Design of coated fuel particles for a hybrid fusion-fission system

Designs have been developed for coated ThO₂ fuel particles to be used in a hybrid fusion-fission system that could be operated without reprocessing. The fresh fertile fuel particle would first be cycled through the blanket of a fusion reactor to breed ²³³U, which would then be ‘burned’ in a thermal fission reactor. The depleted fuel would then be refreshed in a second pass through the fusion reactor, and the process above repeated as many times as feasible. Designs of coated particles for up to three cycles through the hybrid system of reactors have been developed. The outer structural layer for these particles is made from vapor-deposited silicon carbide, because of its remarkable dimensional stability under fast neutron irradiation, and an inner layer of porous pyrocarbon is used to accommodate the buildup of gaseous reaction products inside the particle. The production of gaseous emission products from the interaction of high-energy fusion neutrons with coating materials and with the oxygen in the kernel contributes significantly to pressure vessel stresses in these coatings, whereas gaseous fission products alone dominate in conventional thermal reactors. The most stringent design for the three-cycle particle is identical in fuel loading to the reference fertile particle for an HTGR, which would constitute an ideal hybrid partner for the fusion reactor. Consideration is also given to coated-particle designs for the containment of the bred tritium used to fuel the D-T fusion reactor.     

Reprocessing free nuclear fuel production via fusion fission hybrids

Fusion fission hybrids, driven by a copious source of fusion neutrons can open qualitatively “new” cycles for transmuting nuclear fertile material into fissile fuel. A totally reprocessing-free (ReFree) Th²³²–U²³³ conversion fuel cycle is presented. Virgin fertile fuel rods are exposed to neutrons in the hybrid, and burned in a traditional light water reactor, without ever violating the integrity of the fuel rods. Throughout this cycle (during breeding in the hybrid, transport, as well as burning of the fissile fuel in a water reactor) the fissile fuel remains a part of a bulky, countable, ThO₂ matrix in cladding, protected by the radiation field of all fission products. This highly proliferation-resistant mode of fuel production, as distinct from a reprocessing dominated path via fast breeder reactors (FBR), can bring great acceptability to the enterprise of nuclear fuel production, and insure that scarcity of naturally available U²³⁵ fuel does not throttle expansion of nuclear energy. It also provides a reprocessing free path to energy security for many countries. Ideas and innovations responsible for the creation of a high intensity neutron source are also presented.     

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.     

Reaction balance and efficiency analysis of a D-D fusion/fission hybrid with satellite D-3He reactors

Selected reactor physics and isotope balance characteristics of a fusion hybrid supported D-³He satellite nuclear energy system are formulated and investigated. The system consists of two types of reactors: a parent D-fueled fusion device and a number of smaller reactors optimized for D-³He fusion. The parent hybrid station breeds the helium-3 for the satellites and also breeds fissile fuel for an existing fission reactor economy. Various hybrid operational regimes are examined in order to determine favorable reactorQ values and effective fusion and fission efficiencies. A number of analytical correlations between power output, plasma energetics, blanket neutronics, breeding capacity, and energy conversion cycles are established and evaluated. Numerical examples of performance parameters such as fission-to-fusion power, overall conversion efficiency, and the ratio of satellite to parent fusion power are presented. The range of reactor efficiencies is elucidated as affected by the internal plasma power balances. As an upper bound based on optimistic injection and direct conversion efficiencies, we find the D-³He satellite system power output attaining at best 1/3 of the parent fusion power.     

Possibility of long-term nuclear power development based on today’s technology

The possibility of long-term nuclear power development with a uranium fuel cycle based on ²³⁸U burnup and todays industrial technology is investigated. It is shown that such development is possible with fast reactors, including with sodium coolant. In this case, incomplete fuel reprocessing is admissable in a closed fuel cycle employing a pyroelectrochemical technology, which allows some fission products and actinides to be present in the fresh fuel prepared for reloading after reprocessing. These fission products and actinides can be burned in a reactor, thereby decreasing the quantity of radioactive wastes compared with the complete reprocessing with chemical separation of the fuel elements and decreasing the radiation load on the environment.Translated from Atomnaya Ènergiya, Vol. 97, No. 4, pp. 252–260, October, 2004.     

