14 MeV neutron yields from D-T operation of the MIT Cockcroft-Walton accelerator

As part of our fusion-product diagnostic development program, we have begun a series of experiments with 14 MeV neutrons generated in a Cockcroft-Walton accelerator. Two different detectors have been used to measure the neutron yield: a silicon SBD and a Cu foil. The energy of the emitted neutrons has been determined by using two spectrometers: the SBD and a³He proportional counter. The reaction rate is monitored, with about ±5% accuracy, by detecting the particles from D + T n +. The neutron yields obtained from the Si detector and the Cu activation had associated uncertainties of about ±15% and agreed well with the predicted values from measurements.     

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.     

Internal friction in uranium

Studies have been made of the temperature dependence of internal friction and the shear modulus in uranium. The internal friction in uranium depends on the heat treatment and is reduced after annealing in the ß and regions. During polymorphous transformations the internal friction changes its value isothermally. The transitions ß and ß are accompanied by a reduction in the internal friction and ß and ß by an increase in the internal friction. Each polymorphous modification of uranium in the temperature ranges for its existence has its own value of internal friction.     

Influence of the structure and properties of uranium on its behavior during irradiation

Data are presented on the variations in dimensions and form of uranium specimens during irradiation. It is shown that by regulating the composition of the uranium and treatment conditions (degree of deformation in the-region and heat-treatment conditions), in consequence of variation in grain size and texture, it is possible to vary within wide limits the magnitude of surface distortion due to irradiation and the value of G_i.A study has been made of the dependence of the variation in grain size of quenched uranium, as well as hardness, tensile strength, and yield strength, on the iron, silicon, and aluminum content of uranium. The cooling rate and content of these impurities influence the critical point of the transformation on quenching; for example, for a cooling rate of 400C/ sec and a silicon content of 0.05%, the critical point of the transformation drops to 530C.Experimental results show a creep acceleration during irradiation (nv = 6·10¹² neutr/cm²·sec) of 50–100 times, i.e., by 1.5–2 orders for textured uranium and uranium with disoriented structure. The rate of creep of uranium with a disoriented structure is connected to the burnup rate.The results are given of tensile tests made on uranium directly in the reactor. Even after remaining a short time in the neutron field (up to 1 hour), the percentage elongation is diminished somewhat and the tensile strength is increased.The following assisted in the experimental work: A. G. Lanin, V. M. Teplinskaia, V. K. Zakharova, L. N. Protsenko, V. N. Golovanova and K. A. Borisov.     

Analysis of the observed resonance-width distributions of U233 and Pu239

As a result of the strong fluctuations of fission and reduced neutron widths, a significant number of resonances occur with such a small reduced height 2gF_n ⁰/ that they will not be noticed experimentally. If the fraction of transmitted resonances is considerable, then this should lead to the following effects which might be observable in an experiment: 1) change of the distribution function of the reduced neutron and fission widths, which is particularly sharply manifested for _n ⁰/<_n ⁰> 1 and _f/<_f> 1; 2)appearance of correlation between the form of the neutron distribution b of fission channels; 3) appearance in the total cross section and fission cross section of a background which is approximately proportional to . All the effects mentioned are manifested for the U²³³ nucleus. For Pu²³⁹ this effect is smal! and the observed values are welI described by X² distributions.Translated from Atomnaya Énergiya, Vol. 17, No. 1, pp. 22–27, July, 1964     

Influence of neutron spectrum on formation of233U from232Th

Conclusion The results of the calculations are of interest from two points of view: First, the main physical characteristics of the²³²Th²³³U process are presented as a function of the flux density and the energy spectrum of thethermal neutrons, making it possible to evaluate the efficiency of one reactor design or other; second, the decrease in the²³²Th consumption and the increase in the neutral consumption R as grows give reason to hope that the problem of economic optimization for the²³²Th²³³U process will have a nontrivial solution.Translated from Atomnaya Énergiya, Vol. 56, No. 5, pp. 320–322, May, 1984.     

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.     

