Graphite-moderated,gas-cooled,and water-moderated,water-cooled reactors as power units in nuclearelectric power stations

The present article reviews a number of papers submitted at the Second International Conference on the Peaceful Uses of Atomic Energy bearing on water-cooled, water-moderated, graphite-moderated, and gas-cooled reactors abroad.The basic characteristics of all of the operational power reactors, as well as of the high-power, graphite-gas water-cooled, and water-moderated reactors built abroad are cited in the article.Differences in the pathways and means of development of nuclear power in different countries are given due acknowledgement. In Britain, for example, after the first generation reactors (Calder Hall type) were built, advanced second generation reactors (Hinkley Point, etc.) were introduced, and intensive studies are now underway on third generation reactors. The ultimate purpose of the research in progress is the development of a high- temperature reactor operating in a unit with the associated gas turbine. A transition in the USA from the classical pressurized-water reactor concept (PWR) to that of the boiling water reactor with direct steam feed to the turbine is planned.The continuous improvement in the efficiency of graphite-moderated, gas-cooled reactors and water-cooled, water-moderated reactors and the reduction in capital costs per unit rated power and in cost of power developed is shown.     

Operating experience with the nuclear propulsion plant on the icebreaker “LENIN”

Summary Long-term operation of a nuclear-fueled steam-generating plant on board the icebreaker has provided a basis for a comprehensive assessment of the plant performance under a variety of sailing and weather conditions.The basic layout and makeup of the nuclear power plant has proven highly successful, and the standby equipment provided for has been completely satisfactory.No overexposures of service personnel occurred while the icebreaker was in operation. The nuclear plant proved so reliable that the nuclear plant compartment required only one visit a day to inspect the equipment installed there.The first experiment in building and operating a nuclear maritime power plant, on the icebreaker LENIN, has been crowned with complete success. The technical feasibility of building high-power nuclear icebreakers for service on the Northern Sea Route has been confirmed.The experimental neutron-physics characteristics of the reactor cores on the icebreaker were reported by staff members N. A. Lazukov and A. K. Sledzyuk. Control panel logbooks and cruise reports kept by the operating staff of the icebreaker LENIN were utilized in preparing this paper.Translated from Atomnaya Énergiya, Vol. 17, No. 5, pp. 349–359, November, 1964Paper No. 313 presented by the USSR delegation at the Third International Conference on the Peaceful Uses of Atomic Energy, Geneva, 1964.     

Report of the Integrated Program Planning Activity for the U.S. Department of Energy Fusion Energy Sciences Program

This report of the Integrated Program Planning Activity (IPPA) has been prepared in response to a recommendation by the Secretary of Energy Advisory Board that, Given the complex nature of the fusion effort, an integrated program planning process is an absolute necessity. We therefore undertook this activity to integrate the various elements of the program, to improve communication and performance accountability across the program, and to show the interconnectedness and interdependency of the diverse parts of the national Fusion Energy Sciences Program. This report is based on the September 1999 Fusion Energy Sciences Advisory Committee's (FESAC) report Priorities and Balance within the Fusion Energy Sciences Program. In its December 5, 2000, letter to the Director of the Office of Science, the FESAC reaffirmed the validity of the September 1999 report and stated that the IPPA presents a framework and process to guide the achievement of the 5-year goals listed in the 1999 report. The report also outlines a process for establishing a database for the fusion research program that will indicate how each research element fits into the overall program. This database will also include near-term milestones associated with each research element and will facilitate assessments of the balance within the program at different levels.     

A probability approach to estimating the safety of a reactor installation

Krasnaya Zvezda Scientific-Designer Bureau. Central Institute of Atomic Energy, All-Union Research Institute of Atomic Energy. Translated from Atomnaya Énergiya, Vol. 74, No. 1, pp. 38–42, January, 1993.     

The nuclear reactor circulation circuit as a radiation source

The solution of the problem of circulation circuits with a single radioisotope, which has been found earlier [1], is applied to the general case where several radioisotopes having radioactive progeny are formed in the substance to be activated. The problems of the absolute maximum circuit power and the consumption of neutrons per unit power for a number of elements which can be used as substances to be activated in the circuit are considered. From among them, the most promising are indium and its alloys.Special attention is paid to a circulation circuit where the substance to be activated contains a fissionable isotope (uranium circuit). It is shown that the specific power of such a circuit, all other conditions being equal, is considerably lower than the specific power of circuits with metallic indium or its alloys. As a particular case of a uranium circuit, the circulation from the reactor into the radiation unit,and the reverse,of fuel elements which have not burned up completely is considered. It is shown that, in this case, the power of the unit can be increased two- to fourfold in comparison with the power of a unit, which makes a single use of completely burned-up fuel elements.     