Recycling of uranium from spent RBMK fuel in the nuclear fuel cycle

It is shown that there is promise in using the uranium product obtained by reprocessing spent nuclear fuel from RBMK reactors as a non-initial fuel source for thermal reactors. A technical path for spent nuclear fuel from RBMK reactors is proposed: radiochemical reprocessing and obtaining oxides of recycled uranium. Oxides of the category RBMK-poor are packed and then stored in a near-surface storage facility; oxides of the category RBMK-rich are fluoridated, and UF₆ is fed into separation production for additional enrichment to the required content of ²³⁵U. Additional advantages of recycled RBMK uranium as a source of non-initial ²³⁵U are the low content of ²³²U and the relatively low activity of spent fuel, which simplifies its reprocessing.     

End product economics and fusion research program priorities

It is shown that deuterium based fusion fuels and reactors based on them face severe technological disadvantages in comparison with fission based systems as power sources for central station electric power plants. The author postulates the most plausible deuterium based fusion reactor consistent with the physics of the fusion reaction itself and compares this reactor (called OMR-DT) with existing fission reactors. Since neutrons are the main problem in fusion, the author suggests that a great deal more effort should be given to the study of non-Maxwellian plasmas with the emphasis on neutron-free fuel cycles. The author also suggests that the deuterium based fusion driver may play its best role as a fissile fuel producer.     

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.     

Suitability of Small Scale Linear Systems for a Fission–Fusion Reactor, Breeder, and Waste Transmutation

To date the magnetic fusion effort has been almost entirely devoted to only one application, that being a multi gigawatt central station electric plant. Given the already well established fission based industry, the likelihood that fusion will have any impact on curbing the current carbon-based energy demand in the 21st century is slim. This paper advocates that the first and primary use of fusion neutrons should be as the driver for a sub-critical fission nuclear energy system—a fission–fusion hybrid reactor. This system can also be utilized to transmute long-lived radioactive wastes, and breed fissile nuclear fuel for several additional fission reactors. A small-scale fusion system based on a reciprocating fusion cycle employing the magneto-kinetic compression of the Field Reversed Configuration (FRC) is particularly well suited for this application. The characteristics of this fusion neutron driver and the potential for transmutation of long-lived nuclear wastes and breeding of fissile nuclear fuel in a blanket are presented.     

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.     

Neutronic analysis of a thorium fusion breeder with enhanced protection against nuclear weapon proliferation

The fissile breeding capability of a (D,T) fusion-fission (hybrid) reactor fueled with thorium is analyzed to provide nuclear fuel for light water reactors (LWRs). Three different fertile material compositions are investigated for fissile fuel breeding: (1) ThO₂; (2) ThO₂ denaturated with 10% natural-UO₂ and (3) ThO₂ denaturated with 10% LWR spent fuel. Two different coolants (pressurized helium and Flibe ‘Li₂BeF₄’) are selected for the nuclear heat transfer out of the fissile fuel breeding zone. Depending on the type of the coolant in the fission zone, fusion power plant operation periods between 30 and 48 months are evaluated to achieve a fissile fuel enrichment quality between 3 and 4%, under a first-wall fusion neutron energy load of 5 MW/m² and a plant factor of 75%. Flibe coolant is superior to helium with regard to fissile fuel breeding. During a plant operation over four years, enrichment grades between 3.0 and 5.8% are calculated for different fertile fuel and coolant compositions. Fusion breeder with ThO₂ produces weapon grade ²³³U. The denaturation of the ²³³U fuel is realized with a homogenous mixture of 90% ThO₂ with 10% natural-UO₂ as well as with 10% LWR spent nuclear fuel. The homogenous mixture of 90% ThO₂ with 10% natural-UO₂ can successfully denaturate ²³³U with ²³⁸U. The uranium component of the mixture remains denaturated over the entire plant operation period of 48 months. However, at the early stages of plant operation, the generated plutonium component is of weapon grade quality. The plutonium component can be denaturated after a plant operation period of 24 and 30 months in Flibe cooled and helium cooled blankets, respectively. On the other hand, the homogenous mixture of 90% ThO₂ with 10% LWR spent nuclear fuel remains non-prolific over the entire period for both, uranium and plutonium components. This is an important factor with regard to international safeguarding.     

