A Review on the Potential Use of Austenitic Stainless Steels in Nuclear Fusion Reactors

Various engineering materials; austenitic stainless steels, ferritic/martensitic steels, vanadium alloys, refractory metals and composites have been suggested as candidate structural materials for nuclear fusion reactors. Among these structural materials, austenitic steels have an advantage of extensive technological database and lower cost compared to other non-ferrous candidates. Furthermore, they have also advantages of very good mechanical properties and fission operation experience. Moreover, modified austenitic stainless (Ni and Mo free) have relatively low residual radioactivity. Nevertheless, they can’t withstand high neutron wall load which is required to get high power density in fusion reactors. On the other hand, a protective flowing liquid wall between plasma and solid first wall in these reactors can eliminate this restriction. This study presents an overview of austenitic stainless steels considered to be used in fusion reactors.     

Transmutation missions for fusion neutron sources

There are a number of potential neutron transmutation missions (destruction of long-lived radioisotopes in spent nuclear fuel, ‘disposal’ of surplus weapons-grade plutonium, ‘breeding’ of fissile nuclear fuel) that perhaps best can be performed in sub-critical nuclear reactors driven by a neutron source. The requirements on a tokamak fusion neutron source for such transmutation missions are significantly less demanding than for commercial electrical power production. A tokamak fusion neutron source based on the current physics and technology database (ITER design base) would meet the needs of the spent nuclear fuel transmutation mission; the technical issue would be achieving ≥50% availability, which would require advances in component reliability and in steady-state physics operation.     

Preliminary Comparison of Radioactive Waste Disposal Cost for Fusion and Fission Reactors

The environmental and economic impact of radioactive waste (radwaste) generated from fusion power reactors using five types of structural materials and a fission reactor has been evaluated and compared. Possible radwaste disposal scenario of fusion radwaste in Japan is considered. The exposure doses were evaluated for the skyshine of gamma-ray during the disposal operation, groundwater migration scenario during the institutional control period of 300 years and future site use scenario after the institutional period. The radwaste generated from a typical light water fission reactor was evaluated using the same methodology as for the fusion reactors. It is found that radwaste from the fusion reactors using F82H and SiC/SiC composites without impurities could be disposed by the shallow land disposal presently applied to the low level waste in Japan. The disposal cost of radwaste from five fusion power reactors and a typical light water reactor were roughly evaluated and compared.     

Nuclear power technologies at the stage of sustainable nuclear power development

It is not simple to solve the problem of competitiveness of nuclear power technologies in evolutionary upgrading the conventional nuclear power plants (NPP) such as light water reactors (LWR), which requires high expenditure for safety. Moreover, the existing LWRs cannot provide nuclear power (NP) for a long time (hundreds of years) because the efficiency of use of natural uranium is low and closing the nuclear fuel cycle (NFC) for those reactors is not expedient.The highlighted problem can be solved in the way of use of innovative nuclear power technology in which natural uranium power potential is used effectively and the intrinsic conflict between economic and safety requirements has been essentially mitigated.The technology that is most available and practically demonstrated is the use of reactors SVBR-100 — small power multi-purpose modular fast reactors (100 MWe) cooled by lead-bismuth coolant (LBC). This technology has been mastered for nuclear submarines’ reactors in Russia.High technical and economical parameters of the NPP based on RF SVBR-100 are determined from the fact that the potential energy stored in LBC per a volume unit is the lowest.The compactness of the reactor facility SVBR-100 that results from integral arrangement of the primary circuit equipment allows realizing renovation of power-units LWRs, the vessels’ lifetime of which has been expired. So due to this fact, high economical efficiency can be obtained.The paper also validates the economical advantage of launching the uranium-fueled fast reactors with further changeover to the closed NFC with use of plutonium extracted from the own spent nuclear fuel in comparison with launching fast reactors directly with on uranium-plutonium fuel on the basis of plutonium extraction from spent nuclear fuel of LWRs.     

Study on (n,t) Reactions of Zr, Nb and Ta Nuclei

The world faces serious energy shortages in the near future. To meet the world energy demand, the nuclear fusion with safety, environmentally acceptability and economic is the best suited. Fusion is attractive as an energy source because of the virtually inexhaustible supply of fuel, the promise of minimal adverse environmental impact, and its inherent safety. Fusion will not produce CO₂ or SO₂ and thus will not contribute to global warming or acid rain. Furthermore, there are not radioactive nuclear waste problems in the fusion reactors. Although there have been significant research and development studies on the inertial and magnetic fusion reactor technology, there is still a long way to go to penetrate commercial fusion reactors to the energy market. Because, tritium self-sufficiency must be maintained for a commercial power plant. For self-sustaining (D-T) fusion driver tritium breeding ratio should be greater than 1.05. And also, the success of fusion power system is dependent on performance of the first wall, blanket or divertor systems. So, the performance of structural materials for fusion power systems, understanding nuclear properties systematic and working out of (n,t) reaction cross sections are very important. Zirconium (Zr), Niobium (Nb) and Tantal (Ta) containing alloys are important structural materials for fusion reactors, accelerator-driven systems, and many other fields. In this study, (n,t) reactions for some structural fusion materials such as ^{88,90,92,94,96}Zr, ^93,94,95Nb and ^179,181Ta have been investigated. The calculated results are discussed andcompared with the experimental data taken from the literature.     

