteristics is presented in terms of required photon flux density, a useful parameter to consider during the development laser ignition systems.

Reactor transients tests for SNR fuel elements in HFR reactor

International Nuclear Information System (INIS)

Plitz, H.

1989-01-01

In HFR reactor, fuel pins of LMFBR reactors are putted in irradiation specimen capsules cooled with sodium for reactor transients tests. These irradiation capsules are instrumented and the experiences realized until this day give results on: - Fuel pins subjected at a continual variation of power - melting fuel - axial differential elongation of fuel pins

Material test reactor fuel research at the BR2 reactor

Energy Technology Data Exchange (ETDEWEB)

Dyck, Steven Van; Koonen, Edgar; Berghe, Sven van den [Institute for Nuclear Materials Science, SCK-CEN, Boeretang, Mol (Belgium)

2012-03-15

The construction of new, high performance material test reactor or the conversion of such reactors' core from high enriched uranium (HEU) to low enriched uranium (LEU) based fuel requires several fuel qualification steps. For the conversion of high performance reactors, high density dispersion or monolithic fuel types are being developed. The Uranium-Molybdenum fuel system has been selected as reference system for the qualification of LEU fuels. For reactors with lower performance characteristics, or as medium enriched fuel for high performance reactors, uranium silicide dispersion fuel is applied. However, on the longer term, the U-Mo based fuel types may offer a more efficient fuel alternative and-or an easier back-end solution with respect to the silicide based fuels. At the BR2 reactor of the Belgian nuclear research center, SCK-CEN in Mol, several types of fuel testing opportunities are present to contribute to such qualification process. A generic validation test for a selected fuel system is the irradiation of flat plates with representative dimensions for a fuel element. By flexible positioning and core loading, bounding irradiation conditions for fuel elements can be performed in a standard device in the BR2. For fuel element designs with curved plates, the element fabrication method compatibility of the fuel type can be addressed by incorporating a set of prototype fuel plates in a mixed driver fuel element of the BR2 reactor. These generic types of tests are performed directly in the primary coolant flow conditions of the BR2 reactor. The experiment control and interpretation is supported by detailed neutronic and thermal-hydraulic modeling of the experiments. Finally, the BR2 reactor offers the flexibility for irradiation of full size prototype fuel elements, as 200mm diameter irradiation channels are available. These channels allow the accommodation of various types of prototype fuel elements, eventually using a dedicated cooling loop to provide the

Research reactors for power reactor fuel and materials testing - Studsvik's experience

International Nuclear Information System (INIS)

Grounes, M.

1998-01-01

Presently Studsvik's R2 test reactor is used for BWR and PWR fuel irradiations at constant power and under transient power conditions. Furthermore tests are performed with defective LWR fuel rods. Tests are also performed on different types of LWR cladding materials and structural materials including post-irradiation testing of materials irradiated at different temperatures and, in some cases, in different water chemistries and on fusion reactor materials. In the past, tests have also been performed on HTGR fuel and FBR fuel and materials under appropriate coolant, temperature and pressure conditions. Fuel tests under development include extremely fast power ramps simulating some reactivity initiated accidents and stored energy (enthalpy) measurements. Materials tests under development include different types of in-pile tests including tests in the INCA (In-Core Autoclave) facility .The present and future demands on the test reactor fuel in all these cases are discussed. (author)

Volume ignition of laser driven fusion pellets and double layer effects

International Nuclear Information System (INIS)

Cicchitelli, L.; Eliezer, S.; Goldsworthy, M.P.; Green, F.; Hora, H.; Ray, P.S.; Stening, R.J.; Szichman, H.

1988-01-01

The realization of an ideal volume compression of laser-irradiated fusion pellets opens the possibility for an alternative to spark ignition proposed for many years for inertial confinement fusion. A re-evaluation of the difficulties of the central spark ignition of laser driven pellets is given. The alternative volume compression theory, together with volume burn and volume ignition, have received less attention and are re-evaluated in view of the experimental verification generalized fusion gain formulas, and the variation of optimum temperatures derived at self-ignition. Reactor-level DT fusion with MJ-laser pulses and volume compression to 50 times the solid-state density are estimated. Dynamic electric fields and double layers at the surface and in the interior of plasmas result in new phenomena for the acceleration of thermal electrons to suprathermal electrons. Double layers also cause a surface tension which stabilizes against surface wave effects and Rayleigh-Taylor instabilities. (author)

A Preliminary Analysis of Reactor Performance Test (LOEP) for a Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Kim, Hyeonil; Park, Su-Ki [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of)

2015-10-15

The final phase of commissioning is reactor performance test, which is to prove the integrated performance and safety of the research reactor at full power with fuel loaded such as neutron power calibration, Control Absorber Rod/Second Shutdown Rod drop time, InC function test, Criticality, Rod worth, Core heat removal with natural mechanism, and so forth. The last test will be safety-related one to assure the result of the safety analysis of the research reactor is marginal enough to be sure about the nuclear safety by showing the reactor satisfies the acceptance criteria of the safety functions such as for reactivity control, maintenance of auxiliaries, reactor pool water inventory control, core heat removal, and confinement isolation. After all, the fuel integrity will be ensured by verifying there is no meaningful change in the radiation levels. To confirm the performance of safety equipment, loss of normal electric power (LOEP), possibly categorized as Anticipated Operational Occurrence (AOO), is selected as a key experiment to figure out how safe the research reactor is before turning over the research reactor to the owner. This paper presents a preliminary analysis of the reactor performance test (LOEP) for a research reactor. The results showed how different the transient between conservative estimate and best estimate will look. Preliminary analyses have shown all probable thermal-hydraulic transient behavior of importance as to opening of flap valve, minimum critical heat flux ratio, the change of flow direction, and important values of thermal-hydraulic parameters.

Real time simulator for material testing reactor

Energy Technology Data Exchange (ETDEWEB)

Takemoto, Noriyuki; Imaizumi, Tomomi; Izumo, Hironobu; Hori, Naohiko; Suzuki, Masahide [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan); Ishitsuka, Tatsuo; Tamura, Kazuo [ITOCHU Techno-Solutions Corp., Tokyo (Japan)

2012-03-15

Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)

Real time simulator for material testing reactor

International Nuclear Information System (INIS)

Takemoto, Noriyuki; Imaizumi, Tomomi; Izumo, Hironobu; Hori, Naohiko; Suzuki, Masahide; Ishitsuka, Tatsuo; Tamura, Kazuo

2012-01-01

Japan Atomic Energy Agency (JAEA) is now developing a real time simulator for a material testing reactor based on Japan Materials Testing Reactor (JMTR). The simulator treats reactor core system, primary and secondary cooling system, electricity system and irradiation facility systems. Possible simulations are normal reactor operation, unusual transient operation and accidental operation. The developed simulator also contains tool to revise/add facility in it for the future development. (author)

Investigation on the ignition, thermal acceleration and characteristic temperatures of coal char combustion

International Nuclear Information System (INIS)

Zhang, Bin; Fu, Peifang; Liu, Yang; Yue, Fang; Chen, Jing; Zhou, Huaichun; Zheng, Chuguang

2017-01-01

Highlights: • A new thermal model and measuring method for the ignition temperature are proposed. • Ignition occurs in a region but not a point with ambient conditions changing. • Ignition region is measured from the minimum to maximum ignition temperature. • T_i_g_,_m_a_x of coal char in TG-DSC is in line with the ignition temperature of EFR. - Abstract: Through using a new thermal analysis model and a method of coal/char combustion, the minimum ignition temperature and minimum ignition heat of three different ranks of pulverized coal char were measured by simultaneous Thermogravimetry and Differential Scanning Calorimetry (TG-DSC) experiments. The results show that the ignition of coal char occurs in the range between the minimum ignition temperature and the inflection-point temperature. The thermal acceleration and its gradient G_T increase with increasing heating rate and decrease with increasing coal char rank. The higher the G_T of the coal char, the more easily the ignition occurs and more rapidly the burning and burnout occur. The data show that the G_T of coal char of SLH lignite is 1.6 times more than that of coal char of ZCY bituminous and JWY anthracite in ignition zone, and 3.4 times in burning zone. The characteristic temperatures increase with increasing temperature of prepared char, heating rate and char rank. Moreover, the T_i_g_,_m_a_x calculated in DSC experiment is approximately in line with the ignition temperature obtained in the entrained flow reactor, which demonstrates the feasibility of the proposed theory.

Pulse heating and ignition for off-centre ignited targets

International Nuclear Information System (INIS)

Mahdy, A.I.; Takabe, H.; Mima, K.

1999-01-01

An off-centre ignition model has been used to study the ignition conditions for laser targets related to the fast ignition scheme. A 2-D hydrodynamic code has been used, including alpha particle heating. The main goal of the study is the possibility of obtaining a high gain ICF target with fast ignition. In order to determine the ignition conditions, samples with various compressed core densities having different spark density-radius product (i.e. areal density) values were selected. The study was carried out in the presence of an external heating source, with a constant heating rate. A dependence of the ignition conditions on the heating rate of the external pulse is demonstrated. For a given set of ignition conditions, our simulation showed that an 11 ps pulse with 17 kJ of injected energy into the spark area was required to achieve ignition for a compressed core with a density of 200 g/cm 3 and 0.5 g/cm 2 spark areal density. It is shown that the ignition conditions are highly dependent on the heating rate of the external pulse. (author)

Advanced Instrumentation for Transient Reactor Testing

Energy Technology Data Exchange (ETDEWEB)

Corradini, Michael L.; Anderson, Mark; Imel, George; Blue, Tom; Roberts, Jeremy; Davis, Kurt

2018-01-31

Transient testing involves placing fuel or material into the core of specialized materials test reactors that are capable of simulating a range of design basis accidents, including reactivity insertion accidents, that require the reactor produce short bursts of intense highpower neutron flux and gamma radiation. Testing fuel behavior in a prototypic neutron environment under high-power, accident-simulation conditions is a key step in licensing nuclear fuels for use in existing and future nuclear power plants. Transient testing of nuclear fuels is needed to develop and prove the safety basis for advanced reactors and fuels. In addition, modern fuel development and design increasingly relies on modeling and simulation efforts that must be informed and validated using specially designed material performance separate effects studies. These studies will require experimental facilities that are able to support variable scale, highly instrumented tests providing data that have appropriate spatial and temporal resolution. Finally, there are efforts now underway to develop advanced light water reactor (LWR) fuels with enhanced performance and accident tolerance. These advanced reactor designs will also require new fuel types. These new fuels need to be tested in a controlled environment in order to learn how they respond to accident conditions. For these applications, transient reactor testing is needed to help design fuels with improved performance. In order to maximize the value of transient testing, there is a need for in-situ transient realtime imaging technology (e.g., the neutron detection and imaging system like the hodoscope) to see fuel motion during rapid transient excursions with a higher degree of spatial and temporal resolution and accuracy. There also exists a need for new small, compact local sensors and instrumentation that are capable of collecting data during transients (e.g., local displacements, temperatures, thermal conductivity, neutron flux, etc.).

Proposal of world network on material testing reactors

International Nuclear Information System (INIS)

Takemoto, Noriyuki; Izumo, Hironobu; Hori, Naohiko; Ishitsuka, Etsuo; Ishihara, Masahiro

2011-01-01

Establishment of an international cooperation system of worldwide testing reactor network (world network) is proposed in order to achieve efficient facility utilization and provide high quality irradiation data by role sharing of irradiation tests with materials testing reactors in the world. As for the first step, mutual understanding among materials testing reactors is thought to be necessary. From this point, an international symposium on materials testing reactors (ISMTR) was held to construct the world network from 2008, and a common understanding of world network has begun to be shared. (author)

Analytical solutions for prediction of the ignition time of wood particles based on a time and space integral method

International Nuclear Information System (INIS)

Haseli, Y.; Oijen, J.A. van; Goey, L.P.H. de

2012-01-01

Highlights: ► A simple model for prediction of the ignition time of a wood particle is presented. ► The formulation is given for both thermally thin and thermally thick particles. ► Transition from thermally thin to thick regime occurs at a critical particle size. ► The model is validated against a numerical model and various experimental data. - Abstract: The main idea of this paper is to establish a simple approach for prediction of the ignition time of a wood particle assuming that the thermo-physical properties remain constant and ignition takes place at a characteristic ignition temperature. Using a time and space integral method, explicit relationships are derived for computation of the ignition time of particles of three common shapes (slab, cylinder and sphere), which may be characterized as thermally thin or thermally thick. It is shown through a dimensionless analysis that the dimensionless ignition time can be described as a function of non-dimensional ignition temperature, reactor temperature or external incident heat flux, and parameter K which represents the ratio of conduction heat transfer to the external radiation heat transfer. The numerical results reveal that for the dimensionless ignition temperature between 1.25 and 2.25 and for values of K up to 8000 (corresponding to woody materials), the variation of the ignition time of a thermally thin particle with K and the dimensionless ignition temperature is linear, whereas the dependence of the ignition time of a thermally thick particle on the above two parameters obeys a quadratic function. Furthermore, it is shown that the transition from the regime of thermally thin to the regime of thermally thick occurs at K cr (corresponding to a critical size of particle) which is found to be independent of the particle shape. The model is validated by comparing the predicted and the measured ignition time of several wood particles obtained from different sources. Good agreement is achieved which

REACTION KINETICS SELF-PROPOGATION REGIME DURING PRE-IGNITION PERIOD

Directory of Open Access Journals (Sweden)

D. D. Polishchuk

2015-11-01

Full Text Available Self-propagation high temperature synthesis (SHS technological regulations application is mainly limited by transformation processes taking place in the pre-ignition period. Zn-S, Zn-Se, Ti-C and 3Ni-Al small sample systems ignition experimental study was carried out under heating conditions in inert atmosphere with temperature values T = 1200K.It was shown that at this temperature level a chemical reaction can be initiated, turning into a self-sustaining mode. Wherein the reaction limiting factors can be mass transfer processes. Ignition temperatures were determined and plotted via the samples size. A physical ignition model was developed assuming the pre-ignition period limiting reaction Arrhenius law.The inverse combustion problem solution made it possible to calculate the low-temperature (T = 800 ÷ 1200K reaction kinetic constant values. Comparison thus obtained values  with the known data of other researchers showed their good agreement.Activation energy values for the Zn-S system were used to calculate the heat wave propagation speed. This value appeared to coincide with experimental values.Obtained results analysis leads to the conclusion about the availability and justification for the proposed method of express-analysis of presupposed, but previously not studied SHS systems. The results thus obtained allow us to estimate conditions for the SHS technology implementation, the reactor characteristic sizes and the thermal wave’s propagation speed.

The National Ignition Facility Project

International Nuclear Information System (INIS)

Paisner, J.A.; Campbell, E.M.; Hogan, W.J.

1994-01-01

The mission of the National Ignition Facility is to achieve ignition and gain in ICF targets in the laboratory. The facility will be used for defense applications such as weapons physics and weapons effect testing, and for civilian applications such as fusion energy development and fundamental studies of matter at high temperatures and densities. This paper reviews the design, schedule and costs associated with the construction project

Experimental investigations of the minimum ignition energy and the minimum ignition temperature of inert and combustible dust cloud mixtures

Energy Technology Data Exchange (ETDEWEB)

Addai, Emmanuel Kwasi, E-mail: emmanueladdai41@yahoo.com; Gabel, Dieter; Krause, Ulrich

2016-04-15

Highlights: • Ignition sensitivity of a highly flammable dust decreases upon addition of inert dust. • Minimum ignition temperature of a highly flammable dust increases when inert concentration increase. • Minimum ignition energy of a highly flammable dust increases when inert concentration increase. • The permissible range for the inert mixture to minimize the ignition risk lies between 60 to 80%. - Abstract: The risks associated with dust explosions still exist in industries that either process or handle combustible dust. This explosion risk could be prevented or mitigated by applying the principle of inherent safety (moderation). This is achieved by adding an inert material to a highly combustible material in order to decrease the ignition sensitivity of the combustible dust. The presented paper deals with the experimental investigation of the influence of adding an inert dust on the minimum ignition energy and the minimum ignition temperature of the combustible/inert dust mixtures. The experimental investigation was done in two laboratory scale equipment: the Hartmann apparatus and the Godbert-Greenwald furnace for the minimum ignition energy and the minimum ignition temperature test respectively. This was achieved by mixing various amounts of three inert materials (magnesium oxide, ammonium sulphate and sand) and six combustible dusts (brown coal, lycopodium, toner, niacin, corn starch and high density polyethylene). Generally, increasing the inert materials concentration increases the minimum ignition energy as well as the minimum ignition temperatures until a threshold is reached where no ignition was obtained. The permissible range for the inert mixture to minimize the ignition risk lies between 60 to 80%.

Experimental investigations of the minimum ignition energy and the minimum ignition temperature of inert and combustible dust cloud mixtures

International Nuclear Information System (INIS)

Addai, Emmanuel Kwasi; Gabel, Dieter; Krause, Ulrich

2016-01-01

Highlights: • Ignition sensitivity of a highly flammable dust decreases upon addition of inert dust. • Minimum ignition temperature of a highly flammable dust increases when inert concentration increase. • Minimum ignition energy of a highly flammable dust increases when inert concentration increase. • The permissible range for the inert mixture to minimize the ignition risk lies between 60 to 80%. - Abstract: The risks associated with dust explosions still exist in industries that either process or handle combustible dust. This explosion risk could be prevented or mitigated by applying the principle of inherent safety (moderation). This is achieved by adding an inert material to a highly combustible material in order to decrease the ignition sensitivity of the combustible dust. The presented paper deals with the experimental investigation of the influence of adding an inert dust on the minimum ignition energy and the minimum ignition temperature of the combustible/inert dust mixtures. The experimental investigation was done in two laboratory scale equipment: the Hartmann apparatus and the Godbert-Greenwald furnace for the minimum ignition energy and the minimum ignition temperature test respectively. This was achieved by mixing various amounts of three inert materials (magnesium oxide, ammonium sulphate and sand) and six combustible dusts (brown coal, lycopodium, toner, niacin, corn starch and high density polyethylene). Generally, increasing the inert materials concentration increases the minimum ignition energy as well as the minimum ignition temperatures until a threshold is reached where no ignition was obtained. The permissible range for the inert mixture to minimize the ignition risk lies between 60 to 80%.

JENDL-3.3 thermal reactor benchmark test

International Nuclear Information System (INIS)

Akie, Hiroshi

2001-01-01

Integral tests of JENDL-3.2 nuclear data library have been carried out by Reactor Integral Test WG of Japanese Nuclear Data Committee. The most important problem in the thermal reactor benchmark testing was the overestimation of the multiplication factor of the U fueled cores. With several revisions of the data of 235 U and the other nuclides, JENDL-3.3 data library gives a good estimation of multiplication factors both for U and Pu fueled thermal reactors. (author)

Irradiation Testing Vehicles for Fast Reactors from Open Test Assemblies to Closed Loops

Energy Technology Data Exchange (ETDEWEB)

Sienicki, James J. [Argonne National Lab. (ANL), Argonne, IL (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-12-15

A review of irradiation testing vehicle approaches and designs that have been incorporated into past Sodium-Cooled Fast Reactors (SFRs) or envisioned for incorporation has been carried out. The objective is to understand the essential features of the approaches and designs so that they can inform test vehicle designs for a future U.S. Fast Test Reactor. Fast test reactor designs examined include EBR-II, FFTF, JOYO, BOR-60, PHÉNIX, JHR, and MBIR. Previous designers exhibited great ingenuity in overcoming design and operational challenges especially when the original reactor plant’s mission changed to an irradiation testing mission as in the EBRII reactor plant. The various irradiation testing vehicles can be categorized as: Uninstrumented open assemblies that fit into core locations; Instrumented open test assemblies that fit into special core locations; Self-contained closed loops; and External closed loops. A special emphasis is devoted to closed loops as they are regarded as a very desirable feature of a future U.S. Fast Test Reactor. Closed loops are an important technology for irradiation of fuels and materials in separate controlled environments. The impact of closed loops on the design of fast reactors is also discussed in this report.

Development and Testing of a Green-Propellant Micro-Hybrid Thruster with Electrostatic Ignition

Science.gov (United States)

Whitmore, Stephen A.; Judson, Michael D.

2012-01-01

As early as 1937 German scientists at Peenemunde experimented with highly unstable fuel blends of nitrous oxide (N2O) and ethanol. These early tests mostly resulted in explosions and destroyed rocket engines. More recently several companies have developed experimental nitrous oxide fuel blends (NOFB) with Isp exceeding 300 sec. Although NOFBx has recently been cleared for tests on the International Space Station, this propellant remains highly experimental and has not been cleared for commercial transport by the US DOT. Recent work by Karabeyoglu et al. has raised concerns about the safety risks of mixing hydrocarbons with N2O. Liquid oxidizer/fuel blends are highly explosive and require extreme care in transport and servicing. By adding small amounts of a liquid organic fuel such as alcohol or a hydrocarbon, the odds of an explosive decomposition event are significantly increased.iv The proposed solution mitigates the explosion hazards of NOFB by separating the oxidizer from the hydrocarbon fuel formed as of a small cylindrical section of ABS thermoplastic. As N2O vapor flows across the grain segment, current enters a 1000 VDC high-tension lead in the ABS fuel grain and produces an inductive spark that vaporizes a small amount of the material. The ablated fuel vapor plus residual energy from the spark seed a localized exothermic N2O dissociation that produces sufficient heat to initiate combustion. The process is also effective when gaseous oxygen is used. A low TRL (2-3) prototype demonstrating the feasibility of controlled hydrocarbon-seeding was recently tested at Utah State University.v The unit features a miniature 2.5 cm ABS fuel grain fabricated using a Stratasys Dimension 3-D printer. The 9-N thruster was pulse-fired up to 27 consecutive times on a single ABS grain segment. Ignition was achieved by as little as 12-15 Joules energy input. This value is contrasted with the typical 30-minute pre-heat requirement for the ECAPS LMP-103S ADN-based monopropellant

Evaluation of neutronic characteristics of in-pile test reactor for fast reactor safety research

Energy Technology Data Exchange (ETDEWEB)

Uto, N.; Ohno, S.; Kawata, N. [Power Reactor and Nuclear Fuel Development Corp., Oarai, Ibaraki (Japan). Oarai Engineering Center

1996-09-01

An extensive research program has been carried out at the Power Reactor and Nuclear Fuel Development Corporation for the safety of future liquid-metal fast breeder reactors to be commercialized. A major part of this program is investigation and planning of advanced safety experiments conducted with a new in-pile safety test facility, which is larger and more advanced than any of the currently existing test reactors. Such a transient safety test reactor generally has unique neutronic characteristics that require various studies from the reactor physics point of view. In this paper, the outcome of the neutronics study is highlighted with presenting a reference core design concept and its performance in regard to the safety test objectives. (author)

Ignition Regime for Fusion in a Degenerate Plasma

International Nuclear Information System (INIS)

Son, S.; Fisch, N.J.

2005-01-01

We identify relevant parameter regimes in which aneutronic fuels can undergo fusion ignition in hot-ion degenerate plasma. Because of relativistic effects and partial degeneracy, the self-sustained burning regime is considerably larger than previously calculated. Inverse bremsstrahlung plays a major role in containing the reactor energy. We solve the radiation transfer equation and obtain the contribution to the heat conductivity from inverse bremsstrahlung

The National Ignition Facility Project

International Nuclear Information System (INIS)

Paisner, J.A.; Campbell, E.M.; Hogan, W.J.

1994-01-01

The mission of the National Ignition Facility is to achieve ignition and gain in inertial confinement fusion targets in the laboratory. The facility will be used for defense applications such as weapons physics and weapons effects testing, and for civilian applications such as fusion energy development and fundamental studies of matter at high temperatures and densities. This paper reviews the design, schedule, and costs associated with the construction project

Experimental investigations of the minimum ignition energy and the minimum ignition temperature of inert and combustible dust cloud mixtures.

Science.gov (United States)

Addai, Emmanuel Kwasi; Gabel, Dieter; Krause, Ulrich

2016-04-15

The risks associated with dust explosions still exist in industries that either process or handle combustible dust. This explosion risk could be prevented or mitigated by applying the principle of inherent safety (moderation). This is achieved by adding an inert material to a highly combustible material in order to decrease the ignition sensitivity of the combustible dust. The presented paper deals with the experimental investigation of the influence of adding an inert dust on the minimum ignition energy and the minimum ignition temperature of the combustible/inert dust mixtures. The experimental investigation was done in two laboratory scale equipment: the Hartmann apparatus and the Godbert-Greenwald furnace for the minimum ignition energy and the minimum ignition temperature test respectively. This was achieved by mixing various amounts of three inert materials (magnesium oxide, ammonium sulphate and sand) and six combustible dusts (brown coal, lycopodium, toner, niacin, corn starch and high density polyethylene). Generally, increasing the inert materials concentration increases the minimum ignition energy as well as the minimum ignition temperatures until a threshold is reached where no ignition was obtained. The permissible range for the inert mixture to minimize the ignition risk lies between 60 to 80%. Copyright © 2016 Elsevier B.V. All rights reserved.

Ignition and flame spread properties of wood, elaborated during a new test method based on convective heat flux

DEFF Research Database (Denmark)

Sørensen, Lars Schiøtt; Poulsen, Annemarie

2007-01-01

Ignition and flame spread properties on selected types of wood are elaborated. The tests are established in a new test setup in which the test specimen can be fixed in different angles due to a horizontal level. The heat exposing the test objects is arranged as a convective flux. This principle...

Prospects for Tokamak Fusion Reactors

International Nuclear Information System (INIS)

Sheffield, J.; Galambos, J.

1995-01-01

This paper first reviews briefly the status and plans for research in magnetic fusion energy and discusses the prospects for the tokamak magnetic configuration to be the basis for a fusion power plant. Good progress has been made in achieving fusion reactor-level, deuterium-tritium (D-T) plasmas with the production of significant fusion power in the Joint European Torus (up to 2 MW) and the Tokamak Fusion Test Reactor (up to 10 MW) tokamaks. Advances on the technologies of heating, fueling, diagnostics, and materials supported these achievements. The successes have led to the initiation of the design phases of two tokamaks, the International Thermonuclear Experimental Reactor (ITER) and the US Toroidal Physics Experiment (TPX). ITER will demonstrate the controlled ignition and extended bum of D-T plasmas with steady state as an ultimate goal. ITER will further demonstrate technologies essential to a power plant in an integrated system and perform integrated testing of the high heat flux and nuclear components required to use fusion energy for practical purposes. TPX will complement ITER by testing advanced modes of steady-state plasma operation that, coupled with the developments in ITER, will lead to an optimized demonstration power plant

Advanced Test Reactor National Scientific User Facility

International Nuclear Information System (INIS)

Marshall, Frances M.; Benson, Jeff; Thelen, Mary Catherine

2011-01-01

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

Advanced Test Reactor National Scientific User Facility

Energy Technology Data Exchange (ETDEWEB)

Frances M. Marshall; Jeff Benson; Mary Catherine Thelen

2011-08-01

The Advanced Test Reactor (ATR), at the Idaho National Laboratory (INL), is a large test reactor for providing the capability for studying the effects of intense neutron and gamma radiation on reactor materials and fuels. The ATR is a pressurized, light-water, high flux test reactor with a maximum operating power of 250 MWth. The INL also has several hot cells and other laboratories in which irradiated material can be examined to study material irradiation effects. In 2007 the US Department of Energy (DOE) designated the ATR as a National Scientific User Facility (NSUF) to facilitate greater access to the ATR and the associated INL laboratories for material testing research by a broader user community. This paper highlights the ATR NSUF research program and the associated educational initiatives.

Ignition of mercury-free high intensity discharge lamps

International Nuclear Information System (INIS)

Czichy, M; Mentel, J; Awakowicz, P; Hartmann, T

2008-01-01

To achieve a better understanding of the ignition behaviour of D4 lamps for automotive headlights the ignition of mercury-free metal iodide test lamps characterized by a high xenon pressure, a small electrode distance and small electrode-wall distances is investigated. The ignition of these lamps is dominated by a high voltage requirement. Nevertheless lamps are found that show a surprisingly low ignition voltage. Electrical measurements and simultaneous optical observations of the ultra-fast streamer processes show that the breakdown takes place in two different modes. One of the ignition modes which requires a high ignition voltage is characterized by a breakdown in the volume between the electrode tips. The other mode is characterized by streamer discharges along the wall. In this case the cathode, its base and the wall around is involved in the ignition process and the lamp breaks down at low voltages

Lawson concepts and criticality in DT fusion reactors

International Nuclear Information System (INIS)

Lartigue, J.G.

1987-01-01

The original Lawson concepts (amplification factor R and parameter nτ) as well as their applications in DT reactors are discussed in two cases: the ignition regime and the subignition regime in a self-sufficient plant. The modified Lawson factor or internal amplification factor R a (a function of alpha power) is proposed as a means to measure the ignition level reached by the plasma, in a more precise way than that given by the collective parameter (nτkT). The self-sufficiency factor (δ) is proposed as a means to measure the plant self-sufficiency, δ being more significant than the traditional Q factor. It is stated that the ignition regime (R a = 1) is equivalent to a critical state (energy equilibrium); then, the corresponding critical mass concept is proposed. The analysis of the R a relationship with temperature (kT), (nτ), and recirculating factor (var-epsilon) gives the conditions for the reactor to reach ignition or for the plant to reach self-sufficiency; it also shows that an approach to ignition is not improved by heating from 50 to 100 KeV

A polar-drive shock-ignition design for the National Ignition Facility

Energy Technology Data Exchange (ETDEWEB)

Anderson, K. S.; McKenty, P. W.; Collins, T. J. B.; Craxton, R. S.; Delettrez, J. A.; Marozas, J. A.; Skupsky, S.; Shvydky, A. [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623 (United States); Betti, R. [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623 (United States); Fusion Science Center, University of Rochester, Rochester, New York 14623 (United States); Departments of Mechanical Engineering and Physics, University of Rochester, Rochester, New York 14627 (United States); Hohenberger, M.; Theobald, W.; Lafon, M.; Nora, R. [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623 (United States); Fusion Science Center, University of Rochester, Rochester, New York 14623 (United States)

2013-05-15

Shock ignition [R. Betti et al., Phys. Rev. Lett. 98, 155001 (2007)] is being pursued as a viable option to achieve ignition on the National Ignition Facility (NIF). Shock-ignition target designs use a high-intensity laser spike at the end of a low-adiabat assembly pulse to launch a spherically convergent strong shock to ignite the hot spot of an imploding capsule. A shock-ignition target design for the NIF is presented. One-dimensional simulations indicate an ignition threshold factor of 4.1 with a gain of 58. A polar-drive beam-pointing configuration for shock-ignition experiments on the NIF at 750 kJ is proposed. The capsule design is shown to be robust to the various one- and two-dimensional effects and nonuniformities anticipated on the NIF. The target is predicted to ignite with a gain of 38 when including all anticipated levels of nonuniformity and system uncertainty.

The Ignition Target for the National Ignition Facility

International Nuclear Information System (INIS)

Atherton, L J; Moses, E I; Carlisle, K; Kilkenny, J

2007-01-01

The National Ignition Facility (NIF) is a 192 beam Nd-glass laser facility presently under construction at Lawrence Livermore National Laboratory (LLNL) for performing inertial confinement fusion (ICF) and experiments studying high energy density (HED) science. When completed in 2009, NIF will be able to produce 1.8 MJ, 500 TW of ultraviolet light for target experiments that will create conditions of extreme temperatures (>10 8 K), pressures (10-GBar) and matter densities (> 100 g/cm 3 ). A detailed program called the National Ignition Campaign (NIC) has been developed to enable ignition experiments in 2010, with the goal of producing fusion ignition and burn of a deuterium-tritium (DT) fuel mixture in millimeter-scale target capsules. The first of the target experiments leading up to these ignition shots will begin in 2008. Targets for the National Ignition Campaign are both complex and precise, and are extraordinarily demanding in materials fabrication, machining, assembly, cryogenics and characterization. An overview of the campaign for ignition will be presented, along with technologies for target fabrication, assembly and metrology and advances in growth and x-ray imaging of DT ice layers. The sum of these efforts represents a quantum leap in target precision, characterization, manufacturing rate and flexibility over current state-of-the-art

Relating the octane numbers of fuels to ignition delay times measured in an ignition quality tester (IQT)

KAUST Repository

Naser, Nimal

2016-09-21

A methodology for estimating the octane index (OI), the research octane number (RON) and the motor octane number (MON) using ignition delay times from a constant volume combustion chamber with liquid fuel injection is proposed by adopting an ignition quality tester. A baseline data of ignition delay times were determined using an ignition quality tester at a charge pressure of 21.3 bar between 770 and 850 K and an equivalence ratio of 0.7 for various primary reference fuels (PRFs, mixtures of isooctane and n-heptane). Our methodology was developed using ignition delay times for toluene reference fuels (mixtures of toluene and n-heptane). A correlation between the OI and the ignition delay time at the initial charge temperature enabled the OI of non-PRFs to be predicted at specified temperatures. The methodology was validated using ignition delay times for toluene primary reference fuels (ternary mixtures of toluene, iso-octane, and n-heptane), fuels for advanced combustion engines (FACE) gasolines, and certification gasolines. Using this methodology, the RON, the MON, and the octane sensitivity were estimated in agreement with values obtained from standard test methods. A correlation between derived cetane number and RON is also provided. (C) 2016 Elsevier Ltd. All rights reserved.

