The Rayleigh-Taylor instability in the spherical pinch

The spherical pinch (SP) concept is an outgrowth of the inertial confinement model (ICF). Unlike the ICF, where instabilities, especially the Rayleigh-Taylor instability, have been studied extensively, the instability study of the spherical pinch has just begun. The Rayleigh-Taylor instability is investigated for the first time in the SP in the present work. By using the simple condition for the Rayleigh-Taylor instability p·<0 (density and pressure gradients have opposite direction), we have qualitatively identified the regions for development of instabilities in the SP. It is found that the explosion phase (central discharge) is stable and instabilities take place in the imploding phase. However, the growth rate for the instability is not in exponential form, and the appearance of the Rayleigh-Taylor instability does not prevent the main shock wave from converging to the center of the sphere.     

Interaction of imploding shock waves with expanding central plasma in spherical pinch experiments: Simulation analysis

In the spherical pinch scheme, the hot D-T plasma produced in the center of the high pressure spherical vessel is confined by means of imploding shock waves launched from the periphery of the vessel for a time sufficiently long to achieve break-even conditions for plasma fusion. Theoretical studies on spherical pinch made so far have been limited up to the conditions of substantial expansion of the central plasma and the well-defined time delay between the creation of central plasma and the launching of the peripheral shock which led to the conclusion that, in realistic situations of SP experiments, negative time delays should be adopted, i.e., the launching of the imploding shock wave should precede the formation of the central plasma. However, the interaction of converging shock wave with the central plasma causing an additional heating and compression of the central plasma favoring plasma fusion conditions was not taken into account. Starting from the hydrodynamic equations of the system, the proposed simulation code deals with the propagation of converging shock waves and its interaction with the expanding central plasma. Considering the above-mentioned interaction in a self-consistent manner, the temporal evolution of temperature of central plasma is studied. Some results of the numerical simulation on the dynamics of shock wave propagation are also compared with the predictions of point strong explosing theory.     

