Radiation-damage in molybdenum-rhenium alloys for space reactor applications

Various Mo-Re alloys are attractive candidates for use as fuel cladding and core structural materials in spacecraft reactor applications. Molybdenum alloys with rhenium contents of 41-47.5% (wt%), in particular, have good creep resistance and ductility in both base metal and weldments. However, irradiation-induced changes such as transmutation and radiation-induced segregation could lead to precipitation and, ultimately, radiation-induced embrittlement. The objective of this work is to evaluate the performance of Mo-41Re and Mo-47.5Re after irradiation at space reactor relevant temperatures. Tensile specimens of Mo-41Re and Mo-47.5Re alloys were irradiated to ∼0.7 displacements per atom (dpa) at 1073, 1223, and 1373 K and ∼1.4 dpa at 1073 K in the High Flux Isotope Reactor at Oak Ridge National Laboratory. Following irradiation, the specimens were strained to failure at a rate of 1 × 10⁻³ s⁻¹ in vacuum at the irradiation temperature. In addition, unirradiated specimens and specimens aged for 1100 h at each irradiation temperature were also tested. Fracture mode of the tensile specimens was determined. The tensile tests and fractography showed severe embrittlement and IG failure with increasing temperatures above 1100 K, even at the lowest fluence. This high temperature embrittlement is likely the result of irradiation-induced changes such as transmutation and radiation-induced segregation. These factors could lead to precipitation and, ultimately, radiation-induced embrittlement. The objective of this work is to examine the irradiation-induced degradation for these Mo-Re alloys under neutron irradiation.     

A dosimetric system for prolonged manned space flights

Institute of Biomedical Problems. Central Institute of Physics Research, Hungarian Academy of Sciences. Nuclear Research Center, Fontenay-aux-Roses, France. Translated from Atomnaya Énergiya, Vol. 71, No. 3, pp. 237–242, September, 1991.     

Testing of programs for substantiation of the nuclear safety of space reactors

Russian Scientific Center “Kurchatovskii Institut.” Translated from Atomnaya énergiya, Vol. 77, No. 5, pp. 326–329, November, 1994.     

ITER operational space for full plasma current H-mode operation

Sensitivity studies performed as part of the ITER IO design review highlighted a very stiff dependence of the maximum Q attainable on the machine parameters. In particular, in the considered range, the achievable Q scales with I_p^{^}4. As a consequence, the achievement of the ITER objective of Q = 10 requires the machine to be routinely operated at a nominal current I_p of 15 MA, and at full toroidal field BT of 5.3 T. This paper analyses the capabilities of the poloidal field (PF) system (including the central solenoid) of ITER against realistic full current plasma scenarios. An exploration of the ITER operational space for the 15 and 17 MA inductive scenario is carried out. An extensive analysis includes the evaluation of margins for the closed loop shape control action. The overall results of this analysis indicate that the control of a 15 MA plasma in ITER is likely to be adequate in the range of li 0.7–0.9 whereas, for a 17 MA plasma, control capabilities are strongly reduced. The ITER operational space, provided by the reference pre-2008 PF system, was rather limited if compared to the range of parameters normally observed in present experiment. Proposals for increasing the current and field limits on PF2, PF5 and PF6, adjustment on the number of turns in some of the PF coils, changes to the divertor dome geometry, to the conductor of PF6 to Nb3Sn, moving PF6 radially and/or vertically are described and evaluated in the paper. Some of them have been included in 2008 ITER revised configuration.     

Thermoemission reactor-converters for nuclear power units in outer space

Conclusion Thermoemission nuclear power units with built-in generators in the nuclear reactor core can be regarded as a promising source of electric power for supplying the needs of space equipment for various purposes with a wide range of electric power demands over a long service life and with acceptable mass-limit characteristics.This article is a variant of a report presented at the Sixth Symposium on Nuclear Power in Outer Space (Albuquerque, January 1989).Translated from Atomnaya Énergiya, Vol. 66, No. 6, pp. 374–377, June, 1989.     

High-temperature heat pipes for power modules of space nuclear power systems

The results of an experimental analysis of the technology for producing high-temperature heat pipes and choosing their structure so as to ensure a long service life and high performance characteristics required for the power modules of space nuclear power systems with an out-of-core energy conversion system are examined. It is estimated that the operation of molybdenum-lithium heat pipes, fabricated using the technology developed, for the power modules of such nuclear power systems will be stable for >10⁵ h. 2 figures, 1 table, 13 references. State Enterprise Krasnaya Zvezda. Translated from Atomnaya énergiya, Vol. 89, No. 1, pp. 82–86, July, 2000.     