Requirements for ohmic ignition

An analysis of ohmic ignition criteria is presented, giving the requirements onT, n, andn/j in a form easily applicable to various confinement assumptions. For circular cross-section NeoAlcator tokamaks with Spitzer resistivity, a value ofB ₂ a approximately equal to 250 T²m is required. The outstanding uncertainties in schemes to lower this value are how much increase in current density is achievable by plasma shaping and what the exact NeoAlcator coefficient is.     

Investigation of the γ-rays accompanying thermal-neutron capture by certain rare-earth nuclei

Using a scintillation spectrometer measurements have been made of the spectra of -rays accompanying thermal-neutron capture in a number of nuclei.A number of intense lines have been found below 300 kev in the -ray spectra for thermal-neutron capture in europium, gadolinium, dysprosium, holmium, erbium, thulium, hafnium and tantalum. Lines corresponding to 4⁺2⁺ and 2⁺0⁺ transitions between rotational levels'of Er¹⁶⁸ and Hf¹⁷⁸ were found in the erbium and hafnium spectra. The intensity of these transitions corresponds to 0.5–0.8 photons per capture event.The authors are indebted to Academician I. V. Kurchatov for his interest in the work and to Professor L. V. Groshev, V. M. Strutinskii and D. P. Grechukhin for a number of valuable comments and Professor I. A. Zaozerskii for kindly furnishing the rare-earth samples.We wish to express our gratitude to G. P. Mel'nikov for providing reliable operation of the electronic apparatus.     

Thin-foil electrochemical cells: High-sensitivity fusion tests andin-situ ion beam measurements of deuterium loading

Electrochemical cells constructed with a thin Pd or Ti foil electrode mounted at one wall of the cell have been used both to test for the existence of cold fusion and to measure directly DPd loading ratios in an operating cell. The first type of experiment used a surface-barrier particle detector positioned a few millimeters from the foil to provide a very sensitive monitor for possible fusion-generated protons at 3.02 MeV. The detection limit for this arrangement is estimated to be 10^–24 fusions/deuterium/s, assuming a bulk fusion effect. These experiments included cells with 5- and 25-m-thick Pd foils, 10-m Ti foils, parallel experiments with 0.1M LiOD (heavy water) in one cell and LiOH (light water) in another, current densities up to 0.5 A/cm², and run times as long as 22 days. No evidence for fusion products was seen. The second type of experiment using these cells, both as an adjunct to the fusion tests and to provide new information, was the use of external beam nuclear reaction analysis to monitor directly the loading and unloading of deuterium in the foil of an operating cell. Using a 1.5-MeV³He ion beam in air, the deuterium in the outer 2 m of the exposed Pd foil was measured for the first time using the D(³He,p) nuclear reaction. The maximum DPd ratios observed using this technique were 0.8–0.9.     

D-T experiments on TFTR

Temperatures, densities and confinement of deuterium plasmas confined in tokamaks have been achieved within the last decade that are approaching those required for a D-T reactor. As a result, the unique phenomena present in a D-T reactor plasma (D-T plasma confinement, alpha confinement, alpha heating and possible alpha driven instabilities) can now be studied in the laboratory. Recent experiments on the Tokamak Fusion Test Reactor (TFTR) have been the first magnetic fusion experiments to study plasmas with reactor fuel concentrations of tritium. The injection of 20 MW of tritium and 14 MW of deuterium neutral beams into the TFTR produced a plasma with a T/D density ratio of 1 and yielded a maximum fusion power of 9.2 MW. The fusion power density in the core of the plasma was 1.8 MW m^–3 approximating that expected in a D-T fusion reactor. In other experiments TFTR has produced 6.4 MJ of fusion energy in one pulse satisfying the original 1976 goal of producing 1 to 10 MJ of fusion energy per pulse. A TFTR plasma with T/D density ratio of 1 was found to have 20% higher energy confinement time than a comparable D plasma, indicating a confinement scaling with average ion mass, A, of _E. The core ion temperature increased from 30 keV to 37 keV due to a 35% improvement of ion thermal conductivity. Using the electron thermal conductivity from a comparable deuterium plasma, about 50% of the electron temperature increase from 9 keV to 10.6 keV can be attributed to electron heating by the alpha particles. At fusion power levels of 7.5 MW, fluctuations at the Toroidal Alfvén Eigenmode frequency were observed by the fluctuation diagnostics. However, no additional alpha loss due to the fluctuations was observed. These D-T experiments will continue over a broader range of parameters and higher power levels.Work supported by U.S. Department of Energy Contract No. DE-AC02-76-CHO-3073.     