The need and prospects for improved fusion reactors

Conceptual fusion reactor studies over the past 10–15 yr have projected systems that may be too large, complex, and costly to be of commercial interest. One main direction for improved fusion reactors points toward smaller, higher-power-density approaches. First-order economic issues (i.e., unit direct cost and cost of electricity) are used to support the need for more compact fusion reactors. The results of a number of recent conceptual designs of reversed-field pinch, spheromak, and tokamak fusion reactors are summarized as examples of more compact approaches. While a focus has been placed on increasing the fusion-power-core mass power density beyond the minimum economic threshold of 100–200 kWe/tonne, other means by which the overall attractiveness of fusion as a long-term energy source are also addressed.Nomenclature a Plasma minor radius at outboard equatorial plane (m) - A Plasma aspect ratioR _T /a - AC Annual charges ($/yr) - b Plasma minor radius in vertical direction (m) - B Magentic field at plasma or blanket (T) - B _c Magnetic field at the coil (T) - B Toroidal magnetic field (T) - B Poloidal magnetic field (T) - BOP Balance of plant - C Coil - COE Cost of electricity (mills/kWeh) - CRFPR Compact RFP reactor - CT Compact torus (FRC or spheromak) - c _FPC Unit cost of fusion power core ($/kg) - DC Direct cost ($) - DZP Dense Z-pinch - E Escalation rate (1/yr) - EDC Escalation during construction ($) - ET Elongated tokamak - F Annual fuel charges ($/yr) - FC Component of UDC not strongly dependent or FPC size ($/kWe) - FW First wall - FPC Fusion power core - f _Aux Fraction of gross electric power recirculated to BOP - f ₁ (IC+IDC+EDC)/DC - f ₂ (O&M + SCR + F)/AC - IC Indirect cost ($) - IDC Interest during construction ($) - I _w Neutron first-wall loading (MW/m²) - i Toroidal plasma current (MA) - j Plasma current density, I/a² - k _B Boltzmann constant, 1.602(10)^–16 (J/keV) - LWR Light-water (fission) reactor - MPD Mass power density 1000P_E/M_FPC (kWe/tonne) - M _N Blanket energy multiplication of 14.1-MeV neutron energy - M _FPC Mass of fusion power core (tonne) - n Plasma density (m^–3) or toroidal MHD mode number - O&M Annual operating and maintenance cost ($/yr) - p _f Plant availability factor - PFD Poloidal field dominated (CTs, RFP, DZP) - P Construction time (yr) - P_TH Thermal power (MWt) - P _E Net electric power (1-)P _ET (MWe) - P_ET Total gross electric power (MWe) - p_f Fusion power (MW) - q Tokamak safety factor (B /B _gq )(a/R _T ) - q _e EngineeringQ value, 1/e - R _T Major toroidal radius (m) - RFP Reversed-field pinch - RPE Reactor plant equipment (Account 22) - S Shield - SCR Annual spare component cost ($/yr) - SSR Second stability region for the tokamak - S/T/H Stellarator/torsatron/heliotron - ST Spherical tokamak or spherical torus - T Plasma temperature (keV) - TDC Total direct cost ($) - TOC Total overnight cost ($) - UDC Unit direct cost,TDC/10 ³ P _E ($/kWe) - V _p Plasma volume (m³) - W _p Plasma energy (GJ) - W _B Magnetic field energy (GJ) - Magnetic utilization efficiency, 2nk_BT/(B ²/2₀) - ₀ Permeability of free space, 4(10)^–7 H/m - XE Plasma confinement efficiency, a²/4_E - _e Plasma energy confinement time - _p Overall plant efficiency, _TH(1-) - _TH Thermal conversion efficiency - _FPC AverageFPC mass density (tonne/m³) - Plasma vertical elongation factor,b/a - Thickness of allFPC engineering structure surround plasma (m) - Total recirculating power fraction, (P _ET-P _E)/P _ET, or inverse aspect ratioa/R _T This work was performed under the auspices of USDOE, Office of Fusion Energy.     

The reversed-field-pinch (RFP) fusion neutron source: A conceptual design

The conceptual design of an ohmically heated, reversed-field pinch (RFP) operating at 5-MW/m² steady-state DT fusion neutron wall loading and 124-MW total fusion power is presented. These results are useful in projecting the development of a cost effective, low-input-power (206 MW) source of DT neutrons for large-volume (10 m³), high-fluence (3.4 MW yr/m²) fusion nuclear materials and technology testing.Work supported by U.S. DOE.     

Numerical comparison between the ICF and the ICF-spherical pinch

The spherical pinch concept is an outgrowth of the inertial confinement model. The salient feature of the spherical pinch concept is the creation of a hot plasma in the center of a sphere.^(1,2) This plasma is then compressed by a strong shock wave launched from the periphery of the vessel by an imploded plasma acting as a driver. This scheme, reveals that convergence of the shock, which is the main feature of all inertial confinement schemes, is a component of the spherical pinch model. The reasons for classifying the spherical pinch as a particular ICF model and designating it as a ICF-SP are given here. The fluid mechanics and high-temperature hydrodynamics of the spherical pinch can be briefly described as an explosion within an implosion. The structure of the shock wave for such explosion within an implosion and for, an implosion alone is determined by solving numerically the governing equations of the phenomena. We present here a detailed computational comparison of the inertial confinement model and the spherical pinch in terms of density, pressure, temperature, confinement time, total accumulated number of neutrons, and time-resolved neutron flux from reactions in deuterium-tritium mixture. It is shown that temperature, confinement time, and total number of neutrons for the ICF-Spherical Pinch improve upon the classical ICF.     