RELAP5/3.2 assessment against low pressure onset of flow instability in parallel heated channels

Best Estimate computer codes have been, so far, developed for safety analysis of nuclear power plants and were extensively validated against a large set of separate effects and integral test facilities experimental data relevant to such kind of reactors. Their application to research reactors is not fully straightforward. Modelling problems generally emerge when applying existing models to low pressure and more particularly to subcooled flow boiling situations. The objective of the present work is to investigate the RELAP5/3.2 system code capabilities in predicting phenomena that could be encountered under abnormal research reactor’s operating conditions. For this purpose, the separate effect related to the static onset of flow instability is investigated. The cases considered herein are the flow excursion tests performed at the Oak Ridge National Laboratory thermal hydraulic test loop (THTL) as well as some representative Whittle and Forgan (W & F) experiments. The simulation results are presented and the capabilities of RELAP5/Mod 3.2 in predicting this critical phenomenon are discussed.     

Basic design guideline for the preliminary engineering design of PIE facilities in IFMIF/EVEDA

The international fusion materials irradiation facility (IFMIF) is an accelerator-based intense 14 MeV neutron source for testing fusion reactor materials. Under broader approach (BA) agreement between EURATOM and Japan, the engineering validation and engineering design activity (EVEDA) were started from 2007. The IFMIF needs the post irradiation examination (PIE) facilities to generate a materials irradiation database for the design and licensing of fusion DEMO reactors. In this study we examined and discussed about the safety such as remote handling, hot cell design, and the equipments and apparatus of hot cells, and we summarized a basic design guideline for the preliminary engineering design of the PIE facilities.     

Recent applications of small-angle neutron scattering in the characterization of irradiated steels for nuclear technologies

Small-angle neutron scattering (SANS) is a powerful experimental tool to investigate the microstructural evolution under irradiation in steels for fission and future fusion reactor systems. We present recent SANS results concerning the modelling of helium bubble growth in F82H-mod. steel implanted with α-particles and the dose dependence of microstructural radiation damage in Eurofer-97 steel for fusion reactors irradiated at 250 °C. The discussion of these results is focussed on the quality of the metallurgical information obtained by such SANS measurements and consequently on their usefulness also for engineering and design purposes.     

Realistic seismic qualification using the updated finite element model for in-core components of reactors

It is now mandatory to seismically qualify the safety-related structures and components used in the nuclear power plants. Among several qualification approaches the qualification by the analysis using finite element (FE) method is the most common approach used in practice. However, the estimated dynamic behaviour by FE model of a structure is known to show significant deviations from the dynamic behaviour of the ‘as-installed’ structure in many cases. Considering such limitations, few researchers have advocated re-qualification of such structures after their installation at site to enhance the confidence in qualification vis-à-vis plant safety. For such an exercise, validation of FE model with experimental modal data is important. A validated FE model can be obtained by the model updating methods in conjugation with the in situ experimental modal data. Such a model can then be used for qualification. However, for the reactor in-core components such a modal testing and FE model updating may not be straightforward. Hence, the complication involved in the reliable seismic qualification of in-core components and the advantage of using the FE model updating has been brought out in the paper through an example of a typical in-core component—a perforated horizontal tube recently installed in a nuclear reactor in India.     

Damage parameters of structural materials in fusion environment compared to fission reactor irradiation

Calculations were performed to quantify the damage parameters in the leading candidate structural and plasma facing materials when used in magnetic and inertial confinement fusion systems and when irradiated in fission reactors. The structural materials considered are ferritic steel, austenitic steel, vanadium alloy and SiC/SiC composite. Plasma facing materials included beryllium, tungsten, and carbon fiber composites. Atomic displacement damage and gas production rates are greatly influenced by the neutron energy spectrum. For the same neutron wall loading, atomic displacement damage is slightly lower in inertial fusion systems than in magnetic fusion systems but gas production is about a factor of 2 lower. In addition, much lower gas production is obtained in samples irradiated in fission reactors. The results help guide irradiation experiments in fission reactors to properly simulate the damage environment in fusion systems and facilitate extrapolating to the expected material performance in fusion systems.     