The dynamics of nuclear power stations

This paper makes a comparison of the results of eXperimental and theoretical studies that have been carried out on the properties of the engineering model of the Beloyarskii atomic electric station under construction in the USSR, which uses nuclear superheating of the steam. It is shown that a number of the simplifying assumptions are correct which are often used in discussing the dynamics of nuclear power stations.The results of the studies may be used to make a theoretical analysis of the dynamic properties of several types of nuclear power installations, as well as in analyzing and synthesizing the optimum control system.Notation q() specific heat load, referred to length of segment, kcal/hour · m - f(x) distribution function of specific heat load along the length of segment - () heat transfer coefficient, including the thermal resistence of the fuel element, kcal/m² · hour · degree - t_f.e. (x, ) the current value of fuel element temperature, averaged over the corss section, degrees C - t(x, t) current value of coolant temperature, degrees C - p perimeter of fuel element, bathed by coolant, m - m weight of metal per unit length of fuel element kg/m - C_M heat capacity of metal and fuel element, kcal/kg · degree - i(x, ) current value of heat content of coolant, kcal/kg - specific gravity of coolant, kg/m³ - S live cross section of fuel element, m² - D(x, ) current value of flow of steam phase, kg/hour - G(x, ) current value of the flow of water phase, kg/hour - (x, ) current value of the fraction of the cross section occupied by steam - , specific gravity of water and steam at saturation temperature, kg/m² - i, i heat content of water and steam at saturation temperature, kcal/kg - t_S() saturation temperature, degrees C - P_i() pressure in i-th segment, kg/m² - l height, determining the level pressure between segments, m - g acceleration of gravity, m/hour² - w_i() coolant velocity at the i-th segment, m/hour - D_i() steam flow at the i-th segment of the superheating circuit, kg/hour - V_i volume of i-th segment of the superheating circuit, m³ - mean steam temperature at the i-th segment for the superheating circuit, degrees C - k₁,k₂,k₃,k₄ constant coefficients - N/N₀ relative power change in the evaporating channels, % - P_I, P_II pressure change in the first and second loops, atm - t_sps, t_fw change in temperature of superheated steam and feed water, respectively, degrees C Translated from Atomnaya Énergiya, Vol. 15, No. 2, pp. 115–120, August, 1963     

Radiative capture cross sections of tellurium isotopes as a function of neutron energies up to 1.5 keV

The radiative capture cross sections of tellurium isotopes have been measured as a function of neutron energies up to 1.5 keV. The values of ₀ and of the spin levels, as well as _n ⁰/D have been found for the even-odd isotopes Te¹²³ and Te¹²⁵.Translated from Atomnaya Énergiya, Vol. 14, No. 3, pp. 264–272, March, 1963.This is the second of a series of papers on the neutron cross sections of separated isotopes (for the first paper, on molybdenum isotopes, see ZhÉTP44, No. 4 (1963).     

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.     

