Natural Cooling Research Articles

Thermal Performance of Developed Coating Material as Cool Pavement Material for Tropical Regions

The amount of solar energy emitted back from conventional asphalt pavement contributes to the phenomenon of urban heat island (UHI), and the current work was focused on studying the thermal behavior of asphalt pavement developed as a coating material from wasted tile. The surface temperature of asphalt surface with and without coating material for ambient temperature and underground soil temperature was studied. Results showed that the application of developed coating materials could reduce the surface temperature of asphalt pavement up to 4.4°C. Rainfall also played a significant role as a natural coolant during the experimental run by maintaining the surface temperature continuously for 2 days after the rainfall event, with an average surface temperature of 41.9°C. The underground soil temperature of the coated surface showed reduction in the range of 0.8-1.2°C throughout the day. The coating materials achieved their efficiency in reducing the emitted radiation only during noontime and the results show that the developed coating materials have true potential to serve as cool pavement to combat UHI effects.

Low-power nuclear engineering for heat production

The paper shows the expediency and importance of the development of low-power nuclear engineering as well as feasibility indices of an up-to-date nuclear power plant intended for regional energy production. A high reliability of the vessel-type boiling reactor with a natural coolant circulation is shown under various operating conditions of a nuclear heat production plant.

Boiling water vessel reactors for nuclear district heating

The thermal and economic performance indices of a present-day low-capacity reactor facility with a boiling-water vessel reactor are presented. The operation of a reactor with natural coolant circulation is highly reliable in different operating regimes of a nuclear heat and power plant.

Comparative study of minor actinide transmutation in sodium and lead-cooled fast reactor cores

This paper shows that lead-cooled and sodium-cooled fast reactors (LFRs and SFRs) can preferentially consume minor actinides without burning plutonium, both in homogeneous and in heterogeneous mode. The former approach consists of admixing about 5% of minor actinides (MAs) into uranium–plutonium fuels in the core and using a limited number of thermalising pins consisting of UZrH 1.6. These are needed to keep the negative Doppler feedback larger than the positive coolant reactivity coefficient. Our Monte Carlo burn-up calculations showed that a 600 MW e LFR self-breeder without blankets can burn an average of around 67 kg annually of MAs with a reactivity swing of only about −0.7$ per year. The reactivity swing of a corresponding 600 MW e SFR is more than three times larger due to the poorer breeding and half the critical mass in comparison to the LFR. However, when axial and radial blankets loaded with 10% MAs are added, the SFR burns 25% more MAs (131 kg/yr) and breeds 30% more Pu (150 kg/yr) than an equally sized LFR. When only the blankets are loaded with MAs, the SFR breeds 30% more Pu (198 kg/yr) and still burns about 60 kg a year of MAs. However, in terms of severe accident behaviour, the LFR, with its superior natural coolant circulation and larger heat capacity, has definite advantages.

Possibility, admissability, and desirability of implementing pulsed stochastic regimes in boiling-water reactors with natural coolant circulation

A one-dimensional integral momentum model for the boiling-water channel and the point kinetic equations are used to study the conditions and mechanisms for the appearance of temporary pulses of stochastic regimes and regimes with pulsed turbulence of the neutron field in water-moderated water-cooled boiling-water reactors with natural circulation of the coolant. These self-excited oscillatory regimes occur as a result of a negative steam effect of reactivity. They consist of the fact that neutron bursts with random intensity occur in a reactors in random time intervals. Such regimes make it possible to increase substantially the power and size of boiling-water reactors. The methods proposed for controlling pulsed chaos make it possible to excite and suppress stochastic regimes and to regulate the amplitude of self-excited oscillations so that it does not exceed admissable limits. Small external perturbations have no effect on the characteristics of pulsed chaos. The results are obtained by numerical calculation of the self-excited oscillations.

208 流動性のある潜熱蓄冷材の開発(O.S.2-2 企業・大学等における熱工学研究の最前線(2))(オーガナイズドセッション2 企業・大学等における熱工学研究の最前線)

Ethanol water solution is a very important and common substance, but there are so few data available at present for describing the behaviour of the ethanol water solution at low temperatures. Ethanol water solution is important in that it has the potential to succeed as a coolant or thermal energy storage substance because it satisfies the following requirements : it is fluidic at low temperatures and it provides a balance between the nature and environment. These features have made the ethanol solution attractive for the purposes such as : utilizing low-temperature energy of LNG, rapid cooling of food-stuffs and use as a natural coolant. As the ethanol solution is a binary solution, it also can have a solid-liquid phase. The author has proposed this solid-liquid phase be used as a thermal energy storage substance because it has a superior fluidity.