Safety aspects and environmental impact of an inertial confinement fusion reactor

Releases into the environment of radioactive materials contained in heavy ion fusion (HIF) reactor plants must be prevented by similar safety design concepts as they are applied to present fission converter (e.g. LWR's) and breeder reactors (LMFBR's). This study identifies significant safety aspects of inertial confinement fusion power plant concepts and relates them to the more familiar basis of knowledge about the safety and the hazards of other advanced nuclear power reactor systems such as the LMFBR. Assessments of doses to be expected after the release of tritium from HIF reactor plants — normally and accidentally — are performed and compared with dose limits and with doses resulting from facilities of the fission fuel cycle. Needs for safety related research and development specifically for inertial confinement fusion as well as for the modelling of the various exposure pathways due to released tritium are pointed out.     

基于MCNP和ORIGEN的熔盐快堆燃耗分析计算

熔盐堆是6种第4代先进核能系统中唯一使用液态燃料设计的反应堆型,其堆芯一回路中循环流动的熔盐既是燃料,也是冷却剂。这一特征在省去燃料元件加工制造步骤的同时,也使得熔盐堆能进行在线处理和在线添料的操作。因此,传统固态反应堆燃耗分析程序不再适用于熔盐堆。本文以熔盐快堆(MSFR)为分析对象,基于物理分析程序MCORE(MCNP+ORIGEN),将上述熔盐堆特点考虑进去,开发出能进行熔盐堆燃耗分析的MCORE-MS。初步分析表明,233 U-started模式下,熔盐在线处理可有效降低堆芯熔盐中裂变产物的含量,提高中子经济性。MSFR运行过程中能够一直保持负的温度反应性系数。     

A Compact Fusion–Fission Hybrid Reactor

Up today, two hyper research projects to achieve nuclear fusion energy exist; inertial confinement fusion (ICF) driven by laser, called national ignition facility (NIF) and magnetic confinement fusion the international thermonuclear experimental reactor (ITER) project. In reaching the required temperature and pressure, to ignite nuclear fusion reactor, is technologically complex and economically expensive. Thus, a breakthrough and a short cut, other alternative methods should be considered. Pulsed power ICF driver with repetitive pulse operation, mainly dense plasma focus (DPF) machines for high yield fusion neutrons could be taken as drivers for the fission blanket operation. The setup can be a cost-effective and efficient. In this article, we consider a set of two medium energy sizes DPF to produce simultaneously dense plasma columns, operating as thermonuclear plasma driver, to pierce the pellet target for external nuclear fusion reactions. These DPFs produce sufficient fast neutrons for the fission process in the neutral uranium or thorium and/or weak enriched uranium blanket. The drive systems and the concept for delivering thermonuclear plasma to pellets target in the magnetic free zone of central region will be presented. The feasibility of such fusion–fission hybrid reactor will be discussed.     