Calculating the transient temperature field in a reactor tube and the thermoelastic stresses in the fuel element cladding

Approximate analytic methods are given for calculating the transient temperature field in the fuel elements and the coolant temperatures at any point along the reactor tube, as well as the transient thermoelastic stresses in the cladding of a cylindrical fuel element. The coolant temperature at the input to the tube is constant, and the coolant undergoes no changes in state of aggregation. The approximate methods are illustrated by examples.Results are given, for comparison, of accurate calculations of the same examples made with a rapid calculating machine.List of symbols time - r; z coordinates (radius, distance along tube) - r₁; r₂ internal and external radii of fuel element cladding respectively - H total active length of fuel element - a₁; ₁;c _{1 1} coefficients of temperature conductivity, heat conductivity, specific heat capacity and specific gravity of fissionable material respectively - a₂; ₂; C_p2; ₂ cladding parameters - a; ; c_p; coolant parameters - mean cladding radius - f:f₂ cross-sectional area of tube for coolant and cladding respectively - w coolant velocity - coefficient of heat release to coolant - t (r, ); (); () fuel temperature, mean temperature over cross section of cladding, and coolant temperature at pointz. along tube respectively - qv() specific volume of coolant at pointz - values averaged overz - quantities at the initial instant of time - ₃ delay time - _n time required for coolant to go from z=0 to the point in question     

Evolution of defects in micropellet fuel during the fabrication of HTGR compacts

The stress distribution arising in micropellets and cylindrical fuel compacts during fabrication, the stress concentration in micropellets located near the surface of a compact, and the evolution of defects in micropellets as a function of the type of stress state are investigated.It has been found that an ensemble of micropellets with a large number of particles contains a continuous spectrum of defects in the range 10^–4–10² µm. Mechanical stresses engender evolution of the defects according to the scheme accumulation of microdefects microcracks cracks through defects.Recommendations are formulated for lowering the number of defects in micropellets during deposition of coatings on the micropellets and compaction.Research Institute of the Luch Scientific and Industrial Association.__________Translated from Atomnaya Énergiya, Vol. 98, No. 1, pp. 44–50, January, 2005.     

Study of the zirconium apex op the Zr-Ta-Nb phase diagram

Metallographical examination thermal analysis and electrical resistance measurements have been applied to a study of the zirconium apex, up to 82% zirconium and a temperature of 1200C, of the ternary system Zr-Ta-Nb, with limited solubility of tantalum and niobium in -zirconium ( phase), limited solubility and complete solubility of niobium in -zirconium, with eutectoid decomposition of the \ solid solution and three-phase eutectoid equilibrium + between - and-zirconium. In the investigated portion of the Zr-Ta-Nb phase diagram, the following phase regions were found: a) two one-phase regions and ; b) three two-phase regions + , + and + : c) one three-phase region + + ; the region contracts as the temperature falls below 1200 C.The solubility of tantalum and niobium in -zirconium in the system Zr—Ta—Nb is about 0.5%. On passing from Zr—Ta to Zr—Nb, the + and + regions are displaced toward lower temperature and high niobium concentrations; the boundaries of the + and + + regions are lowered from 790 for Zr—Ta to 612 C for Zr-Nb. Passing between the + and + regions is a binary eutectoid line which, from Zr-Ta to Zr-Nb is displaced toward lower temperatures and higher niobium concentrations. The solubility of niobium in ot zirconium in the Zr-Nb system is about 0.5%by weight. Eutectoid decomposition in the Zr-Ta system shifts the maximum of the martensitic-like transformation to the left and results in an increase in the stability of the phase at room temperature in quenched alloys.     

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.     

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)     

High power pulsed lasers

Pulsed high power lasers can deliver sufficient energy on inertial fusion time scales (0.1–10 ns) to heat and compress DT fuel to fusion reaction conditions. Several laser systems have been examined for application to the fusion problem. Examples are Ndglass, CO₂, KrF, and I₂, etc. A great deal of developmental effort has been applied to the Ndglass laser and the CO₂ gas laser systems. These systems now deliver >10⁴ kJ and >20×10¹² W to inertial fusion targets. The Nova Ndglass laser is being constructed to provide >200 kJ and >200×10¹² W of 1 m radiation for fusion experimentation in the mid-1980s. For inertial fusion target gain, >100 times the laser input, it is expected that the laser must deliver 3–5 MJ of energy on the 10–20 ns time scale. This paper reviews the developments in laser technology and outlines approaches to construction of a 3–5 MJ driver.