Scaling laws for spherical pinch experiments

Spherical pinch experiments are characterized by a central discharge in a spherical vessel followed by an inductive discharge in the vessel's peripheral shell gas. An analysis is carried out of the evolution of the imploding shock waves produced by the shell explosion in order to find out if the central discharge can be contained and compressed by the converging shocks, so as to maintain its temperature for a time sufficiently long for breakeven. The analytical model adopted is essentially that of the recent paper of Ahlborn and Key (Plasma Phys. 23: 435, 1981). One finds that the converging shocks are indeed capable of containing and compressing the central plasma. In addition, if the central spark reaches the critical temperatureT _L = 2.58 keV by the deposition of an energy density of 1.86×10⁸ J·g^?1, the scaling law required in order to contain such a plasma for breakeven isρ ₀ R(E_s/M_s)^1/2 ? 1.96×10⁶, whereρ ₀ is the initial fill gas density,R is the radius of the spherical vessel, andE _s is the energy deposited in the peripheral shell massM _s . The general applicability of the model to other fusion devices based on the implosion principle is discussed.     

Ignition in a tokamak reactor with INTOR-like parameters

The requirements for ignition in a tokamak reactor with INTOR-like parameters were studied using a one-dimensional transport code. With empirical electron energy diffusivity _e , ignition was obtained with 60–75 MW of neutral beam injection at a volume average pressure ratio =4–5% under a variety of conditions. Changing _e gave ignition at the same if the plasma minor radius varied asa _e ^1/2 . The maximum impurity concentration which allows ignition was found to be comparable to that for the much simpler case of a homogeneous plasma with radiative losses only. In long pulse simulations with efficient helium pumping, the maximum toroidal field ripple which allowed ignition was 2.0% (peak-to-peak) at the plasma edge. Ignition was maintained with over 99% recycling of helium ash using 5% less than maximum ripple.     

Paramagnetic effect under the influence of high-frequency pressure and electron paramagnetic resonance in plasma

The interaction of uhf fields ( = 2· 10¹⁰ sec^–1) in a space resonator containing dense plasma (n 10¹³ – 10¹⁴ cm^–3) in a steady magnetic field was studied experimentally. Under the influence ofhf pressure a paramagnetic current arises in the plasma; the associated effect of an increase in the static magnetic field inside the plasma agrees closely with the calculated relation.For _H/ = 0.5 paramagnetic resonance of the electrons takes place; this leads to a sharp rise in plasma pressure p₀, up to =8p₀/H₀ ²0.2.Translated from Atomnaya Énergiya, Vol. 20, No. 5, pp. 401–407, May, 1966.     

Constraints on very-high-β, edge-stabilized tandem-mirror reactors

There are several tandem-mirror schemes which propose a very high and edge stabilization for the center-cell plasma ( being the ratio of the plasma pressure to the vacuum magnetic-field pressure). While the exact criteria for the edge stabilization are uncertain, it is possible to analyze the option space in which a very-high- mirror reactor would operate. The primary physics constraints on such a reactor are the energy balance at ignition, the buildup of He⁴ ash and the hot-particle( ^hot ), and the need for adiabatic conservation of the hot-particle gyro-orbits in the axial field gradients at the center-cell ends. There are also engineering constraints on the allowable wall loading and plant size. In this paper, a wall-stabilized tandem-mirror reactor is analyzed and is found to be an attractive device requiring low center-cell vacuum fields (of the order of 2 to 3 tesla). A primary requirement is that the plasma edge have a thermal conductivity near classical values.     

Containment of plasma in a trap with combined magnetic field

The containing properties of an adiabatic trap with a magnetic field increasing in the longitudinal and radial directions are investigated. This field is obtained from a combination of the ordinary mirror field configuration (main field H0) and the field of a system of current-carrying conductors laid parallel to the axis of the trap (stabilizing field H). The conductors are placed uniformly in azimuth around the side walls. The trap is filled with plasma of density n10⁹–10¹⁰ cm^–3 and proton energy T_i5eV (T_e20 eV). The plasma lifetime is measured as a function of H. and the neutral gas pressure. From the results obtained, it is concluded that such combined fields ensure stable containment of the plasma, unbroken by magnetohydrodynamic instabilities [at any rate for = nI/(H²/8) 10^–4]. The stabilization of the instability is confirmed by analysis of the plasma oscillations for various values of H. The disintegration of the plasma is determined by the charge exchange of fast ions in the residual gas; the maximum containment time which can be achieved is 0.06 sec for p = 7.10^–9mmHg. A qualitative picture of the plasma density over the radius of the trap is obtained.Translated from Atomnaya Énergiya, Vol. 17, No. 5, pp. 366–375, November, 1964     