Radial effects in heating and thermal stability of a sub-ignited tokamak

International Nuclear Information System (INIS)

Fuchs, V.; Shoucri, M.M.; Thibaudeau, G.; Harten, L.; Bers, A.

1982-02-01

The existence of thermally stable sub-ignited equilibria of a tokamak reactor, sustained in operation by a feedback-controlled supplementary heating source, is demonstrated. The establishment of stability depends on a number of radially non-uniform, nonlinear processes whose effect is analyzed. One-dimensional (radial) stability analyses of model transport equations, together with numerical results from a 1-D transport code, are used in studying the heating of DT-plasmas in the thermonuclear regime. Plasma core supplementary heating is found to be a thermally more stable process than bulk heating. In the presence of impurity line radiation, however, core-heated temperature profiles may collapse, contracting inward from the limiter, the result of an instability caused by the increasing nature of the radiative cooling rate, with decreasing temperature. Conditions are established for the realization of a sub-ignited high-Q, toroidal reactor plasma with appreciable output power

Introduction of CFD Analysis for Consideration of Fuel Ignition Test

OpenAIRE

岡崎,航介; 吉田,肇; 入澤,優麿; 櫻庭,隆貴

2014-01-01

Not a few fires in the engine room of ships are caused from the ignition of spilled oil on the hot part of machines such as turbo-chargers, boilers etc. Authors have continued to investigate ignition and combustion phenomenon of oils on the hot surface of a metal plate with various conditions of the surface and many kinds of fuels for understanding of the mechanism of fires in engine rooms. However, it could not be well-considered about flow of the air in the reaction region because of the de...

Thermo-kinetic instabilities in model reactors. Examples in experimental tests

Science.gov (United States)

Lavadera, Marco Lubrano; Sorrentino, Giancarlo; Sabia, Pino; de Joannon, Mara; Cavaliere, Antonio; Ragucci, Raffaele

2017-11-01

The use of advanced combustion technologies (such as MILD, LTC, etc.) is among the most promising methods to reduce emission of pollutants. For such technologies, working temperatures are enough low to boost the formation of several classes of pollutants, such as NOx and soot. To access this temperature range, a significant dilution as well as preheating of reactants is required. Such conditions are usually achieved by a strong recirculation of exhaust gases that simultaneously dilute and pre-heat the fresh reactants. These peculiar operative conditions also imply strong fuel flexibility, thus allowing the use of low calorific value (LCV) energy carriers with high efficiency. However, the intersection of low combustion temperatures and highly diluted mixtures with intense pre-heating alters the evolution of the combustion process with respect to traditional flames, leading to features such as the susceptibility to oscillations, which are undesirable during combustion. Therefore, an effective use of advanced combustion technologies requires a thorough analysis of the combustion kinetic characteristics in order to identify optimal operating conditions and control strategies with high efficiency and low pollutant emissions. The present work experimentally and numerically characterized the ignition and oxidation processes of methane and propane, highly diluted in nitrogen, at atmospheric pressure, in a Plug Flow Reactor and a Perfectly Stirred Reactor under a wide range of operating conditions involving temperatures, mixture compositions and dilution levels. The attention was focused particularly on the chemistry of oscillatory phenomena and multistage ignitions. The global behavior of these systems can be qualitatively and partially quantitatively modeled using the detailed kinetic models available in the literature. Results suggested that, for diluted conditions and lower adiabatic flame temperatures, the competition among several pathways, i.e. intermediate- and

Shock Tube Ignition Delay Data Affected by Localized Ignition Phenomena

KAUST Repository

Javed, Tamour

2016-12-29

Shock tubes have conventionally been used for measuring high-temperature ignition delay times ~ O(1 ms). In the last decade or so, the operating regime of shock tubes has been extended to lower temperatures by accessing longer observation times. Such measurements may potentially be affected by some non-ideal phenomena. The purpose of this work is to measure long ignition delay times for fuels exhibiting negative temperature coefficient (NTC) and to assess the impact of shock tube non-idealities on ignition delay data. Ignition delay times of n-heptane and n-hexane were measured over the temperature range of 650 – 1250 K and pressures near 1.5 atm. Driver gas tailoring and long length of shock tube driver section were utilized to measure ignition delay times as long as 32 ms. Measured ignition delay times agree with chemical kinetic models at high (> 1100 K) and low (< 700 K) temperatures. In the intermediate temperature range (700 – 1100 K), however, significant discrepancies are observed between the measurements and homogeneous ignition delay simulations. It is postulated, based on experimental observations, that localized ignition kernels could affect the ignition delay times at the intermediate temperatures, which lead to compression (and heating) of the bulk gas and result in expediting the overall ignition event. The postulate is validated through simple representative computational fluid dynamic simulations of post-shock gas mixtures which exhibit ignition advancement via a hot spot. The results of the current work show that ignition delay times measured by shock tubes may be affected by non-ideal phenomena for certain conditions of temperature, pressure and fuel reactivity. Care must, therefore, be exercised in using such data for chemical kinetic model development and validation.

TR-EDB: Test Reactor Embrittlement Data Base, Version 1

Energy Technology Data Exchange (ETDEWEB)

Stallmann, F.W.; Wang, J.A.; Kam, F.B.K. [Oak Ridge National Lab., TN (United States)

1994-01-01

The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results.

TR-EDB: Test Reactor Embrittlement Data Base, Version 1

International Nuclear Information System (INIS)

Stallmann, F.W.; Wang, J.A.; Kam, F.B.K.

1994-01-01

The Test Reactor Embrittlement Data Base (TR-EDB) is a collection of results from irradiation in materials test reactors. It complements the Power Reactor Embrittlement Data Base (PR-EDB), whose data are restricted to the results from the analysis of surveillance capsules in commercial power reactors. The rationale behind their restriction was the assumption that the results of test reactor experiments may not be applicable to power reactors and could, therefore, be challenged if such data were included. For this very reason the embrittlement predictions in the Reg. Guide 1.99, Rev. 2, were based exclusively on power reactor data. However, test reactor experiments are able to cover a much wider range of materials and irradiation conditions that are needed to explore more fully a variety of models for the prediction of irradiation embrittlement. These data are also needed for the study of effects of annealing for life extension of reactor pressure vessels that are difficult to obtain from surveillance capsule results

Reliability test for reactor internals rejuvenation technology

International Nuclear Information System (INIS)

Uchiyama, Junichi

1998-01-01

41 transparencies were presented on the subject of 'Reliability test for reactor internals rejuvenation technology'. The items presented give an introduction on the management of plant life in Japan and introduce the Nuclear Power Engineering Corporation (NUPEC). The question of what reliability tests for rejuvenation of reactor internals are is discussed in some detail and an outline of each test is given. Altogether six methods to rejuvenate reactor internals are presented, two of which have already been applied to actual plants. The presentation was supported by many detailed drawings and images

Space reactor fuel element testing in upgraded TREAT

International Nuclear Information System (INIS)

Todosow, M.; Bezler, P.; Ludewig, H.; Kato, W.Y.

1993-01-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ∼60--80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ∼100 MW/L may be achievable

Ground testing of an SP-100 prototypic reactor

International Nuclear Information System (INIS)

Motwani, K.; Pflasterer, G.R.; Upton, H.; Lazarus, J.D.; Gluck, R.

1988-01-01

SP-100 is a space power system which is being developed by GE to meet future space electrical power requirements. The ground testing of an SP-100 prototypic reactor system will be conducted at the Westinghouse Hanford Company site located at Richland, Washington. The objective of this test is to demonstrate the performance of a full scale prototypic reactor system, including the reactor, control system and flight shield. The ground test system is designed to simulate the flight operating conditions while meeting all the necessary nuclear safety requirements in a gravity environment. The goal of the reactor ground test system is to establish confidence in the design maturity of the SP-100 space reactor power system and resolve the technical issues necessary for the development of a flight mission design

Space reactor fuel element testing in upgraded TREAT

Science.gov (United States)

Todosow, Michael; Bezler, Paul; Ludewig, Hans; Kato, Walter Y.

1993-01-01

The testing of candidate fuel elements at prototypic operating conditions with respect to temperature, power density, hydrogen coolant flow rate, etc., is a crucial component in the development and qualification of nuclear rocket engines based on the Particle Bed Reactor (PBR), NERVA-derivative, and other concepts. Such testing may be performed at existing reactors, or at new facilities. A scoping study has been performed to assess the feasibility of testing PBR based fuel elements at the TREAT reactor. Initial results suggests that full-scale PBR elements could be tested at an average energy deposition of ˜60-80 MW-s/L in the current TREAT reactor. If the TREAT reactor was upgraded to include fuel elements with a higher temperture limit, average energy deposition of ˜100 MW/L may be achievable.

Results of assembly test of HTTR reactor internals

International Nuclear Information System (INIS)

Maruyama, S.; Saikusa, A.; Shiozawa, S.; Tsuji, N.; Miki, T.

1996-01-01

The assembly test of the HTTR actual reactor internals had been carried out at the works, prior to their installation in the actual reactor pressure vessel(RPV) at the construction site. The assembly test consists of several items such as examining fabricating precision of each component and alignment of piled-up structures, measuring circumferential coolant velocity profile in the passage between the simulated RPV and the reactor internals as well as under the support plates, measuring by-pass flow rate through gaps between the reactor internals, and measuring the binding force of the core restraint mechanism. Results of the test showed good performance of the HTTR reactor internals. Installation of the reactor internals in the actual RPV was started at the construction site of HTTR in April, 1995. In the installation process, main items of the assembly test at the works were repeated to investigate the reproducibility of installation. (author). 5 refs, 11 figs

Ignition characteristics of 2-methyltetrahydrofuran: An experimental and kinetic study

KAUST Repository

Tripathi, Rupali

2016-10-15

The present paper elucidates oxidation behavior of 2-methyltetrahydrofuran (2-MTHF), a novel second-generation biofuel. New experimental data sets for 2-MTHF including ignition delay time measurements in two different combustion reactors, i.e. rapid compression machine and high-pressure shock tube, are presented. Measurements for 2-MTHF/oxidizer/diluent mixtures were performed in the temperature range of . 639-1413 K, at pressures of 10, 20, and 40 bar, and at three different equivalence ratios of 0.5, 1.0, and 2.0. A detailed chemical kinetic model describing both low-and high-temperature chemistry of 2-MTHF was developed and validated against new ignition delay measurements and already existing flame species profiles and ignition delay measurements. The mechanism provides satisfactory agreement with the experimental data. For identifying key reactions at various combustion conditions and to attain a better understanding of the combustion behavior, reaction path and sensitivity analyses were performed.

The National Ignition Facility (NIF): A path to fusion energy

International Nuclear Information System (INIS)

Moses, Edward I.

2008-01-01

Fusion energy has long been considered a promising, clean, nearly inexhaustible source of energy. Power production by fusion micro-explosions of inertial confinement fusion (ICF) targets has been a long-term research goal since the invention of the first laser in 1960. The National Ignition Facility (NIF) is poised to take the next important step in the journey by beginning experiments researching ICF ignition. Ignition on NIF will be the culmination of over 30 years of ICF research on high-powered laser systems such as the Nova laser at Lawrence Livermore National Laboratory (LLNL) and the OMEGA laser at the University of Rochester, as well as smaller systems around the world. NIF is a 192-beam Nd-glass laser facility at LLNL that is more than 90% complete. The first cluster of 48 beams is operational in the laser bay, the second cluster is now being commissioned, and the beam path to the target chamber is being installed. The Project will be completed in 2009, and ignition experiments will start in 2010. When completed, NIF will produce up to 1.8 MJ of 0.35-μm light in highly shaped pulses required for ignition. It will have beam stability and control to higher precision than any other laser fusion facility. Experiments using one of the beams of NIF have demonstrated that NIF can meet its beam performance goals. The National Ignition Campaign (NIC) has been established to manage the ignition effort on NIF. NIC has all of the research and development required to execute the ignition plan and to develop NIF into a fully operational facility. NIF will explore the ignition space, including direct drive, 2ω ignition, and fast ignition, to optimize target efficiency for developing fusion as an energy source. In addition to efficient target performance, fusion energy requires significant advances in high-repetition-rate lasers and fusion reactor technology. The Mercury laser at LLNL is a high-repetition-rate Nd-glass laser for fusion energy driver development. Mercury

Application of Alcohols to Dual - Fuel Feeding the Spark-Ignition and Self-Ignition Engines

Directory of Open Access Journals (Sweden)

Stelmasiak Zdzisław

2014-10-01

Full Text Available This paper concerns analysis of possible use of alcohols for the feeding of self - ignition and spark-ignition engines operating in a dual- fuel mode, i.e. simultaneously combusting alcohol and diesel oil or alcohol and petrol. Issues associated with the requirements for application of bio-fuels were presented with taking into account National Index Targets, bio-ethanol production methods and dynamics of its production worldwide and in Poland. Te considerations are illustrated by results of the tests on spark- ignition and self- ignition engines fed with two fuels: petrol and methanol or diesel oil and methanol, respectively. Te tests were carried out on a 1100 MPI Fiat four- cylinder engine with multi-point injection and a prototype collector fitted with additional injectors in each cylinder. Te other tested engine was a SW 680 six- cylinder direct- injection diesel engine. Influence of a methanol addition on basic operational parameters of the engines and exhaust gas toxicity were analyzed. Te tests showed a favourable influence of methanol on combustion process of traditional fuels and on some operational parameters of engines. An addition of methanol resulted in a distinct rise of total efficiency of both types of engines at maintained output parameters (maximum power and torque. In the same time a radical drop in content of hydrocarbons and nitrogen oxides in exhaust gas was observed at high shares of methanol in feeding dose of ZI (petrol engine, and 2-3 fold lower smokiness in case of ZS (diesel engine. Among unfavourable phenomena, a rather insignificant rise of CO and NOx content for ZI engine, and THC and NOx - for ZS engine, should be numbered. It requires to carry out further research on optimum control parameters of the engines. Conclusions drawn from this work may be used for implementation of bio-fuels to feeding the combustion engines.

FASTER test reactor preconceptual design report summary

Energy Technology Data Exchange (ETDEWEB)

Grandy, C. [Argonne National Lab. (ANL), Argonne, IL (United States); Belch, H. [Argonne National Lab. (ANL), Argonne, IL (United States); Brunett, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Jin, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Mohamed, W. [Argonne National Lab. (ANL), Argonne, IL (United States); Moisseytsev, A. [Argonne National Lab. (ANL), Argonne, IL (United States); Passerini, S. [Argonne National Lab. (ANL), Argonne, IL (United States); Sienicki, J. [Argonne National Lab. (ANL), Argonne, IL (United States); Sumner, T. [Argonne National Lab. (ANL), Argonne, IL (United States); Vilim, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Hayes, Steven [Argonne National Lab. (ANL), Argonne, IL (United States)

2016-02-29

The FASTER reactor plant is a sodium-cooled fast spectrum test reactor that provides high levels of fast and thermal neutron flux for scientific research and development. The 120MWe FASTER reactor plant has a superheated steam power conversion system which provides electrical power to a local grid allowing for recovery of operating costs for the reactor plant.

Power balance in an Ohmically heated fusion reactor

International Nuclear Information System (INIS)

Christiansen, J.P.; Roberts, K.V.

1982-01-01

A simplified power-balance equation (zero-dimensional model) is used to study the performance of an Ohmically heated fusion reactor with emphasis on a pulsed reversed-field pinch concept (RFP). The energy confinement time tausub(E) is treated as an adjustable function, and empirical tokamak scaling laws are employed in the numerical estimates, which are supplemented by 1-D ATHENE code calculations. The known heating rates and energy losses are represented by the net energy replacement time tausub(W), which is exhibited as a surface in density (n) and temperature (T) space with a saddle point (nsub(*), Tsub(*)), the optimum ignition point. It is concluded that i) ignition by Ohmic heating is more practicable for the RFP reactor than for a tokamak reactor with the same tausub(E), (ii) if at fixed current the minor radius can be reduced or at fixed minor radius the current can be increased, then it is found that Ohmic ignition becomes more likely when present tokamak scaling laws are used. More definitive estimates require, however, a knowledge of tausub(E), which can only be obtained by establishing a reliable set of experimental RFP scaling laws and, in particular, by extending RFP experiments closer to the reactor regime. (author)

Coil-On-Plug Ignition for LOX/Methane Liquid Rocket Engines in Thermal Vacuum Environments

Science.gov (United States)

Melcher, John C.; Atwell, Matthew J.; Morehead, Robert L.; Hurlbert, Eric A.; Bugarin, Luz; Chaidez, Mariana

2017-01-01

A coil-on-plug ignition system has been developed and tested for Liquid Oxygen (LOX) / liquid methane rocket engines operating in thermal vacuum conditions. The igniters were developed and tested as part of the Integrated Cryogenic Propulsion Test Article (ICPTA), previously tested as part of the Project Morpheus test vehicle. The ICPTA uses an integrated, pressure-fed, cryogenic LOX/methane propulsion system including a reaction control system (RCS) and a main engine. The ICPTA was tested at NASA Glenn Research Center's Plum Brook Station in the Spacecraft Propulsion Research Facility (B-2) under vacuum and thermal vacuum conditions. In order to successfully demonstrate ignition reliability in the vacuum conditions and eliminate corona discharge issues, a coil-on-plug ignition system has been developed. The ICPTA uses spark-plug ignition for both the main engine igniter and the RCS. The coil-on-plug configuration eliminates the conventional high-voltage spark plug cable by combining the coil and the spark-plug into a single component. Prior to ICPTA testing at Plum Brook, component-level reaction control engine (RCE) and main engine igniter testing was conducted at NASA Johnson Space Center (JSC), which demonstrated successful hot-fire ignition using the coil-on-plug from sea-level ambient conditions down to 10(exp.-2) torr. Integrated vehicle hot-fire testing at JSC demonstrated electrical and command/data system performance. Lastly, Plum Brook testing demonstrated successful ignitions at simulated altitude conditions at 30 torr and cold thermal-vacuum conditions at 6 torr. The test campaign successfully proved that coil-on-plug technology will enable integrated LOX/methane propulsion systems in future spacecraft.

Coil-On-Plug Ignition for Oxygen/Methane Liquid Rocket Engines in Thermal-Vacuum Environments

Science.gov (United States)

Melcher, John C.; Atwell, Matthew J.; Morehead, Robert L.; Hurlbert, Eric A.; Bugarin, Luz; Chaidez, Mariana

2017-01-01

A coil-on-plug ignition system has been developed and tested for Liquid Oxygen (LOX)/liquid methane (LCH4) rocket engines operating in thermal vacuum conditions. The igniters were developed and tested as part of the Integrated Cryogenic Propulsion Test Article (ICPTA), previously tested as part of the Project Morpheus test vehicle. The ICPTA uses an integrated, pressure-fed, cryogenic LOX/LCH4 propulsion system including a reaction control system (RCS) and a main engine. The ICPTA was tested at NASA Glenn Research Center's Plum Brook Station in the Spacecraft Propulsion Research Facility (B-2) under vacuum and thermal vacuum conditions. A coil-on-plug ignition system has been developed to successfully demonstrate ignition reliability at these conditions while preventing corona discharge issues. The ICPTA uses spark plug ignition for both the main engine igniter and the RCS. The coil-on-plug configuration eliminates the conventional high-voltage spark plug cable by combining the coil and the spark plug into a single component. Prior to ICPTA testing at Plum Brook, component-level reaction control engine (RCE) and main engine igniter testing was conducted at NASA Johnson Space Center (JSC), which demonstrated successful hot-fire ignition using the coil-on-plug from sea-level ambient conditions down to 10(exp -2) torr. Integrated vehicle hot-fire testing at JSC demonstrated electrical and command/data system performance. Lastly, hot-fire testing at Plum Brook demonstrated successful ignitions at simulated altitude conditions at 30 torr and cold thermal-vacuum conditions at 6 torr. The test campaign successfully proved that coil-on-plug technology will enable integrated LOX/LCH4 propulsion systems in future spacecraft.

Tritium and ignition target management at the National Ignition Facility.

Science.gov (United States)

Draggoo, Vaughn

2013-06-01

Isotopic mixtures of hydrogen constitute the basic fuel for fusion targets of the National Ignition Facility (NIF). A typical NIF fusion target shot requires approximately 0.5 mmoles of hydrogen gas and as much as 750 GBq (20 Ci) of 3H. Isotopic mix ratios are specified according to the experimental shot/test plan and the associated test objectives. The hydrogen isotopic concentrations, absolute amounts, gas purity, configuration of the target, and the physical configuration of the NIF facility are all parameters and conditions that must be managed to ensure the quality and safety of operations. An essential and key step in the preparation of an ignition target is the formation of a ~60 μm thick hydrogen "ice" layer on the inner surface of the target capsule. The Cryogenic Target Positioning System (Cryo-Tarpos) provides gas handling, cyro-cooling, x-ray imaging systems, and related instrumentation to control the volumes and temperatures of the multiphase (solid, liquid, and gas) hydrogen as the gas is condensed to liquid, admitted to the capsule, and frozen as a single spherical crystal of hydrogen in the capsule. The hydrogen fuel gas is prepared in discrete 1.7 cc aliquots in the LLNL Tritium Facility for each ignition shot. Post-shot hydrogen gas is recovered in the NIF Tritium Processing System (TPS). Gas handling systems, instrumentation and analytic equipment, material accounting information systems, and the shot planning systems must work together to ensure that operational and safety requirements are met.

Intrinsic reaction kinetics of coal char combustion by direct measurement of ignition temperature

International Nuclear Information System (INIS)

Kim, Ryang-Gyoon; Jeon, Chung-Hwan

2014-01-01

A wire heating reactor that can use a synchronized experimental method was developed to obtain the intrinsic kinetics of large coal char particles ranging in size from 0.4 to 1 mm. This synchronization system consists of three parts: a thermocouple wire for both heating and direct measurement of the particle temperature, a photodetector sensor for determining ignition/burnout points by measuring the intensity of luminous emission from burning particles, and a high-speed camera–long-distance microscope for observing and recording the movement of luminous zone directly. Coal char ignition was found to begin at a spot on the particle's external surface and then moved across the entire particle. Moreover, the ignition point determined according to the minimum of dT/dt is a spot point and not a full growth point. The ignition temperature of the spot point rises as the particle diameter increases. A spot ignition model, which describes the ignition in terms of the internal conduction and external/internal oxygen diffusion, was then developed to evaluate the intrinsic kinetics and predict the ignition temperature of the coal char. Internal conduction was found to be important in large coal char particles because its effect becomes greater than that of oxygen diffusion as the particle diameter increases. In addition, the intrinsic kinetics of coal char obtained from the spot ignition model for two types of coal does not differ significantly from the results of previous investigators. -- Highlights: • A novel technique was used to measure the coal char particle temperature. • The ignition point determined from a dT/dt minimum is a spot ignition point. • A spot ignition model was suggested to analyze the intrinsic reaction kinetics of coal char. • Internal conduction has to be considered in order to evaluate the intrinsic kinetics for larger particle (above 1 mm)

Ignition Delay of Combustible Materials in Normoxic Equivalent Environments

Science.gov (United States)

McAllister, Sara; Fernandez-Pello, Carlos; Ruff, Gary; Urban, David

2009-01-01

Material flammability is an important factor in determining the pressure and composition (fraction of oxygen and nitrogen) of the atmosphere in the habitable volume of exploration vehicles and habitats. The method chosen in this work to quantify the flammability of a material is by its ease of ignition. The ignition delay time was defined as the time it takes a combustible material to ignite after it has been exposed to an external heat flux. Previous work in the Forced Ignition and Spread Test (FIST) apparatus has shown that the ignition delay in the currently proposed space exploration atmosphere (approximately 58.6 kPa and32% oxygen concentration) is reduced by 27% compared to the standard atmosphere used in the Space Shuttle and Space Station. In order to determine whether there is a safer environment in terms of material flammability, a series of piloted ignition delay tests using polymethylmethacrylate (PMMA) was conducted in the FIST apparatus to extend the work over a range of possible exploration atmospheres. The exploration atmospheres considered were the normoxic equivalents, i.e. reduced pressure conditions with a constant partial pressure of oxygen. The ignition delay time was seen to decrease as the pressure was reduced along the normoxic curve. The minimum ignition delay observed in the normoxic equivalent environments was nearly 30% lower than in standard atmospheric conditions. The ignition delay in the proposed exploration atmosphere is only slightly larger than this minimum. Interms of material flammability, normoxic environments with a higher pressure relative to the proposed pressure would be desired.

Laser Ignition Technology for Bi-Propellant Rocket Engine Applications

Science.gov (United States)

Thomas, Matthew E.; Bossard, John A.; Early, Jim; Trinh, Huu; Dennis, Jay; Turner, James (Technical Monitor)

2001-01-01

The fiber optically coupled laser ignition approach summarized is under consideration for use in igniting bi-propellant rocket thrust chambers. This laser ignition approach is based on a novel dual pulse format capable of effectively increasing laser generated plasma life times up to 1000 % over conventional laser ignition methods. In the dual-pulse format tinder consideration here an initial laser pulse is used to generate a small plasma kernel. A second laser pulse that effectively irradiates the plasma kernel follows this pulse. Energy transfer into the kernel is much more efficient because of its absorption characteristics thereby allowing the kernel to develop into a much more effective ignition source for subsequent combustion processes. In this research effort both single and dual-pulse formats were evaluated in a small testbed rocket thrust chamber. The rocket chamber was designed to evaluate several bipropellant combinations. Optical access to the chamber was provided through small sapphire windows. Test results from gaseous oxygen (GOx) and RP-1 propellants are presented here. Several variables were evaluated during the test program, including spark location, pulse timing, and relative pulse energy. These variables were evaluated in an effort to identify the conditions in which laser ignition of bi-propellants is feasible. Preliminary results and analysis indicate that this laser ignition approach may provide superior ignition performance relative to squib and torch igniters, while simultaneously eliminating some of the logistical issues associated with these systems. Further research focused on enhancing the system robustness, multiplexing, and window durability/cleaning and fiber optic enhancements is in progress.

Simultaneous measurement of ignition energy and current signature for brush discharges

International Nuclear Information System (INIS)

Fast, Lars; Andersson, Birgitta; Smallwood, Jeremy; Holdstock, Paul; Paasi, Jaakko

2011-01-01

Accurate prediction of the probability of ignition arising from charged insulators is a crucial element of risk assessment in process industry. Incendiary brush discharges can occur when a large or grounded conductor approaches a charged insulator in the presence of a flammable atmosphere. This paper describes ignition tests based on an IEC standard method and simultaneously recorded temporal distribution of current released in the discharges, using a discharge probe integrated with the ignition probe. Ignition and non-ignition results are compared with peak discharge current and charge transferred in the discharge. No clear ignition threshold was found for either of these parameters. No major differences were found between igniting and non-igniting waveforms.

Closed Loop In-Reactor Assembly (CLIRA): a fast flux test facility test vehicle

International Nuclear Information System (INIS)

Oakley, D.J.

1978-01-01

The Closed Loop In-Reactor Assembly (CLIRA) is a test vehicle for in-core material and fuel experiments in the Fast Flux Test Facility (FFTF). The FFTF is a fast flux nuclear test reactor operated for the Department of Energy (DOE) by Westinghouse Hanford Company in Richland, Washington. The CLIRA is a removable/replaceable part of the Closed Loop System (CLS) which is a sodium coolant system providing flow and temperature control independent of the reactor coolant system. The primary purpose of the CLIRA is to provide a test vehicle which will permit testing of nuclear fuels and materials at conditions more severe than exist in the FTR core, and to isolate these materials from the reactor core

Ignition and Inertial Confinement Fusion at The National Ignition Facility

International Nuclear Information System (INIS)

Moses, E.

2009-01-01

The National Ignition Facility (NIF), the world's largest and most powerful laser system for inertial confinement fusion (ICF) and for studying high-energy-density (HED) science, is now operational at Lawrence Livermore National Laboratory (LLNL). The NIF is now conducting experiments to commission the laser drive, the hohlraum and the capsule and to develop the infrastructure needed to begin the first ignition experiments in FY 2010. Demonstration of ignition and thermonuclear burn in the laboratory is a major NIF goal. NIF will achieve this by concentrating the energy from the 192 beams into a mm 3 -sized target and igniting a deuterium-tritium mix, liberating more energy than is required to initiate the fusion reaction. NIF's ignition program is a national effort managed via the National Ignition Campaign (NIC). The NIC has two major goals: execution of DT ignition experiments starting in FY2010 with the goal of demonstrating ignition and a reliable, repeatable ignition platform by the conclusion of the NIC at the end of FY2012. The NIC will also develop the infrastructure and the processes required to operate NIF as a national user facility. The achievement of ignition at NIF will demonstrate the scientific feasibility of ICF and focus worldwide attention on laser fusion as a viable energy option. A laser fusion-based energy concept that builds on NIF, known as LIFE (Laser Inertial Fusion Energy), is currently under development. LIFE is inherently safe and can provide a global carbon-free energy generation solution in the 21st century. This paper describes recent progress on NIF, NIC, and the LIFE concept.

Performance Characterization and Auto-Ignition Performance of a Rapid Compression Machine

Directory of Open Access Journals (Sweden)

Hao Liu

2014-09-01

Full Text Available A rapid compression machine (RCM test bench is developed in this study. The performance characterization and auto-ignition performance tests are conducted at an initial temperature of 293 K, a compression ratio of 9.5 to 16.5, a compressed temperature of 650 K to 850 K, a driving gas pressure range of 0.25 MPa to 0.7 MPa, an initial pressure of 0.04 MPa to 0.09 MPa, and a nitrogen dilution ratio of 35% to 65%. A new type of hydraulic piston is used to address the problem in which the hydraulic buffer adversely affects the rapid compression process. Auto-ignition performance tests of the RCM are then performed using a DME–O2–N2 mixture. The two-stage ignition delay and negative temperature coefficient (NTC behavior of the mixture are observed. The effects of driving gas pressure, compression ratio, initial pressure, and nitrogen dilution ratio on the two-stage ignition delay are investigated. Results show that both the first-stage and overall ignition delays tend to increase with increasing driving gas pressure. The driving gas pressure within a certain range does not significantly influence the compressed pressure. With increasing compression ratio, the first-stage ignition delay is shortened, whereas the second-stage ignition delay is extended. With increasing initial pressure, both the first-stage and second-stage ignition delays are shortened. The second-stage ignition delay is shortened to a greater extent than that of the first-stage. With increasing nitrogen dilution ratio, the first-stage ignition delay is shortened, whereas the second-stage is extended. Thus, overall ignition delay presents different trends under various compression ratios and compressed pressure conditions.

Theory of hydro-equivalent ignition for inertial fusion and its applications to OMEGA and the National Ignition Facility

Energy Technology Data Exchange (ETDEWEB)

Nora, R.; Betti, R.; Bose, A.; Woo, K. M.; Christopherson, A. R.; Meyerhofer, D. D. [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623-1299 (United States); Fusion Science Center, University of Rochester, 250 East River Road, Rochester, New York 14623-1299 (United States); Department of Physics and/or Mechanical Engineering, University of Rochester, 250 East River Road, Rochester, New York 14623-1299 (United States); Anderson, K. S.; Shvydky, A.; Marozas, J. A.; Collins, T. J. B.; Radha, P. B.; Hu, S. X.; Epstein, R.; Marshall, F. J.; Sangster, T. C. [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623-1299 (United States); McCrory, R. L. [Laboratory for Laser Energetics, University of Rochester, 250 East River Road, Rochester, New York 14623-1299 (United States); Department of Physics and/or Mechanical Engineering, University of Rochester, 250 East River Road, Rochester, New York 14623-1299 (United States)

2014-05-15

The theory of ignition for inertial confinement fusion capsules [R. Betti et al., Phys. Plasmas 17, 058102 (2010)] is used to assess the performance requirements for cryogenic implosion experiments on the Omega Laser Facility. The theory of hydrodynamic similarity is developed in both one and two dimensions and tested using multimode hydrodynamic simulations with the hydrocode DRACO [P. B. Radha et al., Phys. Plasmas 12, 032702 (2005)] of hydro-equivalent implosions (implosions with the same implosion velocity, adiabat, and laser intensity). The theory is used to scale the performance of direct-drive OMEGA implosions to the National Ignition Facility (NIF) energy scales and determine the requirements for demonstrating hydro-equivalent ignition on OMEGA. Hydro-equivalent ignition on OMEGA is represented by a cryogenic implosion that would scale to ignition on the NIF at 1.8 MJ of laser energy symmetrically illuminating the target. It is found that a reasonable combination of neutron yield and areal density for OMEGA hydro-equivalent ignition is 3 to 6 × 10{sup 13} and ∼0.3 g/cm{sup 2}, respectively, depending on the level of laser imprinting. This performance has not yet been achieved on OMEGA.

