Radiation Conditions during Decontamination of an Experimental Facility for Reprocessing Nuclear Fuel

The basic characteristics of the changes occuring in the radiation conditions in an experimental facility for reprocessing spent nuclear fuel during decontamination of the facility after reprocessing irradiated uranium and uranium–plutonium BOR-60 reactor fuel and plutonium tetrafluoride into dioxide by the pyroelectrochemical method are examined. An expression is obtained experimentally for calculating the decrease in the power of photon radiation from contaminated surfaces as a function of the number of decontamination cycles. It is shown that for one-time processing of the surfaces of stainless steel equipment by the two-bath method the decontamination coefficient varies over the range 2.5–25 for emitters and 5–30 for and , emitters.     

Optimized fissile and fusile breeding in a laser-fusion fissile-enrichment flux trap blanket

Optimization of fissile and fusile production in the SOLASE-H laser-fusion fissile-enrichment fuel-factory blanket is carried out. The objective is maximizing fissile breeding with the constraints of maintaining self-sufficiency in tritium production, and realistically accounting in the modeling for structural and coolant compositions and configurations imposed by the thermal-hydraulic and mechanical designs. The effect of radial and axial blanket zone thicknesses on fusile and fissile breeding is studied using a procedure which modifies the zones' effective optical thicknesses, rather than the actual three-dimensional geometrical configurations. A tritium yield per source neutron of 1.08 and a Th (n, ) reaction yield per source neutron of 0.43 can be obtained in such a concept, where ThO₂ Zircaloy-clad fuel assemblies for light water reactors (LWRs) are enriched in the²³³U isotope by irradiating them in a lead flux trap. This corresponds to 0.77 kg/[MW(th)-year] of fissile fuel production, and 1.94 years of irradiation in the fusion reactor to attain an average 3 w/o fissile enrichment in the fuel assemblies. For a once-through LWR cycle, a support ratio of 2–3 is estimated. However, with fuel recycling, more attractive support ratios of 4–6 may be attainable for a conversion ratio of 0.55, and of 5–8 for a conversion ratio of 0.70. These estimates are lower than those reported, around 20, for related designs.     

Pilot plant: An affordable step toward fusion power

Progress is reported on a study to define a pilot plant to demonstrate the production of high grade heat in a fusion power plant configuration at the lowest possible capital cost. We are considering several driven reactor tokamak designs with fusion power production levels in the 15–50 MWth range, using demountable copper coils. We conclude that it is acceptable for such facilities to be net consumers of electricity as a trade-off to achieve low capital cost, which we estimate to be in the $1 billion range. These designs are based on currently accepted physics models. Even lower cost designs may be possible, if we depart somewhat from the current physics database.     

Use of the calculated-expense method for choosing the characteristics of a fast reactor

The article describes how a calculated-expense formula can be used for analyzing the economics of the fuel cycle of a fast plutonium reactor. The proposed formula takes account of the down time in the fuel cycle for the fissionable material of the active zone and the blankets. It gives the results of the calculations, which are used for investigating how the fuel component of the calculated expenses depends on the power density and the flattening of the active zone of a reactor with an electrical power of 1000 MW.Translated from Atomnaya Énergiya, Vol. 21, No. 5, pp. 360–363, November, 1966.     

Reactivity effects in a BN-350 reactor

Conclusions The experiments were the first check on values of the temperature and power reactivity effects and the depletion effect for a fast power reactor. These results and their theoretical analysis enable us to estimate the reactivity balance in a BN-350 reactor, and to check and refine the methods of calculation. The models and methods used for calculations in designing fast power reactors turned out to be effective and gave quite satisfactory results.Calculations of the temperature and power reactivity effects somewhat underestimate their values (by 15%). The largest error in the temperature effect comes from the Doppler and sodium components, and in the power effect from the Doppler effect and the indeterminacy in temperature field calculations.Calculations give a fairly good estimate of the depletion effect also, although in this case obvious care has to be exercised in choosing the method of calculation. This fact is connected with the necessity of correctly allowing for nonuniformity of fuel depletion and the accumulation of plutonium.Investigation of the reactivity effect for prolonged use of the reactor is of great practical and theoretical interest. The effect of depletion and change in structure of the fuel pellets on the temperature and power effects, the nonlinear nature of the power effect, and the influence of plutonium accumulation on the reactivity effects in the reactor, all these are questions of reactor physics the study of which allows us to increase the accuracy of physical calculations in designing fast power reactors.Translated from Atomnaya Énergiya, Vol. 42, No. 1, pp. 3–8, January, 1977.     

