On some methods for raising the power level of reactors with gaseous coolants

The present article is concerned with certain methods for raising the power level of reactors with gaseous coolants: additional cooling of the gas ahead of the gas compressor, increasing the pressure in the loop, and profiling the coolant flow through the reactor.Equations for calculating the theoretical thermodynamic cycle for a reactor with a gaseous coolant, the coolant temperature at the downstream end of the reactor, and the profiling of the coolant flow are derived.     

Hydrodynamic characteristics of lead and lead-bismuth coolants

The results of an analysis of the effect of the physical properties of lead and lead-bismuth coolants on the hydrodynamic characteristics and the results of experimental investigations of the particulars of the hydrodynamic flows of these coolants in application to loops with fast reactors are presented. It is shown that cavitation, in the conventional meaning of this word, cannot arise in flow part of vane pumps pumping lead and lead-bismuth coolants in a reactor loop. It is confirmed that a gas gap can form between the surface of a heavy liquid-metal coolant flow and the channel walls not wetted by it. The results of experimental studies of the rupture of a column of heavy liquid-metal coolant and detachment of a centrifugal pump flow, probably because of the appearance of gas cavitation, are presented.     

Heavy Liquid Metal Coolant – Lead–Bismuth and Lead – Technology

Such heavy coolants as lead and a eutectic alloy of lead with bismuth while having favorable thermophysical and technological properties are comparatively corrosive for structure materials and can be contaminated by solid impurities during the operation of the system.For systems with heavy coolants to operate for a long period of time, the structural materials which come into contact with the alloy must possess corrosion resistance and the coolant and inner surfaces of the equipment in the loop must have the required purity. As a result of solving these problems, a scientific basis has been developed for handling heavy coolants, new structural materials have been chosen or developed, and methods and devices have been developed for monitoring and regulating coolant quality and for removing impurities from the coolant and surfaces in the loop.     

水冷聚变堆主回路活化产物源项计算分析

水冷聚变堆中结构材料活化腐蚀产物和冷却剂活化产物是正常运行工况下的最主要放射性来源,也是反应堆运行及维护过程中工作人员辐照剂量的直接来源。本文使用CATE V2.1程序对国际热核聚变实验堆(International Thermonuclear Experimental Reactor,ITER)LIM-OBB(Limiter-Out-Board Baffle)冷却回路的活化腐蚀产物和水活化产物进行模拟计算,并根据CATE模拟得到的放射性活度通过点核积分程序分别计算正常运行1.2 a及停堆15 d的剂量率。计算结果表明,反应堆运行期间冷却剂活化产物比活度和剂量率远大于结构材料活化腐蚀产物,而停堆后冷却剂活化产物迅速衰变完,结构材料的活化腐蚀产物成为比活度和剂量率的主要来源。     

Feasibility study of applying the passive safety system concept to fusion–fission hybrid reactor

The fusion–fission hybrid reactor can produce energy, breed nuclear fuel, and handle the nuclear waste, etc., with the fusion neutron source striking the subcritical blanket. The passive safety system consists of passive residual heat removal system, passive safety injection system and automatic depressurization system was adopted into the fusion–fission hybrid reactor in this paper. Modeling and nodalization of primary loop, partial secondary loop and passive core cooling system for the fusion–fission hybrid reactor using relap5 were conducted and small break LOCA on cold leg was analyzed. The results of key transient parameters indicated that the actuation of passive safety system could mitigate the accidental consequence of the 4-inch cold leg small break LOCA on cold leg in the early time effectively. It is feasible to apply the passive safety system concept to fusion–fission hybrid reactor. The minimum collapsed liquid level had great increase if doubling the volume of CMTs to increase its coolant injection and had no increase if doubling the volume of ACCs.     

Nuclear Performance of Two-Phase Flow Cooling D-T Fusion Reactor Blankets

The aim of the present paper is to analyze the nuclear performance of a typical D-T fusion reactor blanket cooled by two-phase flow, and, in particular, the dependence of tritium breeding ratio (TBR), nuclear heating and neutron (energy) leakage on design variations such as the volume fraction γ occupied by coolant materials.