Theory of hydro-equivalent ignition for inertial fusion and its applications to OMEGA and the National Ignition Facility

International Nuclear Information System (INIS)

Nora, R.; Betti, R.; Bose, A.; Woo, K. M.; Christopherson, A. R.; Meyerhofer, D. D.; Anderson, K. S.; Shvydky, A.; Marozas, J. A.; Collins, T. J. B.; Radha, P. B.; Hu, S. X.; Epstein, R.; Marshall, F. J.; Sangster, T. C.; McCrory, R. L.

2014-01-01

The theory of ignition for inertial confinement fusion capsules [R. Betti et al., Phys. Plasmas 17, 058102 (2010)] is used to assess the performance requirements for cryogenic implosion experiments on the Omega Laser Facility. The theory of hydrodynamic similarity is developed in both one and two dimensions and tested using multimode hydrodynamic simulations with the hydrocode DRACO [P. B. Radha et al., Phys. Plasmas 12, 032702 (2005)] of hydro-equivalent implosions (implosions with the same implosion velocity, adiabat, and laser intensity). The theory is used to scale the performance of direct-drive OMEGA implosions to the National Ignition Facility (NIF) energy scales and determine the requirements for demonstrating hydro-equivalent ignition on OMEGA. Hydro-equivalent ignition on OMEGA is represented by a cryogenic implosion that would scale to ignition on the NIF at 1.8 MJ of laser energy symmetrically illuminating the target. It is found that a reasonable combination of neutron yield and areal density for OMEGA hydro-equivalent ignition is 3 to 6 × 10 13 and ∼0.3 g/cm 2 , respectively, depending on the level of laser imprinting. This performance has not yet been achieved on OMEGA

The national ignition facility (NIF) : A path to fusion energy

International Nuclear Information System (INIS)

Moses, E. I.

2007-01-01

Fusion energy has long been considered a promising clean, nearly inexhaustible source of energy. Power production by fusion micro-explosions of inertial confinement fusion (ICF) targets has been a long term research goal since the invention of the first laser in 1960. The NIF is poised to take the next important step in the journey by beginning experiments researching ICF ignition. Ignition on NIF will be the culmination of over thirty years of ICF research on high-powered laser systems such as the Nova laser at LLNL and the OMEGA laser at the University of Rochester as well as smaller systems around the world. NIF is a 192 beam Nd-glass laser facility at LLNL that is more than 90% complete. The first cluster of 48 beams is operational in the laser bay, the second cluster is now being commissioned, and the beam path to the target chamber is being installed. The Project will be completed in 2009 and ignition experiments will start in 2010. When completed NIF will produce up to 1.8 MJ of 0.35 μm light in highly shaped pulses required for ignition. It will have beam stability and control to higher precision than any other laser fusion facility. Experiments using one of the beams of NIF have demonstrated that NIF can meet its beam performance goals. The National Ignition Campaign (NIC) has been established to manage the ignition effort on NIF. NIC has all of the research and development required to execute the ignition plan and to develop NIF into a fully operational facility. NIF will explore the ignition space, including direct drive, 2ω ignition, and fast ignition, to optimize target efficiency for developing fusion as an energy source. In addition to efficient target performance, fusion energy requires significant advances in high repetition rate lasers and fusion reactor technology. The Mercury laser at LLNL is a high repetition rate Nd-glass laser for fusion energy driver development. Mercury uses state-o-the art technology such as ceramic laser slabs and light

Preliminary Options Assessment of Versatile Irradiation Test Reactor

Energy Technology Data Exchange (ETDEWEB)

Sen, Ramazan Sonat [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-01-01

The objective of this report is to summarize the work undertaken at INL from April 2016 to January 2017 and aimed at analyzing some options for designing and building a versatile test reactor; the scope of work was agreed upon with DOE-NE. Section 2 presents some results related to KNK II and PRISM Mod A. Section 3 presents some alternatives to the VCTR presented in [ ] as well as a neutronic parametric study to assess the minimum power requirement needed for a 235U metal fueled fast test reactor capable to generate a fast (>100 keV) flux of 4.0 x 1015 n /cm2-s at the test location. Section 4 presents some results regarding a fundamental characteristic of test reactors, namely displacement per atom (dpa) in test samples. Section 5 presents the INL assessment of the ANL fast test reactor design FASTER. Section 6 presents a summary.

Proceedings of the international symposium on materials testing reactors

International Nuclear Information System (INIS)

Ishihara, Masahiro; Kawamura, Hiroshi

2009-01-01

This report is the Proceedings of the International Symposium on Materials Testing Reactors hosted by Japan Atomic Energy Agency (JAEA). The symposium was held on July 16 to 17, 2008, at the Oarai Research and Development Center of JAEA. This symposium was also held for the 40th anniversary ceremony of Japan Materials Testing Reactor (JMTR) from achieving its first criticality. The objective of the symposium is to exchange the information on current status, future plan and so on among each testing reactors for the purpose of mutual understanding. There were 138 participants from Argentina, Belgium, France, Indonesia, Kazakhstan, Korea, the Russian Federation, Sweden, the United State, Vietnam and Japan. The symposium was divided into four technical sessions and three topical sessions. Technical sessions addressed the general topics of 'status and future plan of materials testing reactors', 'material development for research and testing reactors', irradiation technology (including PIE technology)' and 'utilization with materials testing reactors', and 21 presentations were made. Also the topical sessions addressed 'establishment of strategic partnership', 'management on re-operation work at reactor trouble' and 'basic technology for neutron irradiation tests in MTRs', and panel discussion was made. The 21 of the presented papers are indexed individually. (J.P.N.)

Study of ignition in a high compression ratio SI (spark ignition) methanol engine using LES (large eddy simulation) with detailed chemical kinetics

International Nuclear Information System (INIS)

Zhen, Xudong; Wang, Yang

2013-01-01

Methanol has been recently used as an alternative to conventional fuels for internal combustion engines in order to satisfy some environmental and economical concerns. In this paper, the ignition in a high compression ratio SI (spark ignition) methanol engine was studied by using LES (large eddy simulation) with detailed chemical kinetics. A 21-species, 84-reaction methanol mechanism was adopted to simulate the auto-ignition process of the methanol/air mixture. The MIT (minimum ignition temperature) and MIE (minimum ignition energy) are two important properties for designing safety standards and understanding the ignition process of combustible mixtures. The effects of the flame kernel size, flame kernel temperature and equivalence ratio were also examined on MIT, MIE and IDP (ignition delay period). The methanol mechanism was validated by experimental test. The simulated results showed that the flame kernel size, temperature and energy dramatically affected the values of the MIT, MIE and IDP for a methanol/air mixture, the value of the ignition delay period was not only related to the flame kernel energy, but also to the flame kernel temperature. - Highlights: • We used LES (large eddy simulation) coupled with detailed chemical kinetics to simulate methanol ignition. • The flame kernel size and temperature affected the minimum ignition temperature. • The flame kernel temperature and energy affected the ignition delay period. • The equivalence ratio of methanol–air mixture affected the ignition delay period

The Test Reactor Embrittlement Data Base (TR-EDB)

International Nuclear Information System (INIS)

Stallmann, F.W.; Kam, F.B.K.; Wang, J.A.

1993-01-01

The Test Reactor Embrittlement Data Base (TR-EDB) is part of an ongoing program to collect test data from materials irradiations to aid in the research and evaluation of embrittlement prediction models that are used to assure the safety of pressure vessels in power reactors. This program is being funded by the US Nuclear Regulatory Commission (NRC) and has resulted in the publication of the Power Reactor Embrittlement Data Base (PR-EDB) whose second version is currently being released. The TR-EDB is a compatible collection of data from experiments in materials test reactors. These data contain information that is not obtainable from surveillance results, especially, about the effects of annealing after irradiation. Other information that is only available from test reactors is the influence of fluence rates and irradiation temperatures on radiation embrittlement. The first version of the TR-EDB will be released in fall of 1993 and contains published results from laboratories in many countries. Data collection will continue and further updates will be published

Progress Toward Ignition on the National Ignition Facility

International Nuclear Information System (INIS)

Kauffman, R.L.

2011-01-01

The principal approach to ignition on the National Ignition Facility (NIF) is indirect drive. A schematic of an ignition target is shown in Figure 1. The laser beams are focused through laser entrance holes at each end of a high-Z cylindrical case, or hohlraum. The lasers irradiate the hohlraum walls producing x-rays that ablate and compress the fuel capsule in the center of the hohlraum. The hohlraum is made of Au, U, or other high-Z material. For ignition targets, the hohlraum is ∼0.5 cm diameter by ∼1 cm in length. The hohlraum absorbs the incident laser energy producing x-rays for symmetrically imploding the capsule. The fuel capsule is a ∼2-mm-diameter spherical shell of CH, Be, or C filled with DT fuel. The DT fuel is in the form of a cryogenic layer on the inside of the capsule. X-rays ablate the outside of the capsule, producing a spherical implosion. The imploding shell stagnates in the center, igniting the DT fuel. NIC has overseen installation of all of the hardware for performing ignition experiments, including commissioning of approximately 50 diagnostic systems in NIF. The diagnostics measure scattered optical light, x-rays from the hohlraum over the energy range from 100 eV to 500 keV, and x-rays, neutrons, and charged particles from the implosion. An example of a diagnostic is the Magnetic Recoil Spectrometer (MRS) built by a collaboration of scientists from MIT, UR-LLE, and LLNL shown in Figure 2. MRS measures the neutron spectrum from the implosion, providing information on the neutron yield and areal density that are metrics of the quality of the implosion. Experiments on NIF extend ICF research to unexplored regimes in target physics. NIF can produce more than 50 times the laser energy and more than 20 times the power of any previous ICF facility. Ignition scale hohlraum targets are three to four times larger than targets used at smaller facilities, and the ignition drive pulses are two to five times longer. The larger targets and longer

Tests for validation of fast neutron reactors safety

International Nuclear Information System (INIS)

Nagata, T.; Yamashita, H.

2001-01-01

Japanese scientific research and design enterprises in cooperation with industrial and power generating corporations implement a project on creating a fast neutron reactor of the ultimate safety. One of the basic expected results from such a development is creation of a reactor core structure that is able to eliminate recriticality occurrence in the course of reactor accident involving fuel melting. One of the possible ways to solve this problem is to include pipes (meant for specifying directed (controlled) molten fuel relocation) into fuel assembly structure. In the course of conduction and subsequent implementation of such a design the basic issue is to experimentally confirm the operating capacity of FA having such a structure and that is called FAIDUS. Within EAGLE Project on experimental basis of IAE NNC RK an activity has been started on preparation and conduction of out-of-pile and in-pile tests. During tests a sodium coolant will be used. Studies are conducted by NNC RK in cooperation with the Japanese corporations JAPC and JNC. Basic objective of out-of-pile tests was to obtain preliminary information on fuel relocation behavior under conditions simulating accident involving melting of core consisting of FAIDUS FA, which will help to clarify simulation criteria and to develop the most optimum structure of the experimental channel for reactor experiments conduction. The basic objective of in-pile tests was the experimental confirmation of operating capacity of FAIDUS FA model under reactor conditions. According to the program two tests are planned to be performed at IGR reactor: tests for validation of fast neutron reactor safety, and out-of-pile tests at EAGLE experimental facility without sodium coolant

Fast ignition: Physics progress in the US fusion energy program and prospects for achieving ignition

International Nuclear Information System (INIS)

Key, M.; Andersen, C.; Cowan, T.

2003-01-01

Fast ignition (FI) has significant potential advantages for inertial fusion energy and it is therefore being studied as an exploratory concept in the US fusion energy program. FI is based on short pulse isochoric heating of pre-compressed DT by intense beams of laser accelerated MeV electrons or protons. Recent experimental progress in the study of these two heating processes is discussed. The goal is to benchmark new models in order to predict accurately the requirements for full-scale fast ignition. An overview is presented of the design and experimental testing of a cone target implosion concept for fast ignition. Future prospects and conceptual designs for larger scale FI experiments using planned high energy petawatt upgrades of major lasers in the US are outlined. A long-term road map for FI is defined. (author)

The National Ignition Facility and Industry

Science.gov (United States)

Harri, J. G.; Paisner, J. A.; Lowdermilk, W. H.; Boyes, J. D.; Kumpan, S. A.; Sorem, M. S.

1994-09-01

The mission of the National Ignition Facility is to achieve ignition and gain in inertial confinement fusion targets in the laboratory. The facility will be used for defense applications such as weapons physics and weapons effects testing, and for civilian applications such as fusion energy development and fundamental studies of matter at high temperatures and densities. The National Ignition Facility construction project will require the best of our construction industries and its success will depend on the best products offered by hundreds of the nation's high technology companies. Three-fourths of the construction costs will be invested in industry. This article reviews the design, cost and schedule, and required industrial involvement associated with the construction project.

The National Ignition Facility and industry

International Nuclear Information System (INIS)

Harri, J.G.; Lowdermilk, W.H.; Paisner, J.A.; Boyes, J.D.; Kumpan, S.A.; Sorem, M.S.

1994-01-01

The mission of the National Ignition Facility is to achieve ignition and gain in inertial confinement fusion targets in the laboratory. The facility will be used for defense applications such as weapons physics and weapons effects testing, and for civilian applications such as fusion energy development and fundamental studies of matter at high temperatures and densities. The National Ignition Facility construction project will require the best of national construction industries and its success will depend on the best products offered by hundreds of the nation's high technology companies. Three-fourths of the construction costs will be invested in industry. This article reviews the design, cost and schedule, and required industrial involvement associated with the construction project

ZTI: Preliminary characterization of an ignition class reversed-field pinch

International Nuclear Information System (INIS)

Bathke, C.G.; Krakowski, R.A.; Miller, R.L.; Werley, K.A.

1990-01-01

A preliminary cost-optimized conceptual design of an intermediate-step, ignition-class RFP device (ZTI) for the study of alpha-particle physics in a DT plasma is reported. The ZTI design reflects potentially significant cost savings relative to similar ignition-class tokamaks for device parameters that reside on the path to a viable commercial RFP reactor. Reductions in both device costs and number of steps to commercialization portend a significantly reduced development cost for fusion. The methodology and result and coupling realistic physics, engineering, and cost models through a multi-dimensional optimizer are reported for ZTI, which is a device that would follow the 2--4 MA ZTH on a approx-gt 1996--98 timescale. 15 refs., 7 figs., 2 tabs

Thermionic nuclear reactor systems

International Nuclear Information System (INIS)

Kennel, E.B.

1986-01-01

Thermionic nuclear reactors can be expected to be candidate space power supplies for power demands ranging from about ten kilowatts to several megawatts. The conventional ''ignited mode'' thermionic fuel element (TFE) is the basis for most reactor designs to date. Laboratory converters have been built and tested with efficiencies in the range of 7-12% for over 10,000 hours. Even longer lifetimes are projected. More advanced capabilities are potentially achievable in other modes of operation, such as the self-pulsed or unignited diode. Coupled with modest improvements in fuel and emitter material performance, the efficiency of an advanced thermionic conversion system can be extended to the 15-20% range. Advanced thermionic power systems are expected to be compatible with other advanced features such as: (1) Intrinsic subcritically under accident conditions, ensuring 100% safety upon launch abort; (2) Intrinsic low radiation levels during reactor shutdown, allowing manned servicing and/or rendezvous; (3) DC to DC power conditioning using lightweight power MOSFETS; and (4) AC output using pulsed converters

Influence of interface conditions on laser diode ignition of pyrotechnic mixtures: application to the design of an ignition device

Energy Technology Data Exchange (ETDEWEB)

Opdebeck, Frederic; Gillard, Philippe [Laboratoire Energetique Explosions et Structures de l' Universite d' Orleans, 63 boulevard de Lattre de Tassigny, 18020 cedex, Bourges (France); Radenac, D' Erwann [Laboratoire de combustion et de detonique, ENSMA, BP 109, 86960 cedex, Futuroscope (France)

2003-01-01

This paper treats of numerical modelling which simulates the laser ignition of pyrotechnic mixtures. The computation zone is divided into two fields. The first is used to take account of the heat loss with the outside. It can represent an optical fibre or a sapphire protective porthole. The second field represents the reactive tablet which absorbs the laser diode's beam. A specific feature of the model is that it incorporates a thermal contact resistance R{sub c} between the two computation fields. Through knowledge of the thermal, optical and kinetic properties, this code makes it possible to compute the ignition conditions. The latter are defined by the energy E{sub 50} and the time t{sub i} of ignition of any pyrotechnic mixture and for various ignition systems.This work was validated in the case of an ignition system consisting of a laser diode with an optical lens re-focussing system. The reactive tablet contains 62% by mass of iron and 38% by mass of KClO{sub 4}. Its porosity is 25.8%. After an evaluation of the laser's coefficient of absorption, the variations of the ignition parameters E{sub 50} and t{sub i} are studied as a function of the thermal contact resistance R{sub c}. Temperature profiles are obtained as a function of time and for various values of the thermal contact resistance R{sub c}. More fundamental observations are made concerning the position of the hot spot corresponding to priming. From this study, which concerns the heat exchange between the two media, several practical conclusions are given concerning the design of an ignition device. By evaluation of the thermal contact resistance R{sub c}, comparison with test results becomes possible and the results of the computations are in reasonable agreement with the test measurements. (authors)

EBR-2 [Experimental Breeder Reactor-2], IFR [Integral Fast Reactor] prototype testing programs

International Nuclear Information System (INIS)

Lehto, W.K.; Sackett, J.I.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.

1990-01-01

The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development. (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

Compact torsatron reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Carreras, B.A.; Lynch, V.E.; Tolliver, J.S.; Sviatoslavsky, I.N.

1988-05-01

Low-aspect-ratio torsatron configurations could lead to compact stellarator reactors with R 0 = 8--11m, roughly one-half to one-third the size of more conventional stellarator reactor designs. Minimum-size torsatron reactors are found using various assumptions. Their size is relatively insensitive to the choice of the conductor parameters and depends mostly on geometrical constraints. The smallest size is obtained by eliminating the tritium breeding blanket under the helical winding on the inboard side and by reducing the radial depth of the superconducting coil. Engineering design issues and reactor performance are examined for three examples to illustrate the feasibility of this approach for compact reactors and for a medium-size (R 0 ≅ 4 m,/bar a/ /approx lt/ 1 m) copper-coil ignition experiment. 26 refs., 11 figs., 7 tabs

Licensing experience of the HTR-10 test reactor

International Nuclear Information System (INIS)

Sun, Y.; Xu, Y.

1996-01-01

A 10MW high temperature gas-cooled test reactor (HTR-10) is now being projected by the Institute of Nuclear Energy Technology within China's National High Technology Programme. The Construction Permit of HTR-10 was issued by the Chinese nuclear licensing authority around the end of 1994 after a period of about one year of safety review of the reactor design. HTR-10 is the first high temperature gas-cooled reactor (HTGR) to be constructed in China. The purpose of this test reactor project is to test and demonstrate the technology and safety features of the advanced modular high temperature reactor design. The reactor uses spherical fuel elements with coated fuel particles. The reactor unit and the steam generator unit are arranged in a ''side-by-side'' way. Maximum fuel temperature under the accident condition of a complete loss of coolant is limited to values much lower than the safety limit set for the fuel element. Since the philosophy of the technical and safety design of HTR-10 comes from the high temperature modular reactor design, the reactor is also called the Test Module. HTR-10 represents among others also a licensing challenge. On the one side, it is the first helium reactor in China, and there are less licensing experiences both for the regulator and for the designer. On the other side, the reactor design incorporates many advanced design features in the direction of passive or inherent safety, and it is presently a world-wide issue how to treat properly the passive or inherent safety design features in the licensing safety review. In this presentation, the licensing criteria of HTR-10 are discussed. The organization and activities of the safety review for the construction permit licensing are described. Some of the main safety issues in the licensing procedure are addressed. Among these are, for example, fuel element behaviour, source term, safety classification of systems and components, containment design. The licensing experiences of HTR-10 are of

Theoretical study for ICRF sustained LHD type p-11B reactor

International Nuclear Information System (INIS)

Watanabe, Tsuguhiro

2003-04-01

This is a summary of the workshop on 'Theoretical Study for ICRF Sustained LHD Type p- 11 B Reactor' held in National Institute for Fusion Science (NIFS) on July 25, 2002. In the workshop, study of LHD type D- 3 He reactor is also reported. A review concerning the advanced nuclear fusion fuels is also attached. This review was reported at the workshop of last year. The development of the p- 11 B reactor research which uses the LHD magnetic field configuration has been briefly summarized in section 1. In section 2, an integrated report on advanced nuclear fusion fuels is given. Ignition conditions in a D- 3 He helical reactor are summarized in section 3. 0-dimensional particle and power balance equations are solved numerically assuming the ISS95 confinement law including a confinement factor (γ HH ). It is shown that high average beta plasma confinement, a large confinement factor (γ HH > 3) and the hot ion mode (T i /T e > 1.4) are necessary to achieve the ignition of the D- 3 He helical reactor. Characteristics of ICRF sustained p- 11 B reactor are analyzed in section 4. The nuclear fusion reaction rate is derived assuming a quasilinear plateau distribution function (QPDF) for protons, and an ignition condition of p- 11 B reactor is shown to be possible. The 3 of the presented papers are indexed individually. (J.P.N.)

Shock ignition of thermonuclear fuel: principles and modelling

International Nuclear Information System (INIS)

Atzeni, S.; Ribeyre, X.; Schurtz, G.; Schmitt, A.J.; Canaud, B.; Betti, R.; Perkins, L.J.

2014-01-01

Shock ignition is an approach to direct-drive inertial confinement fusion (ICF) in which the stages of compression and hot spot formation are partly separated. The fuel is first imploded at a lower velocity than in conventional ICF. Close to stagnation, an intense laser spike drives a strong converging shock, which contributes to hot spot formation. Shock ignition shows potentials for high gain at laser energies below 1 MJ, and could be tested on the National Ignition Facility or Laser MegaJoule. Shock ignition principles and modelling are reviewed in this paper. Target designs and computer-generated gain curves are presented and discussed. Limitations of present studies and research needs are outlined. (special topic)

Loss on Ignition Furnace Acceptance and Operability Test Procedure

Energy Technology Data Exchange (ETDEWEB)

JOHNSON, D.C.

2000-06-01

The purpose of this Acceptance Test Procedure and Operability Test Procedure (ATP/OTP)is to verify the operability of newly installed LOI equipment, including a model 1608FL CM{trademark} Furnace, a dessicator, and balance. The operability of the furnace will be verified. The arrangement of the equipment placed in Glovebox 157-3/4 to perform Loss on Ignition (LOI) testing on samples supplied from the Thermal Stabilization line will be verified. In addition to verifying proper operation of the furnace, this ATP/OTP will also verify the air flow through the filters, verify a damper setting to establish and maintain the required differential pressure between the glovebox and the room pressure, and test the integrity of the newly installed HEPA filter. In order to provide objective evidence of proper performance of the furnace, the furnace must heat 15 crucibles, mounted on a crucible rack, to 1000 C, according to a program entered into the furnace controller located outside the glovebox. The glovebox differential pressure will be set to provide the 0.5 to 2.0 inches of water (gauge) negative pressure inside the glovebox with an airflow of 100 to 125 cubic feet per minute (cfm) through the inlet filter. The glovebox inlet Glfilter will he flow tested to ensure the integrity of the filter connections and the efficiency of the filter medium. The newly installed windows and glovebox extension, as well as all disturbed joints, will be sonically tested via ultra probe to verify no leaks are present. The procedure for DOS testing of the filter is found in Appendix A.

Comprehensive study of ignition and combustion of single wooden particles

DEFF Research Database (Denmark)

Momenikouchaksaraei, Maryam; Yin, Chungen; Kær, Søren Knudsen

2013-01-01

How quickly large biomass particles can ignite and burn out when transported into a pulverized-fuel (pf) furnace and suddenly exposed to a hot gas flow containing oxygen is very important in biomass co-firing design and optimization. In this paper, the ignition and burnout of the largest possible...... for all the test conditions. As the particle is further heated up and the volume-weighted average temperature reaches the onset of rapid decomposition of hemicellulose and cellulose, a secondary homogeneous ignition occurs. The model-predicted ignition delays and burnout times show a good agreement...... with the experimental results. Homogeneous ignition delays are found to scale with specific surface areas while heterogeneous ignition delays show less dependency on the areas. The ignition and burnout are also affected by the process conditions, in which the oxygen concentration is found to have a more pronounced...

Testing of research reactor fuel in the high flux reactor (Petten)

International Nuclear Information System (INIS)

Guidez, J.; Markgraf, J.W.; Sordon, G.; Wijtsma, F.J.; Thijssen, P.J.M.; Hendriks, J.A.

1999-01-01

The two types of fuel most frequently used by the main research reactors are metallic: highly enriched uranium (>90%) and silicide low enriched uranium ( 3 . However, a need exists for research on new reactor fuel. This would permit some plants to convert without losses in flux or in cycle length and would allow new reactor projects to achieve higher possibilities especially in fluxes. In these cases research is made either on silicide with higher density, or on other types of fuel (UMo, etc.). In all cases when new fuel is proposed, there is a need, for safety reasons, to test it, especially regarding the mechanical evolution due to burn-up (swelling, etc.). Initially, such tests are often made with separate plates, but lately, using entire elements. Destructive examinations are often necessary. For this type of test, the High Flux Reactor, located in Petten (The Netherlands) has many specific advantages: a large core, providing a variety of interesting positions with high fluence rate; a downward coolant flow simplifies the engineering of the device; there exists easy access with all handling possibilities to the hot-cells; the high number of operating days (>280 days/year), together with the high flux, gives a possibility to reach quickly the high burn-up needs; an experienced engineering department capable of translating specific requirements to tailor-made experimental devices; a well equipped hot-cell laboratory on site to perform all necessary measurements (swelling, γ-scanning, profilometry) and all destructive examinations. In conclusion, the HFR reactor readily permits experimental research on specific fuels used for research reactors with all the necessary facilities on the Petten site. (author)

Conceptual design of a moving-ring reactor

International Nuclear Information System (INIS)

Smith, A.C.; Carlson, G.A.; Ashworth, C.P.

1986-01-01

A design of a prototype moving-ring reactor was completed, and a development plan for a pilot reactor is outlined. The fusion fuel is confined in current-carrying rings of magnetically field-reversed plasma (''compact toroids''). The plasma rings, formed by a coaxial plasma gun, undergo adiabatic magnetic compression to ignition temperature while they are being injected into the reactor's burner section. The cylindrical burner chamber is divided into three ''burn stations.'' Separator coils and a slight axial guide field gradient are used to shuttle the ignited toroids rapidly from one burn station to the next, pausing for one-third of the total burn time at each station. Deuterium-tritium- 3 He ice pellets refuel the rings at a rate that maintains constant radiated power. The fusion power per ring is approx. =105.5 MW. The burn time to reach a fusion energy gain of Q = 30 is 5.9 s

DEALS: a maintainable superconducting magnet system for tokamak fusion reactors

International Nuclear Information System (INIS)

Hseih, S.Y.; Danby, G.; Powell, J.R.

1979-01-01

The feasibility of demountable superconducting magnet systems has been examined in a design study of a DEALS [Demountable Externally Anchored Low Stress] TF magnet for an HFITR [High Field Ignition Test Reactor] Tokamak device. All parts of the system appear feasible, including the demountable superconducting joints. Measurements on small scale prototype joints indicate that movable pressure contact joints exhibit acceptable electrical, mechanical, and cryogenic performance. Such joints permit a relatively simple support structure and are readily demountable. Assembly and disassembly sequences are described whereby any failed portion of the magnet, or any part of the reactor inside the TF coils can be removed and replaced if necessary

An Idea of 20% test of the Initial Core Reactor Physics

International Nuclear Information System (INIS)

Roh, Kyung Ho; Yang, Sung Tae; Jung, Ji Eun

2012-01-01

Many tests have been performed on the OPR1000 and APR1400 before commercial operation. Some of these tests were performed at reactor power levels of 20% and 50%. The CPC (Core Protection Calculator) power distribution test is one of these tests. It is performed to assure the reliability of the Core Protection Calculator System (CPCS). Through this test, SAM1 is calculated using the snapshots2. The test takes about nine hours at a reactor power level of 20% and about thirty hours at a reactor power level of 50%. SAM used at each reactor power level is as follows: 1. Reactor power of 0% ∼ 20%: Designed SAM (DSAM) 2. Reactor power of 20% ∼ 50%: SAM calculated (C-SAM) at a reactor power of 20% 3. Reactor power 50% ∼ End of Cycle : SAM calculated at a reactor power of 50% As mentioned earlier, SAM is calculated and punched into CPC to assure the reliability of CPCS. Therefore, CPC is operated having penalties with D-SAM until3 reaching a reactor power of 20%. That is, the penalty of CPC will be removed when SAM is calculated and punched into the CPC at a reactor power of 20%. But these penalties are considered to be removed after a reactor power of 50% test in order to maintain the conservatism of the CPC. This is done because the final values calculated using C-SAM, in contrast to those calculated using SAM, a reactor power of 50%, are not correct. This paper began from an idea, 'If so, what would happen if we removed the CPC power distribution test at a reactor power of 20%?'

Reactor operator screening test experiences

International Nuclear Information System (INIS)

O'Brien, W.J.; Penkala, J.L.; Witzig, W.F.

1976-01-01

When it became apparent to Duquesne Light Company of Pittsburgh, Pennsylvania, that the throughput of their candidate selection-Phase I training-reactor operator certification sequence was something short of acceptable, the utility decided to ask consultants to make recommendations with respect to candidate selection procedures. The recommendation implemented was to create a Nuclear Training Test that would predict the success of a candidate in completing Phase I training and subsequently qualify for reactor operator certification. The mechanics involved in developing and calibrating the Nuclear Training Test are described. An arbitration decision that resulted when a number of International Brotherhood of Electrical Workers union employees filed a grievance alleging that the selection examination was unfair, invalid, not job related, inappropriate, and discriminatorily evaluated is also discussed. The arbitration decision favored the use of the Nuclear Training Test

The ignition and burning behaviour of sodium metal in air

International Nuclear Information System (INIS)

Newman, R.N.

1983-01-01

The ignition and combustion of sodium, both in terms of the fundamental chemistry and also with reference to its use as the heat transfer fluid of a fast breeder reactor are reviewed. The combustion chemistry and the scientific mechanisms of possible fire extinguishants are compared with the burning of hydrocarbon fluids. Quantitative data produced by various agencies in the world in their pursuit of commercial fast reactor technology is provided. Both practical and theoretical studies have been carried out, some on a large scale, mainly in the field of spray fires and pool fires. Vapour combustion, passive and active fire extinction and possible corrosion damage to structures are discussed. (U.K.)

Reactor physics tests of TRIGA Mark-II Reactor in Ljubljana

International Nuclear Information System (INIS)

Ravnik, M.; Mele, I.; Trkov, A.; Rant, J.; Glumac, B.; Dimic, V.

2008-01-01

TRIGA Mark-II Reactor in Ljubljana was recently reconstructed. The reconstruction consisted mainly of replacing the grid plates, the control rod mechanisms and the control unit. The standard type control rods were replaced by the fuelled follower type, the central grid location (A ring) was adapted for fuel element insertion, the triangular cutouts were introduced in the upper plate design. However, the main novelty in reactor physics and operational features of the reactor was the installation of a pulse rod. Having no previous operational experience in pulsing, a detailed and systematic sequence of tests was defined in order to check the predicted design parameters of the reactor with measurements. The following experiments are treated in this paper: initial criticality, excess reactivity measurements, control rod worth measurement, fuel temperature distribution, fuel temperature reactivity coefficient, pulse parameters measurement (peak power, prompt energy, peak temperature). Flux distributions in steady state and pulse mode were measured as well, however, they are treated only briefly due to the volume of the results. The experiments were performed with completely fresh fuel of 12 w% enriched Standard Stainless Steel type. The core configuration was uniform (one fuel element type, including fuelled followers) and compact (no irradiation channels or gaps), as such being particularly convenient for testing the computer codes for TRIGA reactor calculations. Comparison of analytical predictions, obtained with WIMS, SLXTUS, TRIGAP and PULSTRI codes to measured values showed agreement within the error of the measurement and calculation. The paper has the following contents: 1. Introduction; 2. Steady State Experiments; 2.1. Core loading and critical experiment; 2.2. Flux range determination for tests at zero power; 2.3. Digital reactivity meter checkout; 2.4. Control rod worth measurements; 2.5. Excess reactivity measurement; 2.6. Thermal power calibration; 2

Reliability tests for reactor internals replacement technology

International Nuclear Information System (INIS)

Fujimaki, K.; Uchiyama, J.; Ohtsubo, T.