Nonelectric Applications of Fusion

This is the final report of a panel set up by the U.S. Department of Energy (DOE) Fusion Energy Sciences Advisory Committee (FESAC) in response to a charge letter from Dr. James Decker, Acting Director of the DOE Office of Science. In that letter, Dr. Decker asked FESAC to consider whether the Fusion Energy Sciences program should broaden its scope and activities to include non-electric applications of intermediate-term fusion devices. This report, submitted to FESAC July 31, 2003, and subsequently approved by them (Appendix B), presents FESAC's response to that charge.     

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     

Uranium and its alloys

A short resume is given of the results of investigations of the physico-mechanical properties of uranium and its alloys, based on reports given at the Second International Conference on the Peaceful Uses of Atomic Energy, Geneva, 1958. The effect of irradiation on uranium and, in particular, the phenomenon of gas swelling are examined. Some examples are given of -phase and -phase uranium alloys, including alloys of the fissium type. The temperature limits for the use of solid nuclear fuel based on uranium are considered. In conclusion, we give some information on the techniques for metallographic investigation of irradiated and unirradiated uranium.     

The “18 Year” Alternative: A Business Proposal to Develop Commercial Fusion Energy by 2021—The Fusion Power Corporation

This policy essay asserts that the 35 year plan recently adopted by the U.S. Department of Energy's Fusion Energy Sciences Advisory Committee is too risk averse and too costly. An alternative 18 year schedule is proposed. All dollar amounts shown below are undiscounted, and are only intended to be indicative of approximate future costs.     

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.     

Optimal Power Smoothing Method for the SHERHAN Computer Code

In the past, a parametric method based on the golden section method was used to smooth the energy-release field for the SHERHAN computer code, developed using an improved method of heterogeneous reactor theory. In the present paper, a different method is presented for solving the problem of optimal smoothing of the power. The new method is based on the method of linear programming. Some advantages of the new method are demonstrated.     

Recommendations on the Nature and Level of U.S. Participation in the International Thermonuclear Experimental Reactor (ITER) Extension of the Engineering Design Activities

The Department of Energy (DOE) Office of Energy Research chartered through the Fusion Energy Sciences Advisory Committee (FESAC) a panel to address the topic of U.S. participation in an ITER construction phase, assuming the ITER Parties decide to proceed with construction. Given that there is expected to be a transition period of 3 to 5 years between the conclusion of the Engineering Design Activities (EDA) and the possible construction start, the DOE Office of Energy Research expanded the charge to include the U.S. role in an interim period between the EDA and construction.This panel has heard presentations and received input from a wide cross-section of parties with an interest in the fusion program. The panel concluded it could best fulfill its responsibility under this charge by considering the fusion energy science and technology portion of the U.S. program in its entirely. Accordingly, the panel is making some recommendations for optimum use of the transition period considering the goals of the fusion program and budget pressures.     

Some characteristics ofγ-ray scattering at the boundary of two media

The characteristics of -ray scattering (0.33 E 1,25 MeV) at the boundary between two media were experimentally investigated. It was established that the scattered radiation was produced mainly by a narrow beam of primary radiation. The effective scattering area was contained within a region with a radius r 10h (for h z). The contribution to the total energy from radiation scattered by the medium increased for angles of incidence of the primary narrow beam up to 85–88 °, and then decreased monotonically.An estimate was made of the effective dimensions of the area from which scattered quanta were emitted into the backward half-space when an isotropie source was located at the surface of the scatterer.The energy flux of the scattered radiation was measured with a gas counter which had a practically constant sensitivity for -rays of different energies.The authors are deeply grateful to Prof. O. I. Leipunskii and to A. S. Strelkov for assistance and for discussions of the results.     

End product economics and fusion research program priorities

It is shown that deuterium based fusion fuels and reactors based on them face severe technological disadvantages in comparison with fission based systems as power sources for central station electric power plants. The author postulates the most plausible deuterium based fusion reactor consistent with the physics of the fusion reaction itself and compares this reactor (called OMR-DT) with existing fission reactors. Since neutrons are the main problem in fusion, the author suggests that a great deal more effort should be given to the study of non-Maxwellian plasmas with the emphasis on neutron-free fuel cycles. The author also suggests that the deuterium based fusion driver may play its best role as a fissile fuel producer.     

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.     