Development of a high energy pulsed plasma simulator for the study of liquid lithium trenches

To simulate detrimental events in a tokamak and provide a test-stand for a liquid-lithium infused trench (LiMIT) device [1], a pulsed plasma source utilizing a theta pinch in conjunction with a coaxial plasma accelerator has been developed. The plasma is characterized using a triple Langmuir probe, optical methods, and a calorimeter. Clear advantages have been observed with the application of a coaxial plasma accelerator as a pre-ionization source. The experimental results of the plasma gun in conjunction with the existing theta pinch show a significant improvement from the previous energy deposition by a factor of 14 or higher, resulting in a maximum energy and heat flux of 0.065 ± 0.002 MJ/m² and 0.43 ± 0.01 GW/m². A few ways to further increase the plasma heat flux for LiMIT experiments are discussed.     

Techniques for measuring the alpha-particle distribution in magnetically confined plasmas

Methods are proposed for measuring the alpha-particle distribution in magnetically confined fusion plasmas using neutral-atom doping beams, ultraviolet spectroscopy, and neutral particle detectors. In the first method, single charge exchange reactions, A⁰+He²⁺A⁺ +(He⁺)^*, are used to populate then=2 andn=3 levels of He⁺. The ultraviolet photons from the decaying excited states are Doppler shifted by 5–10 Å from those produced by the thermalized alpha-particle ash. In the second method, double charge exchange reactions, A⁰+He²⁺A²⁺+He⁰, enable fast neutralized alpha particles to escape from the plasma and be detected by neutral particle analyzers. These methods are distinguished from similar techniques of observing plasma impurities in that, in principle, they allow a determination of the dependence of the distribution function on energy and pitch angle, as well as on spatial position. Detector configurations are analyzed, count rates are estimated, and the detector feasibility is discussed. A preliminary analysis of the feasibility of the required neutral beams is presented, and exploratory experiments on existing devices are suggested.     

Ion beam propagation and focusing

Inertial confinement fusion with ion beams requires the efficient delivery of high energy (1 MJ), high power (100 TW) ion beams to a small fusion target. The propagation and focusing of such beams is the subject of this paper. Fundamental constraints on ion beam propagation and focusing are discussed, and ion beam propagation modes are categorized. For light ion fusion (LIF), large currents (2–33 MA) of moderate energy (3–50 MeV) ions of low atomic number (1A12) must be directed to a target of radius 1 cm. The development of pulsed power ion diodes for LIF is discussed, and the necessity for virtually complete charge neutralization during transport and focusing is emphasized. Fornear-term LIF experiments, the goal is to produce pellet ignition without the standoff needed for the ultimate reactor application. Ion diodes for use on Sandia National Laboratories Particle Beam Fusion Accelerators PBFA-I (2–4 MV, 1 MJ, 30 TW, operational) and PBFA-II (2–16 MV, 3.5 MJ, 100 TW, scheduled for operation in 1985) are discussed. Ion beam transport from these diodes to the pellet is examined in reference to the power brightness . While values of =2–5 TW/cm²/sr have been achieved to date, a value of 100 TW/cm²/sr is needed for breakeven. Research is now directed toward increasing , and means already exist (e.g., scaling to higher voltages, enhanced ion diode current densities, and bunching), which indicate that the required goal should be attainable. Forfar-term LIF applications, the goal is to produce net energy gain with standoff suitable for a reactor. This may be achieved by ion beam transport in preformed, current-carrying plasma channels. Channel transport research is discussed, including experiments with wire-initiated, wall-initiated, and laser-initiated discharge channels, all of which have demonstrated transport with high efficiency (50–100%). Alternate approaches to LIF are also discussed, including comoving electron beam schemes and a neutralized beam scheme. For heavy ion fusion (HIF), moderate currents (10 kA) of high energy (10 GeV) ions of high atomic number (A200) must be directed to a target of radius 0.3 cm. Conventional accelerator drivers for HIF are noted. For a baseline HIF reactor system, the optimum transport mode for low charge state beams is ballistic transport in near vacuum (10^–4–10^–3 Torr lithium), although a host of other possibilities exists. Development of transport modes suitable for higher charge state HIF beams may ultimately result in more economical HIF accelerator schemes. Alternate approaches to HIF are also discussed which involve collective effects accelerators. The status of the various ion beam transport and focusing modes for LIF and HIF are summarized, and the directions of future research are indicated.     