2000-01-01

Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for replacing reactor internals, which was directed at preventive maintenance before damage and repair after damage for the aging degradation. The project has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995, and it follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the whole test plans and the test results for the BWR reactor internals replacement methods; core shroud, ICM housing, and CRD Housing and stub tube. The test results have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

Development of key technologies in DPSSL system for fast-ignition, laser fusion reactor - FIREX, HALNA, and protection of final optics

International Nuclear Information System (INIS)

Norimatsu, T.; Azechi, H.; Fujimoto, Y.; Jitsuno, T.; Kanabe, T.; Kodama, R.; Kondo, K.; Miyanaga, N.; Nagatomo, H.; Nakatsuka, M.; Shiraga, H.; Tanaka, K.A.; Tsubakimoto, K.; Yamanaka, M.; Yasuhara, R.; Izawa, Y.; Kawashima, T.; Kurita, T.; Matsumoto, O.; Tsuchiya, Y.; Sekine, T.; Kan, H.

2005-01-01

A critical path to a laser fusion power plant is construction of a reliable, efficient, high repetitive energy driver including the relation with the reactor environment. At ILE, Osaka University, FIREX project has been proposed and the phase I to show heating of compressed fuel to 5 keV has started with construction of the FIREX laser. This project will demonstrate physics of fast ignition and elemental studies are carried out to obtain persuasive data to find the path to the goal. A diode-laser-pumped, solid-state-laser (DPSSL) HALNA-10 succeeded in operation of 7.5J output power at 10 Hz rep-rate. Contamination of final optics by metal vapor was studied using a 1/10 model of the beam duct. The result indicated that contamination can be controlled with high speed shutters and a low pressure buffer gas. (author)

Reactor group constants and benchmark test

Energy Technology Data Exchange (ETDEWEB)

Takano, Hideki [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

2001-08-01

The evaluated nuclear data files such as JENDL, ENDF/B-VI and JEF-2 are validated by analyzing critical mock-up experiments for various type reactors and assessing applicability for nuclear characteristics such as criticality, reaction rates, reactivities, etc. This is called Benchmark Testing. In the nuclear calculations, the diffusion and transport codes use the group constant library which is generated by processing the nuclear data files. In this paper, the calculation methods of the reactor group constants and benchmark test are described. Finally, a new group constants scheme is proposed. (author)

Data on test results of vessel cooling system of high temperature engineering test reactor

International Nuclear Information System (INIS)

Saikusa, Akio; Nakagawa, Shigeaki; Fujimoto, Nozomu; Tachibana, Yukio; Iyoku, Tatsuo

2003-02-01

High Temperature Engineering Test Reactor (HTTR) is the first graphite-moderated helium gas cooled reactor in Japan. The rise-to-power test of the HTTR started on September 28, 1999 and thermal power of the HTTR reached its full power of 30 MW on December 7, 2001. Vessel Cooling System (VCS) of the HTTR is the first Reactor Cavity Cooling System (RCCS) applied for High Temperature Gas Cooled Reactors. The VCS cools the core indirectly through the reactor pressure vessel to keep core integrity during the loss of core flow accidents such as depressurization accident. Minimum heat removal of the VCS to satisfy its safety requirement is 0.3MW at 30 MW power operation. Through the performance test of the VCS in the rise-to-power test of the HTTR, it was confirmed that the VCS heat removal at 30 MW power operation was higher than 0.3 MW. This paper shows outline of the VCS and test results on the VCS performance. (author)

Numerical simulation of a liquid droplet combustion experiment focusing on ignition process

International Nuclear Information System (INIS)

Yamaguchi, Akira; Tajima, Yuji

1999-11-01

SPHINCS (Sodium Fire phenomenology IN multi-Cell System) computer program has been developed for the safety analysis of sodium fire accident in a Fast Breeder Reactor. The program can deal with spray combustion and pool surface combustion. In this report the authors investigate a single droplet combustion phenomena focusing on an ignition process. The spray combustion model of SPHINCS is as follows. The liquid droplet-burning rate after ignition is based on the D-square law and a diffusion flame assumption. Before the droplet is ignited, the burning rate is evaluated by mass flux of oxidizer gases. Forced convection effect that skews the sphere shape of the flame zone surrounding a droplet is taken into consideration. It enhances the burning rate. The chemical equilibrium theory is used to determine the resultant fraction of reaction products of Na-O 2 -H 2 O system. It is noted that users have to give an ignition temperature based on empirical evidences. According to this model, it is obvious that a smaller liquid droplet with higher initial temperature tends to burn more easily. What is observed in a recent experiment is that the smallest liquid droplet (2mm diameter) did not ignited of itself and larger droplets (3.7mm and 4.5mm diameter) burnt at 300degC initial temperature. The current model for liquid droplet combustion cannot predict the experimental results. Therefore, in the present study, a surface reaction model has been developed to predict the ignition process. The model has been used to analyze a combustion experiment of a stationary liquid droplet. The authors investigate the validity of the physical modeling of the liquid droplet combustion and surface reaction. It has been found, as the results, that the model can predict the influence of the initial temperature on the temperature lower limit for spontaneous ignition and ignition delay time. Also investigated is the influence of the moisture on the ignition phenomena. From the present study, it has

Tests of vacuum interrupters for the Tokamak Fusion Test Reactor

International Nuclear Information System (INIS)

Warren, R.; Parsons, M.; Honig, E.; Lindsay, J.

1979-04-01

The Tokamak Fusion Test Reactor (TFTR) project at Princeton University requires the insertion of a resistor in an excited ohmic-heating coil circuit to produce a plasma initiation pulse (PIP). It is expected that the maximum duty for the switching system will be an interruption of 24 kA with an associated recovery voltage of 25 kV. Vacuum interrupters were selected as the most economical means to satisfy these requirements. However, it was felt that some testing of available systems should be performed to determine their reliability under these conditions. Two interrupter systems were tested for over 1000 interruptions each at 24 kA and 25 kV. One system employed special Westinghouse type WL-33552 interrupters in a circuit designed by LASL. This circuit used a commercially available actuator and a minimum size counterpulse bank and saturable reactor. The other used Toshiba type VGB2-D20 interrupters actuated by a Toshiba mechanism in a Toshiba circuit using a larger counterpulse bank and saturable reactor

Drop-in capsule testing of plutonium-based fuels in the Advanced Test Reactor

International Nuclear Information System (INIS)

Chang, G.S.; Ryskamp, J.M.; Terry, W.K.; Ambrosek, R.G.; Palmer, A.J.; Roesener, R.A.

1996-09-01

The most attractive way to dispose of weapons-grade plutonium (WGPu) is to use it as fuel in existing light water reactors (LWRs) in the form of mixed oxide (MOX) fuel - i.e., plutonia (PuO[sub 2]) mixed with urania (UO[sub 2]). Before U.S. reactors could be used for this purpose, their operating licenses would have to be amended. Numerous technical issues must be resolved before LWR operating licenses can be amended to allow the use of MOX fuel. The proposed weapons-grade MOX fuel is unusual, even relative to ongoing foreign experience with reactor-grade MOX power reactor fuel. Some demonstration of the in- reactor thermal, mechanical, and fission gas release behavior of the prototype fuel will most likely be required in a limited number of test reactor irradiations. The application to license operation with MOX fuel must be amply supported by experimental data. The Advanced Test Reactor (ATR) at the Idaho National Engineering Laboratory (INEL) is capable of playing a key role in the irradiation, development, and licensing of these new fuel types. The ATR is a 250- MW (thermal) LWR designed to study the effects of intense radiation on reactor fuels and materials. For 25 years, the primary role of the ATR has been to serve in experimental investigations for the development of advanced nuclear fuels. Both large- and small-volume test positions in the ATR could be used for MOX fuel irradiation. The ATR would be a nearly ideal test bed for developing data needed to support applications to license LWRs for operation with MOX fuel made from weapons-grade plutonium. Furthermore, these data can be obtained more quickly by using ATR instead of testing in a commercial LWR. Our previous work in this area has demonstrated that it is technically feasible to perform MOX fuel testing in the ATR. This report documents our analyses of sealed drop-in capsules containing plutonium-based test specimens placed in various ATR positions

Automated testing of reactor protection instrumentation made easy

International Nuclear Information System (INIS)

Iborra, A.; De Marcos, F.; Pastor, J.A.; Alvarez, B.; Jimenez, A.; Mesa, E.; Alsonso, L.; Regidor, J.J.

1997-01-01

Maintenance and testing of reactor protection systems is an important cause of unplanned reactor trips. Automated testing is the answer because it minimises test times and reduces human error. The GAMA I system, developed and implemented at Vandellos II in Spain, has the added advantage that it uses visual programming, which means that changing the software does not need specialist programming skills. (author)

Loss on Ignition Furnace Acceptance and Operability Test Procedure

International Nuclear Information System (INIS)

JOHNSTON, D.C.

2000-01-01

The purpose of this Acceptance Test Procedure and Operability Test Procedure (ATP/OTP)is to verify the operability of newly installed Loss on Ignition (LOI) equipment, including a model 1608FL CMTM Furnace, a dessicator, and balance. The operability of the furnace will be verified. The arrangement of the equipment placed in Glovebox 157-3/4 to perform LOI testing on samples supplied from the Thermal Stabilization line will be verified. In addition to verifying proper operation of the furnace, this ATP/OTP will also verify the air flow through the filters, verify a damper setting to establish and maintain the required differential pressure between the glovebox and the room pressure, and test the integrity of the newly installed HEPA filter. In order to provide objective evidence of proper performance of the furnace, the furnace must heat 15 crucibles, mounted on a crucible rack, to 1000 C, according to a program entered into the furnace controller located outside the glovebox. The glovebox differential pressure will be set to provide the 0.5 to 2.0 inches of water (gauge) negative pressure inside the glovebox with an expected airflow of 100 to 125 cubic feet per minute (cfm) through the inlet filter. The glovebox inlet G1 filter will be flow tested to ensure the integrity of the filter connections and the efficiency of the filter medium. The newly installed windows and glovebox extension, as well as all disturbed joints, will be sonically tested via ultra probe to verify no leaks are present. The procedure for DOS testing of the filter is found in Appendix A

Technology Development of a Fiber Optic-Coupled Laser Ignition System for Multi-Combustor Rocket Engines

Science.gov (United States)

Trinh, Huu P.; Early, Jim; Osborne, Robin; Thomas, Matthew E.; Bossard, John A.

2002-01-01

This paper addresses the progress of technology development of a laser ignition system at NASA Marshall Space Flight Center (MSFC). The first two years of the project focus on comprehensive assessments and evaluations of a novel dual-pulse laser concept, flight- qualified laser system, and the technology required to integrate the laser ignition system to a rocket chamber. With collaborations of the Department of Energy/Los Alamos National Laboratory (LANL) and CFD Research Corporation (CFDRC), MSFC has conducted 26 hot fire ignition tests with lab-scale laser systems. These tests demonstrate the concept feasibility of dual-pulse laser ignition to initiate gaseous oxygen (GOX)/liquid kerosene (RP-1) combustion in a rocket chamber. Presently, a fiber optic- coupled miniaturized laser ignition prototype is being implemented at the rocket chamber test rig for future testing. Future work is guided by a technology road map that outlines the work required for maturing a laser ignition system. This road map defines activities for the next six years, with the goal of developing a flight-ready laser ignition system.

Study of the pitting effects during the pre-ignition plasma–propellant interaction process

International Nuclear Information System (INIS)

Hang, Yuhua; Li, Xingwen; Wu, Jian; Jia, Shenli; Zhao, Weiyu; Murphy, Anthony B

2016-01-01

The propellant ignition mechanism has become a central issue in the electrothermal chemical (ETC) launch technology, and the pre-ignition plasma–propellant interactions are critical in determining the ignition characteristics. In this work, both an open-air ablation test and an interrupted burning test are conducted for three different propellants. A fused silica window, which is transparent in all relevant wavelengths, is utilized to investigate the role of the plasma radiation. Surface pitting of the propellants after interaction with the plasma is analyzed using a scanning electron microscope (SEM). The effect of pits on the plasma ignition is then studied and a possible formation mechanism of pits is proposed. The input heat flux and the surface temperature of the propellants are obtained by solving a pre-ignition plasma–propellant interaction model. The results shed light on the pre-ignition plasma ignition mechanisms and will assist in the development of propellants for an ETC launcher. (paper)

The National Ignition Facility Project. Revision 1

International Nuclear Information System (INIS)

Paisner, J.A.; Campbell, E.M.; Hogan, W.J.

1994-01-01

The mission of the National Ignition Facility is to achieve ignition and gain in inertial confinement fusion targets in the laboratory. The facility will be used for defense applications such as weapons physics and weapons effects testing, and for civilian applications such as fusion energy development and fundamental studies of matter at high temperatures and densities. This paper reviews the design, schedule, and costs associated with the construction project

Hydrogen/Oxygen Reactions at High Pressures and Intermediate Temperatures: Flow Reactor Experiments and Kinetic Modeling

DEFF Research Database (Denmark)

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

A series of experimental and numerical investigations into hydrogen oxidation at high pressures and intermediate temperatures has been conducted. The experiments were carried out in a high pressure laminar flow reactor at 50 bar pressure and a temperature range of 600–900 K. The equivalence ratio......, the mechanism is used to simulate published data on ignition delay time and laminar burning velocity of hydrogen. The flow reactor results show that at reducing, stoichiometric, and oxidizing conditions, conversion starts at temperatures of 750–775 K, 800–825 K, and 800–825 K, respectively. In oxygen atmosphere......, ignition occurs at the temperature of 775–800 K. In general, the present model provides a good agreement with the measurements in the flow reactor and with recent data on laminar burning velocity and ignition delay time....

Dual coil ignition system

Energy Technology Data Exchange (ETDEWEB)

Huberts, Garlan J.; Qu, Qiuping; Czekala, Michael Damian

2017-03-28

A dual coil ignition system is provided. The dual coil ignition system includes a first inductive ignition coil including a first primary winding and a first secondary winding, and a second inductive ignition coil including a second primary winding and a second secondary winding, the second secondary winding connected in series to the first secondary winding. The dual coil ignition system further includes a diode network including a first diode and a second diode connected between the first secondary winding and the second secondary winding.

Operation and control of high density tokamak reactors

International Nuclear Information System (INIS)

Attenberger, S.E.; McAlees, D.G.

1976-01-01

The incentive for high density operation of a tokamak reactor is discussed. The plasma size required to attain ignition is determined. Ignition is found to be possible in a relatively small system provided other design criteria are met. These criteria are described and the technology developments and operating procedures required by them are outlined. The parameters for such a system and its dynamic behavior during the operating cycle are also discussed

Conceptual design for simulator of irradiation test reactors

International Nuclear Information System (INIS)

Takemoto, Noriyuki; Ohto, Tsutomu; Magome, Hirokatsu; Izumo, Hironobu; Hori, Naohiko

2012-03-01

A simulator of irradiation test reactors has been developed since JFY 2010 for understanding reactor behavior and for upskilling in order to utilize a nuclear human resource development (HRD) and to promote partnership with developing countries which have a plan to introduce nuclear power plant. The simulator is designed based on the JMTR, one of the irradiation test reactors, and it simulates operation, irradiation tests and various kinds of accidents caused by the reactor and irradiation facility. The development of the simulator is sponsored by the Japanese government as one of the specialized projects of advanced research infrastructure in order to promote basic as well as applied researches. The training using the simulator will be started for the nuclear HRD from JFY 2012. This report summarizes the result of the conceptual design of the simulator in JFY 2010. (author)

Heat Pipe Reactor Dynamic Response Tests: SAFE-100 Reactor Core Prototype

Science.gov (United States)

Bragg-Sitton, Shannon M.

2005-01-01

The SAFE-I00a test article at the NASA Marshall Space Flight Center was used to simulate a variety of potential reactor transients; the SAFEl00a is a resistively heated, stainless-steel heat-pipe (HP)-reactor core segment, coupled to a gas-flow heat exchanger (HX). For these transients the core power was controlled by a point kinetics model with reactivity feedback based on core average temperature; the neutron generation time and the temperature feedback coefficient are provided as model inputs. This type of non-nuclear test is expected to provide reasonable approximation of reactor transient behavior because reactivity feedback is very simple in a compact fast reactor (simple, negative, and relatively monotonic temperature feedback, caused mostly by thermal expansion) and calculations show there are no significant reactivity effects associated with fluid in the HP (the worth of the entire inventory of Na in the core is .tests, the point kinetics model was based on core thermal expansion via deflection measurements. It was found that core deflection was a strung function of how the SAFE-100 modules were fabricated and assembled (in terms of straightness, gaps, and other tolerances). To remove the added variable of how this particular core expands as compared to a different concept, it was decided to use a temperature based feedback model (based on several thermocouples placed throughout the core).

Tests of the RBMK-1500 reactor fuel assemblies in the Leningrad reactor

International Nuclear Information System (INIS)

Aden, V.C.; Varovin, I.A.; Vorontsov, B.A.

1981-01-01

Test of fuel assemblies of the RBMK-1500 reactor is conducted in the reactor of the Leningrad NPP unit 2 for proving the calculational values of critical power of the RBMK-1500 reactor fuel assemblies adopted in design. The experiment presupposes the maximal approximation of the fuel assembly operation parameters to the calculational critical parameters without bringing into the mode of heat transfer crisis. The experiments are carried out at 500, 850 and 900 MW(el) of the reactor. The maximal channel power made up 472 kW at 20.5 t/h coolant flow rate and 49% mass steam content at the outlet of the channel. It was concluded that there was supply up to the heat transfer crisis in all the investigated modes. Data of temperature measurings of the fuel element cans, readings of the devices of the failure control system of the fuel element cans and external inspection of the assemblies after the tests testify to it [ru

Review of Transient Fuel Test Results at Sandia National Laboratories and the Potential for Future Fast Reactor Fuel Transient Testing in the Annular Core Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Wright, Steven A.; Pickard, Paul S.; Parma, Edward J.; Vernon, Milton E.; Kelly, John; Tikare, Veena [Sandia National Laboratories, Org 6872 MS-1146, PO Box 5800 Albuquerque, New Mexico 87185 (United States)

2009-06-15

Reactor driven transient tests of fast reactor fuels may be required to support the development and certification of new fuels for Fast Reactors. The results of the transient fuel tests will likely be needed to support licensing and to provide validation data to support the safety case for a variety of proposed fast fuel types and reactors. In general reactor driven transient tests are used to identify basic phenomenology during reactor transients and to determine the fuel performance limits and margins to failure during design basis accidents such as loss of flow, loss of heat sink, and reactivity insertion accidents. This paper provides a summary description of the previous Sandia Fuel Disruption and Transient Axial Relocation tests that were performed in the Annular Core Research Reactor (ACRR) for the U.S. Nuclear Regulatory Commission almost 25 years ago. These tests consisted of a number of capsule tests and flowing gas tests that used fission heating to disrupt fresh and irradiated MOX fuel. The behavior of the fuel disruption, the generation of aerosols and the melting and relocation of fuel and cladding was recorded on high speed cinematography. This paper will present videos of the fuel disruption that was observed in these tests which reveal stark differences in fuel behavior between fresh and irradiated fuel. Even though these tests were performed over 25 years ago, their results are still relevant to today's reactor designs. These types of transient tests are again being considered by the Advanced Fuel Cycle Initiative to support the Global Nuclear Energy Partnership because of the need to perform tests on metal fuels and transuranic fuels. Because the Annular Core Research Reactor is the only transient test facility available within the US, a brief summary of Sandia's continued capability to perform these tests in the ACRR will also be provided. (authors)

Reliability tests for reactor internals rejuvenation technology

International Nuclear Information System (INIS)

Fujimaki, Katsumi; Hitoki, Yoichi; Otsubo, Toru; Uchiyama, Junichi

1998-01-01

Structural damage due to aging degradation of LWR reactor internals has been reported in several nuclear plants. NUPEC has started a project to test the reliability of the technology for rejuvenating reactor internals which has been funded by the Ministry of International Trade and Industry (MITI) of Japan since 1995. The project follows the policy of a report that the MITI has formally issued in April 1996 summarizing the countermeasures to be considered for aging nuclear plants and equipment. This paper gives an outline of the test plans and results which are directed at preventive maintenance before damage and repair after damage for reactor internals aging degradation. The test results for the replacement methods of ICM housing and BWR core shroud have shown that the methods were reliable and the structural integrity was appropriate based on the evaluation. (author)

Implications of fusion results for a reactor: a proposed next step device-JIT

International Nuclear Information System (INIS)

Rebut, P.H.

1989-01-01

Simulations with a critical-temperature model have been made of proposed future devices (NET, ITER, JIT, etc.). These show that only machines with a current capability of ∼ 30MA have a sufficient ignition domain to cope with more realistic operating conditions (i.e. taking into account sawteeth effects, impurity dilution and semi-continuous operation). The importance of dilution and Bremsstrahlung radiation are clearly demonstrated; a mean temperature > 7keV is required for ignition. This prevents higher field, lower current devices from reaching ignition. Transient operations with monster sawteeth or H-mode allow such devices (>30MA) to reach ignition at lower density without additional heating. To investigate the problems of a controlled burning plasma for days in semi-continuous operation, the plasma of the next-step tokamak should be similar in size and performance to an energy producing reactor. The scientific and technical aims of such a machine should be to study burning plasma, test wall technology, provide a test-bed for breeding blankets and most importantly to demonstrate the potential and viability of fusion as an energy source. The main design characteristics of a Thermonuclear Furnace-JIT-dedicated to these objectives are presented. Watercooled copper magnets are used to benefit from proven technology. A single-null divertor configuration ensures helium exhaust and possibly benefits from an H-mode to reach the ignition domain. The X-point position relative to the dump plates would be swept to limit wall loading

Experiences in stability testing of boiling water reactors

International Nuclear Information System (INIS)

March-Leuba, J.; Otaduy, P.J.

1986-01-01

The purpose of this paper is to summarize experiences with boiling water reactor (BWR) stability testing using noise analysis techniques. These techniques have been studied over an extended period of time, but it has been only recently that they have been well established and generally accepted. This paper contains first a review of the problem of BWR neutronic stability, focusing on its physical causes and its effects on reactor operation. The paper also describes the main techniques used to quantify, from noise measurements, the reactor's stability in terms of a decay ratio. Finally, the main results and experiences obtained from the stability tests performed at the Dresden and the Browns Ferry reactors using noise analysis techniques are summarized

Safety requirements, facility user needs, and reactor concepts for a new Broad Application Test Reactor

International Nuclear Information System (INIS)

Ryskamp, J.M.; Liebenthal, J.L.; Denison, A.B.; Fletcher, C.D.

1992-07-01

This report describes the EG ampersand G Laboratory Directed Research and Development Program (LDRD) Broad Application Test Reactor (BATR) Project that was conducted in fiscal year 1991. The scope of this project was divided into three phases: a project process definition phase, a requirements development phase, and a preconceptual reactor design and evaluation phase. Multidisciplinary teams of experts conducted each phase. This report presents the need for a new test reactor, the project process definition, a set of current and projected regulatory compliance and safety requirements, a set of facility user needs for a broad range of projected testing missions, and descriptions of reactor concepts capable of meeting these requirements. This information can be applied to strategic planning to provide the Department of Energy with management options

Fast reactor safety testing in Transient Reactor Test (TREAT) in the 1980s

International Nuclear Information System (INIS)

Wright, A.E.; Dutt, D.S.; Harrison, L.J.

1990-01-01

Several series of fast reactor safety tests were performed in TREAT during the 1980s. These focused on the transient behavior of full-length oxide fuels (US reference, UK reference, and US advanced design) and on modern metallic fuels. Most of the tests addressed fuel behavior under transient overpower or loss-of-flow conditions. The test series were the PFR/TREAT tests; the RFT, TS, CDT, and RX series on oxide fuels; and the M series on metallic fuels. These are described in terms of their principal results and relevance to analyses and safety evaluation. 4 refs., 3 tabs

Operating experiences since rise-to-power test in high temperature engineering test reactor (HTTR)

International Nuclear Information System (INIS)

Tochio, Daisuke; Watanabe, Shuji; Motegi, Toshihiro; Kawano, Shuichi; Kameyama, Yasuhiko; Sekita, Kenji; Kawasaki, Kozo

2007-03-01

The rise-to-power test of the High Temperature Engineering Test Reactor (HTTR) was actually started in April 2000. The rated thermal power of 30MW and the rated reactor outlet coolant temperature of 850degC were achieved in the middle of Dec. 2001. After that, the reactor thermal power of 30MW and the reactor outlet coolant temperature of 950degC were achieved in the final rise-to-power test in April 2004. After receiving the operation licensing at 850degC, the safety demonstration tests have conducted to demonstrate inherent safety features of the HTGRs as well as to obtain the core and plant transient data for validation of safety analysis codes and for establishment of safety design and evaluation technologies. This paper summarizes the HTTR operating experiences for six years from start of the rise-to-power test that are categorized into (1) Operating experiences related to advanced gas-cooled reactor design, (2) Operating experiences for improvement of the performance, (3) Operating experiences due to fail of system and components. (author)

Dark Matter Ignition of Type Ia Supernovae.

Science.gov (United States)

Bramante, Joseph

2015-10-02

Recent studies of low redshift type Ia supernovae (SN Ia) indicate that half explode from less than Chandrasekhar mass white dwarfs, implying ignition must proceed from something besides the canonical criticality of Chandrasekhar mass SN Ia progenitors. We show that 1-100 PeV mass asymmetric dark matter, with imminently detectable nucleon scattering interactions, can accumulate to the point of self-gravitation in a white dwarf and collapse, shedding gravitational potential energy by scattering off nuclei, thereby heating the white dwarf and igniting the flame front that precedes SN Ia. We combine data on SN Ia masses with data on the ages of SN Ia-adjacent stars. This combination reveals a 2.8σ inverse correlation between SN Ia masses and ignition ages, which could result from increased capture of dark matter in 1.4 vs 1.1 solar mass white dwarfs. Future studies of SN Ia in galactic centers will provide additional tests of dark-matter-induced type Ia ignition. Remarkably, both bosonic and fermionic SN Ia-igniting dark matter also resolve the missing pulsar problem by forming black holes in ≳10  Myr old pulsars at the center of the Milky Way.

Material and geometry options and performance characteristics for a test reactor

International Nuclear Information System (INIS)

Jahshan, S.N.; Fletcher, C.D.; Terry, W.K.

1993-01-01

For the past 3 yr, an Idaho National Engineering Laboratory (INEL) design team has studied design options for a new test reactor to provide continued testing services after several aging test reactors in the United States are decommissioned. This new reactor, the Broad Application Test Reactor (BATR), would also fill other currently unmet needs, such as medical isotope production and space reactor component testing. Consideration of user needs, safety requirements, developmental uncertainties, and other factors led to the selection of an evolutionary design with plate fuel and several independently cooled test loops. The fuel would be cooled by light water, but most neutron moderation would come from heavy water or beryllium. The BATR design was tentatively scaled to the Advanced Test Reactor (ATR), an existing reactor at INEL: The power output of BATR is 250 MW(thermal), and the active core heights is 1 m. For safety in loss-of-flow events, the coolant flows upward through the core. The BATR design has one large test loop (with a test space diameter of 15.0 cm) along the central axis of the core and six smaller test loops (with test space diameters of 8.0 cm) centered at 6-deg azimuthal intervals on a 24.71-cm-diam circle around the central core axis

The behavior of fission products during nuclear rocket reactor tests

International Nuclear Information System (INIS)

Bokor, P.C.; Kirk, W.L.; Bohl, R.J.

1991-01-01

Fission product release from nuclear rocket propulsion reactor fuel is an important consideration for nuclear rocket development and application. Fission product data from the last six reactors of the Rover program are collected in this paper to provide as basis for addressing development and testing issues. Fission product loss from the fuel will depend on fuel composition and reactor design and operating parameters. During ground testing, fission products can be contained downstream of the reactor. The last Rover reactor tested, the Nuclear Furnance, was mated to an effluent clean-up system that was effective in preventing the discharge of fission products into the atmosphere

EBR-2 [Experimental Breeder Reactor-2] test programs

International Nuclear Information System (INIS)

Sackett, J.I.; Lehto, W.K.; Lindsay, R.W.; Planchon, H.P.; Lambert, J.D.B.; Hill, D.J.

1990-01-01

The Experimental Breeder Reactor-2 (EBR-2) is a sodium cooled power reactor supplying about 20 MWe to the Idaho National Engineering Laboratory (INEL) grid and, in addition, is the key component in the development of the Integral Fast Reactor (IFR). EBR-2's testing capability is extensive and has seen four major phases: (1) demonstration of LMFBR power plant feasibility, (2) irradiation testing for fuel and material development, (3) testing the off-normal performance of fuel and plant systems and (4) operation as the IFR prototype, developing and demonstrating the IFR technology associated with fuel and plant design. Specific programs being carried out in support of the IFR include advanced fuels and materials development, advanced control system development, plant diagnostics development and component testing. This paper discusses EBR-2 as the IFR prototype and the associated testing programs. 29 refs

Effect of biomass blending on coal ignition and burnout during oxy-fuel combustion

Energy Technology Data Exchange (ETDEWEB)

B. Arias; C. Pevida; F. Rubiera; J.J. Pis [Instituto Nacional del Carbon, CSIC, Oviedo (Spain)

2008-09-15

Oxy-fuel combustion is a GHG abatement technology in which coal is burned using a mixture of oxygen and recycled flue gas, to obtain a rich stream of CO{sub 2} ready for sequestration. An entrained flow reactor was used in this work to study the ignition and burnout of coals and blends with biomass under oxy-fuel conditions. Mixtures of CO{sub 2}/O{sub 2} of different concentrations were used and compared with air as reference. A worsening of the ignition temperature was detected in CO{sub 2}/O{sub 2} mixtures when the oxygen concentration was the same as that of the air. However, at an oxygen concentration of 30% or higher, an improvement in ignition was observed. The blending of biomass clearly improves the ignition properties of coal in air. The burnout of coals and blends with a mixture of 79%CO{sub 2}-21%O{sub 2} is lower than in air, but an improvement is achieved when the oxygen concentration is 30 or 35%. The results of this work indicate that coal burnout can be improved by blending biomass in CO{sub 2}/O{sub 2} mixtures. 26 refs., 7 figs., 1 tab.

Spontaneous ignition in afterburner segment tests at an inlet temperature of 1240 K and a pressure of 1 atmosphere with ASTM jet-A fuel

Science.gov (United States)

Schultz, D. F.; Branstetter, J. R.

1973-01-01

A brief testing program was undertaken to determine if spontaneous ignition and stable combustion could be obtained in a jet engine afterburning operating with an inlet temperature of 1240 K and a pressure of 1 atmosphere with ASTM Jet-A fuel. Spontaneous ignition with 100-percent combustion efficiency and stable burning was obtained using water-cooled fuel spraybars as flameholders.

Radiation analysis of the CIT (Compact Ignition Tokamak) pellet injector system and its impact on personnel access

Energy Technology Data Exchange (ETDEWEB)

Selcow, E.C.; Stevens, P.N.; Gomes, I.C.; Gomes, L.M.

1987-01-01

Conceptual design of the Compact Ignition Tokamak (CIT) is near completion. This short-pulse ignition experiment is planned to follow the operations of the Tokamak Fusion Test Reactor (TFTR) at the Princeton Plasma Physics Laboratory. The high neutron wall loadings, /approximately/4-5 MW/m/sup 2/, associated with the operation of this device require that neutronics-related issues be considered in the overall system design. Radiation shielding is required for the protection of device components and personnel. A close-in igloo shield has been designed around the periphery of the tokamak structure, and the entire experiment is housed in a circular test cell facility with a radius of /approximately/12 m. The most critical radiation concern in the CIT design process relates to the numerous penetrations in the device. This paper discusses the impact of a major penetration on the design and operations of the CIT pellet injection system. The pellet injector is a major component, which has a line-of-sight penetration through the igloo and test cell wall. All current options for maintenance of the injector require personnel access. A nuclear analysis has been performed to determine the feasibility of hands-on access. Results indicate that personnel access to the pellet injector glovebox is possible. 10 refs., 3 figs., 3 tabs.