Nuclear district heating plants AST-500. Present status and prospects for future in Russia

Development activity in the field of nuclear reactors for district heating purposes was initiated in the ex-USSR in the middle 1970s. It was stimulated by a growing deficiency of fossil fuels needed to provide for the heating of large cities in the European regions of the country. Construction of two pilot twin-reactor nuclear district stations was started in the early 1980s in the cities of Gorky and Voronezh. Nuclear heating reactor AST-500, developed by OKB Mechanical Engineering, has a number of unique engineering features, e.g. integral lay-out of primary system's components, natural coolant circulation, steam-gas self-pressurizing system, additional safety vessel around the reactor unit, etc. All these specific features, which were adopted in order to assure enhanced safety of the reactor plant, required fulfilment of an extensive R&D program. Description of the reference AST-500 design features is given in this paper, as well as respective R&D activities carried out for the design solutions and safety validation. Currently, activity in the field is directed to AST-500 design upgrading and development and also to a whole series of heating-only and co-generation reactor plants ranging from 30 to 600 MW(th) for the new generation nuclear stations.

Bilibinsakya Nuclear Power Plant: 23 years operation in the specific conditions of the Russian far north-east

The Bilibinskaya Nuclear Power Plant (BNPP) is situated near the town of Bilibino in the Chukotsk Autonomous District. BNPP has been in operation from early 1974 and has proved to be a highly efficient use of nuclear power in the far north-east conditions of Russia. BNPP is designed as a nuclear heat and power co-generation plant. Specific features of the location required solving many unique and difficult technical problems. The correctness of these solutions has been confirmed by operational experience. BNPP consists of four power units of the same type. The rated electric capacity of the plant is of 48 MW with simultaneous heat production of 78 MW. The capacity factor of BNPP was about 85% and the availability factor was about 90–92% during the period of stability in Russian economics (up to 1991). At that time the electric power prime cost was 1.3–1.5 times lower than that from organic fuel power sources located in the most favorable conditions (in terms of organic fuel cost) for this region on the Arctic Ocean coast. The heat prime cost was 2–2.5 times lower than that in Bilibino's heating plant that was fuelled by imported organic fuel. At present this difference has increased even more in favor of BNPP. Analysis of accidents to the safety systems operations and their failure has revealed ‘inherent’ safety features of the plant. It is ensured by a combination of the following factors: negative feedback of physical characteristics; natural coolant circulation; tubular fuel elements in contact with the graphite moderator whose thermal capacity is considerably higher than that of the fuel elements; a considerable margin in graphite moderator and fuel elements working temperature as compared to that at which a bulk loss of tightness of the external fuel element cladding occurs; and relatively small dimensions of the core.

Distribution of coolant flow rate in a steam generator in natural coolant circulation regimes

The large hydraulic nonuniformity of steam generator pipes operating in parallel in the natural coolant circulation regime results in a lower efficiency of the heat-transfer surface during emergency cooldown of the reactor plant, and it limits the operational possibilities, specifically, for using this regime at partial power levels. It is obvious that circulation reversal in the pipes of steam generators in the natural circulation regime can have an unfavorable influence on individual structural elements of steam generators as a result of additional temperature stresses appearing in the metal.

Integral nuclear power reactor with natural coolant circulation. Investigation of passive RHR system

The development of a small power (up to 240 MWe) integral PWR for nuclear co-generation power plants has been carried out. The distinctive features of this advanced reactor are: primary circuit arrangement in a single pressure vessel; natural coolant circulation; passive safety systems with self-activated control devices; use of a second (guard) vessel housing the reactor; favourable conditions for the most severe accident management. A passive steam condensing channel has been developed which is activated by the direct action of the primary circuit pressure without an automatic controlling action or manual intervention for emergency cooling of an integral reactor with an in-built pressurizer. In an emergency situation as pressure rises in the reactor a self-activated device blows out non-condensable gases from the condenser tube bundle and returns them in the steam-condensing mode of the operation with the returing primary coolant condensate into the reactor. The thermo-physical test facility is constructed and the experimental development of the steam-condensing channels is performed aiming at the verification of mathematical models for these channels operation in integral reactors both at loss-of-heat removal and LOCA accidents.

Separation system of a new-generation boiling-water vessel reactor

Steam separation design is considered with respect to the construction of new BWRs in the Russian Federation. Two steam separator designs for simplified boiling water reactors (SBWRs) in nuclear power plants are briefly described. The effects of separation system and draft section construction on reactor vessel dimensions are investigated to determine the optimal design for an enhanced-safety BWR with natural coolant circulation. Three systems were examined: (1) free steam volume, secondary louvered separator, and common draft section; (2) submerged sheet with no-bubbling attachments, secondary louvered separator, and common draft section; and (3) submerged sheet with no-bubbling attachments (or on-axis centrifugal separators), additional steam volume, and draft pipe/mixing chamber draft section. The calculated reactor vessel inner diameter and height reactor corresponding to the three separation systems were determined to be: (1) 7m and 21.4m, (2) 6.7m and 17.75m, and (3) 6.7m and 14.25m. Separation designs (1) and (2) were examined for a 500 MW(e) reactor; design (3) was determined to be appropriate for unit power levels up to 300 MW(e). 9 refs., 3 figs.