Effect of a Boron Screen on the Sensitivity of Neutron Fission Chambers

It is suggested that absorbing screens with ¹⁰B be used to maintain constant sensitivity under prolonged irradiation of fission chambers with natural uranium. The transmission factor T (E) of boron screens with various thicknesses ( = 0.1–2 ge/cm²) for a wide neutron energy range and attenuation of a spectrum of the type _e/E are estimated. The group and average group constants of the transmission factor of boron are calculated for neutron fluxes in 25 energy groups of the neutron cross sections library.The contribution of ²³⁸U and ²³⁵U to the signal of a fission chamber with natural uranium is analyzed as a function of the boron screen thickness. ²³⁹Pu accumulation and ²³⁸U burnup are estimated using ²³⁸U group capture cross sections, ²³⁸U and ²³⁹Pu fission cross sections, and the group values T (E)E/E obtained by the authors. It is shown that in the absence of a boron screen for thermal-neutron fluence 10¹⁷ cm^–2 the sensitivity of a fission chamber with natural uranium increases as a result of the formation of ²³⁹Pu. A boron screen with = 1 g/cm² makes it possible to maintain the sensitivity of the fission chamber constant up to thermal-neutron fluence 5·10²² cm^–2.     

The high-density Z-pinch as a pulsed fusion neutron source for fusion nuclear technology and materials testing

The dense Z-pinch (DZP) is one of the earliest and simplest plasma heating and confinement schemes. Recent experimental advances based on plasma initiation from hair-like (10s m in radius) solid hydrogen filaments have so far not encountered the usually devastating MHD instabilities that plagued early DZP experimenters. These encouraging results along with the debut of a number of proof-of principle, high-current (1–2 MA in 10–100 ns) experiments have prompted consideration of the DZP as a pulsed source of DT fusion neutrons of sufficient strength (S _N 10¹⁹ n/s) to provide uncollided neutron fluxes in excess ofI _w = 5–10 MW/m² over test volumes of 10–30 liters or greater. While this neutron source would be pulsed (100s ns pulse widths, 10–100 Hz pulse rate), giving flux time compressions in the range 10⁵–10⁶, its simplicity, near-term feasibility, low cost, high-Q operation, and relevance to fusion systems thatmay provide a pulsed commercial end-product, e.g., inertial confinement or the DZP itself, together create the impetus for preliminary consideration as a neutron source for fusion nuclear technology and materials testings. The results of a preliminary parametric systems study (focusing primarily on physics issues), conceptual design, and cost vs. performance analyses are presented. The DZP promises an inexpensive and efficient means to provide pulsed DT neutrons at an average rate in excess of 10¹⁹ n/s, with neutron currents I_w10 MW/m² over volumes V_exp 30 liter using single-pulse technologies that differ little from those being used in present-day experiments.Work supported by U.S. DOE.     