Configuration space control for Wendelstein 7-X

The Wendelstein 7-X stellarator (W7-X) is a superconducting fusion experiment, presently under construction at the Greifswald branch of the Max-Planck-Institut für Plasmaphysik. W7-X is a device with high geometrical complexity due to the close packing of the components in the cryostat and their complex 3D shape e.g. of the superconducting coils. The tasks of configuration space control are to ensure that all these components do not collide with each other under a set of defined configurations, i.e. at the time of assembly, at 4 K or for various coil currents. To fulfill these tasks sophisticated tools and procedures were developed and implemented within the realm of a newly founded division that focuses on design, configuration control and configuration management.     

Work on nuclear power plants for space

Conclusions It should be noted that the space problems gave an unprecedented impetus to improving computational methods and design of reactors and shielding and the development of the latest technology. This is not surprising, since the designs required that the mass of the nuclear power plant be determined to within several kilograms. The placement of every kilogram of a satellite in orbit costs several thousands of dollars. The development of nuclear rocket motors and the first thermionic systems "Topaz" in the world was a very important achievement in space nuclear technology. Unfortunately, as sometimes happens in practice, this achievement was far ahead of its time and remains unused. Translated from Atomnaya énergiya, Vol. 80, No. 5, pp. 357–361, May, 1996.     

Open-cycle multi-megawatt MHD space nuclear power facility

The results of calculations of the characteristics and development of a scheme and technical make-up of an open-cycle space power facility based on a high-temperature nuclear reactor for a nuclear rocket motor and a 20 MW Faraday MHD generator are presented. A heterogeneous channel-vessel IVG-1 reactor, which heated hydrogen to 3100 K, with the pressure at the exit from the reactor core up to 5 MPa, burn rate 5 kg/sec, and thermal power up to 220 MW is examined. The main parameters of the MHD generator are determined: Cs seed fraction 20%, stopping pressure at the entrance 2 MPa, electric conductivity ≈ 30 S/m, Mach number ≈ 0.7, magnetic induction 6 T, electric power 20 MW, specific energy extraction ∼4 MJ/kg. The construction of the scheme of a MHD facility with zero-moment exhaust of the working body and its main characteristics are presented. Translated from Atomnaya énergiya, Vol. 105, No. 3, pp. 135–144, September, 2008.     

Development and improvement of thermoelectric generator modules for a space nuclear power system

Ways to improve the components of a high-temperature silicon-germanium thermoelectric module with radial-annular geometry of thermoelements for use in space nuclear power systems with specific heat fluxes 30–70 W/cm² at temperatures ∼1200 K are presented. Certain structural elements, which make a thermoelectric converter realistic, such as an electrical insulation unit and a thermojunction which damps stresses, have been developed. The reasons for the losses of thermoelectric efficiency in previously developed thermoelectric modules based on a silicon-germanium alloy are analyzed. The processes responsible for the degradation of thermoelectric modules and determining their service life are determined. Preliminary experiments show that all ways enumerated for lowering the losses and increasing the stability of thermoelectric modules consisting of silicon-germanium alloy are realizable. 7 references. Sukhumi Physicotechnical Institute. Translated from Atomnaya énergiya, Vol. 89, No. 1, pp. 74–78, July, 2000.     

Choice of parameters for current conversion systems in high-power space nuclear power systems

The energy-mass and operational characteristics of a nuclear electrorocket propulsion system based on thermionic reactor-converters largely depend on the parameters of the current converter and the onboard cable network. The decrease of the specific mass of a nuclear electrorocket propulsion system due to the choice of the temperature regime of the current converter and the onboard cable network as well as the frequency of the supply voltage is examined. Investigations have shown that increasing the working temperature of the onboard cable network and current converter makes it possible to decrease the specific mass of the nuclear electrorocket propulsion system by 33–83%, as a result of which the useful load fraction of the spacecraft increases from tens of percent to a factor of 2–3, 1 figure, 6 references. Translated from Atomnaya énergiya, Vol. 89, No. 1, pp. 78–81, July, 2000.     