The value of γ plays a central role in determining the nuclear performance of the blanket considered. The TBR and nuclear heating decrease with decreasing γ while the inverse trend is found for the leakage from the blanket. To obtain the TBR greater than unity would require γ at least 30%. The feasibility of the two-phase flow cooling concept for D-T reactor blankets is contigent upon finding the way of taking advantage of the many good features associated with the flow, even at such γ.     

Effect of circulation velocity on radiation conversion of an organic coolant at high temperatures

The authors have studied the decomposition of certain organic coolants in the field of radiation from a nuclear reactor at 250–520°C with residence times of the order of tenths of a second. They found that with these short times the rate of radiation-thermal decomposition depends on the circulation speed (residence time). These results are explained on the hypothesis that these times are comparable with the decomposition times of aromatic radicals which are formed in these conditions. The effect studied is important in assessing the feasibility of improving the stability of an organic substance in a field of radiation, and in particular in the use of organic moderator-coolants in nuclear reactors.Translated from Atomnaya Énergiya, Vol. 22, No. 5, pp. 378–384, May 1967.     

Lead-alloy coolant technology and materials – technology readiness level evaluation

Lead and lead–bismuth eutectic heavy liquid metal coolants are under wide-ranging international investigation and development for advanced nuclear systems for energy production and waste transmutation (reactor-based or accelerator-driven). This report reviews the major supporting international R&D programs, the key advances in the main areas of coolant technology and materials, the state of technology, and the strategic directions for further development. Based on this review, we conservatively evaluate the technological readiness level (TRL) for programmatic and industrial applications in high-temperature advanced reactors to be 7, “one-dimensional engineering-scale demonstration”, or the first level in the proof-of-performance category. A 3-D engineering-scale integral test and demonstration facility should be the next step toward the realization of a test and demonstration nuclear system (reactor or accelerator-driven). The recent success of MEGAPIE, a 1 MW class lead–bismuth eutectic spallation target operating at the Paul Scherrer Institute signals that for such applications of short to intermediate durations at moderate temperatures, the TRL is close to 9, meaning the technology is nearly ready for deployment.     

Effect of Using Thorium Molten Salts on the Neutronic Performance of PACER

Utilization of nuclear explosives can produce a significant amount of energy which can be converted into electricity via a nuclear fusion power plant. An important fusion reactor concept using peaceful nuclear explosives is called as PACER which has an underground containment vessel to handle the nuclear explosives safely. In this reactor, Flibe has been considered as a working coolant for both tritium breeding and heat transferring. However, the rich neutron source supplied from the peaceful nuclear explosives can be used also for fissile fuel production. In this study, the effect of using thorium molten salts on the neutronic performance of the PACER was investigated. The computations were performed for various coolants bearing thorium and/or uranium-233 with respect to the molten salt zone thickness in the blanket. Results pointed out that an increase in the fissile content of the salt increased the neutronic performance of the reactor remarkably. In addition, higher energy production was obtained with thorium molten salts compared to the pure mode of the reactor. Moreover, a large quantity of ²³³U was produced in the blanket in all cases.     

Temperature distribution in mixed ThO2–UO2 fuel rods located in blanket of an inertial fusion energy breeder

Investigations of neutronic analysis and temperature distribution in fuel rods located in a blanket driven ICF (Inertial Confinement Fusion) have been performed for various mixed fuels and coolants under a first wall load of 5 MW/m². The fuel rods containing ThO₂ and UO₂ mixed by various mixing methods for achieving a flat fission power density are replaced in the blanket and cooled with different coolants; natural lithium, flibe, eutectic lithium and helium for the nuclear heat transfer. It is assumed that surface temperature of the fuel rod increases linearly from 500 °C (at top) to 700 °C (at bottom) during cooling fuel zone. Neutronic and temperature distribution calculations have been performed by MCNP4B Code and HEATING7, respectively. In the blanket fueled with pure UO₂ and cooled with helium, M (fusion energy multiplication ratio) increases to 3.9 due to uranium having higher fission cross-section than thorium. The high fission energy released in this blanket, therefore, causes proportionally increasing of temperature in the fuel rods to 823 °C. However, the M is 2.00 in the blanket fueled with pure ThO₂ and cooled with eutectic lithium because of more capture reaction than fission reaction. Maximum and minumum values of TBR (tritium breeding ratio) being one of main neutronic paremeters for a fusion reactor are 1.07 and 1.45 in the helium and the natural lithium coolant blanket, respectively. These consequences bring out that the investigated reactor can produce substantial electricity in situ during breeding fissile fuel and can be self-sufficient in the tritium required for the DT fusion driver in all cases of mixed fuels and coolant types. Quasi-constant fission power density profiles in FFB (fissile fuel breeding) zone are obtained by parabolically increasing mixture fraction of UO₂ in radial and axial directions for all coolant types. Such as, in the helium coolant blanket and the case of PMF (parabolically mixed fuel), Γ (peek-to-average fission power density ratio) of the blanket is reduced to 1.1, and the maximum temperatures of the fuel rods in radial direction of the FFB zone are also quasi-constant. At the same time, in the case of PMF, for all coolant types, the temperature profiles in the radial direction of the fuel rods rise proportionally with surface temperature from the top to the bottom of fuel rods in the axial direction. In other words, for each radial temperature profile in the axial direction, temperature differences between centerline and surface of the fuel rods are quasi-constant. According to the coolant types, these temperature diffences vary between 30 and 45 °C.     