A performance comparison of lithium,helium, and flibe cooled tube/header fusion blankets

The objective of this study is to provide a comparison of thermal-hydraulic and structural performance of lithium, helium, and flibe cooled fusion blankets based on a tube/header geometry in a liquid lithium breeder. Type 316 stainless steel and TZM are considered as representative near-term and long-term, high temperature blanket structural materials, respectively, to show the potentials of each coolant. The flibe-TZM system has the best characteristics, while lithium-316SS, helium-316SS, and helium-TZM are comparable but definitely more limited in operating conditions. These results suggest that molten salt-refractory metal systems deserve more attention.Nomenclature a radial direction half-width of region cooled by single tube (m) - A A=_st/_cD - A _w first wall area (m²) - b azimuthal half-width of region cooled by single tube (m) - B magnetic field strength (T) - C _p specific heat of coolant (J/kg-°C) - C ₁ pumping power ratio - D _h ,D _t header and cooling tube diameter (m) - E _F energy deposited in the blanket region per fusion neutron, determined from neutronic calculations; 15.2 MeV used in this study - F _c allowance factor in pressure loss calculations for lithium system - h heat transfer coefficient (W/m²-°C) - Ha Hartmann number,Ha=BD _c /gm - J the ratio of percent change of first wall loading to percent change of a design parameter - K _c ,K _Li,K _s thermal conductivity of coolant, lithium, and structure (W/m-°C) - L major on-axis circumference of reactor (m) - M blanket energy multiplication factor,M=E _F /14.1 - n number of coolant tubes per header - N number of blanket modules (or headers) azimuthally - N _t total number of coolant tubes - Nu Nusselt number,Nu = hD_t/K_c - P coolant pressure (Pa) - P header and total pressure loss (Pa) - P _r Prandtl number - q _w first wall neutron energy loading (W/m²) - q average volumetric heat generation rate in the blanket (W/m³) - q(r) volumetric heat generation rate in blanket (W/m³) - r radial distance from first wall (m) - r _e radial position of the tube close to the hottest spot in the lithium pool - R gas constant - R _w first wall radius (m) - S defined by Eq. (25) - t _t ,t _h coolant tube and header tube thickness (m) - ¯T average coolant temperature (°C) - T _in inlet temperature (°C) - T _Li,max maximum lithium pool temperature (°C) - T _w,max maximum tube temperature (°C) - T _c coolant temperature rise across blanket (°C) - T _F film temperature rise (°C) - T _m temperature rise between coolant tube and maximum in pool (°C) - T _w wall temperature rise (°C) - U _h coolant velocity at header inlet for lithium system (m/s) - U _t coolant velocity in coolant tubes (m/s) - U _h ,max maximum inlet velocity for the lithium system, given by Eq. (13) - W _s surface heat flux in coolant tube (W/m²) - V _m voltage drop across the tube in flibe system (V) - V _t total blanket volume (m³) - X axial length of coolant tubes (m) - X _e entry and exit tube length in flibe system (m) - Z radial thickness of blanket (m) - _c , _s fraction of blanket volume occupied by coolant and structural material (exclusive of header region) - ratio of the minimum value ofq(r) to q, 0.4 - coolant viscosity (kg/m-s) - fiction coefficient - coolant density (kg/m³) - _t tube density (m^–3) - _c , _s electrical conductivity of coolant and structure (1/-m) - _h hoop stress (Pa) - _y structural material design yield stress limit (Pa)     

Radiochemical studies of nuclear reactions leading to the formation of π mesons

A study of nuclear reactions leading to the formation of -mesons is of value for understanding nuclear forces. A radiochemical method is used for detecting such reactions and this makes it possible to determine the cross section of reactions of very low probability.Reactions of the type (p,=) and (p,p) have been studied with the nuclei of elements of average atomic weight. It was found that their cross section equals 10^–30 cm²/nucl. at proton energies of 480–660 Mev. A noticeable increase in the cross section was observed in the range of energies 110–480 Mev. Such reactions could not be detected for nuclei of heavy elements. The products of the (p,¦2⁺) reaction with copper could hot be identified.The possibility of a (p.=) reaction in the interaction of a particle with a complex of nucleons in a nucleus is discussed. The information obtained gives a fuller understanding of the processes occurring when high-energy particles react with complex nuclei than can be obtained from the theory of nucleon-nucleon collisions.     