Possibilities of Deep Burnup of Americium and Neptunium in the Core of a Fast Sodium Reactor

One-time deep burnup of actinides (up to burnup >90% h.a.) followed by burial in geological formations without reprocessing is of great interest. In this paper, a method is proposed for deep burnup of americium and neptunium in special assemblies, containing actinides in an inert (stone-like) matrix and a strong moderator (zirconium hydride). Placing 130 such assemblies into a BN-800 core with oxide fuel permits 90–95% h.a. burnup in a realistic time – 2–3 runs (2.5–3.5 yr). Such a reactor permits utilizing up to 110 kg of americium and neptunium per year.     

Fission-suppressed blankets for fissile fuel breeding fusion reactors

Two blanket concepts for deuterium-tritium (DT) fusion reactors are presented which maximize fissile fuel production while at the same time suppress fission reactions. By suppressing fission reactions, the reactor will be less hazardous, and therefore easier to design, develop, and license. A fusion breeder operating a given nuclear power level can produce much more fissile fuel by suppressing fission reactions. The two blankets described use beryllium for neutron multiplication. One blanket uses two separate circulating molten salts: one salt for tritium breeding and the other salt for U-233 breeding. The other uses separate solid forms of lithium and thorium for breeding and helium for cooling.Nuclear power is the sum of fusion (D + T 14 MeV neutron+ 3.5 MeV alpha) power plus additional power from neutron-induced reactions in the blanket.     

Reactivity Behavior During RBMK Startup

During startup of an RBMK reactor, the reactivity varies from –(4–7)_eff to 0–0.1_eff. Positive reactivity is introduced locally – by extracting control rods. Since the physical dimensions of an RBMK reactor are large, a local change in the properties can produce a large change in the spatial distribution of the neutron flux in the core. The possible range of variation of the reactivity of a subcritical and a critical reactor with one control rod extracted is analyzed for the actual states of the power-generating units of a nuclear power plant with RBMK reactors. It is shown that the extraction of some rods in an RBMK reactor in subcritical and critical states can increase the reactivity by 1_eff or more.     

Using thermionic nuclear power systems with a thermal reactor for interorbital space flights

Conclusion The use of a thermionic NPS with a thermal reactor in space technology to supply power to the RMPS offers broad possibilities for interorbital delivery of payloads while using comparatively cheap launch rockets to place spacecraft in a fixed orbit. The flight time from a fixed to a geostationary orbit ranges from several months to half a year, and the mass of the payload in a geostationary orbit for optimal RMPS parameters may reach 7–8 tons (not counting the mass of the NPS).It should be noted that after the flight is completed, the NPS can serve as a source of electrical power for spacecraft in geostationary orbit.Red Star Scientific-Production Organization. Translated from Atomnaya Énergiya, Vol. 70, No. 4, pp. 221–224, April, 1991.     

Study of thorium–uranium based molten salt blanket in a fusion–fission hybrid reactor

Not only solid fuels, but also liquid fuels can be used for the fusion–fission symbiotic reactor blanket. The operational record of the molten salt reactor with F–Li–Be was very successful, so the F–Li–Be blanket was chosen for research. The molten salt has several features which are suited for the fusion–fission applications.The fuel material uranium and thorium were dissolved in the F–Li–Be molten salt. A combined program, COUPLE, was used for neutronics analysis of the molten salt blanket. Several cases have been calculated and compared. Not only the influence of the different fuels have been studied, but also the thickness of the molten salt, and the concentration of the ⁶Li in the molten salt.Preliminary studies indicate that when thorium–uranium–plutonium fuels were added into a F–Li–Be molten salt blanket and with a component of 71% LiF–2% BeF₂–13.5% ThF₄–8.5% UF₄–5% PuF₃, and also with the molten salt thickness of 40 cm and natural concentration of ⁶Li, the appropriate blanket energy multiplication factor and TBR can be obtained.The result shows that thorium–uranium molten salt can be used in the blanket of a fusion–fission symbiotic reactor. The research on the molten salt blanket must be valuable for the design of fusion–fission symbiotic reactor.     