The feasibility of cooling heavy-water reactors with supercritical fluids

It appears technically feasible to use supercritical carbon dioxide as a coolant for a CANDU-type reactor. A new supercritical loop is proposed in which the reactor is cooled by a single-phase fluid pumped in a high density liquid-like state. The supercritical fluid-cooled reactor has the advantage of gas-cooled reactors of avoiding dryout, and of liquid-cooled reactors of low coolant-circulation power. By eliminating dryout, the maximum operating temperature of the fuel sheath can be increased to 1021°F (550°C) for existing Canadian fuel bundles, with a coolant exit temperature of 855°F (458°C) producing steam comparable to that of conventional fossil-fuel plants. Since the reactor coolant exit temperature from the steam generator may be as high as 280°F (138°C) low-pressure steam may also be produced. A new dual-reheat cycle is proposed with an ideal overall plant efficiency of 33%, comparable to the Pickering generating station.     

Cross-comparison of fast reactor concepts with various coolants

Four fast reactor concepts using lead (LFR), liquid salt, NaCl-KCl-MgCl₂ (LSFR), sodium (SFR), and supercritical CO₂ (GFR) coolants are compared. Since economy of scale and power conversion system compactness are the same by virtue of the consistent 2400 MWt rating and use of the S-CO₂ power conversion system, the achievable plant thermal efficiency, core power density and core specific powers become the dominant factors. The potential to achieve the highest efficiency among the four reactor concepts can be ranked from highest to lowest as follows: (1) GFR, (2) LFR and LSFR, and (3) SFR. Both the lead- and salt-cooled designs achieve about 30% higher power density than the gas-cooled reactor, but attain power density 3 times smaller than that of the sodium-cooled reactor. Fuel cycle costs are favored for the sodium reactor by virtue of its high specific power of 65 kW/kgHM compared to the lead, salt and gas reactor values of 45, 35, and 21 kW/kgHM, respectively. In terms of safety, all concepts can be designed to accommodate the unprotected limiting accidents through passive means in a self-controllable manner. However, it does not seem to be a preferable option for the GFR where the active or semi-passive approach will likely result in a more economic and reliable plant. Lead coolant with its superior neutronic characteristics and the smallest coolant temperature reactivity coefficient is easiest to design for self-controllability, while the LSFR requires special reactivity devices to overcome its large positive coolant temperature coefficient. The GFR required a special core design using BeO diluent and a supercritical CO₂ reflector to achieve negative coolant void worth—one of the conditions necessary for inherent shutdown following large LOCA. Protected accidents need to be given special attention in the LSFR and LFR due to the small margin to freezing of their coolants, and to a lesser extent in the SFR.     