Spark Ignition Characteristics of a L02/LCH4 Engine at Altitude Conditions

Science.gov (United States)

Kleinhenz, Julie; Sarmiento, Charles; Marshall, William

2012-01-01

The use of non-toxic propellants in future exploration vehicles would enable safer, more cost effective mission scenarios. One promising "green" alternative to existing hypergols is liquid methane/liquid oxygen. To demonstrate performance and prove feasibility of this propellant combination, a 100lbf LO2/LCH4 engine was developed and tested under the NASA Propulsion and Cryogenic Advanced Development (PCAD) project. Since high ignition energy is a perceived drawback of this propellant combination, a test program was performed to explore ignition performance and reliability versus delivered spark energy. The sensitivity of ignition to spark timing and repetition rate was also examined. Three different exciter units were used with the engine s augmented (torch) igniter. Propellant temperature was also varied within the liquid range. Captured waveforms indicated spark behavior in hot fire conditions was inconsistent compared to the well-behaved dry sparks (in quiescent, room air). The escalating pressure and flow environment increases spark impedance and may at some point compromise an exciter s ability to deliver a spark. Reduced spark energies of these sparks result in more erratic ignitions and adversely affect ignition probability. The timing of the sparks relative to the pressure/flow conditions also impacted the probability of ignition. Sparks occurring early in the flow could trigger ignition with energies as low as 1-6mJ, though multiple, similarly timed sparks of 55-75mJ were required for reliable ignition. An optimum time interval for spark application and ignition coincided with propellant introduction to the igniter and engine. Shifts of ignition timing were manifested by changes in the characteristics of the resulting ignition.

Spark Ignition Characteristics of a LO2/LCH4 Engine at Altitude Conditions

Science.gov (United States)

Kleinhenz, Julie; Sarmiento, Charles; Marshall, William

2012-01-01

The use of non-toxic propellants in future exploration vehicles would enable safer, more cost effective mission scenarios. One promising "green" alternative to existing hypergols is liquid methane/liquid oxygen. To demonstrate performance and prove feasibility of this propellant combination, a 100lbf LO2/LCH4 engine was developed and tested under the NASA Propulsion and Cryogenic Advanced Development (PCAD) project. Since high ignition energy is a perceived drawback of this propellant combination, a test program was performed to explore ignition performance and reliability versus delivered spark energy. The sensitivity of ignition to spark timing and repetition rate was also examined. Three different exciter units were used with the engine's augmented (torch) igniter. Propellant temperature was also varied within the liquid range. Captured waveforms indicated spark behavior in hot fire conditions was inconsistent compared to the well-behaved dry sparks (in quiescent, room air). The escalating pressure and flow environment increases spark impedance and may at some point compromise an exciter.s ability to deliver a spark. Reduced spark energies of these sparks result in more erratic ignitions and adversely affect ignition probability. The timing of the sparks relative to the pressure/flow conditions also impacted the probability of ignition. Sparks occurring early in the flow could trigger ignition with energies as low as 1-6mJ, though multiple, similarly timed sparks of 55-75mJ were required for reliable ignition. An optimum time interval for spark application and ignition coincided with propellant introduction to the igniter and engine. Shifts of ignition timing were manifested by changes in the characteristics of the resulting ignition.

Current and prospective fuel test programmes in the MIR reactor

Energy Technology Data Exchange (ETDEWEB)

Izhutov, A.L.; Burukin, A.V.; Iljenko, S.A.; Ovchinnikov, V.A.; Shulimov, V.N.; Smirnov, V.P. [State Scientific Centre of Russia Research Institute of Atomic Reactors, Ulyanovsk region (Russian Federation)

2007-07-01

MIR reactor is a heterogeneous thermal reactor with a moderator and a reflector made of metal beryllium, it has a channel-type design and is placed in a water pool. MIR reactor is mainly designed for testing fragments of fuel elements and fuel assemblies (FA) of different nuclear power reactor types under normal (stationary and transient) operating conditions as well as emergency situations. At present six test loop facilities are being operated (2 PWR loops, 2 BWR loops and 2 steam coolant loops). The majority of current fuel tests is conducted for improving and upgrading the Russian PWR fuel, these tests involve issues such as: -) long term tests of short-size rods with different modifications of cladding materials and fuel pellets; -) further irradiation of power plant re-fabricated and full-size fuel rods up to achieving 80 MW*d/kg U; -) experiments with leaking fuel rods at different burnups and under transient conditions; -) continuation of the RAMP type experiments at high burnup of fuel; and -) in-pile tests with simulation of LOCA and RIA type accidents. Testing of the LEU (low enrichment uranium) research reactor fuel is conducted within the framework of the RERTR programme. Upgrading of the gas cooled and steam cooled loop facilities is scheduled for testing the HTGR fuel and sub-critical water-cooled reactor, correspondingly. The present paper describes the major programs of the WWER high burn-up fuel behavior study in the MIR reactor, capabilities of the applied techniques and some results of the performed irradiation tests. (authors)

Prechamber Compression-Ignition Engine Performance

Science.gov (United States)

Moore, Charles S; Collins, John H , Jr

1938-01-01

Single-cylinder compression-ignition engine tests were made to investigate the performance characteristics of prechamber type of cylinder head. Certain fundamental variables influencing engine performance -- clearance distribution, size, shape, and direction of the passage connecting the cylinder and prechamber, shape of prechamber, cylinder clearance, compression ratio, and boosting -- were independently tested. Results of motoring and of power tests, including several typical indicator cards, are presented.

The advanced test reactor strategic evaluation program

International Nuclear Information System (INIS)

Buescher, B.J.

1989-01-01

Since the Chernobly accident, the safety of test reactors and irradiation facilities has been critically evaluated from the public's point of view. A systematic evaluation of all safety, environmental, and operational issues must be made in an integrated manner to prioritize actions to maximize benefits while minimizing costs. Such a proactive program has been initiated at the Advanced Test Reactor (ATR). This program, called the Strategic Evaluation Program (STEP), is being conducted for the ATR to provide integrated safety and operational reviews of the reactor against the standards applied to licensed commercial power reactors. This has taken into consideration the lessons learned by the US Nuclear Regulatory Commission (NRC) in its Systematic Evaluation Program (SEP) and the follow-on effort known as the Integrated Safety Assessment Program (ISAP). The SEP was initiated by the NRC to review the designs of older operating nuclear power plants to confirm and document their safety. The ATR STEP objectives are discussed

Turbulent spark-jet ignition in SI gas fuelled engine

Directory of Open Access Journals (Sweden)

Pielecha Ireneusz

2017-01-01

Full Text Available The article contains a thermodynamic analysis of a new combustion system that allows the combustion of stratified gas mixtures with mean air excess coefficient in the range 1.4-1.8. Spark ignition was used in the pre-chamber that has been mounted in the engine cylinder head and contained a rich mixture out of which a turbulent flow of ignited mixture is ejected. It allows spark-jet ignition and the turbulent combustion of the lean mixture in the main combustion chamber. This resulted in a two-stage combustion system for lean mixtures. The experimental study has been conducted using a single-cylinder test engine with a geometric compression ratio ε = 15.5 adapted for natural gas supply. The tests were performed at engine speed n = 2000 rpm under stationary engine load when the engine operating parameters and toxic compounds emissions have been recorded. Analysis of the results allowed to conclude that the evaluated combustion system offers large flexibility in the initiation of charge ignition through an appropriate control of the fuel quantities supplied into the pre-chamber and into the main combustion chamber. The research concluded with determining the charge ignition criterion for a suitably divided total fuel dose fed to the cylinder.

Theoretical study for ICRF sustained LHD type p-{sup 11}B reactor

Energy Technology Data Exchange (ETDEWEB)

Watanabe, Tsuguhiro (ed.)

2003-04-01

This is a summary of the workshop on 'Theoretical Study for ICRF Sustained LHD Type p-{sup 11}B Reactor' held in National Institute for Fusion Science (NIFS) on July 25, 2002. In the workshop, study of LHD type D-{sup 3}He reactor is also reported. A review concerning the advanced nuclear fusion fuels is also attached. This review was reported at the workshop of last year. The development of the p-{sup 11}B reactor research which uses the LHD magnetic field configuration has been briefly summarized in section 1. In section 2, an integrated report on advanced nuclear fusion fuels is given. Ignition conditions in a D-{sup 3}He helical reactor are summarized in section 3. 0-dimensional particle and power balance equations are solved numerically assuming the ISS95 confinement law including a confinement factor ({gamma}{sub HH}). It is shown that high average beta plasma confinement, a large confinement factor ({gamma}{sub HH} > 3) and the hot ion mode (T{sub i}/T{sub e} > 1.4) are necessary to achieve the ignition of the D-{sup 3}He helical reactor. Characteristics of ICRF sustained p-{sup 11}B reactor are analyzed in section 4. The nuclear fusion reaction rate < {sigma}{upsilon} > is derived assuming a quasilinear plateau distribution function (QPDF) for protons, and an ignition condition of p-{sup 11}B reactor is shown to be possible. The 3 of the presented papers are indexed individually. (J.P.N.)

Compact toroid refueling of reactors

International Nuclear Information System (INIS)

Gouge, M.J.; Hogan, J.T.; Milora, S.L.; Thomas, C.E.

1988-04-01

The feasibility of refueling fusion reactors and devices such as the International Thermonuclear Engineering Reactor (ITER) with high-velocity compact toroids is investigated. For reactors with reasonable limits on recirculating power, it is concluded that the concept is not economically feasible. For typical ITER designs, the compact toroid fueling requires about 15 MW of electrical power, with about 5 MW of thermal power deposited in the plasma. At these power levels, ideal ignition (Q = ∞) is not possible, even for short-pulse burns. The pulsed power requirements for this technology are substantial. 6 ref., 1 figs

Flow Friction or Spontaneous Ignition?

Science.gov (United States)

Stoltzfus, Joel M.; Gallus, Timothy D.; Sparks, Kyle

2012-01-01

"Flow friction," a proposed ignition mechanism in oxygen systems, has proved elusive in attempts at experimental verification. In this paper, the literature regarding flow friction is reviewed and the experimental verification attempts are briefly discussed. Another ignition mechanism, a form of spontaneous combustion, is proposed as an explanation for at least some of the fire events that have been attributed to flow friction in the literature. In addition, the results of a failure analysis performed at NASA Johnson Space Center White Sands Test Facility are presented, and the observations indicate that spontaneous combustion was the most likely cause of the fire in this 2000 psig (14 MPa) oxygen-enriched system.

Automated reactor protection testing saves time and avoids errors

International Nuclear Information System (INIS)

Raimondo, E.

1990-01-01

When the Pressurized Water Reactor units in the French 900MWe series were designed, the instrumentation and control systems were equipped for manual periodic testing. Manual reactor protection system testing has since been successfully replaced by an automatic system, which is also applicable to other instrumentation testing. A study on the complete automation of process instrumentation testing has been carried out. (author)

Nuclear start-up, testing and core management of the Fast Test Reactor (FTR)

International Nuclear Information System (INIS)

Bennett, R.A.; Daughtry, J.W.; Harris, R.A.; Jones, D.H.; Nelson, J.V.; Rawlins, J.A.; Rothrock, R.B.; Sevenich, R.A.; Zimmerman, B.D.

1980-01-01

Plans for the nuclear start-up, low and high power physics testing, and core management of the Fast Test Reactor (FTR) are described. Owing to the arrangement of the fuel-handling system, which permits continuous instrument lead access to experiments during refuelling, it is most efficient to load the reactor in an asymmetric fashion, filling one-third core sectors at a time. The core neutron level will be monitored during this process using both in-core and ex-core detectors. A variety of physics tests are planned following the core loading. Because of the experimental purpose of the reactor, these tests will include a comprehensive characterization programme involving both active and passive neutron and gamma measurements. Following start-up tests, the FTR will be operated as a fast neutron irradiation facility, to test a wide variety of fast reactor core components and materials. Nuclear analyses will be made prior to each irradiation cycle to confirm that the planned arrangement of standard and experimental components satisfies all safety and operational constraints, and that all experiments are located so as to achieve their desired irradiation environment. (author)

Energy transport requirements for tokamak reactors in the second ballooning stability regime

International Nuclear Information System (INIS)

Potok, R.E.; Bromberg, L.; Cohn, D.R.

1986-01-01

The authors present an analysis of ignition confinement constraints on a tokamak reactor operating in the second regime of ballooning stability. This regime is characterized by flat plasma pressure profiles, with a sharp pressure gradient near a conducting first wall at the plasma edge. The energy confinement time is determined by transport processes across the pressure gradient region. The authors have found that the required transport needed for ignition in the edge region is very close to the value predicted by neoclassical ion conductivity scaling. Only by carefully tailoring the conductivity scaling across the flux coordinate were the authors able to match both the kink stability and ignition requirements. With optimistic assumptions, R/sub o/ ≅ 7 m appears to be the minimum major radius for an economical tokamak reactor in the second ballooning stability regime. This paper presents a base design case at R/sub o/ = 7 m, and shows how the reactor design varies with changes in major radius, ion transport scaling, and electron transport scaling

Tokamak reactor startup power

International Nuclear Information System (INIS)

Weldon, D.M.; Murray, J.G.

1983-01-01

Tokamak startup with ohmic heating (OH)-induced voltages requires rather large voltages and power supplies. On present machines, with no radiofrequency (rf)-assist provisions, hundreds of volts have been specified for their designs. With the addition of electron cyclotron resonant heating (ECRH) assist, the design requirements have been lowered. To obtain information on the cost and complexity associated with this ECRH-assisted, OH-pulsed startup voltage for ignition-type machines, a trade-off study was completed. The Fusion Engineering Device (FED) configuration was selected as a model because information was available on the structure. The data obtained are applicable to all tokamaks of this general size and complexity, such as the Engineering Test Reactor

Stability of Ignition Transients

OpenAIRE

V.E. Zarko

1991-01-01

The problem of ignition stability arises in the case of the action of intense external heat stimuli when, resulting from the cut-off of solid substance heating, momentary ignition is followed by extinction. Physical pattern of solid propellant ignition is considered and ignition criteria available in the literature are discussed. It is shown that the above mentioned problem amounts to transient burning at a given arbitrary temperature distribution in the condensed phase. A brief survey...

Present status and future perspective of research and test reactors in JAERI

International Nuclear Information System (INIS)

Baba, Osamu; Kaieda, Keisuke

1999-01-01

Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfil a major role in the study of nuclear energy and fundamental research. At present, four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR), are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has reached first criticality and is waiting for the power-up test. This paper introduce these reactors and describe their present operational status. The recent tendency of utilization and future perspectives are also reported. (author)

Present status and future perspective of research and test reactors in JAERI

Energy Technology Data Exchange (ETDEWEB)

Baba, Osamu [Japan Atomic Energy Research Inst., Oarai, Ibaraki (Japan). Oarai Research Establishment; Kaieda, Keisuke

1999-08-01

Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfil a major role in the study of nuclear energy and fundamental research. At present, four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR), are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has reached first criticality and is waiting for the power-up test. This paper introduce these reactors and describe their present operational status. The recent tendency of utilization and future perspectives are also reported. (author)

Ignition probabilities for Compact Ignition Tokamak designs

International Nuclear Information System (INIS)

Stotler, D.P.; Goldston, R.J.

1989-09-01

A global power balance code employing Monte Carlo techniques had been developed to study the ''probability of ignition'' and has been applied to several different configurations of the Compact Ignition Tokamak (CIT). Probability distributions for the critical physics parameters in the code were estimated using existing experimental data. This included a statistical evaluation of the uncertainty in extrapolating the energy confinement time. A substantial probability of ignition is predicted for CIT if peaked density profiles can be achieved or if one of the two higher plasma current configurations is employed. In other cases, values of the energy multiplication factor Q of order 10 are generally obtained. The Ignitor-U and ARIES designs are also examined briefly. Comparisons of our empirically based confinement assumptions with two theory-based transport models yield conflicting results. 41 refs., 11 figs

Study for improvement of performance of the test and research reactors

Energy Technology Data Exchange (ETDEWEB)

Sakurai, Fumio [Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan). Tokai Research Establishment

1999-03-01

Current utilization needs for the test and research reactors become more advanced and diversified along with the advance of nuclear science and technology. Besides, the requested safety for the research and test reactors grows strictly every year as well as a case of the power reactors. Under this circumstance, every effort to improve reactor performance including its safety is necessary to be sustained for allowing more effective utilization of the test and research reactors as experimental apparatus for advanced researches. In this study, the following three themes i.e., JMTR high-performance fuel element, evaluation method of fast neutron irradiation dose in the JMTR, evaluation method of performance of siphon break valve as core covering system for water-cooled test and research reactors, were investigated respectively from the views of improvement of core performance as a neutron source, utilization performance as an experimental apparatus, and safety as a reactor plant. (author)

Review of the general atomic experimental fusion power reactor initial conceptual design

International Nuclear Information System (INIS)

Baker, C.C.; Sager, P.H. Jr.; Harder, C.R.

1976-01-01

The primary objective of the Experimental Power Reactor (EPR) is to provide the necessary interface between physics experiments and the first demonstration power plants. Since economically viable tokamak-type reactors may well have to be very high Q devices (ratio of fusion power out to power into the plasma), it will be essential for a tokamak demonstration reactor to operate at or near ignition conditions. Thus, it is believed that one of the primary objectives of the EPR must be to fully model the behavior of a D-T burning plasma required in the reactor of a demonstration plant. Therefore, a major objective of the EPR should be to achieve ignition conditions. In addition to demonstrating the ability to ignite and control a D-T plasma, it is also desirable that the EPR should produce, or at least demonstrate the ability to produce, a small amount of net electrical power. These objectives should be accomplished at a reasonable cost; this implies achieving a sufficiently high β (ratio of plasma pressure to magnetic field pressure). It is believed that noncircular cross section tokamaks offer the best chance of realizing these objectives. Consequently, noncircular cross sections are a major design feature of the General Atomic EPR

Direct electrical arc ignition of hybrid rocket motors

Science.gov (United States)

Judson, Michael I., Jr.

Hybrid rockets motors provide distinct safety advantages when compared to traditional liquid or solid propellant systems, due to the inherent stability and relative inertness of the propellants prior to established combustion. As a result of this inherent propellant stability, hybrid motors have historically proven difficult to ignite. State of the art hybrid igniter designs continue to require solid or liquid reactants distinct from the main propellants. These ignition methods however, reintroduce to the hybrid propulsion system the safety and complexity disadvantages associated with traditional liquid or solid propellants. The results of this study demonstrate the feasibility of a novel direct electrostatic arc ignition method for hybrid motors. A series of small prototype stand-alone thrusters demonstrating this technology were successfully designed and tested using Acrylonitrile Butadiene Styrene (ABS) plastic and Gaseous Oxygen (GOX) as propellants. Measurements of input voltage and current demonstrated that arc-ignition will occur using as little as 10 watts peak power and less than 5 joules total energy. The motor developed for the stand-alone small thruster was adapted as a gas generator to ignite a medium-scale hybrid rocket motor using nitrous oxide /and HTPB as propellants. Multiple consecutive ignitions were performed. A large data set as well as a collection of development `lessons learned' were compiled to guide future development and research. Since the completion of this original groundwork research, the concept has been developed into a reliable, operational igniter system for a 75mm hybrid motor using both gaseous oxygen and liquid nitrous oxide as oxidizers. A development map of the direct spark ignition concept is presented showing the flow of key lessons learned between this original work and later follow on development.

Conceptual design study of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1986-11-01

Since 1980 the design study has been conducted at JAERI for the Fusion Experimental Reactor (FER) which has been proposed to be the next machine to JT-60 in the Japanese long term program of fusion reactor development. During two years from 1984 to 1985 FER concept was reviewed and redesigned. This report is the summary of the results obtained in the review and redesign activities in 1984 and 85. In the first year FER concept was discussed again and its frame work was reestablished. According to the new frame work the major reactor components of FER were designed. In the second year the whole plant system design including plant layout plan was conducted as well as the more detailed design analysis of the reactor conponents. The newly established frame for FER design is as follows: 1) Plasma : Self-ignition. 2) Operation scenario : Quasi-steady state operation with long burn pulse. 3) Neutron fluence on the first wall : 0.3 MWY/M 2 . 4) Blanket : Non-tritium breeding blanket with test modules for breeding blanket development. 5) Magnets : Superconducting Magnets. (author)

Utilization of fission reactors for fusion engineering testing

International Nuclear Information System (INIS)

Deis, G.A.; Miller, L.G.

1985-01-01

Fission reactors can be used to conduct some of the fusion nuclear engineering tests identified in the FINESSE study. To further define the advantages and disadvantages of fission testing, the technical and programmatic constraints on this type of testing are discussed here. This paper presents and discusses eight key issues affecting fission utilization. Quantitative comparisons with projected fusion operation are made to determine the technical assets and limitations of fission testing. Capabilities of existing fission reactors are summarized and compared with technical needs. Conclusions are then presented on the areas where fission testing can be most useful

Design of high temperature Engineering Test Reactor (HTTR)

International Nuclear Information System (INIS)

Saito, Shinzo; Tanaka, Toshiyuki; Sudo, Yukio

1994-09-01

Construction of High Temperature Engineering Test Reactor (HTTR) is now underway to establish and upgrade basic technologies for HTGRs and to conduct innovative basic research at high temperatures. The HTTR is a graphite-moderated and helium gas-cooled reactor with 30 MW in thermal output and outlet coolant temperature of 850degC for rated operation and 950degC for high temperature test operation. It is planned to conduct various irradiation tests for fuels and materials, safety demonstration tests and nuclear heat application tests. JAERI received construction permit of HTTR reactor facility in February 1990 after 22 months of safety review. This report summarizes evaluation of nuclear and thermal-hydraulic characteristics, design outline of major systems and components, and also includes relating R and D result and safety evaluation. Criteria for judgment, selection of postulated events, major analytical conditions for anticipated operational occurrences and accidents, computer codes used in safety analysis and evaluation of each event are presented in the safety evaluation. (author)

Hot Surface Ignition

OpenAIRE

Tursyn, Yerbatyr; Goyal, Vikrant; Benhidjeb-Carayon, Alicia; Simmons, Richard; Meyer, Scott; Gore, Jay P.

2015-01-01

Undesirable hot surface ignition of flammable liquids is one of the hazards in ground and air transportation vehicles, which primarily occurs in the engine compartment. In order to evaluate the safety and sustainability of candidate replacement fuels with respect to hot surface ignition, a baseline low lead fuel (Avgas 100 LL) and four experimental unleaded aviation fuels recommended for reciprocating aviation engines were considered. In addition, hot surface ignition properties of the gas tu...

Ignition studies of two low-octane gasolines

KAUST Repository

Javed, Tamour

2017-07-24

Low-octane gasolines (RON ∼ 50–70 range) are prospective fuels for gasoline compression ignition (GCI) internal combustion engines. GCI technology utilizing low-octane fuels has the potential to significantly improve well-to-wheel efficiency and reduce the transportation sector\\'s environmental footprint by offsetting diesel fuel usage in compression ignition engines. In this study, ignition delay times of two low-octane FACE (Fuels for Advanced Combustion Engines) gasolines, FACE I and FACE J, were measured in a shock tube and a rapid compression machine over a broad range of engine-relevant conditions (650–1200 K, 20 and 40 bar and ϕ = 0.5 and 1). The two gasolines are of similar octane ratings with anti-knock index, AKI = (RON + MON)/2, of ∼ 70 and sensitivity, S = RON–MON, of ∼ 3. However, the molecular compositions of the two gasolines are notably different. Experimental ignition delay time results showed that the two gasolines exhibited similar reactivity over a wide range of test conditions. Furthermore, ignition delay times of a primary reference fuel (PRF) surrogate (n-heptane/iso-octane blend), having the same AKI as the FACE gasolines, captured the ignition behavior of these gasolines with some minor discrepancies at low temperatures (T < 700 K). Multi-component surrogates, formulated by matching the octane ratings and compositions of the two gasolines, emulated the autoignition behavior of gasolines from high to low temperatures. Homogeneous charge compression ignition (HCCI) engine simulations were used to show that the PRF and multi-component surrogates exhibited similar combustion phasing over a wide range of engine operating conditions.

A numerical study of the influence of ammonia addition on the auto-ignition limits of methane/air mixtures

International Nuclear Information System (INIS)

Van den Schoor, F.; Norman, F.; Vandebroek, L.; Verplaetsen, F.; Berghmans, J.

2009-01-01

In this study the auto-ignition limit of ammonia/methane/air mixtures is calculated based upon a perfectly stirred reactor model with convective heat transfer. The results of four different reaction mechanisms are compared with existing experimental data at an initial temperature of 723 K with ammonia concentrations of 0-20 mol.% and methane concentrations of 2.5-10 mol.%. It is found that the calculation of the auto-ignition limit pressure at constant temperature leads to larger relative deviations between calculated and experimental results than the calculation of the auto-ignition temperature at constant pressure. In addition to the calculations, a reaction path analysis is performed to explain the observed lowering of the auto-ignition limit of methane/air mixtures by ammonia addition. It is found that this decrease is caused by the formation of NO and NO 2 , which enhance the oxidation of methane at low temperatures.

Performance Test for Neutron Detector and Associated System using Research Reactor

Energy Technology Data Exchange (ETDEWEB)

Yang, Seongwoo; Park, Sung Jae; Cho, Man Soon [Korea Atomic Energy Research Institute, Daejeon (Korea, Republic of); Oh, Se Hyun [USERS, Daejeon (Korea, Republic of); Shin, Ho Cheol [KHNP CRI, Daejeon (Korea, Republic of)

2016-10-15

SPND (Self-Powered Neutron Detector) has been developed to extend its lifespan. ENFMS (Ex-Core Flux Monitoring System) of pressurized water reactor has been also improved. After the development and improvement, their performance must be verified under the neutron irradiation environment. We used a research reactor for the performance verification of neutron detector and associated system because the research reactor can meet the neutron flux level of commercial nuclear reactor. In this paper, we report the performance verification method and result for the SPND and ENFMS using the research reactor. The performance tests for the SPND and ENFMS were conducted using UCI TRIGA reactor. The test environment of commercial reactor’s neutron flux level must be required. However, it is difficult to perform the test in the commercial rector due to the constraint of time and space. The research reactor can be good alternative neutron source for the test of neutron detectors and associated system.

Progress of impact ignition

International Nuclear Information System (INIS)

Murakami, M.; Nagatomo, H.; Johzaki, T.

2010-11-01

In impact ignition scheme, a portion of the fuel (the impactor) is accelerated to a super-high velocity, compressed by convergence, and collided with a precompressed main fuel. This collision generates shock waves in both the impactor and the main fuel. Since the density of the impactor is generally much lower than that of the main fuel, the pressure balance ensures that the shock-heated temperature of the impactor is significantly higher than that of the main fuel. Hence, the impactor can reach ignition temperature and thus become an igniter. Here we report major new results on recent impact ignition research: (1) A maximum velocity ∼ 1000 km/s has been achieved under the operation of NIKE KrF laser at Naval Research Laboratory (laser wavelength=0.25μm) in the use of a planar target made of plastic and (2) We have performed two-dimensional simulation for burn and ignition to show the feasibility of the impact ignition. (author)

Tokamak engineering test reactor

International Nuclear Information System (INIS)

Conn, R.W.; Jassby, D.L.

1975-07-01

The design criteria for a tokamak engineering test reactor can be met by operating in the two-component mode with reacting ion beams, together with a new blanket-shield design based on internal neutron spectrum shaping. A conceptual reactor design achieving a neutron wall loading of about 1 MW/m 2 is presented. The tokamak has a major radius of 3.05 m, the plasma cross-section is noncircular with a 2:1 elongation, and the plasma radius in the midplane is 55 cm. The total wall area is 149 m 2 . The plasma conditions are T/sub e/ approximately T/sub i/ approximately 5 keV, and ntau approximately 8 x 10 12 cm -3 s. The plasma temperature is maintained by injection of 177 MW of 200-keV neutral deuterium beams; the resulting deuterons undergo fusion reactions with the triton-target ions. The D-shaped toroidal field coils are extended out to large major radius (7.0 m), so that the blanket-shield test modules on the outer portion of the torus can be easily removed. The TF coils are superconducting, using a cryogenically stable TiNb design that permits a field at the coil of 80 kG and an axial field of 38 kG. The blanket-shield design for the inner portion of the torus nearest the machine center line utilizes a neutron spectral shifter so that the first structural wall behind the spectral shifter zone can withstand radiation damage for the reactor lifetime. The energy attenuation in this inner blanket is 8 x 10 -6 . If necessary, a tritium breeding ratio of 0.8 can be achieved using liquid lithium cooling in the []outer blanket only. The overall power consumption of the reactor is about 340 MW(e). A neutron wall loading greater than 1 MW/m 2 can be achieved by increasing the maximum magnetic field or the plasma elongation. (auth)

Imperfection detection probability at ultrasonic testing of reactor vessels

International Nuclear Information System (INIS)

Kazinczy, F. de; Koernvik, L.Aa.

1980-02-01

The report is a lecture given at a symposium organized by the Swedish nuclear power inspectorate on February 1980. Equipments, calibration and testing procedures are reported. The estimation of defect detection probability for ultrasonic tests and the reliability of literature data are discussed. Practical testing of reactor vessels and welded joints are described. Swedish test procedures are compared with other countries. Series of test data for welded joints of the OKG-2 reactor are presented. Future recommendations for testing procedures are made. (GBn)

Present status of Japan materials testing reactor

Energy Technology Data Exchange (ETDEWEB)

Hori, Naohiko; Kaminaga, Masanori; Kusunoki, Tsuyoshi; Ishihara, Masahiro; Niimi, Motoji; Komori, Yoshihiro; Suzuki, Masahide; Kawamura, Hiroshi [Japan Atomic Energy Agency, Oarai Research and Development Center, Oarai, Ibaraki (Japan)

2012-03-15

The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, a detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

Present status of Japan materials testing reactor

International Nuclear Information System (INIS)

Hori, Naohiko; Kaminaga, Masanori; Kusunoki, Tsuyoshi; Ishihara, Masahiro; Niimi, Motoji; Komori, Yoshihiro; Suzuki, Masahide; Kawamura, Hiroshi

2012-01-01

The Japan Materials Testing Reactor (JMTR) in Japan Atomic Energy Agency (JAEA) is a light water cooled tank type reactor with first criticality in March 1968. Owing to the connection between the JMTR and hot laboratory by a canal, easy re-irradiation tests can be conducted with safe and quick transportation of irradiated samples. The JMTR has been applied to fuel/material irradiation examinations for LWRs, HTGR, fusion reactor and RI production. However, the JMTR operation was once stopped in August 2006, and check and review on the reoperation had been conducted by internal as well as external committees. As a result of the discussion, the JMTR reoperation was determined, and refurbishment works started from the beginning of JFY 2007. The refurbishment works have finished in March 2011 taking four years from JFY 2007. Unfortunately, at the end of the JFY 2010 on March 11, the Great-Eastern-Japan-Earthquake occurred, and functional tests before the JMTR restart, such as cooling system, reactor control system and so on, were delayed by the earthquake. Moreover, a detail inspection found some damages such as slight deformation of the truss structure at the roof of the JMTR reactor building. Consequently, the restart of the JMTR will be delayed from June to next October, 2012. Now, the safety evaluation after the earthquake disaster is being carried out aiming at the restart of the JMTR. The renewed JMTR will be started from JFY 2012 and operated for a period of about 20 years until around JFY 2030. The usability improvement of the JMTR, e.g. higher reactor availability, shortening turnaround time to get irradiation results, attractive irradiation cost, business confidence, is also discussed with users as the preparations for re-operation. (author)

Laser ignition - Spark plug development and application in reciprocating engines

Science.gov (United States)

Pavel, Nicolaie; Bärwinkel, Mark; Heinz, Peter; Brüggemann, Dieter; Dearden, Geoff; Croitoru, Gabriela; Grigore, Oana Valeria

2018-03-01

solutions for positioning of the laser spark plug, i.e. placing it apart from or directly on the engine, are introduced. The path taken from the first solution proposed, to build a compact laser suitable for ignition, to the practical realization of a laser spark plug is described. Results obtained by ignition of automobile test engines, with laser devices that resemble classical spark plugs, are specifically discussed. It is emphasized that technological advances have brought this method of laser ignition close to the application and installation in automobiles powered by gasoline engines. Achievements made in the laser ignition of natural gas engines are outlined, as well as the utilization of laser ignition in other applications. Scientific and technical advances have allowed realization of laser devices with multiple (up to four) beam outputs, but many other important aspects (such as integration, thermal endurance or vibration strength) are still to be solved. Recent results of multi-beam ignition of a single-cylinder engine in a test bench set-up are encouraging and have led to increased research interest in this direction. A fundamental understanding of the processes involved in laser ignition is crucial in order to exploit the technology's full potential. Therefore, several measurement techniques, primarily optical types, used to characterize the laser ignition process are reviewed in this work.