Nuclear design of a heliotron-H fusion power reactor

Nuclear analysis was carried out for the heliotron-H fusion power reactor employing anl=2 helical heliotron field. The neutronics aspects examined were (a) tritium breeding capability, (b) shielding effectiveness for the superconducting magnet (SCM), and (c) induced activity after shutdown. In this reactor design of the heliotron-H, the space available for the blanket and shield is limited due to the reactor geometry. Thus, some parametric survey calculations were performed to satisfy the design requirements. The nucleonic design features of the heliotron-H are as follows. An adequate tritium breeding ratio of 1.17 is obtained when a 10-cm thick Pb neutron multiplier and a 40-cm thick Li₂O breeding blanket are used. In this case, the total nuclear energy deposition is 16.10 MeV per 14.06 MeV incident neutron. The performance of the SCM is assured during 2 yr of continuous operation using a 20-cm thick tungsten shield. Biological dose rate behind the SCM at 1 day after shutdown is too high for hands-on maintenance.     

Neutronic performance of HYLIFE-II fusion reactor using various thorium molten salts

HYLIFE-II is one of the major inertial fusion energy reactor design concepts in which a thick molten salt layer (Flibe = Li₂BeF₄) is injected between the reaction chamber walls and the explosions. Molten salt coolant eliminates the frequent replacement of solid first wall structure during reactors lifetime by decreasing intense neutron flux. This study presents the neutronic analysis of HYLIFE-II fusion reactor using various liquid wall coolants, namely, 75% LiF–25% ThF₄, 75% LiF–24% ThF₄–1% ²³³UF₄ or 75% LiF–23% ThF₄–2% ²³³UF₄. Neutron transport calculations for the evaluation of neutron spectra were conducted with the help of Scale 4.3 by solving the Boltzmann transport equation in S₈–P₃ approximation. The effects of flowing liquid wall thickness and type of coolant on the neutronic performance of the reactor were investigated. Furthermore, radiation damage calculations at the first wall structure with respect to type and thickness of liquid wall were carried out. Numerical results showed that using the flowing liquid wall containing the molten salt, 75% LiF–23% ThF₄–2% UF₄ with a thickness of 70 cm maintained tritium self-sufficiency of the (DT) fusion driver and extended the first wall lifetime to the reactors lifetime (30 full power years). In addition significant amount of high quality fissile fuel was bred through (n, γ) reaction of ²³²Th. Moreover, energy multiplication factor (M) was increased to 12 by high rate fission reactions of ²³³U occurring in the flowing wall. On the other hand, it was concluded that using the other two coolants, 75% LiF–25% ThF₄ or 75% LiF–24% ThF₄–1% ²³³UF₄, as liquid wall did not satisfy the radiation damage and the tritium sufficiency criteria together at any thickness, so that these two coolants were not suitable to improve neutronic performance of HYLIFE-II reactor.     

Spectra of fast reactor neutrons after passing through graphite,lead, and iron

By means of a fast neutron scintillation spectrometer with one hydrogen-containing detector, the spectra of fast reactor neutrons after passing through various thicknesses of lead, graphite, and iron were measured in the range 0.7–11 MeV. The measurements were carried out in a water-moderated water-cooled experimental reactor in barrier geometry. The results of the experiments enabled us to determine the deformation of the neutron spectrum in relation to the penetration through the layers of material, and to calculate the relaxation lengths and the removal Cross sections. These quantities were punished earlier for fission spectrum neutrons in the energy range 0.7^–3 MeV.Translated from Atomnaya Énergiya, Vol. 16, No. 1, pp. 32–40, January, 1964     

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.     

Feasibility study of a magnetic fusion production reactor

A magnetic fusion reactor can produce 10.8 kg of tritium at a fusion power of only 400 MW —an order of magnitude lower power than that of a fission production reactor. Alternatively, the same fusion reactor can produce 995 kg of plutonium. Either a tokamak or a tandem mirror production plant can be used for this purpose; the cost is estimated at about $1.4 billion (1982 dollars) in either case. (The direct costs are estimated at $1.1 billion.) The production cost is calculated to be $22,000/g for tritium and $260/g for plutonium of quite high purity (1%²⁴⁰Pu). Because of the lack of demonstrated technology, such a plant could not be constructed today without significant risk. However, good progress is being made in fusion technology and, although success in magnetic fusion science and engineering is hard to predict with assurance, it seems possible that the physics basis and much of the needed technology could be demonstrated in facilities now under construction. Most of the remaining technology could be demonstrated in the early 1990s in a fusion test reactor of a few tens of megawatts. If the Magnetic Fusion Energy Program constructs a fusion test reactor of approximately 400 MW of fusion power as a next step in fusion power development, such a facility could be used later as a production reactor in a spinoff application. A construction decision in the late 1980s could result in an operating production reactor in the late 1990s. A magnetic fusion production reactor (MFPR) has four potential advantages over a fission production reactor: (1) no fissile material input is needed; (2) no fissioning exists in the tritium mode and very low fissioning exists in the plutonium mode thus avoiding the meltdown hazard; (3) the cost will probably be lower because of the smaller thermal power required; (4) and no reprocessing plant is needed in the tritium mode. The MFPR also has two disadvantages: (1) it will be more costly to operate because it consumes rather than sells electricity, and (2) there is a risk of not meeting the design goals.This paper represents work carried out from 1980 to 1982 and was in draft form in 1982. It was received for publication with only minor editing of its 1982 version, explaining the fact that some of the material is dated.     