基于PXI Express的托卡马克同步数据采集系统设计

随着J-TEXT装置的发展,原有的数据采集系统在稳定性、模块化、采样率等方面已不能满足装置运行的需要,所以需建立一套新的数据采集系统来满足实验需求。本文介绍了基于PXI Express的托卡马克分布式高速同步数据采集系统的设计与实现。系统的采集单元由PXIe机箱NI PXIe-1062Q、PXIe控制器NI PXIe-8133和高速同步数据采集卡NI PXIe-6368组成,兼容ITER CODAC最新标准,具有良好的机械封装性、模块化程度高和高采样率等优点。系统采用同步差分采集方式采集实验数据,并将数据存储于核聚变领域通用的MDSplus数据库中。测试和使用结果表明,系统能在2 MSps采样率下连续稳定工作,可较好地满足装置运行的需要。     

The elastic properties of lithium oxide and their variation with temperature

The energies of the acoustic phonon modes in a single crystal of ⁷Li₂O have been measured in the temperature range 293–1603 K using the technique of inelastic neutron scattering. The slopes of the phonon energy dispersion curves as they approach the Brillouin zone centre give values for the cubic elastic stiffness constants, C_ij. C₁₁ is found to undergo a sharp decrease above ˜ 1350 K similar to that observed in structurally related compounds, such as CaF₂ and SrCl₂, as they undergo a transition to a fast-ion phase. The Reuss and Voigt averaging methods have been used to calculate the temperature dependence of the adiabatic Young's modulus, shear modulus, bulk modulus and Poisson's ratio of polycrystalline Li₂O. Estimates of the corresponding isothermal values are obtained using an expression for the linear thermal expansion coefficient of Li₂O obtained in this work, together with thermodynamic data available from specific heat measurements. These results represent the first experimental data describing the elastic properties of Li₂O at elevated temperatures and are important in predicting the behaviour of this material in its potential role as a tritium breeding blanket material for future fusion reactors.     

Fuel elements of nuclear reactors

On the basis of foreign reports presented at the Second International Conference on the Peaceful Uses of Atomic Energy, Geneva, 1958, the characteristics of construction of fuel elements (FE) and basic data relating to them for a series of reactors are given. Problems on the selection of fuel and construction materials as well as the technology of preparing FE for various types of nuclear reactor are examined.     

Thermo-mechanical analysis of a DEMO divertor cooling finger under the EFREMOV test conditions

A divertor component of the forthcoming DEMO fusion reactor should be able to withstand heat flux loads larger than 10 MW/m². Successful design should withstand high flux loads for a number of load cycles since initially the DEMO reactor is expected to operate in a non-steady-state mode. Computations for evaluating the structural response of the divertor published so far have, however, been based on the stationary approach. A combined computational fluid dynamics and structural model for evaluating the structural response of a divertor under the non-stationary load conditions is therefore developed in this work. Heat transfer coefficients between the helium and inner surface of the thimble are calculated first by solving the helium steady-state flow equations. Spatially distributed heat transfer coefficients are then used as a boundary condition in a non-stationary thermo-mechanical analysis of the divertor. This analysis is performed for a number of load cycles under different surface heat flux levels. The model is validated against the EFREMOV test experimental conditions, designed to be close to reactor operation conditions. Good agreement of the highest temperatures on the tile’s top surface with the experimental data is obtained. The results suggest that there are three critical regions in the design where damage could initialize: (a) the thimble’s inner surface with the highest thermal gradients, (b) the tile’s outer surface and (c) the filler layer of the brazed tile-thimble joint where the temperature is higher than permissible. Post-examination data of experimental specimen confirm these conclusions as cracks were observed in the above mentioned areas (a) and (b), while melting of the layer (c) was also observed.     

Calculations of Cross-Sections and Astrophysical S-factors for the (α,n) Reactions of Some Structural Fusion Materials

Structural material selection in design of fusion reactors is very crucial. These structural materials should satisfy the hard conditions such as high thermo-mechanical stresses, high heat loads and severe radiation damage without compromising on safety considerations. The materials such as titanium, vanadium and chromium are used in the construction of fusion reactors. Therefore, it is important to examine these materials. Obtained results from the nuclear reactions using structural materials can be used for developing of these structural materials. For this reason, in this study cross sections of the ⁴⁶Ti(α,n) ⁴⁹Cr, ⁵⁰Cr(α,n) ⁵³Fe and ⁵¹V(α,n) ⁵⁴Mn reactions have been calculated at 2–20 MeV energy range. In these theoretical calculations, the TALYS 1.8 and NON-SMOKER codes were used. Also, the astrophysical S-factors which describe the possibility of reaction in low energies were calculated. Results of our calculations were checked to the experimental data obtained from EXFOR database.     