A fusion–fission hybrid reactor with water-cooled pressure tube blanket for energy production

A fusion–fission hybrid reactor is proposed to achieve the energy gain of 3000 MW thermal power with self-sustaining tritium. The hybrid reactor is designed based on the plasma conditions and configurations of ITER, as well as the well-developed pressurized light water cooling technologies. For the sake of safety, the pressure tube bundles are employed to protect the first wall from the high pressure of coolant. The spent nuclear fuel discharged from 33GWD/tU Light Water Reactors (LWRs) and natural uranium oxide are taken as driver fuel for energy multiplication. According to thermo-mechanics calculation results, the first wall of 20 mm is safe. The radiation damage analysis indicates that the first wall has a lifetime of more than five years. Neutronics calculations show that the proposed hybrid reactor has high energy multiplication factor, tritium breeding ratio and power density; the fuel cannot reach the level of plutonium required for a nuclear weapon. Thermal-hydraulic analysis indicates that the temperatures of the fuel zone are well below the limited values and a large safety margin is provided.     

Long life small CANDLE-HTGRs with thorium

CANDLE (constant axial shape of neutron flux, nuclide densities and power shape during life of energy producing reactor) burnup strategy is applied to small (30 MWth) block-type high temperature gas-cooled reactors (HTGRs) with thorium fuel. The CANDLE burnup is adopted in this study since it has several promising merits such as simple and safe reactor operation, and the ease of designing a long life reactor core. Burnup performances of thorium fuel (²³³U, ²³²Th)O₂ are investigated for a range of enrichment ⩽15%. Discharged fuel burnup and burning region motion velocity are major parameters of its performances in this study. The reactors with thorium fuel show a better burnup performance in terms of higher discharged fuel burnup and slower burning region motion velocity (longer core lifetime) compared to the reactors with uranium fuel.     

A rationale for fusion economies based on inherent safety

A comparison of the direct capital costs of recent light-water-reactor fission plants and recent magnetic fusion designs indicates that cost reductions by innovation in both the fusion reactor and the fusion balance of the plant (nonreactor part) will be required for fusion direct costs to become competitive. Both future fusion and future fission designs would benefit from modularity, standardization, and inherent safety (passive afterheat removal without damage in loss-of-coolant and loss-of-flow accidents) to improve their utility and public acceptance and to reduce nonreactor balance-of-plant costs, indirect costs, and time-related costs. If this can be achieved, inherently safe fusion and fission designs should have comparable reactor enclosure power densities, of the order of 1 MW_th/m³. With a reduction by a factor of about 2 in the fusion magnet and heating-system cost per watt, such fusion and fission designs should also have comparable direct and total capital costs per watt, provided the fusion plants can be modularly built in somewhat larger units (300 to 600 MW_e) than some of the modular fission units (100 MW_e). With the eventual fusion capital cost per watt thus attaining parity with fission some decades later than the development and deployment of second-generation fission plants, the economic incentive for switching growth to new fusion electric plants would derive from lower fuel-cycle costs (fuel startup, operating, processing, and waste-disposal costs). At the same time, there would also be strong economic incentives to build hybrid fusion plants to supply fissile fuel for the established fission plants,Disclaimer: This document was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor the University of California nor any of their employees, makes any warranty, express or implied, or assumes any legal liability or responsibility for the accuracy, completeness, or usefulness of any information, apparatus, product, or process disclosed, or represents that its use would not infringe privately owned rights. Reference herein to any specific commercial products, process, or service by trade name, trademark, manufacturer, or otherwise, does not necessarily constitute or imply its endorsement, recommendation, or favoring by the United States Government or the University of California. The views and opinions of authors expressed herein do not necessarily state or reflect those of the United States Government or the University of California, and shall not be used for advertising or product endorsement purposes.     

Neutronic analysis of PROMETHEUS reactor fueled with various compounds of thorium and uranium