Possibility of obtaining the effect of self-quenching of power pulses from a pulsed fast reactor

Conclusions When a reactivity modulator of heterogeneous structure is used [1], by an appropriate choice of modulator parameters we can appreciably enhance the self-quenching effect, thus expanding the range of allowable deviations of reactivity of a periodically acting pulsed fast reactor while, at the same time, increasing the nuclear safety of the plant. In the case of a reactor with the parameters of the IBR-2, when the harmonic component of the reactivity has an amplitude of 3·10^–3 and the coefficient of the reactivity parabola is 4·10^–5 cm^–2 the maximum deviation of reactivity which does not yet result in damage to the reactor core is 2·10^–3, which is double that for a modulator of heterogeneous structure without a self-quenching effect and triple that when an ordinary movable reflector is used.Translated from Atomnaya Énergiya, Vol. 52, No. 5, pp. 320–323, May, 1982.     

Radiation Conditions during the Preparation of Irradiated BOR-60 Fuel Assemblies for Reprocessing

The basic stages in the preparation of irradiated BOR-60 reactor fuel for reprocessing are examined. It is determined that during the separation of the fuel part of the fuel elements the coefficient of transfer of ¹³⁷Cs from the fuel into aerosol is 5·10^–6 and for fragmentation the value is 3·10^–5. It is found that the real catching efficiency for aerosol particles caught by a V-05 filter ranges from 42 to 99%. The specific entry of radioactive aerosols into the ventillation center after the first stage of air purification was 0.3 MBq for -emitters and 7.7 MBq for and emitters per 1 kg of reprocessed fuel. The total collective dose formed at the stages of preparation of a large batch of irradiated fuel (four spent fuel asemblies with average burnup 11.4% and a 10.5 to 23.7 yr holding period) for reprocessing was 11.5·10^–3 persons·Sv.     

A low wall-loading DEMO reactor design with high priority for early and reliable realization of a tokamak fusion reactor over the cost performance

Based on scientific databases adopted for designing ITER plasmas and on the advancement of fusion nuclear technology from the recent R&D program, a low wall-loading DEMO fusion reactor has been designed, where high priority has been given to the early and reliable realization of a tokamak fusion plasma over the cost performance. Since the major radius of this DEMO reactor is chosen to be 10 m, plasma ignition is achievable with a low fusion power of 0.8 GW and an operation period of 4–5 hours is available only with inductive current drive. The low ignition power makes it possible to adopt a first wall with an austenitic stainless steel, for which significant databases and operating experience exists, due to its use in the presence of neutron irradiation in fission reactors. In step with development of advanced materials, a step-wise increase of the fusion power seems to be feasible and realistic, because this DEMO reactor has the potential to produce a fusion power of 5 GW.     

Simulation of nuclear-fuel solvent-extraction reprocessing 7. Separation of macroscopic amounts of plutonium and uranium by displacement reextraction of plutonium in reprocessing fast-reactor fuel (section 1)

Conclusions These data have been obtained from simulating displacement reextraction in a system with two inputs (system for reprocessing a fast-reactor fuel), and they show that it is possible to obtain a uranium extract with not more than 100 g Pu/kg U in a counter current system with 18 stages to provide partial separation of the uranium and plutonium by reprocessing of an organic solution containing U⁺ Pu (10 1) with 64% saturation in the sum of the metals to produce reextract containing plutonium with U : Pu3 and over 99.99% extraction of the uranium; this requires 90 g/liter of uranium in the reextractant and the parameters n=2.06–2.00 ( =0.3 M); n=1.82–1.87 ( =0.5 M); n=1.61–1.78 ( =1.0 M).Full data obtained from the simulation are to be found in [5], from which one can extract the parameters of the working state of the extractor and other data on the separation of uranium and plutonium, e.g., for other specifications for the plutonium level in the uranium. The next part of the present study will be concerned with engineering solutions that can extend the range of conditions that provide the appropriate output parameters within specified ranges and thereby improve the reliability in operating the process in the optimal region.Translated from Atomnaya Énergiya, Vol. 47, No. 6, pp. 377–381, December, 1979.