Laser ignited engines: progress, challenges and prospects.

Science.gov (United States)

Dearden, Geoff; Shenton, Tom

2013-11-04

Laser ignition (LI) has been shown to offer many potential benefits compared to spark ignition (SI) for improving the performance of internal combustion (IC) engines. This paper outlines progress made in recent research on laser ignited IC engines, discusses the potential advantages and control opportunities and considers the challenges faced and prospects for its future implementation. An experimental research effort has been underway at the University of Liverpool (UoL) to extend the stratified speed/load operating region of the gasoline direct injection (GDI) engine through LI research, for which an overview of some of the approaches, testing and results to date are presented. These indicate how LI can be used to improve control of the engine for: leaner operation, reductions in emissions, lower idle speed and improved combustion stability.

Equilibrium ignition for ICF capsules

International Nuclear Information System (INIS)

Lackner, K.S.; Colgate, S.A.; Johnson, N.L.; Kirkpatrick, R.C.; Menikoff, R.; Petschek, A.G.

1993-01-01

There are two fundamentally different approaches to igniting DT fuel in an ICF capsule which can be described as equilibrium and hot spot ignition. In both cases, a capsule which can be thought of as a pusher containing the DT fuel is imploded until the fuel reaches ignition conditions. In comparing high-gain ICF targets using cryogenic DT for a pusher with equilibrium ignition targets using high-Z pushers which contain the radiation. The authors point to the intrinsic advantages of the latter. Equilibrium or volume ignition sacrifices high gain for lower losses, lower ignition temperature, lower implosion velocity and lower sensitivity of the more robust capsule to small fluctuations and asymmetries in the drive system. The reduction in gain is about a factor of 2.5, which is small enough to make the more robust equilibrium ignition an attractive alternative

Laser induced plasma methodology for ignition control in direct injection sprays

International Nuclear Information System (INIS)

Pastor, José V.; García-Oliver, José M.; García, Antonio; Pinotti, Mattia

2016-01-01

Highlights: • Laser Induced Plasma Ignition system is designed and applied to a Diesel Spray. • A method for quantification of the system effectiveness and reliability is proposed. • The ignition system is optimized in atmospheric and engine-like conditions. • Higher system effectiveness is reached with higher ambient density. • The system is able to stabilize Diesel combustion compared to auto-ignition cases. - Abstract: New combustion modes for internal combustion engines represent one of the main fields of investigation for emissions control in transportation Industry. However, the implementation of lean fuel mixture condition and low temperature combustion in real engines is limited by different unsolved practical issues. To achieve an appropriate combustion phasing and cycle-to-cycle control of the process, the laser plasma ignition system arises as a valid alternative to the traditional electrical spark ignition system. This paper proposes a methodology to set-up and optimize a laser induced plasma ignition system that allows ensuring reliability through the quantification of the system effectiveness in the plasma generation and positional stability, in order to reach optimal ignition performance. For this purpose, experimental tests have been carried out in an optical test rig. At first the system has been optimized in an atmospheric environment, based on the statistical analysis of the plasma records taken with a high speed camera to evaluate the induction effectiveness and consequently regulate and control the system settings. The same optimization method has then been applied under engine-like conditions, analyzing the effect of thermodynamic ambient conditions on the plasma induction success and repeatability, which have shown to depend mainly on ambient density. Once optimized for selected engine conditions, the laser plasma induction system has been used to ignite a direct injection Diesel spray, and to compare the evolution of combustion

Present status and future perspectives of research and test reactor in Japan

International Nuclear Information System (INIS)

Kaneko, Yoshihiko; Kaieda, Keisuke

2000-01-01

Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfill a major role in the study of nuclear energy and fundamental research. At present four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR) are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has recently reached first criticality and now in the power up test. In 1966, the Kyoto University built the Kyoto University Reactor (KUR) and started its operation for joint use program of the Japanese universities. This paper introduces these reactors and describes their present operational status and also efforts for aging management. The recent tendency of utilization and future perspectives is also reported. (author)

Present status and future perspectives of research and test reactor in Japan

Energy Technology Data Exchange (ETDEWEB)

Kaneko, Yoshihiko [Atomic Energy Research Laboratory, Musashi Institute of Technology, Kawasaki, Kanagawa (Japan); Kaieda, Keisuke [Department of Research Reactor, Japan Atomic Energy Research Inst., Tokai, Ibaraki (Japan)

2000-10-01

Since 1957, Japan Atomic Energy Research Institute (JAERI) has constructed several research and test reactors to fulfill a major role in the study of nuclear energy and fundamental research. At present four reactors, the Japan Research Reactor No. 3 and No. 4 (JRR-3M and JRR-4 respectively), the Japan Materials Testing Reactor (JMTR) and the Nuclear Safety Research Reactor (NSRR) are in operation, and a new High Temperature Engineering Test Reactor (HTTR) has recently reached first criticality and now in the power up test. In 1966, the Kyoto University built the Kyoto University Reactor (KUR) and started its operation for joint use program of the Japanese universities. This paper introduces these reactors and describes their present operational status and also efforts for aging management. The recent tendency of utilization and future perspectives is also reported. (author)

Progress Towards Ignition on the National Ignition Facility

Science.gov (United States)

Edwards, John

2012-10-01

Since completion of the National Ignition Facility (NIF) construction project in March 2009, a wide variety of diagnostics, facility infrastructure, and experimental platforms have been commissioned in pursuit of generating the conditions necessary to reach thermonuclear ignition in the laboratory via the inertial confinement approach. NIF's capabilities and infrastructure include over 50 X-ray, optical, and nuclear diagnostics systems and the ability to shoot cryogenic DT layered capsules. There are two main approaches to ICF: direct drive in which laser light impinges directly on a capsule containing a solid layer of DT fuel, and indirect drive in which the laser light is first converted to thermal X-rays. To date NIF has been conducting experiments using the indirect drive approach, injecting up to 1.8MJ of ultraviolet light (0.35 micron) into 1 cm scale cylindrical gold or gold-coated uranium, gas-filled hohlraums, to implode 1mm radius plastic capsules containing solid DT fuel layers. In order to achieve ignition conditions the implosion must be precisely controlled. The National Ignition Campaign (NIC), an international effort with the goal of demonstrating thermonuclear burn in the laboratory, is making steady progress toward this. Utilizing precision pulse-shaping experiments in early 2012 the NIC achieve fuel rhoR of approximately 1.2 gm/cm^2 with densities of around 600-800 g/cm^3 along with neutron yields within about a factor of 5 necessary to enter a regime in which alpha particle heating will become important. To achieve these results, experimental platforms were developed to carefully control key attributes of the implosion. This talk will review NIF's capabilities and the progress toward ignition, as well as the physics of ignition targets on NIF and on other facilities. Acknowledgement: this work performed under the auspices of the U.S. Department of Energy by Lawrence Livermore National Laboratory under Contract DE-AC52-07NA27344.

the JHR Material Testing Reactor

International Nuclear Information System (INIS)

Roure, C.; Cornu, B.; Berthet, B.; Simon, E.; Estre, N.; Guimbal, P.; Kinnunen, P.; Kotiluoto, P.

2013-06-01

The Jules Horowitz Reactor (JHR) is a European experimental reactor under construction in CEA Cadarache. It will be dedicated to material and fuel irradiation tests, and to medical isotopes production. Non-Destructive nuclear Examinations systems (NDE) will be implemented in pools to analyse the irradiated fuel or tested material in their supporting experimental irradiation devices extracted from the core or its immediate periphery. The Nuclear Measurement Laboratory (NML) of CEA Cadarache is working in collaboration with VTT (Technical Research Centre in Finland) in designing and developing NDE systems implementing gamma-ray spectroscopy and high energy X-ray imaging of the sample and irradiation device. CEA is also designing a neutron radiography system for which NML is working on the detection system. Design studies are performed with Monte Carlo transport codes and specific simulation tools developed by the NML for Xray and neutron imaging. (authors)

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

Advanced Demonstration and Test Reactor Options Study

Energy Technology Data Exchange (ETDEWEB)

Petti, David Andrew [Idaho National Lab. (INL), Idaho Falls, ID (United States); Hill, R. [Argonne National Lab. (ANL), Argonne, IL (United States); Gehin, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States); Heidet, F. [Argonne National Lab. (ANL), Argonne, IL (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Grandy, Christopher [Argonne National Lab. (ANL), Argonne, IL (United States); Qualls, A. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Brown, Nicholas [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Powers, J. [Oak Ridge National Lab. (ORNL), Oak Ridge, TN (United States); Hoffman, E. [Argonne National Lab. (ANL), Argonne, IL (United States); Croson, D. [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2017-01-01

Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power’s share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercialization of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy’s (DOE’s) broader commitment to pursuing an “all of the above” clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate “advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy�

Advanced Demonstration and Test Reactor Options Study

International Nuclear Information System (INIS)

Petti, David Andrew; Hill, R.; Gehin, J.; Gougar, Hans David; Strydom, Gerhard; Heidet, F.; Kinsey, J.; Grandy, Christopher; Qualls, A.; Brown, Nicholas; Powers, J.; Hoffman, E.; Croson, D.

2017-01-01

Global efforts to address climate change will require large-scale decarbonization of energy production in the United States and elsewhere. Nuclear power already provides 20% of electricity production in the United States (U.S.) and is increasing in countries undergoing rapid growth around the world. Because reliable, grid-stabilizing, low emission electricity generation, energy security, and energy resource diversity will be increasingly valued, nuclear power's share of electricity production has a potential to grow. In addition, there are non electricity applications (e.g., process heat, desalination, hydrogen production) that could be better served by advanced nuclear systems. Thus, the timely development, demonstration, and commercialization of advanced nuclear reactors could diversify the nuclear technologies available and offer attractive technology options to expand the impact of nuclear energy for electricity generation and non-electricity missions. The purpose of this planning study is to provide transparent and defensible technology options for a test and/or demonstration reactor(s) to be built to support public policy, innovation and long term commercialization within the context of the Department of Energy's (DOE's) broader commitment to pursuing an 'all of the above' clean energy strategy and associated time lines. This planning study includes identification of the key features and timing needed for advanced test or demonstration reactors to support research, development, and technology demonstration leading to the commercialization of power plants built upon these advanced reactor platforms. This planning study is consistent with the Congressional language contained within the fiscal year 2015 appropriation that directed the DOE to conduct a planning study to evaluate 'advanced reactor technology options, capabilities, and requirements within the context of national needs and public policy to support innovation in nuclear energy'. Advanced reactors are

Ignition on the National Ignition Facility: a path towards inertial fusion energy

International Nuclear Information System (INIS)

Moses, Edward I.

2009-01-01

The National Ignition Facility (NIF), the world's largest and most powerful laser system for inertial confinement fusion (ICF) and experiments studying high-energy-density (HED) science, is nearing completion at Lawrence Livermore National Laboratory (LLNL). NIF, a 192-beam Nd-glass laser facility, will produce 1.8 MJ, 500 TW of light at the third-harmonic, ultraviolet light of 351 nm. The NIF project is scheduled for completion in March 2009. Currently, all 192 beams have been operationally qualified and have produced over 4.0 MJ of light at the fundamental wavelength of 1053 nm, making NIF the world's first megajoule laser. The principal goal of NIF is to achieve ignition of a deuterium-tritium (DT) fuel capsule and provide access to HED physics regimes needed for experiments related to national security, fusion energy and for broader scientific applications. The plan is to begin 96-beam symmetric indirect-drive ICF experiments early in FY2009. These first experiments represent the next phase of the National Ignition Campaign (NIC). This national effort to achieve fusion ignition is coordinated through a detailed plan that includes the science, technology and equipment such as diagnostics, cryogenic target manipulator and user optics required for ignition experiments. Participants in this effort include LLNL, General Atomics, Los Alamos National Laboratory, Sandia National Laboratory and the University of Rochester Laboratory for Energetics (LLE). The primary goal for NIC is to have all of the equipment operational and integrated into the facility soon after project completion and to conduct a credible ignition campaign in 2010. When the NIF is complete, the long-sought goal of achieving self-sustaining nuclear fusion and energy gain in the laboratory will be much closer to realization. Successful demonstration of ignition and net energy gain on NIF will be a major step towards demonstrating the feasibility of inertial fusion energy (IFE) and will likely focus

Ignition on the National Ignition Facility: a path towards inertial fusion energy

Science.gov (United States)

Moses, Edward I.

2009-10-01

The National Ignition Facility (NIF), the world's largest and most powerful laser system for inertial confinement fusion (ICF) and experiments studying high-energy-density (HED) science, is nearing completion at Lawrence Livermore National Laboratory (LLNL). NIF, a 192-beam Nd-glass laser facility, will produce 1.8 MJ, 500 TW of light at the third-harmonic, ultraviolet light of 351 nm. The NIF project is scheduled for completion in March 2009. Currently, all 192 beams have been operationally qualified and have produced over 4.0 MJ of light at the fundamental wavelength of 1053 nm, making NIF the world's first megajoule laser. The principal goal of NIF is to achieve ignition of a deuterium-tritium (DT) fuel capsule and provide access to HED physics regimes needed for experiments related to national security, fusion energy and for broader scientific applications. The plan is to begin 96-beam symmetric indirect-drive ICF experiments early in FY2009. These first experiments represent the next phase of the National Ignition Campaign (NIC). This national effort to achieve fusion ignition is coordinated through a detailed plan that includes the science, technology and equipment such as diagnostics, cryogenic target manipulator and user optics required for ignition experiments. Participants in this effort include LLNL, General Atomics, Los Alamos National Laboratory, Sandia National Laboratory and the University of Rochester Laboratory for Energetics (LLE). The primary goal for NIC is to have all of the equipment operational and integrated into the facility soon after project completion and to conduct a credible ignition campaign in 2010. When the NIF is complete, the long-sought goal of achieving self-sustaining nuclear fusion and energy gain in the laboratory will be much closer to realization. Successful demonstration of ignition and net energy gain on NIF will be a major step towards demonstrating the feasibility of inertial fusion energy (IFE) and will likely focus

Oxidation of aluminum alloy cladding for research and test reactor fuel

Science.gov (United States)

Kim, Yeon Soo; Hofman, G. L.; Robinson, A. B.; Snelgrove, J. L.; Hanan, N.

2008-08-01

The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

Oxidation of aluminum alloy cladding for research and test reactor fuel

Energy Technology Data Exchange (ETDEWEB)

Kim, Yeon Soo [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States)], E-mail: yskim@anl.gov; Hofman, G.L. [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States); Robinson, A.B. [Idaho National Laboratory, P.O. Box 1625, Idaho Falls, ID 83415-6188 (United States); Snelgrove, J.L.; Hanan, N. [Argonne National Laboratory, Nuclear Engineering, 9700 South Cass Avenue, Argonne, IL 60439 (United States)

2008-08-31

The oxide thicknesses on aluminum alloy cladding were measured for the test plates from irradiation tests RERTR-6 and 7A in the ATR (advanced test reactor). The measured thicknesses were substantially lower than those of test plates with similar power from other reactors available in the literature. The main reason is believed to be due to the lower pH (pH 5.1-5.3) of the primary coolant water in the ATR than in the other reactors (pH 5.9-6.5) for which we have data. An empirical model for oxide film thickness predictions on aluminum alloy used as fuel cladding in the test reactors was developed as a function of irradiation time, temperature, surface heat flux, pH, and coolant flow rate. The applicable ranges of pH and coolant flow rates cover most research and test reactors. The predictions by the new model are in good agreement with the in-pile test data available in the literature as well as with the RERTR test data measured in the ATR.

Ultrasonically triggered ignition at liquid surfaces.

Science.gov (United States)

Simon, Lars Hendrik; Meyer, Lennart; Wilkens, Volker; Beyer, Michael

2015-01-01

Ultrasound is considered to be an ignition source according to international standards, setting a threshold value of 1mW/mm(2) [1] which is based on theoretical estimations but which lacks experimental verification. Therefore, it is assumed that this threshold includes a large safety margin. At the same time, ultrasound is used in a variety of industrial applications where it can come into contact with explosive atmospheres. However, until now, no explosion accidents have been reported in connection with ultrasound, so it has been unclear if the current threshold value is reasonable. Within this paper, it is shown that focused ultrasound coupled into a liquid can in fact ignite explosive atmospheres if a specific target positioned at a liquid's surface converts the acoustic energy into a hot spot. Based on ignition tests, conditions could be derived that are necessary for an ultrasonically triggered explosion. These conditions show that the current threshold value can be significantly augmented. Copyright © 2014 Elsevier B.V. All rights reserved.

High frequency ignition arrangement

Energy Technology Data Exchange (ETDEWEB)

Canup, R E

1977-03-03

The invention concerns an HF ignition arrangement for combustion engines with a transistor oscillator. As this oscillator requires a current of 10A, with peak currents up to about 50A, it is not sensible to take this current through the remote ignition switch for switching it on and off. According to the invention the HF high voltage transformer of the ignition is provided with a control winding, which only requires a few milliamps DC and which can therefore be switched via the ignition switch. If the ignition switch is in the 'running' position, then a premagnetising DC current flows through the control winding, which suppresses the oscillation of the oscillator which has current flowing through it, until this current is interrupted by the interruptor contacts controlled by the combustion engine, so that the oscillations of the oscillator start immediately; the oscillator only continues to oscillate during the period during which the interruptor contacts controlled by the machine are open and interrupt the premagnetisation current. The control winding is short circuited in the 'off' position of the ignition switch.

Ignition of PTFE-lined flexible hoses by rapid pressurization with oxygen

Science.gov (United States)

Janoff, Dwight; Bamford, Larry J.; Newton, Barry E.; Bryan, Coleman J.

1989-01-01

A high-volume pneumatic-impact system has been used to test PTFE-lined stainless steel braided hoses, in order to characterize the roles played in the mechanism of oxygen-induced ignition by impact pressure, pressurization rate, and upstream and downstream volumes of the hose. Ignitions are noted to have occurred at impact pressures well below the working pressure of the hoses, as well as at pressurization rates easily obtainable through manual operation of valves. The use of stainless steel hardlines downstream of the hose prevented ignitions at all pressures and pressurization rates; internal observations have shown evidence of shock ionization in the oxygen prior to ignition.

Dynamic ignition regime of condensed system by radiate heat flux

International Nuclear Information System (INIS)

Arkhipov, V A; Zolotorev, N N; Korotkikh, A G; Kuznetsov, V T

2017-01-01

The main ignition characteristics of high-energy materials are the ignition time and critical heat flux allowing evaluation of the critical conditions for ignition, fire and explosive safety for the test solid propellants. The ignition process is typically studied in stationary conditions of heat input at constant temperature of the heating surface, environment or the radiate heat flux on the sample surface. In real conditions, ignition is usually effected at variable time-dependent values of the heat flux. In this case, the heated layer is formed on the sample surface in dynamic conditions and significantly depends on the heat flux change, i.e. increasing or decreasing falling heat flux in the reaction period of the propellant sample. This paper presents a method for measuring the ignition characteristics of a high-energy material sample in initiation of the dynamic radiant heat flux, which includes the measurement of the ignition time when exposed to a sample time varying radiant heat flux given intensity. In case of pyroxyline containing 1 wt. % of soot, it is shown that the ignition times are reduced by 20–50 % depending on the initial value of the radiant flux density in initiation by increasing or decreasing radiant heat flux compared with the stationary conditions of heat supply in the same ambient conditions. (paper)

Improving the proliferation resistance of research and test reactors

International Nuclear Information System (INIS)

Lewis, R.A.

1978-01-01

Elimination, or substantial reduction, of the trade in unirradiated highly-enriched fuel elements for research and test reactors would significantly reduce the proliferation risk associated with the current potential for diversion of these materials. To this end, it is the long-term goal of U.S. policy to fuel all new and existing research and test reactors with uranium of less-than-20% enrichment (but substantially greater than natural) excepting, perhaps, only a small number of high-power, high-performance, reactors. The U.S. development program for enrichment reduction in research and test reactor designs currently using 90-93% enriched uranium is based on the practical criterion that enrichment reduction should not cause significant flux performance (flux per unit power) or burnup performance degradation relative to the unmodified reactor design. To first order, this implies the requirement that the 235 U loading in the reduced-enrichment fuel elements be the same as the 235 U loading in the 90-93% enriched fuel elements. This can be accomplished by substitution of higher uranium density fuel technology for currently-used fuel technology in the fuel meat volume of the current fuel element design and/or by increasing the usable fuel meat volume. For research and test reactors of power greater than 5-10 megawatts, fuel technology does not currently exist that would permit enrichment reductions to below 20% utilizing this criterion. A program is now beginning in the U.S. to develop the necessary fuel technology. Currently-proven fuel technology is capable, however, of accommodating enrichment reductions to the 30-45% range (from 90-93%) for many reactors in the 5-50MW range. Accordingly the U.S. is proposing to convert existing reactors (and new designs) in the 5-50MW range from the use of highly-enriched fuel to the use of 30-45% enriched fuel, and reactors of less that about 5MW to less-than-20% enrichment, wherever this can be done without significant

Reactor vessel dismantling at the high flux materials testing reactor Petten

International Nuclear Information System (INIS)

Tas, A.; Teunissen, G.

1986-01-01

The project of replacing the reactor vessel of the high flux materials testing reactor (HFR) originated in 1974 when results of several research programs confirmed severe neutron embrittlement of aluminium alloys suggesting a limited life of the existing facility. This report describes the dismantling philosophy and organisation, the design of special underwater equipment, the dismantling of the reactor vessel and thermal column, and the conditioning and shielding activities resulting in a working area for the installation of the new vessel with no access limitations due to radiation. Finally an overview of the segmentation, waste disposal and radiation exposure is given. The total dismantling, segmentation and conditioning activities resulted in a total collective radiation dose of 300 mSv. (orig.) [de

Reaching ignition in the tokamak

International Nuclear Information System (INIS)

Furth, H.P.

1985-06-01

This review covers the following areas: (1) the physics of burning plasmas, (2) plasma physics requirements for reaching ignition, (3) design studies for ignition devices, and (4) prospects for an ignition project

Robustness studies of ignition targets for the National Ignition Facility in two dimensions

International Nuclear Information System (INIS)

Clark, Daniel S.; Haan, Steven W.; Salmonson, Jay D.

2008-01-01

Inertial confinement fusion capsules are critically dependent on the integrity of their hot spots to ignite. At the time of ignition, only a certain fractional perturbation of the nominally spherical hot spot boundary can be tolerated and the capsule still achieve ignition. The degree to which the expected hot spot perturbation in any given capsule design is less than this maximum tolerable perturbation is a measure of the ignition margin or robustness of that design. Moreover, since there will inevitably be uncertainties in the initial character and implosion dynamics of any given capsule, all of which can contribute to the eventual hot spot perturbation, quantifying the robustness of that capsule against a range of parameter variations is an important consideration in the capsule design. Here, the robustness of the 300 eV indirect drive target design for the National Ignition Facility [Lindl et al., Phys. Plasmas 11, 339 (2004)] is studied in the parameter space of inner ice roughness, implosion velocity, and capsule scale. A suite of 2000 two-dimensional simulations, run with the radiation hydrodynamics code LASNEX, is used as the data base for the study. For each scale, an ignition region in the two remaining variables is identified and the ignition cliff is mapped. In accordance with the theoretical arguments of Levedahl and Lindl [Nucl. Fusion 37, 165 (1997)] and Kishony and Shvarts [Phys. Plasmas 8, 4925 (2001)], the location of this cliff is fitted to a power law of the capsule implosion velocity and scale. It is found that the cliff can be quite well represented in this power law form, and, using this scaling law, an assessment of the overall (one- and two-dimensional) ignition margin of the design can be made. The effect on the ignition margin of an increase or decrease in the density of the target fill gas is also assessed

In-core materials testing under LWR conditions in the Halden reactor

International Nuclear Information System (INIS)

Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A.

2002-01-01

The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

In-core materials testing under LWR conditions in the Halden reactor

Energy Technology Data Exchange (ETDEWEB)

Bennett, P.J.; Hauso, E.; Hoegberg, N.W.; Karlsen, T.M.; McGrath, M.A. [OECD Halden Reactor Project (Norway)

2002-07-01

The Halden boiling water reactor (HBWR) has been in operation since 1958. It is a test reactor with a maximum power of 18 MW and is cooled and moderated by boiling heavy water, with a normal operating temperature of 230 C and a pressure of 34 bar. In the past 15 years increasing emphasis has been placed on materials testing, both of in-core structural materials and fuel claddings. These tests require representative light water reactor (LWR) conditions, which are achieved by housing the test rigs in pressure flasks that are positioned in fuel channels in the reactor and connected to dedicated water loops, in which boiling water reactor (BWR) or pressurised water reactor (PWR) conditions are simulated. Understanding of the in-core behaviour of fuel or reactor materials can be greatly improved by on-line measurements during power operation. The Halden Project has performed in-pile measurements for a period of over 35 years, beginning with fuel temperature measurements using thermocouples and use of differential transformers for measurement of fuel pellet or cladding dimensional changes and internal rod pressure. Experience gained over this period has been applied to on-line instrumentation for use in materials tests. This paper gives details of the systems used at Halden for materials testing under LWR conditions. The techniques used to provide on-line data are described and illustrative results are presented. (authors)

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Kinsey, J. [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-03-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

High Temperature Gas-Cooled Test Reactor Point Design: Summary Report

Energy Technology Data Exchange (ETDEWEB)

Sterbentz, James William [Idaho National Lab. (INL), Idaho Falls, ID (United States); Bayless, Paul David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Nelson, Lee Orville [Idaho National Lab. (INL), Idaho Falls, ID (United States); Gougar, Hans David [Idaho National Lab. (INL), Idaho Falls, ID (United States); Strydom, Gerhard [Idaho National Lab. (INL), Idaho Falls, ID (United States)

2016-01-01

A point design has been developed for a 200-MW high-temperature gas-cooled test reactor. The point design concept uses standard prismatic blocks and 15.5% enriched uranium oxycarbide fuel. Reactor physics and thermal-hydraulics simulations have been performed to characterize the capabilities of the design. In addition to the technical data, overviews are provided on the technology readiness level, licensing approach, and costs of the test reactor point design.

Situation of test and research reactors' spent fuels

International Nuclear Information System (INIS)

Shimizu, Kenichi; Uchiyama, Junzo; Sato, Hiroshi

1996-01-01

The U.S. DOE decided a renewal Off-Site Fuel Policy for stopping to spread a highly enriched uranium which was originally enriched at the U.S., the policy declared that to receive all HEU spent fuels from Test and Research reactors in all the world. In Japan, under bilateral agreement of cooperation between the government of the United States and the government of Japan concerning peaceful uses of nuclear energy, the highly enriched uranium of Test and Research Reactors' fuels was purchased from the U.S. and the fuels had been manufactured in Japan, America, Germany and France. On the other hand, a former president of the U.S. J. Carter proposed that to convert the fuels from HEU to LEU concerning a nonproliferation of nuclear materials in 1978, and Japan absolutely supported this policy. Under this condition, the U.S. stopped to receive the spent fuels from the other countries concerning legal action to the Off-Site Fuels Policy. As a result, the spent fuels are increasing, and to cross to each reactor's storage capacity, and if this policy start, a faced crisis of Test and Research Reactors will be avoided. (author)

Action Memorandum for the Engineering Test Reactor under the Idaho Cleanup Project

Energy Technology Data Exchange (ETDEWEB)

A. B. Culp

2007-01-26

This Action Memorandum documents the selected alternative for decommissioning of the Engineering Test Reactor at the Idaho National Laboratory under the Idaho Cleanup Project. Since the missions of the Engineering Test Reactor Complex have been completed, an engineering evaluation/cost analysis that evaluated alternatives to accomplish the decommissioning of the Engineering Test Reactor Complex was prepared adn released for public comment. The scope of this Action Memorandum is to encompass the final end state of the Complex and disposal of the Engineering Test Reactor vessol. The selected removal action includes removing and disposing of the vessel at the Idaho CERCLA Disposal Facility and demolishing the reactor building to ground surface.

Ignition tuning for the National Ignition Campaign

Directory of Open Access Journals (Sweden)

Landen O.

2013-11-01

Full Text Available The overall goal of the indirect-drive inertial confinement fusion [1] tuning campaigns [2] is to maximize the probability of ignition by experimentally correcting for likely residual uncertainties in the implosion and hohlraum physics [3] used in our radiation-hydrodynamic computational models, and by checking for and resolving unexpected shot-to-shot variability in performance [4]. This has been started successfully using a variety of surrogate capsules that set key laser, hohlraum and capsule parameters to maximize ignition capsule implosion velocity, while minimizing fuel adiabat, core shape asymmetry and ablator-fuel mix.

Potential off-normal events and associated radiological source terms for the compact ignition tokamak: Fusion Safety Program

International Nuclear Information System (INIS)

Holland, D.F.; Lyon, R.E.

1987-10-01

The Compact Ignition Tokamak (CIT), the latest step in the United States program to develop the commercial application of fusion power, is designed as the first fusion device to achieve ignition conditions. It is to be constructed near Princeton, New Jersey on the site of the existing Tokamak Fusion Test Reactor (TFTR). To address the environmental impact and public safety concerns, a preliminary analysis was performed of potential off-normal radiological releases. Operational occurrences, natural phenomena, accidents with external origins, and accidents external to the PPPL site were considered as potential sources for off-normal events. Based on an initial screening, events were selected for preliminary analysis. Included in these events were tritium releases from the tritium delivery and recovery system, tritium releases from the torus, releases of activated nitrogen from the test cell or cryostat, seismic events, and shipping accidents. In each case, the design considerations related to the event were reviewed and the release scenarios discussed. Because of the complexity of some of the proposed safety systems, in some cases event trees were used to describe the accident scenarios. For each scenario, the probability was estimated as well as the release magnitude, isotope, chemical form, and release mode. 10 refs., 17 figs., 5 tabs

Accelerated irradiation test of Gundremmingen reactor vessel trepan material

Energy Technology Data Exchange (ETDEWEB)

Hawthorne, J.R. [Materials Engineering Associates, Inc., Lanham, MD (United States)

1992-08-01

Initial mechanical properties tests of beltline trepanned from the decommissioned KRB-A pressure vessel and archive material irradiated in the UBR test reactor revealed a major anomaly in relative radiation embrittlement sensitivity. Poor correspondence of material behavior in test vs. power reactor environments was observed for the weak test orientation (ASTL C-L) whereas correspondence was good for the strong orientation (ASTM C-L). To resolve the anomaly directly, Charpy-V specimens from a low (essentially-nil) fluence region of the vessel were irradiated together with archive material at 279{degrees}C in the UBR test reactor. Properties tests before UBR irradiation revealed a significant difference in 41-J transition temperature and upper shelf energy level between the materials. However, the materials exhibited essentially the same radiation embrittlement sensitivity (both orientations), proving that the anomaly is not due to a basic difference in material irradiation resistances. Possible causes of the original anomaly and the significance to NRC Regulatory Guide 1.99 are discussed.

Accelerated irradiation test of gundremmingen reactor vessel trepan material

International Nuclear Information System (INIS)

Hawthorne, J.R.

1992-08-01

Initial mechanical properties tests of beltline trepanned from the decommissioned KRB-A pressure vessel and archive material irradiated in the UBR test reactor revealed a major anomaly in relative radiation embrittlement sensitivity. Poor correspondence of material behavior in test vs. power reactor environments was observed for the weak test orientation (ASTL C-L) whereas correspondence was good for the strong orientation (ASTM C-L). To resolve the anomaly directly, Charpy-V specimens from a low (essentially-nil) fluence region of the vessel were irradiated together with archive material at 279 degrees C in the UBR test reactor. Properties tests before UBR irradiation revealed a significant difference in 41-J transition temperature and upper shelf energy level between the materials. However, the materials exhibited essentially the same radiation embrittlement sensitivity (both orientations), proving that the anomaly is not due to a basic difference in material irradiation resistances. Possible causes of the original anomaly and the significance to NRC Regulatory Guide 1.99 are discussed

Operation and control of high density tokamak reactors

International Nuclear Information System (INIS)

Attenberger, S.E.; McAlees, D.G.