Space reactor power systems with no single point failures

Nuclear reactor power systems could revolutionize space exploration and support human outpost on the moon and Mars. This paper reviews various energy conversion technologies for use in space reactor power systems and provides estimates of the system's net efficiency and specific power, and the specific area of the radiator. The suitable combinations of the energy conversion technologies and the nuclear reactors, classified based on the coolant type and cooling method, for best system performance and highest specific power, are also discussed. In addition, a number of power system concepts with both static and dynamic energy conversion, but with no single point failures in reactor cooling, energy conversion and heat rejection, and for nominal electrical powers up to 110 kW_e, are presented. The first two power systems employ reactors cooled with lithium and sodium heat pipes, SiGe thermoelectric (TE) and alkali-metal thermal-to-electric conversion (AMTEC), and potassium heat pipes radiators. The reactors heat pipes operate at a fraction of the prevailing capillary or sonic limit, and in the case of a multiple heat pipes failure, those in the adjacent modules remove the additional heat load, thus maintaining the reactor adequately cooled and the power system operating at a reduced power. The third power system employs SiGe TE converters and a liquid metal cooled reactor with a divided core into six sectors that are neurotically and thermally coupled, but hydraulically decoupled. Each sector has a separate energy conversion loop, a heat rejection loop, and a rubidium heat pipes radiator panel. When a core sector experiences a loss-of-coolant, the fission power of the reactor is reduced, and that generated in the sector in question is removed by the circulating coolant in the adjacent sectors. The fourth power system employs a gas cooled reactor with a core divided into three identical sectors, and each sector is coupled to a separate Closed Brayton Cycle (CBC) loop with He-Xe binary mixture (40 g/mol) working fluid, a secondary loop with circulating liquid Nak-78, and two water heat pipes radiator panels.     

Static design of a liquid-salt-cooled pebble bed reactor (LSPBR)

A renewed interest has been raised for liquid-salt-cooled nuclear reactors. The excellent heat transfer properties of liquid-salt coolants provide several benefits, like lower fuel temperatures, higher average coolant temperature, increased core power density and better decay heat removal, and thus higher achievable core power. In order to benefit from the on-line refueling capability of a pebble bed reactor, the liquid salt pebble bed reactor (LSPBR) is proposed. This is a high temperature pebble bed reactor with a fuel design similar to existing HTRs, but using a liquid-salt as coolant. In this paper, the selection criteria for the liquid-salt coolant are described. Based on its neutronic properties, LiF–BeF₂ (flibe) was selected for the LSPBR. Two designs of the LSPBR were considered: a cylindrical core and an annular core with a graphite inner reflector. Coupled neutronic thermal-hydraulic calculations were performed to obtain the steady state power distribution and the corresponding fuel temperature distribution. Calculations were performed to investigate the decay heat removal capability in a protected loss-of-forced cooling accident. The maximum allowable power that can be produced with the LSPBR is hereby determined.     

Lead-cooled system design and challenges in the frame of Generation IV International Forum

The Generation IV International Forum (GIF) Technology Roadmap identified the Lead-cooled Fast Reactor (LFR) as a technology well suited for electricity generation, hydrogen production and actinide management in a closed fuel cycle. One of the most important features of the LFR is the fact that lead is a relatively inert coolant, a feature that conveys significant advantages in terms of safety, system simplification, and the consequent potential for economic performance.In 2004, the GIF LFR Provisional System Steering Committee was organized and began to develop the LFR System Research Plan. The committee selected two pool-type reactor concepts as candidates for international cooperation and joint development in the GIF framework: these are the Small Secure Transportable Autonomous Reactor (SSTAR); and the European Lead-cooled System (ELSY).The high boiling point (1745 °C) of lead has a beneficial impact to the safety of the system, whereas its high melting point (327.4 °C) requires new engineering strategies, especially for In-Service-Inspection and refuelling. Lead, especially at high temperatures, is also relatively corrosive towards structural materials. This necessitates that coolant purity and the level of dissolved oxygen be carefully controlled, in addition to the proper selection of structural materials.For the GIF LFR concepts, lead has been chosen as the coolant rather than Lead-Bismuth Eutectic primarily because of its greatly reduced generation of the alpha-emitting ²¹⁰Po isotope formed in the coolant. This results in significantly reduced levels of radioactive contamination of the coolant while minimizing the effect of decay power in the coolant from such contaminants; an additional consideration is the desire to eliminate dependence on bismuth which might be a limited resource.This paper provides an overview of the historical development of the LFR, a summary of the advantages and challenges associated with heavy liquid metal coolants, and an update of the current status of development of LFR concepts under consideration. The main characteristics of the SSTAR and ELSY systems are summarized, and the current status of design of each system is presented. Because of the significant recent efforts in the ELSY system design, greater emphasis is placed on the ELSY plant, with focus on the technological development and design provisions intended to overcome or alleviate recognized drawbacks to the use of heavy liquid metal coolants. In the case of the SSTAR system for which development has proceeded more slowly, a more limited summary is provided. It is noted that both systems share many of the same research needs and objectives thus providing a strong basis for international collaboration.     