Investigation of the Effect of Radiation and High-Temperature Treatment on the x-Ray Luminescence of Quartz Glasses

The x-ray luminescence of KI, KV, and KU-1 quartz glasses, irradiated with and n– radiation in the dose range 10²–10⁷ Gy and neutron fluence range 10¹⁵–10¹⁷ cm^–2 and subjected to high-temperature annealing in air at 450 and 900°C is investigated. It is shown that the spectra of the nonirradiated and the and n– irradiated glasses of the first two types are a superposition of bands with _max = 410 and 460 nm, which are due to an impurity center initially present in the glasses (_max = 410 nm) and the initial and radiation-generated with dose 10⁶ Gy and fluence 10¹⁶ cm^–2 E' centers (_max = 460 nm). X-Ray luminescence is not observed in nonirradiated KU-1 glasses; a band with _max = 460–470 nm, due to radiation-generated E' centers, appears in the spectra of and n– irradiated glasses. As the radiation dose and the neutron fluence increase, the number of impurity centers decreases and the number of E' centers increases. It is established that the 410 nm band is due to the component of the n– radiation. High-temperature annealing in air at 900°C induces in the spectra new bands with _max = 470 and 520–540 nm, which are believed to be due to interstitial defects of the type O^– and O₂ ^–, formed when oxygen from air diffuses into the glass and localizes in interstices. 6 figures, 7 references.     

Compression, Heating and Fusion of Colliding Plasmoids by a Z-theta Driven Plasma Liner

A potentially promising approach to fusion employs a plasma shell to radially compress two colliding plasmoids. The presence of the magnetic field in the target plasma suppresses the thermal transport to the confining shell, thus lowering the imploding power needed to compress the target to fusion conditions. With the momentum flux being delivered by an imploding plasma shell, many of the difficulties encountered in imploding a solid metal liner are eliminated or minimized. The best plasma for the target in this approach is the FRC. It has demonstrated both high β, and robustness in translation and compression that is demanded for the target plasma. A high density compressed plasmoid is formed by a staged axial and radial compression of two colliding/merging FRCs where the energy that is required for the implosion compression and heating of the magnetized target plasmoid is stored in the kinetic energy of the plasmas used to compress it. An experimental apparatus is being constructed for the demonstration of both the target plasmoid formation as well as the compression of the plasmoid by a plasma liner. It is believed that with the confinement properties and the high β nature of the FRC, combined with the unique approach to be taken, that an nτ_E T _i triple product ∼5 × 10¹⁷ m⁻³ s keV can be achieved.     

Measurement of the Efficiency of a 6LiD Thermal-to-14 MeV Neutron Converter in the Experimental Channel of an IVV-2M Reactor

The results of measurements of the efficiency of a ⁶LiD thermal-to-fast neutron converter for neutrons from DT and ⁶LiT fusion reactions with energy 14 MeV in the experimental channel of an IVV-2M reactor are presented. The first experimental estimates of the conversion coefficients for the corresponding fusion reactions are obtained: K ^D 2.11·10^–4 and K ^Li 1.36·10^–4. The value found in this work for the total conversion coefficient 3.47·10^–4 is approximately 1.7 times greater than the previously measured value and about 20% greater than the maximum computed estimate for a ⁶LiD converter.An experimental apparatus with a ⁶LiD converter in an IVV-2M channel is an accessible, comparatively inexpensive, and unique (for now) source of 14 MeV neutrons that can provide continuous and approximately uniform irradiation of Ø4 × 20 cm samples by neutrons from DT and ⁶LiT fusion 2.7·10¹⁰ sec^–1·cm^–2 for a comparatively long time (500 h).