Autonomous reactor control using model based predictive control for space propulsion applications

Reliable reactor control is important to reactor safety, both in terrestrial and space systems. For a space system, where the time for communication to Earth is significant, autonomous control is imperative. Based on feedback from reactor diagnostics, a controller must be able to automatically adjust to changes in reactor temperature and power level to maintain nominal operation without user intervention. Model-based predictive control (MBPC) is investigated as a potential control methodology for reactor start-up and transient operation in the presence of either a constant or a time varying external source. Bragg-Sitton and Holloway [Bragg-Sitton, S.M., Holloway, J.P., 2004. Reactor start-up and control methodologies. In: El-Genk, M. (Ed.), Proceedings of the Space Technology and Applications International Forum (STAIF-2004), AIP Conference Proceedings 699, pp. 614–622.] assessed the applicability of MBPC to reactor start-up from a cold, zero-power condition in the presence of a time-varying external radiation source, where large fluctuations in the external radiation source can significantly impact a reactor during start-up operations. Here the MBPC algorithm is applied using the point kinetics model to describe the reactor dynamics, with a single group of delayed neutrons and a fast neutron lifetime of 10⁻⁷ s. Controller stability is assessed by carefully considering the dependencies of each component in the defined cost (objective) function and its subsequent effect on the selected “optimal” control maneuvers. Additional analysis demonstrates the effectiveness of the controller when a lower fidelity reactor kinetics model is adopted for the model system versus using a full six-group delayed neutron representation in the point kinetics equations to represent the “real” system operation.     

“Yennisei” space nuclear power system

March 1, 2000 is the 25th anniversary of the power start-up of a prototype of the “Yenisei” space nuclear power watem, which is to supply power to the spacecraft for a direct television roadcasting system, on the stand “1” at the Russian Science Center “Kurchatov Institute.” During the nuclear power tests, which continued for 5000 hours, the, subsequent finishing work, and a study of the state of individual components, it was demonstrated that the unit functions, properly under standard conditions. and the technology for preparing the unit for power tests and a procedure for starting up the unit and for performing power tests and the subsequent final adjustments together with an investigation of the critical component that determine the service life were developed. The results of the final adjustments and the investigations pointed the way to making the required improvements to the power-generating chanels in order to quarantee a service life of at least 15 years and then to increase it to 3 years at subsequent stages of the development work. The history of the development of the “Yenisei” nuclear power system is presented, the basic parameters and the results of comprehensive tests are described, and the international collaboration with the US is discussed. 7 figures, 1 table. Translated from Atomnaya énergiya, Vol. 88, No. 2, pp. 95–108, February, 2000.     

A review of refractory metal alloys and mechanically alloyed-oxide dispersion strengthened steels for space nuclear power systems

Mechanical and thermo-physical properties of refractory metal alloys and mechanically alloyed (MA)-oxide dispersion strengthened (ODS) steels are reviewed and their potential for use in space nuclear reactors is examined. Preferable refractory alloys for use in liquid metal and gas-cooled space reactors include Nb-1%Zr, PWC-11, Mo-TZM, Mo-xRe where x varies from 7% to 44.5%, T-111 and ASTAR-811C. These alloys are heavy, difficult to fabricate, and are not readily available. The advantages of the MA-ODS alloys are: (a) their strength at high temperatures (>1000 K), which decreases slower with temperature than those of niobium and molybdenum alloys; (b) relatively lightweight and less expensive; (c) low swelling and no embrittlement with exposure to high-energy neutrons (>0.1 MeV) up to 10²⁷ n/m²; and (d) high resistance to oxidation and nitration. The few data available on compatibility of MA-ODS alloys with alkali liquid metals up to 1100 K are encouraging, however, additional tests at typical temperatures (1000-1400 K) in space nuclear reactors are needed. The anisotropy of MA-ODS alloys when cold worked, and particularly rolled into tubes, should not hinder their use in space nuclear power systems, in which operation pressure is either near atmospheric or as high as 2 MPa, but joints weldability is an issue.     

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.     

Temperature and burnup reactivities and operational lifetime for the submersion-subcritical, safe space (S4) reactor