In this study, neutronic performance of the DT driven blanket in the PROMETHEUS-H (heavy ion) fueled with different fuels, namely, ThO₂, ThC, UO₂, UC, U₃Si₂ and UN is investigated. Helium is used as coolant, and SiC is used as cladding material to prevent fission products from contaminating coolant and direct contact fuel with coolant in the blanket. Calculations of neutronic data per DT fusion neutron are performed by using SCALE 4.3 Code. M (energy multiplication factor) changes from 1.480 to 2.097 depending on the fuel types in the blanket under resonance-effect. M reaches the highest value in the blanket fueled with UN. Therefore, the investigated reactor can produce substantial electricity in situ. UN has the highest value of ²³⁹Pu breeding capability among the uranium fuels whereas UO₂ has the lowest one. ²³⁹Pu production ratio changes from 0.119 to 0.169 according to the uranium fuel types, and ²³³U production values are 0.125 and 0.140 in the blanket fueled with ThO₂ and ThC under resonance-effect, respectively. Heat production per MW (D,T) fusion neutron load varies from 1.30 to 7.89 W/cm³ in the first row of fissile fuel breeding zone depending on the fuel types. Heat production attains the maximum value in the blanket fueled with UN. Values of TBR (tritium breeding ratio) being one of the most important parameters in a fusion reactor are greater than 1.05 for all type of fuels so that tritium self-sufficiency is maintained for DT fusion driver. Values of peak-to-average fission power density ratio, Γ, are in the range of 1.390 and ∼1.476 depending on the fuel types in the blanket. Values of neutron leakage out of the blanket for all fuels are quite low due to SiC reflector. The maximum neutron leakage is only ∼0.025. Consequently, for all cases, the investigated reactor has high neutronic performance and can produce substantial electricity in situ, fissile fuel and tritium required for (D,T) fusion reaction.     

Numerical Study on Effects of Fuel Mixture Fraction and Li-6 Enrichment on Neutronic Parameters of a Fusion–Fission Hybrid Reactor

This study presents the effects of mixture fractions of nuclear fuels (mixture of fissile–fertile fuels and mixture of two different fertile fuels) and ⁶Li enrichment on the neutronic parameters (the tritium breeding ratio, TBR, the fission rate, FR, the energy multiplication ratio, M, the fissile breeding rate, FBR, the neutron leakage out of blanket, L, and the peak-to-average fission power density ratio, Γ) of a deuterium–tritium (D–T) fusion neutron-driven hybrid blanket. Three different fertile fuels (²³²Th, ²³⁸U and ²⁴⁴Cm), and one fissile fuel (²³⁵U) were selected as the nuclear fuel. Two different coolants (pressurized helium and natural lithium) were used for the nuclear heat transfer out of the fuel zone (FZ). The Boltzmann transport equation was solved numerically for obtaining the neutronic parameters with the help of the neutron transport code XSDRNPM/SCALE4.4a. In addition, these calculations were performed by also using the MCNP4B code. The sub-limits of the mixture fractions and ⁶Li enrichment were determined for the tritium self-sufficiency. The considered hybrid reactor can be operated in a self-sufficiency mode in the cases with the fuel mixtures mixed with a fraction of equal to or greater than these sub-limits. Furthermore, the numerical results show that the fissile fuel breeding and fission potentials of the blankets with the helium coolant are higher than with the lithium coolant.     

“CANDLE” burnup regime after LWR regime

CANDLE (Constant Axial shape of Neutron flux, nuclide densities and power shape During Life of Energy producing reactor) burnup strategy can derive many merits. From safety point of view, the change of excess reactivity along burnup is theoretically zero, and the core characteristics, such as power feedback coefficients and power peaking factor, are not changed along burnup. Application of this burnup strategy to neutron rich fast reactors makes excellent performances. Only natural or depleted uranium is required for the replacing fuels. About 40% of natural or depleted uranium undergoes fission without the conventional reprocessing and enrichment.

If the LWR produced energy of X Joules, the CANDLE reactor can produce about 50X Joules from the depleted uranium left at the enrichment facility for the LWR fuel. If we can say LWRs have produced energy sufficient for full 20 years, we can produce the energy for 1000 years by using the CANDLE reactors with depleted uranium. We need not mine any uranium ore, and do not need reprocessing facility. The burnup of spent fuel becomes 10 times. Therefore, the spent fuel amount per produced energy is also reduced to one-tenth.

The details of the scenario of CANDLE burnup regime after LWR regime will be presented at the symposium.