Bootstrap calculation of ultimate strength temperature maxima for neutron irradiated ferritic/martensitic steels

The dependence of mechanical properties of ferritic/martensitic (F/M) steels on irradiation temperature is of interest because these steels are used as structural materials for fast, fusion reactors and accelerator driven systems. Experimental data demonstrating temperature peaks in physical and mechanical properties of neutron irradiated pure iron, nickel, vanadium, and austenitic stainless steels are available in the literature. A lack of such an information for F/M steels forces one to apply a computational mathematical-statistical modeling methods. The bootstrap procedure is one of such methods that allows us to obtain the necessary statistical characteristics using only a sample of limited size. In the present work this procedure is used for modeling the frequency distribution histograms of ultimate strength temperature peaks in pure iron and Russian F/M steels EP-450 and EP-823. Results of fitting the sums of Lorentz or Gauss functions to the calculated distributions are presented. It is concluded that there are two temperature (at 360 and 390 °C) peaks of the ultimate strength in EP-450 steel and single peak at 390 °C in EP-823.     

Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation’s energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.     

Neutron irradiation studies on low density pan fiber based carbon/carbon composites

Carbon has been extensively used in nuclear reactors and there has been growing interest to develop carbon-based materials for high-temperature nuclear and fusion reactors. Carbon-carbon composite materials as against conventional graphite material are now being looked into as the promising materials for the high temperature reactor due their ability to have high thermal conductivity and high thermal resistance. Research on the development of such materials and their irradiation stability studies are scant. In the present investigations carbon-carbon composite has been developed using polyacrylonitrile (PAN) fiber. Two samples denoted as Sample-1 and Sample-2 have been prepared by impregnation using phenolic resin at pressure of 30 bar for time duration 10 h and 20 h respectively, and they have been irradiated by neutrons. The samples were irradiated in a flux of 10¹² n/cm²/s at temperature of 40 °C. The fluence was 2.52 × 10¹⁶ n/cm². These samples have been characterized by XRD and Raman spectroscopy before and after neutron irradiation. DSC studies have also been carried out to quantify the stored energy release behavior due to irradiation. The XRD analysis of the irradiated and unirradiated samples indicates that the irradiated samples show the tendency to get ordered structure, which was inferred from the Raman spectroscopy. The stored energy with respect to the fluence level was obtained from the DSC. The stored energy from these carbon composites is very less compared to irradiated graphite under ambient conditions.     

The fusion breeder

The fusion breeder is a fusion reactor designed with special blankets to maximize the transmutation by 14 MeV neutrons of uranium-238 to plutonium or thorium to uranium-233 for use as a fuel for fission reactors. Breeding fissile fuels has not been a goal of the U.S. fusion energy program. This paper suggests it is time for a policy change to make the fusion breeder a goal of the U.S. fusion program and the U.S. nuclear energy program. There is wide agreement that many approaches will work and will produce fuel for five equal-sized LWRs, and some approach as many as 20 LWRs at electricity costs within 20% of those at today's price of uranium ($30/lb of U₃O₈). The blankets designed to suppress fissioning, called symbiotes, fusion fuel factories, or just fusion breeders, will have safety characteristics more like pure fusion reactors and will support as many as 15 equal power LWRs. The blankets designed to maximize fast fission of fertile material will have safety characteristics more like fission reactors and will support 5 LWRs. This author strongly recommends development of the fission suppressed blanket type, a point of view not agreed upon by everyone. There is, however, wide agreement that, to meet the market price for uranium which would result in LWR electricity within 20% of today's cost with either blanket type, fusion components can cost severalfold more than would be allowed for pure fusion to meet the goal of making electricity alone at 20% over today's fission costs. Also widely agreed is that the critical-path-item for the fusion breeder is fusion development itself; however, development of fusion breeder specific items (blankets, fuel cycle) should be started now in order to have the fusion breeder by the time the rise in uranium prices forces other more costly choices.     

Fusion Reactors: Radiation Hazards and Risk

The safety advantage of next-step fusion machines, such as ITER, and of future fusion power reactors, comes not so much from the engineered safety features incorporated into the design, as it does from the inherent, or intrinsic, features of the machine/reactor. The key intrinsic feature is the small quantity of radioactive fuel (tritium) and its low health hazard. The radioactive health hazard is small enough that even the largest unmitigated release would produce offsite consequences that are small. Nevertheless, engineered safety features are necessary to reduce the probability of an unmitigated release to a small enough value, such that the resulting risk is within risk targets. The risk targets are established such that the total accident risk is not greater than the risk from normal operation. Risk targets are also useful in determining the performance requirements for the engineered safety features. Such a risk-based approach should be considered when developing fusion safety requirements, taking account of the low radiation hazards involved.