The He–Xe gas-cooled, S4 reactor has a sectored, Mo–14%Re solid core for avoidance of single point failures in reactor cooling and Closed Brayton Cycle (CBC) energy conversion. The reactor core is loaded with UN fuel and each of its three sectors is thermal-hydraulically coupled to a separate CBC loop and radiator panels. The solid core minimizes voids, and the BeO reflectors are designed to easily disassemble upon impact, ensuring that the bare S4 reactor is sufficiently subcriticial when submerged in wet sand or seawater and flooded with seawater, following a launch abort accident. Spectral shift absorber (SSA) additives in the core and thin SSA coatings on the outer surface of the core can also be used to ensure subcriticality in such an accident. This paper investigates the effects of various SSAs (Re, Ir, Eu-151, B-10 and Gd-155) on the temperature and burnup reactivity coefficients and the operating lifetime of the S4 reactor at a steady thermal power of 550 kW. The calculations of the burnup, reactivity feedback coefficient used a mixture of the top 10 light and top 10 heavy fission products plus Sm-149 and are performed for isothermal reactor core and reflector temperatures of 1200 and 900 K. In this fast spectrum space reactor, SSAs markedly increase fuel enrichment and decrease the burnup reactivity coefficient, but only slightly decrease the temperature, reactivity feedback coefficient. With no SSAs, the UN fuel enrichment is lowest (58.5 wt.%), the temperature and burnup reactivity coefficients are the highest (−0.2709 ¢/K and −1.3470 $/at.%), and the estimated operating lifetime is the shortest (7.6 years). The temperature and burnup reactivity coefficients decrease to −0.2649 ¢/K and −1.0230 $/at.%, and the operating lifetime increases to 8.3 years when rhenium additives are used. With europium-151 and gadolinium-155 additions, fuel enrichment (91.5 and 94 wt.%) and operating lifetime (9.9 and 9.8 years) are the highest and both the temperature reactivity feedback coefficient (−0.2382 and −0.2447 ¢/K) and the burnup reactivity coefficient (−0.9073 and −0.8502 $/at.%) are the lowest.     

Preliminary conceptual design of a new moderated reactor utilizing an LEU fuel for space nuclear thermal propulsion

Nuclear Thermal Rocket (NTR) propulsion is a viable and meritorious option for human exploration into deep-space because of its high thrust, improved specific impulse, well established technology, bimodal capability, and enhanced mission safety and reliability. The NTR technology has already been investigated and tested by the United States of America and Russia and the former Soviet Union. The representative Nuclear Engine for Rocket Vehicle Applications (NERVA) type reactors traditionally used Highly Enriched Uranium (HEU) fuels, shaped in hexagonal fuel element geometries because of the importance of making a high power reactor with a minimum size. Although the HEU-NTR designs are the best choice in terms of rocket performance and technical maturity, they inevitably provoke nuclear proliferation obstacles not only for all research and development activities by civilians and non-nuclear weapon states but also for potential commercialization. To overcome the security issues due to HEU, the non-proliferative, small-size NTR engine with low thrust levels of 41 kN–53 kN (9.2 klb_f ∼ 11.9 klb_f), Korea Advanced NUclear Thermal Engine Rocket utilizing a Low-Enriched Uranium fuel (KANUTER-LEU), is being designed for future generations. Its design goals are to make use of an LEU fuel for its fairly compact core, but to minimize the rocket performance sacrifice relative to the traditional HEU-NTRs. To achieve these goals, a new space propulsion reactor is conceptually designed with the key concepts of a high uranium density fuel with resistance against high heating and H₂ corrosion, a thermal neutron spectrum core, and a compact and integrated fuel element core design with protective cooling capability. In addition, a preliminary design study of neutronics and thermal-hydraulics was performed to explore the design space of the new LEU-NTR reactor concept. The result indicates that the innovative reactor concept has great potential, both to implement the use of an LEU fuel and to create comparable rocket performance, compared to the existing HEU-NTR designs.     

Reactivity control options of space nuclear reactors

This paper compares two ex-core control options of the gas-cooled Submersion Subcritical Safe Space (S^4) reactor with a fast neutrons energy spectrum: (a) rotating BeO drums with 120° thin segments of enriched B₄C in the BeO radial reflector; and (b) sliding segments in the BeO radial reflector. Investigated are the effects on the beginning-of-life (BOL) excess reactivity, reactivity depletion rate and operation life, and the spatial neutron flux distributions and fission power profiles in the core. Also investigated is the effect of reducing the thickness of the enriched B₄C segments in the control drums on the BOL excess reactivity, when one or two of the 6 drums are stuck in the shutdown position. Reducing the thickness of the B₄C segments from 0.5 mm to 0.238 mm, with one drums stuck in the shutdown position, increases BOL cold and hot-clean excess reactivity from +$1.71 and +$0.47 to +$2.38 and +$0.89, respectively. These reactivity values are almost identical to those of the reactor with one of the six reflector segments stuck open in the shutdown position. Results also showed that the control options made little difference in the reactor performance. The power peaking in the reactor core with sliding reflector segments is slightly lower and the spatial power profiles are relatively flatter. The operation life of the reactor with a sliding reflector segments control, when operating at a nominal thermal power of 471 kW, is only 22 full power days longer than with rotating drums control.     

Prospects of using thermoelectric nuclear power generators for manufacturing of materials in space

State Enterprises Krasnaya Zvezda. Nuclear Reactor Institute of Kurchatov Institute Scientific Center. Translated from Atomnaya Énergiya, Vol. 75, No. 4, pp. 254-259, October, 1993.