Elimination of yield point phenomenon in mild steel by warm working

Wire and arc additive manufacturing (WAAM) is a promising technology for fabricating large-sized metal parts. However, to use this method efficiently, it is necessary to overcome the problem of severe heat accumulation that reduces manufacturing efficiency and deteriorates material properties. This paper proposes a new method using the in-situ gas cooling (ISGC) for reducing heat accumulation. Three parts were manufactured via the proposed method utilising CO2, N2, and air as cooling gases, respectively. A fourth part was fabricated with natural cooling (NC), without cooling gas. The temperature data measured by thermocouples and infrared pyrometers indicated that compared with NC, ISGC considerably reduced the temperature of the whole part and CO2 had the best cooling effect. ISGC efficiently mitigated heat accumulation, due to the large heat transfer coefficient and the great temperature difference between the high-temperature region behind the torch and the cooling gas. Numerical simulations verified by the experimental data further denoted that in ISGC, high-temperature areas near the molten pool shrank sharply and heat dissipation conditions in the molten pool were improved markedly. ISGC not only improved the surface quality of the parts, but also strengthened the mechanical properties of the parts by obtained fine grains. The maximum ultimate tensile strength (UTS) and yield strength (YS) increased to 483 MPa and 358 MPa, respectively, which represented an increase of 4.1 % and 4.7 % respectively. The average hardness was enhanced to 161.3 HV, representing an improvement of 3.7 %.

Fuel-element testing channel loop with natural coolant circulation

Fuel-element testing channel loop with natural coolant circulation

Performance of Rice Hull Media for Pot Easter Lilies Under Three Forcing Systems1

Abstract Potting media containing parboiled rice hulls as a sand or soil substitute were compared to a soil-sand-peat mix for the production of potted Lilium longiflorum Thunb. cv. Ace under 3 forcing systems: precooling (PC), natural cooling (NC), and controlled temperature forcing (CTF). Plants grown in media containing rice hulls were delayed in flowering (2-3 days) and possessed more flower buds (0.9 buds) independently of forcing system with no effect on stem ht and root dry wt at bloom time. NC and CTF reduced forcing time, and increased stem ht and flower no. over PC. Rice hulls-sand-peat mix increased leaf no. by approx 10 leaves per stem over soil-sand-peat mix under PC or NC but in soil-sand-peat mix, CTF plants had 11 to 15 more leaves per stem than PC and NC plants. Rice hulls in the potting media resulted in lighter wt and improved drainage.

Reactor dynamics calculations for reactor with natural circulation of single phase coolant

REACTOR DYNAMICS CALCULATIONS FOR REACTOR WITH NATURAL CIRCULATION OF SINGLE PHASE COOLANT* V. M. Selivanov, A. A. Gorev, I. I. Sidorova, and V. M. Gribunin UDC 621.039.56 The nonlinear mathematical model of the dynamics of the pr imary loop of a water-cooled watermoderated reactor with natural circulation of single-phase coolant is discussed. The solution of the nonstationary heat t ransfer equations is represented by an expansion in orthogonal Legendre polynomials with t ime-varying coefficients [1]. The introduction of this expansion makes it possible to reduce the entire system of equations describing the dynamics of the reactor to the form dx = Az + ] (x), where A is the (n • n) matrix of constant coefficients; x is the column vector of the phase coordinates; f(x) is the column vector of the nonlinear terms. The model under consideration is utilized in the case f(x) 0 to obtain the frequency response to an experimental arrangemenr simulating a reactor system [2]. It is pointed out that in accordance with the Gyftopoulos criterion [3], an increase in the negative temperature coefficient of reactivity (in absolute magnitude) will result in impaired system stability. Our calculations show that, in the presence of modest power perturbations, a reactor with natural coolant circulation and temperature feedback features is dynamically equivalent to a reactor with forced coolant circulation. 2. 3. LITERATURE CITED S. K. Sanathanan et a l . , Nucl. Sci. Engng., 23, No.2 (1965). V. M. Selivanov et a l . , At. Energ . , 27, 101-~969)o E. Gyftopoulos, Nucl. Sci. Engng., ~ No. 1 (1966)o NOTE ON OPTIMIZATION OF A STATIONARY REFLECTOR IN A PULSED FAST REACTOR B. I. Kuprin and E. P. Shabalin UDC 621.039.51 The pulsed neutron source method is highly recommended as a tool for solving many problems in neutron physics and in the physics of condensed media. Promising neutron sources for this line of research * Translated from Atomnaya ]~nergiya, Vol.31, No. 5, p. 505, November, 1971. Original article submitted submitted January 20, 1971; revision submitted April 22, 1971. t Translated from Ato~rmaya ]~nergiya, Vol. 31, No. 5, pp. 505-506, November, 1971. Original art icle submitted October 15, 1970. 9 1972 Consultants Bureau, a division of Plenum Publishing Corporation, 227 West 17th Street, New York, N. Y. 10011. All rights reserved. This article cannot be reproduced for any purpose whatsoever without permission o f the publisher. A copy of this article is available from the publisher for $15.00.