Integral analysis of large LOCA beyond design basis accident scenario in RBMK-1500

The integral analysis of severe accident scenario for RBMK-1500 was performed using combined approach with RELAP5, RELAP/SCDAPSIM, ASTEC and COCOSYS codes. The performed analysis covered response of the reactor core, the reactor cooling system and the confinement. There were performed several analyses: the first analysis assumed that operators take no action or their actions are not successful to provide the coolant injection to the reactor core; the other analyses were performed to investigate the accident management measures to restore the core cooling at different temperatures of the reactor core. The results of performed analyses showed that the operators have ∼5 h before the ruptures of fuel claddings occur and ∼8 h before the onset of exothermic steam-zirconium reaction. The coolant injection to the reactor core should be restored as soon as possible in order to prevent high hydrogen concentrations in the confinement and significant release of the fission products to the environment.     

Molten salts and nuclear energy production

Molten salts (fluorides or chlorides) were considered near the beginning of research into nuclear energy production. This was initially due to their advantageous physical and chemical properties: good heat transfer capacity, radiation insensitivity, high boiling point, wide range solubility for actinides. In addition it was realised that molten salts could be used in numerous situations: high temperature heat transfer, core coolants with solid fuels, liquid fuel in a molten salt reactor, solvents for spent nuclear solid fuel in the case of pyro-reprocessing and coolant and tritium production in the case of fusion. Molten salt reactors, one of the six innovative concepts chosen by the Generation IV international forum, are particularly interesting for use as either waste incinerators or thorium cycle systems. As the neutron balance in the thorium cycle is very tight, the possibility to perform online extraction of some fission product poisons from the salt is very attractive. In this article the most important questions that must be addressed to demonstrate the feasibility of molten salt reactor will be reviewed.     

Neutronic analysis of an inertial fusion energy breeder with SiCf/SiC

Neutronic performance of a blanket driven ICF (Inertial confinement fusion) neutron based on SiC_f/SiC composite material is investigated for fissile fuel breeding. The investigated blanket is fueled with ThO₂ and cooled with natural lithium or (LiF)₂BeF₂ or Li₁₇Pb₈₃ or ⁴He coolant. MCNP4B Code is used for calculations of neutronic data per DT neutron. Calculations have show that values of TBR (tritium breeding ratio) being one of the main neutronic paremeters of fusion reactors are greater than 1.05 in all type of coolant, and the breeder hybrid reactor is self-sufficient in the tritium required for the DT fusion driver. Calculations show that natural lithium coolant blanket has the highest TBR (1.298) and M (fusion energy multiplication) (2.235), Li₁₇Pb₈₃ coolant blanket has the highest FFBR (fissile fuel breeding ratio) (0.3489) and NNM (net neutron multiplication) (1.6337). ⁴He coolant blanket has also the best Γ (peek-to-average fission power density ratio) (1.711). Values of neutron leakage out of the blanket in all type of coolants are quite low due to SiC reflector and B₄C shielding.     

Neutronic performance of new coolants in a fusion–fission (hybrid) reactor

Selection of coolant used in the fuel zone of a fusion–fission (hybrid) reactor affects the neutronic performance of the blanket much. Recently, two coolants namely, Flinabe and Li₂₀Sn₈₀ have been investigated to use in fusion reactors as tritium breeder and energy carrier due to their advantages of low melting point, low vapor pressure. In this study, neutronic performance of these coolants in a hybrid reactor using Canada Deuterium Uranium Reactor (CANDU) spent fuel was investigated for an operation period of 48 months. And also that of natural lithium and Flibe was also examined for comparison. Neutron transport calculations were conducted on a simple experimental hybrid blanket in a cylindrical geometry with the help of the SCALE4.3 system by solving the Boltzmann transport equation with the XSDRNPM code in 238 neutron groups and a S₈–P₃ approximation.