Report of the Fusion Simulation Project Steering Committee

The Fusion Simulation Project (FSP) is envisioned as a 15 year, $20M/year multi-institutional project to develop a comprehensive simulation capability for magnetic fusion experiments with a focus on the International Thermonuclear Experimental Reactor (ITER). The FSP would be able to contribute to design decisions, experimental planning and performance optimization for ITER, substantially increasing ITERs likelihood of success and its value to the US Fusion Program. The FSP would be jointly supported by the DOE Office of Fusion Energy Sciences and the DOE Office of Advanced Scientific Computing Research. The potential for developing this simulation capability rests on the exponential growth of computer power over the last 50 years, the progress in physics understanding developed by the international fusion program and the continued progress in computational mathematics that enables the use of the new ultra-scale computers to solve difficult mathematical problems. The initial concept for the FSP was developed by the Fusion Energy Sciences Advisory Committee Integrated Simulation and Optimization of Fusion Systems Subcommittee (J. Dahlburg and J. Corones, et al., J. Fusion Energy, 20(4), 135–196.). The DOE asked the FSP Steering Committee to develop a project vision, a governance concept and a roadmap for the FSP. The Committee recommends that the FSP consist of three elements: a production component, a research and integration component, and a software infrastructure component. The key challenge is developing components that bridge the enormous distance and time scales involved with the disparate physics elements of tokamak performance. The committee recommended that this challenge be met through Focused Integration Initiatives that would first seek to integrate different physics packages with disparate distance and time scales. An example is the integration of Radio Frequency (RF) Current Drive and Magnetohydrodynamics (MHD) components to produce an integrated capability to simulate the use of RF current drive to suppress MHD instabilities. This report also defines the requirements for a governance structure. The FSP Steering Committee judged that the project begin with a conceptual design phase lasting one or two years and be followed by a staged ramp-up over a few years to the full funding level.     

Nuclear fusion from crack-generated particle acceleration

Summary In summary, the high-voltages necessary to accelerate deuterons to energies sufficient to produce modest numbers (10⁴–10⁵/sec) of d-d neutrons appears to be possible as a result of cracking or fracture of the metal lattice in the cold fusion experiments.This mechanism requires neither massive electrons nor exotic nuclear reactions to explain the apparent cold fusion d-d neutron production results. Instead, it is possible that high voltage electrostatic fields, known to be associated with cracking, can reside across a crack gap long enough for the deuterons to be accelerated to sufficiently high energy to produce the d-d reactions. Interestingly, the electrostatic acceleration is quite similar to that of laboratory accelerators except for its submicron scale. Clearly, much work is still required to determine whether such a crack-generated acceleration mechanism, a quasi-particle mechanism, some combination of these, or some other, as yet unidentified mechanism is responsible for the nuclear effects seen in cold fusion experiments.Presented at the Workshop on Cold Fusion Phenomena, Sante Fe, New Mexico, May 23–25, 1989.     

On the classical thermal conductivity in a toroidal plasma

In toroidal equilibrium configurations the drift flow of heat causes a redistribution in plasma temperature along the magnetic surfaces. The resulting temperature gradient is equalized by an axial nonmagnetic heat flow, which leads to a certain effective flow of heat across the magnetic surfaces. This additional toroidat heat flow proves to be greater than the magnetic flow, which determines the heat loss from a plasma in a cylindrical geometry, The effective toroidal coefficient of thermal conductivity is calculated for smooth toroidal systems of Tokamak and stellarator types, the latter having a spatial figure-eight form. Expressions are also obtained for the distribution of the electric potential associated with the above-mentioned temperature redistribution.Translated from Atomnaya Énergiya, Vol. 19, No. 2, pp. 120–125, August, 1965     

Resonance scattering of γ-rays in Mg24

Resonance scattering of -rays With energies E₁=1.38 Mev, corresponding to the transition to the ground state in Mg²⁴. have been observed in metallic magnesium. The energy given off by the -ray (E₁=1.38 Mev) in emission and collision with the nucleus, is compensated for by the energy obtained due to recoil associated with the emission of the preceding -ray with an energy E₂ = 2.76 Mev. Using a fast coincidence method and amplitude discrimination, coincidences were recorded between the -rays with energies E₁=1.38 Mev and E₂=2.76 Mev. Scattercrs of magnesium and aluminum were alternately placed in the path of the 1.38 Mev -rays. The source was radioactive Na²⁴ in a water solution of NaOH. At an angle of 120 ° between the -rays a strong attenuatlon of the 1.38 Mev -rays was observed; this is attributed to resonance scattering. When the angle between the -rays was varied by 5 °. the strong attenuation of the flux disappeared. The width of the level at 1.38 Mev in Mg²⁴ has been estimated at > 1.6· in^–4 Mev.