1976-01-01

The incentive for high density operation of a tokamak reactor was discussed. It is found that high density permits ignition in a relatively small, moderately elongated plasma with a moderate magnetic field strength. Under these conditions, neutron wall loadings approximately 4 MW/m 2 must be tolerated. The sensitivity analysis with respect to impurity effects shows that impurity control will most likely be necessary to achieve the desired plasma conditions. The charge exchange sputtered impurities are found to have an important effect so that maintaining a low neutral density in the plasma is critical. If it is assumed that neutral beams will be used to heat the plasma to ignition, high energy injection is required (approximately 250 keV) when heating is accompished at full density. A scenario is outlined where the ignition temperature is established at low density and then the fueling rate is increased to attain ignition. This approach may permit beams with energies being developed for use in TFTR to be successfully used to heat a high density device of the type described here to ignition

Effort to increase an engine performance using electrical ignition system for motor vehicle

Directory of Open Access Journals (Sweden)

I Wayan Bandem Adnyana

2012-11-01

Full Text Available Increasing engine performances using electrical ignition system on motor vehicle. In accordance with the development oftechnology, improvisation of automotive is created in order to increase the performance of engine. The method to increase thisperformance has been done by modify the ignition system, where the conventional method of ignition system which uses contactbreaker substituted by using capacitor. The improvisation of ignition system has been tested by increasing the speed and load onstationary condition. Results show that the improvisation of ignition system by using capacitor increases the effective power andreduce the specific fuel consumption of engine and reduce the gas emission of CO.

Advanced Test Reactor probabilistic risk assessment

International Nuclear Information System (INIS)

Atkinson, S.A.; Eide, S.A.; Khericha, S.T.; Thatcher, T.A.

1993-01-01

This report discusses Level 1 probabilistic risk assessment (PRA) incorporating a full-scope external events analysis which has been completed for the Advanced Test Reactor (ATR) located at the Idaho National Engineering Laboratory

Preliminary design studies on the Broad Application Test Reactor

International Nuclear Information System (INIS)

Terry, W.J.; Terry, W.K.; Ryskamp, J.M.; Jahshan, S.N.; Fletcher, C.D.; Moore, R.L.; Leyse, C.F.; Ottewitte, E.H.; Motloch, C.G.; Lacy, J.M.

1992-08-01

This report describes progress made at the Idaho National Engineering Laboratory during the first three quarters of Fiscal Year (FY) 1992 on the Laboratory-Directed Research and Development (LDRD) project to perform preliminary design studies on the Broad Application Test Reactor (BATR). This work builds on the FY-92 BATR studies, which identified anticipated mission and safety requirements for BATR and assessed a variety of reactor concepts for their potential capability to meet those requirements. The main accomplishment of the FY-92 BATR program is the development of baseline reactor configurations for the two conventional conceptual test reactors recommended in the FY-91 report. Much of the present report consists of descriptions and neutronics and thermohydraulics analyses of these baseline configurations. In addition, we considered reactor safety issues, compared the consequences of steam explosions for alternative conventional fuel types, explored a Molten Chloride Fast Reactor concept as an alternate BATR design, and examined strategies for the reduction of operating costs. Work planned for the last quarter of FY-92 is discussed, and recommendations for future work are also presented

Thermal-hydraulic tests for reactor safety system

International Nuclear Information System (INIS)

Chun, Se Young; Chung, Moon Ki; Baek, Won Pil

2002-05-01

Tests for the safety depressurization system, Sparger adopted for the Korean next generation reactor, APR1400 are carried out for several geometries with the B and C (Blowdown and Condensation) facility in the condition of high temperature and pressure and with a small test facility in the condition of atmospheric temperature and pressure. Tests for the critical heat flux are performed with the RCS(Reactor Coolant System) facility as well as with the Freon CHF Loop in the condition of high temperature and pressure. The atmospheric temperature and pressure facility is utilized for development of the high standard thermal hydraulic measurement technology. The optical method is developed to measure the local thermal-hydraulic behavior for the single and two-phase boiling phenomena

ATFSR: a small torsatron reactor

International Nuclear Information System (INIS)

Houlberg, W.A.; Lacatski, J.T.; Uckan, N.A.

1985-01-01

A small (average minor radius anti a approx. = 1 m), moderate-aspect-ratio torsatron reactor based on the Advanced Toroidal Facility (ATF) is proposed as a starting point for improved stellarator reactor designs. The major limitation of the compact size is the lack of space under the helical coils for the blanket and shield. Neoclassical confinement models for helically trapped particles show that a large electric potential (radial electric field) is necessary to achieve ignition in a device of this size, although high-Q operation is still attainable with more modest potentials

Switch evaluation test system for the National Ignition Facility

International Nuclear Information System (INIS)

Savage, M.E.; Simpson, W.W.; Reynolds, F.D.

1997-01-01

Flashlamp pumped lasers use pulsed power switches to commute energy stored in capacitor banks to the flashlamps. The particular application in which the authors are interested is the National Ignition Facility (NIF), being designed by Lawrence Livermore National Laboratory, Los Alamos National Laboratory, and Sandia National Laboratories (SNL). To lower the total cost of these switches, SNL has a research program to evaluate large closing switches. The target value of the energy switched by a single device is 1.6 MJ, from a 6 mF, 24kV capacitor bank. The peak current is 500 kA. The lifetime of the NIF facility is 24,000 shots. There is no switch today proven at these parameters. Several short-lived switches (100's of shots) exist that can handle the voltage and current, but would require maintenance during the facility life. Other type devices, notably ignitrons, have published lifetimes in excess of 20,000 shots, but at lower currents and shorter pulse widths. The goal of the experiments at SNL is to test switches with the full NIF wave shape, and at the correct voltage. The SNL facility can provide over 500 kA at 24 kV charge voltage. the facility has 6.4 mF total capacitance, arranged in 25 sub-modules. the modular design makes the facility more flexible (for possible testing at lower current) and safer. For pulse shaping (the NIF wave shape is critically damped) there is an inductor and resistor for each of the 25 modules. Rather than one large inductor and resistor, this lowers the current in the pulse shaping components, and raises their value to those more easily attained with lumped inductors and resistors. The authors show the design of the facility, and show results from testing conducted thus far. They also show details of the testing plan for high current switches

Test program for NIS calibration to reactor thermal output in HTTR

International Nuclear Information System (INIS)

Nakagawa, Shigeaki; Shinozaki, Masayuki; Tachibana, Yukio; Kunitomi, Kazuhiko

2000-03-01

Rise-to-power test program for reactor thermal output measurement has been established to calibrate a neutron instrumentation system taking account of the characteristics of the High Temperature Engineering Test Reactor (HTTR). An error of reactor thermal output measurement was evaluated taking account of a configuration of instrumentation system. And the expected dispersion of measurement in the full power operation was evaluated from non-nuclear heat-up of primary coolant up to 213degC. From the evaluation, it was found that an error of reactor thermal output measurement would be less than ±2.0% at the rated power. This report presents the detailed program of rise-to-power test for reactor thermal output measurement and discusses its measurement error. (author)

Physical and engineering aspects of a fusion engineering test facility based on mirror confinement

International Nuclear Information System (INIS)

Kawabe, T.; Hirayama, S.; Hojo, H.; Kozaki, Y.; Yoshikawa, K.

1986-01-01

Controlled fusion research has accomplished great progress in the field of confinement of high-density and high-temperature plasmas and breakeven experiments are expected before the end of the 1980s. Many experiments have been proposed as the next step for fusion research. Among them is the study of ignited plasmas and another is the study of fusion engineering. Some of the important studies in fusion engineering are the integrated test in a fusion reactor environment as well as tests of first-wall materials and of the reactor structures, and test for tritium breeding and blanket modules or submodules. An ideal neutron source for the study of fusion engineering is the deuterium-tritium (D-T) fusion plasma itself. A neutron facility based on a D-T-burning plasma consists of all of the components that a real fusion power reactor would have, so eventually the integrated test for fusion reactor engineering can be done as well as the tests for each engineering component

A Hydrogen Ignition Mechanism for Explosions in Nuclear Facility Piping Systems

Energy Technology Data Exchange (ETDEWEB)

Leishear, Robert A.

2013-09-18

Hydrogen explosions may occur simultaneously with water hammer accidents in nuclear facilities, and a theoretical mechanism to relate water hammer to hydrogen deflagrations and explosions is presented herein. Hydrogen and oxygen generation due to the radiolysis of water is a recognized hazard in pipe systems used in the nuclear industry, where the accumulation of hydrogen and oxygen at high points in the pipe system is expected, and explosive conditions may occur. Pipe ruptures in nuclear reactor cooling systems were attributed to hydrogen explosions inside pipelines, i.e., Hamaoka, Nuclear Power Station in Japan, and Brunsbuettel in Germany. Prior to these accidents, an ignition source for hydrogen was not clearly demonstrated, but these accidents demonstrated that a mechanism was, in fact, available to initiate combustion and explosion. A new theory to identify an ignition source and explosion cause is presented here, and further research is recommended to fully understand this explosion mechanism.

Review: laser ignition for aerospace propulsion

Directory of Open Access Journals (Sweden)

Steven A. O’Briant

2016-03-01

This paper aims to provide the reader an overview of advanced ignition methods, with an emphasis on laser ignition and its applications to aerospace propulsion. A comprehensive review of advanced ignition systems in aerospace applications is performed. This includes studies on gas turbine applications, ramjet and scramjet systems, and space and rocket applications. A brief overview of ignition and laser ignition phenomena is also provided in earlier sections of the report. Throughout the reading, research papers, which were presented at the 2nd Laser Ignition Conference in April 2014, are mentioned to indicate the vast array of projects that are currently being pursued.

Investigation of the loss of forced cooling test by using the high temperature engineering test reactor (HTTR) (Contract research)

International Nuclear Information System (INIS)

Nakagawa, Shigeaki; Takamatsu, Kuniyoshi; Inaba, Yoshitomo; Goto, Minoru; Tochio, Daisuke

2007-09-01

The three gas circulators trip test and the vessel cooling system stop test as the safety demonstration test by using the High Temperature engineering Test Reactor (HTTR) are under planning to demonstrate inherent safety features of High Temperature Gas-cooled Reactor. All three gas circulators to circulate the helium gas as the coolant are stopped to simulate the loss of forced cooling in the three gas circulators trip test. The stop of the vessel cooling system located outside the reactor pressure vessel to remove the residual heat of the reactor core follows the stop of all three gas circulators in the vessel cooling system stop test. The analysis of the reactor transient for such tests and abnormal events postulated during the test was performed. From the result of analysis, it was confirmed that the three gas circulators trip test and the vessel cooling system stop test can be performed within the region of the normal operation in the HTTR and the safety of the reactor facility is ensured even if the abnormal events would occur. (author)

Refurbish research and test reactors corresponding to global age of nuclear energy

International Nuclear Information System (INIS)

Mishima, Kaichiro; Oyama, Yukio; Okamoto, Koji; Yamana, Hajime; Yamaguchi, Akira

2011-01-01

This special article featured arguments for refurbishment of research and test reactors corresponding to global age of nuclear energy, based on the report: 'Investigation of research facilities necessary for future joint usage' issued by the special committee of Atomic Energy Society of Japan (AESJ) in September 2010. It consisted of six papers titled as 'Introduction-establishment of AESJ special committee for investigation', 'State of research and test reactors in Japan', 'State of overseas research and test reactors', 'Needs analysis for research and test reactors', 'Proposal of AESJ special committee' and 'Summary and future issues'. In order to develop human resources and promote research and development needed in global age of nuclear energy, research and test reactors would be refurbished as an Asian regional center of excellence. (T. Tanaka)

An Experimental and Numerical Study of N-Dodecane/Butanol Blends for Compression Ignition Engines

KAUST Repository

Wakale, Anil Bhaurao; Mohamed, Samah; Naser, Nimal; Jaasim, Mohammed; Banerjee, Raja; Im, Hong G.; Sarathy, Mani

2018-01-01

Alcohols are potential blending agents for diesel that can be effectively used in compression ignition engines. This work investigates the use of n-butanol as a blending component for diesel fuel using experiments and simulations. Dodecane was selected as a surrogate for diesel fuel and various concentrations of n-butanol were added to study ignition characteristics. Ignition delay times for different n-butanol/dodecane blends were measured using the ignition quality tester at KAUST (KR-IQT). The experiments were conducted at pressure of 21 and 18 bar, temperature ranging from 703-843 K and global equivalence ratio of 0.85. A skeletal mechanism for n-dodecane and n-butanol blends with 203 species was developed for numerical simulations. The mechanism was developed by combining n-dodecane skeletal mechanism containing 106 species and a detailed mechanism for all the butanol isomers. The new mixture mechanism was validated for various pressure, temperature and equivalence ratio using a 0-D homogeneous reactor model from CHEMKIN for pure base fuels (n-dodecane and butanol). Computational fluid dynamics (CFD) code, CONVERGE was used to further validate the new mechanism. The new mechanism was able to reproduce the experimental results from IQT at different pressure and temperature conditions.

An Experimental and Numerical Study of N-Dodecane/Butanol Blends for Compression Ignition Engines

KAUST Repository

Wakale, Anil Bhaurao

2018-04-03

Alcohols are potential blending agents for diesel that can be effectively used in compression ignition engines. This work investigates the use of n-butanol as a blending component for diesel fuel using experiments and simulations. Dodecane was selected as a surrogate for diesel fuel and various concentrations of n-butanol were added to study ignition characteristics. Ignition delay times for different n-butanol/dodecane blends were measured using the ignition quality tester at KAUST (KR-IQT). The experiments were conducted at pressure of 21 and 18 bar, temperature ranging from 703-843 K and global equivalence ratio of 0.85. A skeletal mechanism for n-dodecane and n-butanol blends with 203 species was developed for numerical simulations. The mechanism was developed by combining n-dodecane skeletal mechanism containing 106 species and a detailed mechanism for all the butanol isomers. The new mixture mechanism was validated for various pressure, temperature and equivalence ratio using a 0-D homogeneous reactor model from CHEMKIN for pure base fuels (n-dodecane and butanol). Computational fluid dynamics (CFD) code, CONVERGE was used to further validate the new mechanism. The new mechanism was able to reproduce the experimental results from IQT at different pressure and temperature conditions.

Modelling auto ignition of hydrogen in a jet ignition pre-chamber

Energy Technology Data Exchange (ETDEWEB)

Boretti, Alberto A. [School of Science and Engineering, University of Ballarat, PO Box 663, Ballarat, Victoria 3353 (Australia)

2010-04-15

Spark-less jet ignition pre-chambers are enablers of high efficiencies and load control by quantity of fuel injected when coupled with direct injection of main chamber fuel, thus permitting always lean burn bulk stratified combustion. Towards the end of the compression stroke, a small quantity of hydrogen is injected within the pre-chamber, where it mixes with the air entering from the main chamber. Combustion of the air and fuel mixture then starts within the pre-chamber because of the high temperature of the hot glow plug, and then jets of partially combusted hot gases enter the main chamber igniting there in the bulk, over multiple ignition points, lean stratified mixtures of air and fuel. The paper describes the operation of the spark-less jet ignition pre-chamber coupling CFD and CAE engine simulations to allow component selection and engine performance evaluation. (author)

Assessing the prospects for achieving double-shell ignition on the National Ignition Facility using vacuum hohlraums

International Nuclear Information System (INIS)

Amendt, Peter; Cerjan, C.; Hamza, A.; Hinkel, D. E.; Milovich, J. L.; Robey, H. F.

2007-01-01

The goal of demonstrating ignition on the National Ignition Facility [J. D. Lindl et al., Phys. Plasmas 11, 339 (2003)] has motivated a revisit of double-shell (DS) targets as a complementary path to the cryogenic baseline approach. Expected benefits of DS ignition targets include noncryogenic deuterium-tritium (DT) fuel preparation, minimal hohlraum-plasma-mediated laser backscatter, low threshold-ignition temperatures (≅4 keV) for relaxed hohlraum x-ray flux asymmetry tolerances, and minimal (two-) shock timing requirements. On the other hand, DS ignition presents several formidable challenges, encompassing room-temperature containment of high-pressure DT (≅790 atm) in the inner shell, strict concentricity requirements on the two shells ( 2 nanoporous aerogels with suspended Cu particles. A prototype demonstration of an ignition DS is planned for 2008, incorporating the needed novel nanomaterials science developments and the required fabrication tolerances for a realistic ignition attempt after 2010

Enhanced in-pile instrumentation at the advanced test reactor

Energy Technology Data Exchange (ETDEWEB)

Rempe, J. L.; Knudson, D. L.; Daw, J. E.; Unruh, T.; Chase, B. M.; Palmer, J.; Condie, K. G.; Davis, K. L. [Idaho National Laboratory, MS 3840, P.O. Box 1625, Idaho Falls, ID 83415 (United States)

2011-07-01

Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper reports results from this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and realtime flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted. (authors)

Enhanced In-Pile Instrumentation at the Advanced Test Reactor

Science.gov (United States)

Rempe, Joy L.; Knudson, Darrell L.; Daw, Joshua E.; Unruh, Troy; Chase, Benjamin M.; Palmer, Joe; Condie, Keith G.; Davis, Kurt L.

2012-08-01

Many of the sensors deployed at materials and test reactors cannot withstand the high flux/high temperature test conditions often requested by users at U.S. test reactors, such as the Advanced Test Reactor (ATR) at the Idaho National Laboratory. To address this issue, an instrumentation development effort was initiated as part of the ATR National Scientific User Facility in 2007 to support the development and deployment of enhanced in-pile sensors. This paper provides an update on this effort. Specifically, this paper identifies the types of sensors currently available to support in-pile irradiations and those sensors currently available to ATR users. Accomplishments from new sensor technology deployment efforts are highlighted by describing new temperature and thermal conductivity sensors now available to ATR users. Efforts to deploy enhanced in-pile sensors for detecting elongation and real-time flux detectors are also reported, and recently-initiated research to evaluate the viability of advanced technologies to provide enhanced accuracy for measuring key parameters during irradiation testing are noted.

Conceptual design of a moving-ring reactor

International Nuclear Information System (INIS)

Smith, A.C. Jr.; Ashworth, C.P.; Abreu, K.E.

1983-01-01

A design of a prototype Moving-Ring Reactor has been completed. The fusion fuel is confined in current-carrying rings of magnetically field-reversed plasma (''compact toroids''). The plasma rings, formed by a coaxial plasma gun, undergo adiabatic magnetic compression to ignition temperature while they are being injected into the reactor's burner section. The cylindrical burner chamber is divided into three ''burn stations''. Separator coils and a slight axial guide-field gradient are used to shuttle the ignited toroids rapidly from one burn station to the next, pausing for one third of the total burn time at each station. D-T- 3 He ice pellets refuel the rings at a rate which maintains constant radiated power. The first wall and tritium breeding blanket designs make credible use of helium cooling, SiC and Li 2 O to minimize structural radioactivity. ''Hands-on'' maintenance is possible on all reactor components outside the blanket. The first wall and blanket are designed to shut the reactor down passively in the event of a loss-of-coolant or loss-of-flow accident. Helium removes heat from the first wall, blanket and shield, and is used in a closed-cycle gas turbine to produce electricity. Energy residing in the plasma ring at the end of the burn is recovered via magnetic expansion. Electrostatic direct conversion is not used in this design. The reactor produces a constant net power of 99 MW(e). (author)

Evaluation of Torsatrons as reactors

International Nuclear Information System (INIS)

Lyon, J.F.; Gulec, K.; Miller, R.L.; El-Guebaly, L.

1994-03-01

Stellarators have significant operational advantages over tokamaks as ignited steady-state reactors. This scoping study, which uses an integrated cost-minimization code that incorporates costing and reactor component models self-consistently with a 1-D energy transport calculation, shows that a torsatron reactor could also be economically competitive with a tokamak reactor. The projected cost of electricity (COE) estimated using the Advanced Reactor Innovation and Evaluation Studies (ARIES) costing algorithms is 65.6 mill/kW(e)h in constant 1992 dollars for a reference 1-GW(e) Compact Torsatron reactor case. The COE is relatively insensitive (<10% variation) over a wide range of assumptions, including variations in the maximum field allowed on the coils, the coil elongation, the shape of the density profile, the beta limit, the confinement multiplier, and the presence of a large loss region for alpha particles. The largest variations in the COE occur for variations in the electrical power output demanded and the plasma-coil separation ratio

Preliminary irradiation test results from the Yankee Atomic Electric Company reactor vessel test irradiation program

International Nuclear Information System (INIS)

Biemiller, E.C.; Fyfitch, S.; Campbell, C.A.

1993-01-01

The Yankee Atomic Electric Company test irradiation program was implemented to characterize the irradiation response of representative Yankee Rowe reactor vessel beltline plate materials and to remove uncertainties in the analysis of existing irradiation data on the Yankee Rowe reactor vessel steel. Plate materials each containing 0.24 w/o copper, but different nickel contents at 0.63 w/o and 0.19 w/o, were heat treated to simulate the Yankee vessel heat treatment (austenitized at 1800 deg F) and to simulate Regulatory Guide 1.99 database materials (austenitized at 1600 deg. F). These heat treatments produced different microstructures so the effect of microstructure on irradiation damage sensitivity could be tested. Because the nickel content of the test plates varied and the copper level was constant, the effect of nickel on irradiation embrittlement was also tested. Correlation monitor material, HSST-02, was included in the program to benchmark the Ford Nuclear Reactor (U. of Michigan Test Reactor) which had never been used for this type of irradiation program. Materials taken from plate surface locations (vs. 1/4T) were included to test whether or not the improved toughness properties of the plate surface layer, resulting from the rapid quench, is maintained after irradiation. If the improved properties are maintained, pressurized thermal shock calculations could utilize this margin. Finally, for one experiment, irradiations were conducted at two irradiation temperatures (500 deg. F and 550 deg. F) to determine the effect of irradiation temperature on embrittlement. The preliminary results of the irradiation program show an increase in T 30 shift of 69 deg. F for a decrease in irradiation temperature of 50 deg. F. The results suggest that for nickel bearing steels, the superior toughness of plate surface material is maintained after irradiation and for the copper content tested, nickel had no apparent effect on irradiation response. No apparent microstructure

Plasma engineering analyses of tokamak reactor operating space

International Nuclear Information System (INIS)

Houlberg, W.; Attenberger, S.E.

1981-01-01

A comprehensive method is presented for analyzing the potential physics operating regime of fusion reactor plasmas with detailed transport codes. Application is made to the tokamak Fusion Engineering Device (FED). The relationships between driven and ignited operation and supplementary heating requirements are examined. The reference physics models give a finite range of density and temperature over which physics objectives can be reached. Uncertainties in the confinement scaling and differences in supplementary heating methods can expand or contract this operating regime even to the point of allowing ignition with the more optimistic models

Cavity temperature and flow characteristics in a gas-core test reactor

Science.gov (United States)

Putre, H. A.

1973-01-01

A test reactor concept for conducting basic studies on a fissioning uranium plasma and for testing various gas-core reactor concepts is analyzed. The test reactor consists of a conventional fuel-element region surrounding a 61-cm-(2-ft-) diameter cavity region which contains the plasma experiment. The fuel elements provide the neutron flux for the cavity region. The design operating conditions include 60-MW reactor power, 2.7-MW cavity power, 200-atm cavity pressure, and an average uranium plasma temperature of 15,000 K. The analytical results are given for cavity radiant heat transfer, hydrogen transpiration cooling, and uranium wire or powder injection.

Instruments for non-destructive evaluation of advanced test reactor inpile tubes

International Nuclear Information System (INIS)

Livingston, R.A.; Beller, L.S.; Edgett, S.M.

1986-01-01

The Advanced Test Reactor is a 250 MW LWR used primarily for irradiation testing of materials contained in inpile tubes that pass through the reactor core. These tubes provided the high pressure and temperature water environment required for the test specimens. The reactor cooling water surrounding the inpile tubes is at much lower pressure and temperature. The structural integrity of the inpile tubes is monitored by routine surveillance to ensure against unplanned reactor shutdowns to replace defective inpile tubes. The improved instruments developed for inpile tube surveillance include a bore profilometer, ultrasonic flaw detetion system and bore diameter gauges. The design and function of these improved instruments is presented

Development of Augmented Spark Impinging Igniter System for Methane Engines

Science.gov (United States)

Marshall, William M.; Osborne, Robin J.; Greene, Sandra E.

2017-01-01

The Lunar Cargo Transportation and Landing by Soft Touchdown (Lunar CATALYST) program is establishing multiple no-funds-exchanged Space Act Agreement (SAA) partnerships with U.S. private sector entities. The purpose of this program is to encourage the development of robotic lunar landers that can be integrated with U.S. commercial launch capabilities to deliver payloads to the lunar surface. NASA can share technology and expertise under the SAA for the benefit of the CATALYST partners. MSFC seeking to vacuum test Augmented Spark Impinging (ASI) igniter with methane and new exciter units to support CATALYST partners and NASA programs. ASI has previously been used/tested successfully at sea-level, with both O2/CH4 and O2/H2 propellants. Conventional ignition exciter systems historically experienced corona discharge issues in vacuum. Often utilized purging or atmospheric sealing on high voltage lead to remedy. Compact systems developed since PCAD could eliminate the high-voltage lead and directly couple the exciter to the spark igniter. MSFC developed Augmented Spark Impinging (ASI) igniter. Successfully used in several sea-level test programs. Plasma-assisted design. Portion of ox flow is used to generate hot plasma. Impinging flows downstream of plasma. Additional fuel flow down torch tube sleeve for cooling near stoichiometric torch flame. Testing done at NASA GRC Altitude Combustion Stand (ACS) facility 2000-lbf class facility with altitude simulation up to around 100,000 ft. (0.2 psia [10 Torr]) via nitrogen driven ejectors. Propellant conditioning systems can provide temperature control of LOX/CH4 up to test article.

Ignition of Aluminum Particles and Clouds

Energy Technology Data Exchange (ETDEWEB)

Kuhl, A L; Boiko, V M

2010-04-07

Here we review experimental data and models of the ignition of aluminum (Al) particles and clouds in explosion fields. The review considers: (i) ignition temperatures measured for single Al particles in torch experiments; (ii) thermal explosion models of the ignition of single Al particles; and (iii) the unsteady ignition Al particles clouds in reflected shock environments. These are used to develop an empirical ignition model appropriate for numerical simulations of Al particle combustion in shock dispersed fuel explosions.

Reactor recirculation pump test loop

International Nuclear Information System (INIS)

Taka, Shusei; Kato, Hiroyuki

1979-01-01

A test loop for a reactor primary loop recirculation pumps (PLR pumps) has been constructed at Ebara's Haneda Plant in preparation for production of PLR pumps under license from Byron Jackson Pump Division of Borg-Warner Corporation. This loop can simulate operating conditions for test PLR pumps with 130 per cent of the capacity of pumps for a 1100 MWe BWR plant. A main loop, primary cooling system, water demineralizer, secondary cooling system, instrumentation and control equipment and an electric power supply system make up the test loop. This article describes the test loop itself and test results of two PLR pumps for Fukushima No. 2 N.P.S. Unit 1 and one main circulation pump for HAZ Demonstration Test Facility. (author)

Operation, test, research and development of the high temperature engineering test reactor (HTTR). FY2003

International Nuclear Information System (INIS)

2005-03-01

The High Temperature Engineering Test Reactor (HTTR) constructed at the Oarai Research Establishment of The Japan Atomic Energy Research Institute (JAERI) is the first high-temperature gas-cooled reactor (HTGR) in Japan, which is a graphite-moderated and helium gas-cooled reactor with 30MW of thermal power. Coolant of helium-gas circulates under the pressure of about 4Mpa, and the reactor inlet and outlet temperature are 395degC and 950degC (maximum), respectively coated particle fuel is used as fuel, and the HTTR core is composed of graphite prismatic blocks. The full power operation of 30MW was attained in December, 2001, and then JAERI received the commissioning license for the HTTR in March, 2002. Since 2002, we have been carrying out rated power operation, safety demonstration tests and several R and Ds, etc., and conducted the high-temperature test operation of 950degC in April, 2004. This report summarizes activities and test results on HTTR operation and maintenance as well as safety demonstration tests and several R and Ds, which were carried out in the fiscal year of 2003 before the high temperature test operation of 950degC. (author)

Ignition parameters and early flame kernel development of laser-ignited combustible gas mixtures

International Nuclear Information System (INIS)

Kopecek, H.; Wintner, E.; Ruedisser, D.; Iskra, K.; Neger, T.

2002-01-01

Full text: Laser induced breakdown of focused pulsed laser radiation, the subsequent plasma formation and thermalization offers a possibility of ignition of combustible gas mixtures free from electrode interferences, an arbitrary choice of the location within the medium and exact timing regardless of the degree of turbulence. The development and the decreasing costs of solid state laser technologies approach the pay-off for the higher complexity of such an ignition system due to several features unique to laser ignition. The feasability of laser ignition was demonstrated in an 1.5 MW(?) natural gas engine, and several investigations were performed to determine optimal ignition energies, focus shapes and laser wavelengths. The early flame kernel development was investigated by time resolved planar laser induced fluorescence of the OH-radical which occurs predominantly in the flame front. The flame front propagation showed typical features like toroidal initial flame development, flame front return and highly increased flame speed along the laser focus axis. (author)

Ignition and burn control in tokamak plasmas

International Nuclear Information System (INIS)

Borrass, K.; Gruber, O.; Lackner, K.; Minardi, E.; Neuhauser, J.; Wilhelm, R.; Wunderlich, R.; Bromberg, L.; Cohn, D.R.

1981-01-01

Different schemes for the control of the thermal instability in an ignited fusion reactor are analysed by zero- and one-dimensional models. Passive stabilization methods considered are ripple-enhanced ion heat conduction, the effect of the major-radius variation of the plasma column in a time-independent vertical field, and the combination of both effects, including the spatial variation of the toroidal-ripple amplitude. Active control methods analysed are high-Q-driven operation and feedback-controlled major-radius variation following different scenarios. One-dimensional analyses taking into account only conductive losses show the existence of a single unstable mode in the energy balance, justifying, under these assumptions, the study of only global control. (author)

Parameter study of a screw-pinch reactor with circular cross-section

International Nuclear Information System (INIS)

Bustraan, M.; Franken, W.M.P.; Klippel, H.Th.; Muysken, M.; Verschuur, K.A.

1977-04-01

In the framework of system studies on pulsed high-β fusion reactors, a parameter study of a reactor based on a screw pinch with a circular cross-section has been performed. The plasma is heated to ignition in two stages. First, the cold plasma is heated by fast implosion in order to guarantee pitch conservation of the inward moving magnetic field lines. The relevant implosion theory has been generalized to a β<1 plasma. In the second stage, an adiabatic compression heats the plasma to the ignition temperature at which point α-particle heating takes over. For stability reasons, β is kept below 0.25. The choice of a particular set of basic parameter values is justified by global design considerations of the reactor. These considerations, e.g. on blanket design and electrotechnical requirements, are presented in some detail. A computer program searches for optimal reactors, i.e. for which at a given thermal output the net efficiency is a maximum. The parameters of a Reference Screw-Pinch Reactor and some other numerical examples are given. The main conclusions are: the net efficiency, although increasing with output energy, is low because of ohmic losses in the compression coil system; the application of sustained fields generated by superconducting coils to reduce these ohmic losses is problematical; a belt-shaped screw pinch in which higher values of β may be reached, improves the net efficiency and alleviates the technical requirements; heating by implosion and adiabatic compression of a plasma with values of β as low as considered here, is inefficient. Therefore, other means of heating the plasma to ignition may be attractive

Role of fission-reactor-testing capabilities in the development of fusion technology

International Nuclear Information System (INIS)

Hsu, P.Y.; Deis, G.A.; Longhurst, G.R.; Miller, L.G.; Schmunk, R.E.; Takata, M.L.; Watts, K.D.

1981-01-01

Testing of fusion materials and components in fission reactors will be increasingly important in the future due to the near-term lack of fusion engineering test devices, and the long-term high demand for testing when fusion reactors become available. Fission testing is capable of filling many gaps in fusion reactor design information, and thus should be aggressively pursued. EG and G Idaho has investigated the application of fission testing in three areas, which are discussed in this paper. First, we investigated radiation damage to magnet insulators. This work is now continuing with the use of an improved test capsule. Second, a study was performed which indicated that a fission-suppressed hybrid blanket module could be effectively tested in a reactor such as the Engineering Test Reactor (ETR), closely reproducing the predicted performance in a fusion environment. Finally, we explored a conceptual design for a fission-based Integrated Test Facility (ITF), which can accommodate entire First Wall/Blanket (FW/B) modules for testing in a nuclear environment, simultaneously satisfying many of the FW/B test requirements. This ITF can provide a cyclic neutron/gamma flux, as well as the necessary module support functions

A comparative experimental study on engine operating on premixed charge compression ignition and compression ignition mode

Directory of Open Access Journals (Sweden)

Bhiogade Girish E.

2017-01-01

Full Text Available New combustion concepts have been recently developed with the purpose to tackle the problem of high emissions level of traditional direct injection Diesel engines. A good example is the premixed charge compression ignition combustion. A strategy in which early injection is used causing a burning process in which the fuel burns in the premixed condition. In compression ignition engines, soot (particulate matter and NOx emissions are an extremely unsolved issue. Premixed charge compression ignition is one of the most promising solutions that combine the advantages of both spark ignition and compression ignition combustion modes. It gives thermal efficiency close to the compression ignition engines and resolves the associated issues of high NOx and particulate matter, simultaneously. Premixing of air and fuel preparation is the challenging part to achieve premixed charge compression ignition combustion. In the present experimental study a diesel vaporizer is used to achieve premixed charge compression ignition combustion. A vaporized diesel fuel was mixed with the air to form premixed charge and inducted into the cylinder during the intake stroke. Low diesel volatility remains the main obstacle in preparing premixed air-fuel mixture. Exhaust gas re-circulation can be used to control the rate of heat release. The objective of this study is to reduce exhaust emission levels with maintaining thermal efficiency close to compression ignition engine.

D-3He fuel cycles for neutron lean reactors

International Nuclear Information System (INIS)

Kernbichler, W.; Miley, G.H.; Heindler, M.

1989-01-01

The intrinsic potential of D-3He as a reactor fuel is investigated for a large range of 3He to D density ratios. A steady-state zero-dimensional reactor model is developed in which much care is attributed to a proper treatment of fast fusion products. Useful ranges of reactor parameters as well as temperature-density windows for driven and ignited operation are identified. Various figures of merit are calculated, such as power densities, net power production, neutron production, tritium load and radiative power. These results suggest several optimistic conclusions about the performance of D-3He as a reactor fuel

The roles of EBR-II and TREAT [Transient Reactor Test] in establishing liquid metal reactor safety

International Nuclear Information System (INIS)

Sackett, J.I.; Lehto, W.K.; Solbrig, C.W.

1990-01-01

This paper examines the role of the Experimental Breeder Reactor II (EBR-II) and Transient Reactor Test (TREAT) facilities in contributing to the understanding and resolution of key safety issues in liquid metal reactor safety during the decade of the 80's. Fuels and materials testing has been carried out to address questions on fuels behavior during steady-state and upset conditions. In addition, EBR-II has conducted plant tests to demonstrate passive response to ATWS events and to develop control and diagnostic strategies for safe operation of advanced LMRs. TREAT and EBR-II complement each other and between them provide a transient testing capability that covers the whole range of concerns during overpower conditions. EBR-II, with use of the special Automatic Control Rod Drive System, can generate power change rates that overlap the lower end of the TREAT capability. 21 refs

Design of ignition targets for the National Ignition Facility

International Nuclear Information System (INIS)

Haan, S.W.; Dittrich, T.R.; Marinak, M.M.; Hinkel, D.E.

1999-01-01

This is a brief update on the work being done to design ignition targets for the National Ignition Facility. Updates are presented on three areas of current activity : improvements in modeling, work on a variety of targets spanning the parameter space of possible ignition targets ; and the setting of specifications for target fabrication and diagnostics. Highlights of recent activity include : a simulation of the Rayleigh-Taylor instability growth on an imploding capsule, done in 3D on a 72degree by 72degree wedge, with enough zones to resolve modes out to 100 ; and designs of targets at 250eV and 350eV, as well as the baseline 300 eV ; and variation of the central DT gas density, which influences both the Rayleigh-Taylor growth and the smoothness of the DT ice layer

Off reactor testings. Technological engineering applicative research

International Nuclear Information System (INIS)

Doca, Cezar

2001-01-01

By the end of year 2000 over 400 nuclear electro-power units were operating world wide, summing up a 350,000 MW total capacity, with a total production of 2,300 TWh, representing 16% of the world's electricity production. Other 36 units, totalizing 28,000 MW, were in construction, while a manifest orientation towards nuclear power development was observed in principal Asian countries like China, India, Japan and Korea. In the same world's trend one find also Romania, the Cernavoda NPP Unit 1 generating electrical energy into the national system beginning with 2 December 1996. Recently, the commercial contract was completed for finishing the Cernavoda NPP Unit 2 and launching it into operation by the end of year 2004. An important role in developing the activity of research and technological engineering, as technical support for manufacturing the CANDU type nuclear fuel and supplying with equipment the Cernavoda units, was played by the Division 7 TAR of the INR Pitesti. Qualification testings were conducted for: - off-reactor CANDU type nuclear fuel; - FARE tools, pressure regulators, explosion proof panels; channel shutting, as well as functional testing for spare pushing facility as a first step in the frame of the qualification tests for the charging/discharging machine (MID) 4 and 5 endings. Testing facilities are described, as well as high pressure hot/cool loops, measuring chains, all of them fulfilling the requirements of quality assurance. The nuclear fuel off-reactor tests were carried out to determine: strength; endurance; impact, pressure fall and wear resistance. For Cernavoda NPP equipment testings were carried out for: the explosion proof panels, pressure regulators, behaviour to vibration and wear of the steam generation tubings, effects of vibration upon different electronic component, channel shutting (for Cernavoda Unit 2), MID operating at 300 and 500 cycles. A number of R and D programs were conducted in the frame of division 7 TAR of INR

Needs for development in nondestructive testing for advanced reactor systems

International Nuclear Information System (INIS)

McClung, R.W.

1978-01-01

The needs for development of nondestructive testing (NDT) techniques and equipment were surveyed and analyzed relative to problem areas for the Liquid-Metal Fast Breeder Reactor, the Molten-Salt Breeder Reactor, and the Advanced Gas-Cooled Reactor. The paper first discusses the developmental needs that are broad-based requirements in nondestrutive testing, and the respective methods applicable, in general, to all components and reactor systems. Next, the requirements of generic materials and components that are common to all advanced reactor systems are examined. Generally, nondestructive techniques should be improved to provide better reliability and quantitativeness, improved flaw characterization, and more efficient data processing. Specific recommendations relative to such methods as ultrasonics, eddy currents, acoustic emission, radiography, etc., are made. NDT needs common to all reactors include those related to materials properties and degradation, welds, fuels, piping, steam generators, etc. The scope of applicability ranges from initial design and material development stages through process control and manufacturing inspection to in-service examination

Moving-ring field-reversed mirror reactor

International Nuclear Information System (INIS)

Smith, A.C. Jr.; Ashworth, C.P.; Abreu, K.E.

1981-01-01

We describe a first prototype fusion reactor design of the Moving-Ring Field-Reversed Mirror Reactor. The fusion fuel is confined in current-carrying rings of magnetically-field-reversed plasma. The plamsa rings, formed by a coaxial plasma gun, are magnetically compressed to ignition temperature while they are being injected into the reactor's burner section. DT ice pellets refuel the rings during the burn at a rate which maintains constant fusion power. A steady train of plasma rings moves at constant speed through the reactor under the influence of a slightly diverging magnetic field. The aluminum first wall and breeding zone structure minimize induced radioactivity; hands-on maintenance is possible on reactor components outside the breeding blanket. Helium removes the heat from the Li 2 O tritium breeding blanket and is used to generate steam. The reactor produces a constant, net power of 376 MW

Nuclear fuels for material test reactors

International Nuclear Information System (INIS)

Ramanathan, L.V.; Durazzo, M.; Freitas, C.T. de

1982-01-01

Experimental results related do the development of nuclear fuels for reactors cooled and moderated by water have been presented cylindrical and plate type fuels have been described in which the core consists of U compouns dispersed in an Al matrix and is clad with aluminium. Fabrication details involving rollmilling, swaging or hot pressing have been described. Corrosion and irradiation test results are also discussed. The performance of the different types of fuels indicates that it is possible to locally fabricate fuel plates with U 3 O 8 +Al cores (20% enriched U) for use in operating Brazilian research reactors. (Author) [pt

A review of experiments and results from the transient reactor test (TREAT) facility

International Nuclear Information System (INIS)

Deitrich, L. W.

1998-01-01

The TREAT Facility was designed and built in the late 1950s at Argonne National Laboratory to provide a transient reactor for safety experiments on samples of reactor fuels. It first operated in 1959. Throughout its history, experiments conducted in TREAT have been important in establishing the behavior of a wide variety of reactor fuel elements under conditions predicted to occur in reactor accidents ranging from mild off normal transients to hypothetical core disruptive accidents. For much of its history, TREAT was used primarily to test liquid-metal reactor fuel elements, initially for the Experimental Breeder Reactor-II (EBR-II), then for the Fast Flux Test Facility (FFTF), the Clinch River Breeder Reactor Plant (CRBRP), the British Prototype Fast Reactor (PFR), and finally, for the Integral Fast Reactor (IFR). Both oxide and metal elements were tested in dry capsules and in flowing sodium loops. The data obtained were instrumental in establishing the behavior of the fuel under off-normal and accident conditions, a necessary part of the safety analysis of the various reactors. In addition, TREAT was used to test light-water reactor (LWR) elements in a steam environment to obtain fission-product release data under meltdown conditions. Studies are now under way on applications of TREAT to testing of the behavior of high-burnup LWR elements under reactivity-initiated accident (RIA) conditions using a high-pressure water loop

Qualification of Magnesium/Teflon/Viton Pyrotechnic Composition Used in Rocket Motors Ignition System

Directory of Open Access Journals (Sweden)

Luciana de Barros

2016-04-01

Full Text Available The application of fluoropolymers in high-energy-release pyrotechnic compositions is common in the space and defense areas. Pyrotechnic compositions of magnesium/Teflon/Viton are widely used in military flares and pyrogen igniters for igniting the solid propellant of a rocket motor. Pyrotechnic components are considered high-risk products as they may cause catastrophic accidents if initiated or ignited inadvertently. To reduce the hazards involved in the handling, storage and transportation of these devices, the magnesium/Teflon/Viton composition was subjected to various sensitivity tests, DSC and had its stability and compatibility tested with other materials. This composition obtained satisfactory results in all the tests, which qualifies it as safe for production, handling, use, storage and transportation.

Operation, test, research and development of the High Temperature Engineering Test Reactor (HTTR). FY2014

International Nuclear Information System (INIS)

2016-02-01

The High Temperature Engineering Test Reactor (HTTR), a graphite-moderated and helium gas-cooled reactor with 30 MW of thermal power, constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency is the first high-temperature gas-cooled reactor (HTGR) in Japan. The HTTR was attained at the full power operation of 30 MW in December 2001 and achieved the 950degC of coolant outlet temperature at outside of the reactor pressure vessel in June 2004. To establish and upgrade basic technologies for HTGRs, we have obtained demonstration test data necessary for several R and Ds, and accumulated operation and maintenance experience of HTGRs throughout the HTTR's operation such as rated power operations, safety demonstration tests and long-term high temperature operations, and so on. In fiscal year 2014, we started to apply the application document of reactor installation license for the HTTR to prove conformity with the new research reactor's safety regulatory requirements taken effect from December 2013. We had been making effort to restart the HTTR which was stopped since the 2011 by the Pacific coast of Tohoku Earthquake. This report summarizes activities and results of HTTR operation, maintenance, and several R and Ds, which were carried out in the fiscal year 2014. (author)

Simplified dynamic simulation of a traveling wave nuclear reactor; Simulacion dinamica simplificada de un reactor nuclear de onda viajera

Energy Technology Data Exchange (ETDEWEB)

Sanchez M, H.; Espinosa P, G. [Universidad Autonoma Metropolitana, Unidad Iztapalapa, Area de Ingenieria en Recursos Energeticos, San Rafael Atlixco No. 186, Col. Vicentina, 09340 Ciudad de Mexico (Mexico); Francois, J. L. [UNAM, Facultad de Ingenieria, Paseo Cuauhnahuac 8532, Jiutepec 62550, Morelos (Mexico); Lopez S, R., E-mail: heribertosanchez7@hotmail.com [ININ, Carretera Mexico-Toluca s/n, 52750 Ocoyoacac, Estado de Mexico (Mexico)

2016-09-15

In this work the nuclear fuel burn wave in a fast traveling wave reactor (TWR) is presented, using the reduced model of the neutron diffusion equation, considering only the axial component, and the equations of the transuranic dynamics of U-Pu and a radionuclide of Pu. Two critical zones of the reactor are considered, one enriched with U-Pu called ignition zone and the other impoverished zone or of U-238, named breeding zone. Occupying Na as refrigerant within TWR, and Fe as structural material; both are present in the ignition and breeding zones. Considering as a fissile material the Pu, since by neutron capture the U is transformed into Pu, thus increasing the quantity of Pu more than that of U; in this way the fuel burn stability with the wave dynamics is understood. The calculation of the results was approached numerically to determine the temporal space evolution of the neutron flux in this system and of the main isotopes involved in the burning process. (Author)

Safety analysis calculations for research and test reactors

Energy Technology Data Exchange (ETDEWEB)

Chen, S Y; MacDonald, R; MacFarlane, D [Argonne National Laboratory, Argonne, IL (United States)

1983-08-01

The goal of the RERTR (Reduced Enrichment in Research and Test Reactor) Program at ANL is to provide technical means for conversion of research and test reactors from HEU (High-Enrichment Uranium) to LEU (Low-Enrichment Uranium) fuels. In exploring the feasibility of conversion, safety considerations are a prime concern; therefore, safety analyses must be performed for reactors undergoing the conversion. This requires thorough knowledge of the important safety parameters for different types of reactors for both HEU and LEU fuel. Appropriate computer codes are needed to predict transient reactor behavior under postulated accident conditions. In this discussion, safety issues for the two general types of reactors i.e., the plate-type (MTR-type) reactor and the rod-type (TRIGA-type) reactor, resulting from the changes associated with LEU vs. HEU fuels, are explored. The plate-type fuels are typically uranium aluminide (UAl{sub x}) compounds dispersed in aluminum and clad with aluminum. Moderation is provided by the water coolant. Self shut-down reactivity coefficients with EU fuel are entirely a result of coolant heating, whereas with LEU fuel there is an additional shut down contribution provided by the direct heating of the fuel due to the Doppler coefficient. In contrast, the rod-type (TRIGA) fuels are mixtures of zirconium hydride, uranium, and erbium. This fuel mixture is formed into rods ( {approx} 1 cm diameter) and clad with stainless steel or Incoloy. In the TRIGA fuel the self-shutdown reactivity is more complex, depending on heating of the fuel rather than the coolant. The two most important mechanisms in providing this feedback are: spectral hardening due to neutron interaction with the ZrH moderator as it is heated and Doppler broadening of resonances in erbium and U-238. Since these phenomena result directly from heating of the fuel, and do not depend on heat transfer to the moderator/coolant, the coefficients are prompt acting. Results of transient

International benchmark on the natural convection test in Phenix reactor

International Nuclear Information System (INIS)

Tenchine, D.; Pialla, D.; Fanning, T.H.; Thomas, J.W.; Chellapandi, P.; Shvetsov, Y.; Maas, L.; Jeong, H.-Y.; Mikityuk, K.; Chenu, A.; Mochizuki, H.; Monti, S.

2013-01-01

Highlights: ► Phenix main characteristics, instrumentation and natural convection test are described. ► “Blind” calculations and post-test calculations from all the participants to the benchmark are compared to reactor data. ► Lessons learned from the natural convection test and the associated calculations are discussed. -- Abstract: The French Phenix sodium cooled fast reactor (SFR) started operation in 1973 and was stopped in 2009. Before the reactor was definitively shutdown, several final tests were planned and performed, including a natural convection test in the primary circuit. During this natural convection test, the heat rejection provided by the steam generators was disabled, followed several minutes later by reactor scram and coast-down of the primary pumps. The International Atomic Energy Agency (IAEA) launched a Coordinated Research Project (CRP) named “control rod withdrawal and sodium natural circulation tests performed during the Phenix end-of-life experiments”. The overall purpose of the CRP was to improve the Member States’ analytical capabilities in the field of SFR safety. An international benchmark on the natural convection test was organized with “blind” calculations in a first step, then “post-test” calculations and sensitivity studies compared with reactor measurements. Eight organizations from seven Member States took part in the benchmark: ANL (USA), CEA (France), IGCAR (India), IPPE (Russian Federation), IRSN (France), KAERI (Korea), PSI (Switzerland) and University of Fukui (Japan). Each organization performed computations and contributed to the analysis and global recommendations. This paper summarizes the findings of the CRP benchmark exercise associated with the Phenix natural convection test, including blind calculations, post-test calculations and comparisons with measured data. General comments and recommendations are pointed out to improve future simulations of natural convection in SFRs

Rupture tests with reactor pressure vessel head models

International Nuclear Information System (INIS)

Talja, H.; Keinaenen, H.; Hosio, E.; Pankakoski, P.H.; Rahka, K.

2003-01-01

In the LISSAC project (LImit Strains in Severe ACcidents), partly funded by the EC Nuclear Fission and Safety Programme within the 5th Framework programme, an extensive experimental and computational research programme is conducted to study the stress state and size dependence of ultimate failure strains. The results are aimed especially to make the assessment of severe accident cases more realistic. For the experiments in the LISSAC project a block of material of the German Biblis C reactor pressure vessel was available. As part of the project, eight reactor pressure vessel head models from this material (22 NiMoCr 3 7) were tested up to rupture at VTT. The specimens were provided by Forschungszentrum Karlsruhe (FzK). These tests were performed under quasistatic pressure load at room temperature. Two specimens sizes were tested and in half of the tests the specimens contain holes describing the control rod penetrations of an actual reactor pressure vessel head. These specimens were equipped with an aluminium liner. All six tests with the smaller specimen size were conducted successfully. In the test with the large specimen with holes, the behaviour of the aluminium liner material proved to differ from those of the smaller ones. As a consequence the experiment ended at the failure of the liner. The specimen without holes yielded results that were in very good agreement with those from the small specimens. (author)

Shock timing technique for the National Ignition Facility

International Nuclear Information System (INIS)

Munro, David H.; Celliers, Peter M.; Collins, Gilbert W.; Gold, David M.; Silva, Luiz B. da; Haan, Steven W.; Cauble, Robert C.; Hammel, Bruce A.; Hsing, Warren W.

2001-01-01

Among the final shots at the Nova laser [Campbell et al., Rev. Sci. Instrum. 57, 2101 (1986)] was a series testing the VISAR (velocity interferometry system for any reflector) technique that will be the primary diagnostic for timing the shocks in a NIF (National Ignition Facility) ignition capsule. At Nova, the VISAR technique worked over the range of shock strengths and with the precision required for the NIF shock timing job--shock velocities in liquid D 2 from 12 to 65 μm/ns with better than 2% accuracy. VISAR images showed stronger shocks overtaking weaker ones, which is the basis of the plan for setting the pulse shape for the NIF ignition campaign. The technique is so precise that VISAR measurements may also play a role in certifying beam-to-beam and shot-to-shot repeatability of NIF laser pulses

Linear induction accelerator requirements for ion fast ignition

International Nuclear Information System (INIS)

Logan, G.

1998-01-01

induction linacs, the purpose of this memo is to explore possible new features and characteristic parameters that induction linacs would need to meet the stringent requirements for beam quality and compression (sufficiently low longitudinal and transverse thermal spread) for ion driven fast ignition. Separately, Ed Lee at LBNL is looking at heavy-ion synchrotrons to meet similar fast ignition requirements. Parameters relating to cost (e.g, total beam-line length and transport quads, total core volt-seconds and power switching) have to be considered in addition to meeting the challenging beam quality requirements for fast ignition compared to conventional HIF. The aim of this preliminary study is to motivate, after critical debate, taking a next step to do more detailed designs, particle simulations, and experimental tests of the most critical accelerator elements and focusing optics, to further assess the feasibility of ion-driven fast ignition

Spark Ignition LPG for Hydrogen Gas Combustion the Reduction Furnace ME-11 Process

International Nuclear Information System (INIS)

Achmad Suntoro

2007-01-01

Reverse engineering method for automatic spark-ignition system of LPG to burn hydrogen gaseous in the reducing process of ME-11 furnace has been successfully implemented using local materials. A qualitative study to the initial behaviour of the LPG flame system has created an idea by modification to install an automatic spark-ignition of the LPG on the reducing furnace ME-11. The automatic spark-ignition system has been tested and proved working well. (author)

Analysis of cyclic variations during mode switching between spark ignition and controlled auto-ignition combustion operations

OpenAIRE

Chen, T; Zhao, H; Xie, H; He, B

2014-01-01

© IMechE 2014. Controlled auto-ignition, also known as homogeneous charge compression ignition, has been the subject of extensive research because of their ability to provide simultaneous reductions in fuel consumption and NOx emissions from a gasoline engine. However, due to its limited operation range, switching between controlled auto-ignition and spark ignition combustion is needed to cover the complete operating range of a gasoline engine for passenger car applications. Previous research...

The National Ignition Facility. The path to ignition and inertial fusion energy

International Nuclear Information System (INIS)

Eric Storm

2010-01-01

Complete text of publication follows. The National Ignition Facility (NIF), the world's largest and most energetic laser system built for studying inertial confinement fusion (ICF) and high-energy-density (HED) science, is now operational at Lawrence Livermore National Laboratory (LLNL). NIF's 192 beams are capable of producing 1.8 MJ and 500 TW of ultraviolet light and are configured to create pressures as high as 100 GB, matter temperatures approaching 10 9 and densities over 1000 g/cm 3 . With these capabis70lities, the NIF will enable exploring scientific problems in strategic defense, basic science and fusion energy. One of the early NIF campaigns is focusing on demonstrating laboratory-scale thermonuclear ignition and burn to produce net fusion energy gains of 10-20 with 1.2 to 1.4 MJ of 0.35 μm light. NIF ignition experiments began late in FY2009 as part of the National Ignition Campaign (NIC). Participants of NIC include LLNL, General Atomics, Los Alamos National Laboratory, Sandia National Laboratory, and the University of Rochester Laboratory for Energetics (LLE) as well as variety of national and international collaborators. The results from these initial experiments show great promise for the relatively near-term achievement of ignition. Capsule implosion experiments at energies up to 1.2 MJ have demonstrated laser energetics, radiation temperatures, and symmetry control that scale to ignition conditions. Of particular importance is the demonstration of peak hohlraum temperatures near 300 eV with low overall backscatter less than 10%. Cryogenic target capability and additional diagnostics are being installed in preparation for layered target deuterium-tritium implosions to be conducted later in 2010. The goal for NIC is to demonstrate a predictable fusion experimental platform by the end of 2012. Successful demonstration of ignition and net energy gain on NIF will be a major step towards demonstrating the feasibility of Inertial Fusion Energy (IFE) and

Auto-Ignition and Combustion of Diesel Fuel in a Constant-Volume Bomb

Science.gov (United States)

Selden, Robert F

1938-01-01

Report presents the results of a study of variations in ignition lag and combustion associated with changes in air temperature and density for a diesel fuel in a constant-volume bomb. The test results have been discussed in terms of engine performance wherever comparisons could be drawn. The most important conclusions drawn from this investigation are: the ignition lag was essentially independent of the injected fuel quantity. Extrapolation of the curves for the fuel used shows that the lag could not be greatly decreased by exceeding the compression-ignition engines. In order to obtain the best combustion and thermal efficiency, it was desirable to use the longest ignition lag consistent with a permissible rate of pressure rise.

Requirements, needs, and concepts for a new broad-application test reactor

International Nuclear Information System (INIS)

Ryskamp, J.M.; Fletcher, C.D.; Denison, A.B.; Liebenthal, J.L.

1992-01-01

For a variety of reasons, including (a) the increasing demands of the 1990s regulatory environment, (b) limited existing test capactiy and capability to satisfy projected future testing missions, and (c) an expected increasing need for nuclear information to support development of advanced reactors, there is a need for requirements and preliminary concepts for a new broad-application test reactor (BATR). These requirements must include consideration not only for a broad range of projected testing missions but also for current and projected regulatory compliance and safety requirements. The requirements will form the basis for development and assessment of preconceptual reactor designs and lead to the identification of key technologies to support the government's long-term strategic and programmatic planning. This paper outlines the need for a new BATR and suggests a few preliminary reactor concepts that can meet that need

Thermal and fast reactor benchmark testing of ENDF/B-6.4

International Nuclear Information System (INIS)

Liu Guisheng

1999-01-01

The benchmark testing for B-6.4 was done with the same benchmark experiments and calculating method as for B-6.2. The effective multiplication factors k eff , central reaction rate ratios of fast assemblies and lattice cell reaction rate ratios of thermal lattice cell assemblies were calculated and compared with testing results of B-6.2 and CENDL-2. It is obvious that 238 U data files are most important for the calculations of large fast reactors and lattice thermal reactors. However, 238 U data in the new version of ENDF/B-6 have not been renewed. Only data of 235 U, 27 Al, 14 N and 2 D have been renewed in ENDF/B-6.4. Therefor, it will be shown that the thermal reactor benchmark testing results are remarkably improved and the fast reactor benchmark testing results are not improved

Conceptual design of fusion experimental reactor (FER)

International Nuclear Information System (INIS)

1984-01-01

Conceptual Design of Fusion Experimental Reactor (FER) of which the objective will be to realize self-ignition with D-T reaction is reported. Mechanical Configurations of FER are characterized with a noncircular plasma and a double-null divertor. The primary aim of design studies is to demonstrate fissibility of reactor structures as compact and simple as possible with removable torus sectors. The structures of each component such as a first-wall, blanket, shielding, divertor, magnet and so on have been designed. It is also discussed about essential reactor plant system requirements. In addition to the above, a brief concept of a steady-state reactor based on RF current drive is also discussed. The main aim, in this time, is to examine physical studies of a possible RF steady-state reactor. (author)

Operation, test, research and development of the high temperature engineering test reactor (HTTR). (FY2005)

International Nuclear Information System (INIS)

2007-03-01

The High Temperature Engineering Test Reactor (HTTR) constructed at the Oarai Research and Development Center of the Japan Atomic Energy Agency (JAEA) is the first high-temperature gas-cooled reactor (HTGR) in Japan, which is a graphite-moderated and helium gas-cooled reactor with 30 MW of thermal power. The full power operation of 30 MW was attained in December, 2001, and then JAERI (JAEA) received the commissioning license for the HTTR in March, 2002. Since 2002, we have been carrying out rated power operation, safety demonstration tests and several R and Ds, etc., and conducted the high-temperature test operation of 950degC in April, 2004. In fiscal 2005 year, periodical inspection and overhaul of reactivity control system were conducted, and safety demonstration tests were promoted. This report summarizes activities and test results on HTTR operation and maintenance as well as safety demonstration tests and several R and Ds, which were carried out in the fiscal year of 2005. (author)

Assessment of the slowly-imploding liner (LINUS) fusion reactor concept

International Nuclear Information System (INIS)

Miller, R.L.; Krakowski, R.A.

1980-01-01

Prospects for the slowly-imploding liner (LINUS) fusion reactor concept are reviewed. The concept envisages the nondestructive, repetitive and reversible implosion of a liquid-metal cylindrical annulus (liner) onto field-reversed DT plasmoids. Adiabatic heating of the plasmoid to ignition at ultra-high magnetic fields results in a compact, high power density fusion reactor with unique solutions to several technological problems and potentially favorable economics

Testing of a transport cask for research reactor spent fuel

International Nuclear Information System (INIS)

Mourao, Rogerio P.; Silva, Luiz Leite da; Miranda, Carlos A.; Mattar Neto, Miguel; Quintana, Jose F.A.; Saliba, Roberto O.; Novara, Oscar E.

2011-01-01

Since the beginning of the last decade three Latin American countries which operate research reactors - Argentina, Brazil and Chile - have been joining efforts to improve the regional capability in the management of spent fuel elements from the reactors operated in the region. As a step in this direction, a packaging for the transport of irradiated fuel from research reactors was designed by a tri-national team and a half-scale model for MTR fuel constructed in Argentina and tested in Brazil. Two test campaigns have been carried out so far, covering both normal conditions of transportation and hypothetical accident conditions. Although the specimen has not successfully performed the tests, its overall performance was considered very satisfactory, and improvements are being introduced to the design. A third test sequence is planned for 2011. (author)

Chinese nuclear heating test reactor and demonstration plant

International Nuclear Information System (INIS)

Wang Dazhong; Ma Changwen; Dong Duo; Lin Jiagui

1992-01-01

In this report the importance of nuclear district heating is discussed. From the viewpoint of environmental protection, uses of energy resources and transport, the development of nuclear heating in China is necessary. The development program of district nuclear heating in China is given in the report. At the time being, commissioning of the 5 MW Test Heating Reactor is going on. A 200 MWt Demonstration Plant will be built. In this report, the main characteristics of these reactors are given. It shows this type of reactor has a high inherent safety. Further the report points out that for this type of reactor the stability is very important. Some experimental results of the driving facility are included in the report. (orig.)

Structural features of the Compact Ignition Tokamak

International Nuclear Information System (INIS)

Citrolo, J.; Brown, G.; Rogoff, P.

1987-01-01

The Compact Ignition Tokamak (CIT) is undergoing preliminary structural design and definitions. It will be relatively inexpensive with ignition capabilities. During the definition phase it was concluded that the TF coil should be assembled from the laminate copper-Inconel plates since copper alone cannot sustain the expected magnetic and thermal loads. An extensive test program is being initiated to investigate the various materials, and their elastic and inelastic response and to develop the constitutive equations required for the selection of design criteria and for the stress analysis of this device. Finite element analysis nonlinear material capabilities are being used to study, predict and correlate the machine behavior

Advanced Test Reactor National Scientific User Facility Partnerships

International Nuclear Information System (INIS)

Marshall, Frances M.; Allen, Todd R.; Benson, Jeff B.; Cole, James I.; Thelen, Mary Catherine

2012-01-01

In 2007, the United States Department of Energy designated the Advanced Test Reactor (ATR), located at Idaho National Laboratory, as a National Scientific User Facility (NSUF). This designation made test space within the ATR and post-irradiation examination (PIE) equipment at INL available for use by researchers via a proposal and peer review process. The goal of the ATR NSUF is to provide researchers with the best ideas access to the most advanced test capability, regardless of the proposer's physical location. Since 2007, the ATR NSUF has expanded its available reactor test space, and obtained access to additional PIE equipment. Recognizing that INL may not have all the desired PIE equipment, or that some equipment may become oversubscribed, the ATR NSUF established a Partnership Program. This program enables and facilitates user access to several university and national laboratories. So far, seven universities and one national laboratory have been added to the ATR NSUF with capability that includes reactor-testing space, PIE equipment, and ion beam irradiation facilities. With the addition of these universities, irradiation can occur in multiple reactors and post-irradiation exams can be performed at multiple universities. In each case, the choice of facilities is based on the user's technical needs. Universities and laboratories included in the ATR NSUF partnership program are as follows: (1) Nuclear Services Laboratories at North Carolina State University; (2) PULSTAR Reactor Facility at North Carolina State University; (3) Michigan Ion Beam Laboratory (1.7 MV Tandetron accelerator) at the University of Michigan; (4) Irradiated Materials at the University of Michigan; (5) Harry Reid Center Radiochemistry Laboratories at University of Nevada, Las Vegas; (6) Characterization Laboratory for Irradiated Materials at the University of Wisconsin-Madison; (7) Tandem Accelerator Ion Beam. (1.7 MV terminal voltage tandem ion accelerator) at the University of Wisconsin

WWER type reactor primary loop imitation on large test loop facility in MARIA reactor

International Nuclear Information System (INIS)

Moldysh, A.; Strupchevski, A.; Kmetek, Eh.; Spasskov, V.P.; Shumskij, A.M.

1982-01-01

At present in Poland in cooperation with USSR a nuclear water loop test facility (WL) in 'MARIA' reactor in Sverke is under construction. The program objective is to investigate processes occuring in WWER reactor under emergency conditions, first of all after the break of the mainprimary loop circulation pipe-line. WL with the power of about 600 kW consists of three major parts: 1) an active loop, imitating the undamaged loops of the WWER reactor; 2) a passive loop assignedfor modelling the broken loop of the WWER reactor; 3) the emergency core cooling system imitating the corresponding full-scale system. The fuel rod bundle consists of 18 1 m long rods. They were fabricated according to the standard WWER fuel technology. In the report some general principles of WWERbehaviour imitation under emergency conditions are given. They are based on the operation experience obtained from 'SEMISCALE' and 'LOFT' test facilities in the USA. A description of separate modelling factors and criteria effects on the development of 'LOCA'-type accident is presented (the break cross-section to the primary loop volume ratio, the pressure differential between inlet and outlet reactor chambers, the pressure drop rate in the loop, the coolant flow rate throuh the core etc.). As an example a comparison of calculated flow rate variations for the WWER-1000 reactor and the model during the loss-of-coolant accident with the main pipe-line break at the core inlet is given. Calculations have been carried out with the use of TECH'-M code [ru