High Power Density Fusion Reactor Research Articles

Using recurrent neural networks for estimation of minor actinides’ transmutation in a high power density fusion reactor

In this paper, recurrent neural networks (RNNs) were presented for the computation of minor actinides’ transmutation with reactor’s operation period. The results of the RNNs implemented for the computation of the change in the atomic density of minor actinides ( 237Np, 241Am, 242Cm, 238Pu, 239Pu) and the results available in the literature obtained by using Scale 4.3 ( Übeyli, 2004) were compared. The results brought out that the proposed RNNs could provide an accurate computation of the atomic densities of minor actinides of the hybrid reactor with respect to operation period of reactor.

05/00654 Transmutation of minor actinides discharged from LMFBR spent fuel in a high power density fusion reactor

05/00654 Transmutation of minor actinides discharged from LMFBR spent fuel in a high power density fusion reactor

04/02772 LWR spent fuel transmutation in a high powerdensity fusion reactor: Sahin, S. and Übeyli, M. Annals of Nuclear Energy, 2004, 31, (8), 871–890

04/02772 LWR spent fuel transmutation in a high powerdensity fusion reactor: Sahin, S. and Übeyli, M. Annals of Nuclear Energy, 2004, 31, (8), 871–890

Transmutation of minor actinides discharged from LMFBR spent fuel in a high power density fusion reactor

Significant amounts of nuclear wastes consisting of plutonium, minor actinides and long lived fission products are produced during the operation of commercial nuclear power plants. Therefore, the destruction of these wastes is very important with respect to public health, environment and also the future of nuclear energy. In this study, transmutation of minor actinides (MAs) discharged from LMFBR spent fuel in a high power density fusion reactor has been investigated under a neutron wall load of 10 MW/m2 for an operation period of 10 years. Also, the effect of MA percentage on the transmutation has been examined. The fuel zone, containing MAs as spheres cladded with W-5Re, has been located behind the first wall to utilize the high neutron flux for transmutation effectively. Helium at 40 atm has been used as an energy carrier. At the end of the operation period, the total burning and transmutation are greater than the total buildups in all investigated cases, and very high burnups (420–470 GWd/tHM) are reached, depending on the MA content. The total transmutation rate values are 906 and 979 kg/GWth year at startup and decrease to 140 and 178 kg/GWth year at the end of the operation for fuel with 10% and 20% MA, respectively. Over an operation period of 10 years, the effective half lives decrease from 2.38, 2.21 and 3.08 years to 1.95, 1.80 and 2.59 years for 237Np, 241Am and 243Am, respectively. Total atomic densities decrease exponentially during the operation period. The reductions in the total atomic densities with respect to the initial ones are 79%, 81%, 82%, 83%, 85% and 86% for 10%, 12%, 14%, 16%, 18% and 20% MAs, respectively.

LWR spent fuel transmutation in a high power density fusion reactor

The prospect of light water reactor (LWR) spent fuel incineration in a high power density fusion reactor has been investigated. The neutron wall load is taken at 10 MW/m 2 and a refractory alloy (W-5Re) is used in the first wall. Neutron transport calculations are conducted over an operation period of 48 months 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 S 8-P 3 approximation. In the neutron rich environment, the tritium breeding ratio remains > 1.05 so that the tritium self-sufficiency is maintained for the fusion reactor. The presence of fissionable nuclear waste fuel in the investigated blanket causes significant energy amplification. The energy multiplication factor is ∼4 at startup and it increases steadily up to 5.55 during power plant operation so that even a modest fusion reactor can supply a significant quantity of electricity. In the course of nuclear waste incineration, most of the fissionable fuel is burnt in situ. In addition to that, excess fissile fuel production enhances the nuclear quality of the nuclear fuel. Starting with an initial cumulative fissile fuel enrichment (CFFE) value of the spent fuel of 2.172%, CFFE can reach 4% after an irradiation period of ∼12 months. Then the spent fuel becomes suitable for a new recharge in an LWR as a regenerated fuel. Further residence in the fusion blanket continues to upgrade the nuclear waste so that after 48 months, CFFE can reach such a high level (9%) that it becomes qualified to be used in a new type of the advanced high temperature reactors for the Generation-IV.

Isotopic Tailoring to Improve the Breeding Capability of the Sn-Li Eutectic in Liquid First Wall Fusion Blanket Concepts

Due to its low vapor pressure, the Sn-Li eutectic has been identified as a potential breeder for deployment in the liquid first wall (FW)/Blanket concepts under investigation in the APEX study for high power density fusion reactors. However, its breeding capability is limited. This drawback is rather improved due to the neutron multiplication via Sn(n,2n) reactions. However, the “local” tritium breeding ratio (TBR) was found to be still on the marginal side (even with 90%Li-6 enrichment). Aside from using a beryllium multiplier, other means to improve the capability of Sn-Li for breeding are explored. In this paper, we show that by isotopic tailoring of Tin in the Sn-Li breeder, one can achieve substantial improvement in TBR in addition to attaining significant reduction in the activation level in this material.

Pebble divertor concept for high power density fusion reactors

The pebble divertor is a new concept applicable to the strike zone with very high heat (>20 MW/m 2) and particle load (>10 24 atoms/m 2s). In this concept, a large number of small pebbles made from refractory materials with multifunctions are used as a moving surface. A marked feature of our pebble divertor concept is the use of multi-layer pebbles. The multi-layer pebble consists of a kernel, a plasma facing layer (PFL) and an intermediate tritium permeation barrier. In the case of the pebble with a diameter of 1–2 mm, it is found that the pebble can resist high heat loads of about 30 MW/m 2. When the PFL of the pebble is made from carbon material, pebbles also perform fuel gas pumping due to the hydrogen gettering characteristic of carbon. From the estimation of the energy conversion in the pebble divertor system, the maximum operation temperature should be larger than about 800 K to obtain a energy gain, which is the ratio of recovery energy in Calnot cycle to the additional heat energy, larger than one in the divertor system.

Experiments on helium pumping of pebble divertor

The pebble divertor concept was proposed for compact and high power density fusion reactors. In this concept, fuel gas and impurities including the helium ash might be pumped out by using gas gettering material at the surface layer of the pebbles. In this study, the helium pumping performance of the pebble divertor is experimentally studied. In particular, the transient release from graphite just after the high particle flux irradiation of 5×10 20 He/m 2s (He +, 2 keV) have been measured with keeping the sample temperature at the maximum temperature in the irradiation. Then the residual retained helium in the sample has been measured by heating up the sample at 873 K. The amount of transiently retained helium is not appreciably different with that of the residual retained helium. From this result in the helium pumping performance, the pebble temperature in the irradiation must be lower than 730 K to satisfy the requirement in steady state operation of A-SSTR.

Pumping performance of pebble drop divertor

The pebble drop divertor concept was proposed for compact and high power density fusion reactors. In this concept, fuel gas can be pumped out by using fuel gas gettering by the surface layer of the pebbles. We are interested in the pumping by using transient hydrogen release from graphite just after ion irradiation. In this study, the pumping performance of the pebble drop divertor by using the transient hydrogen release is experimentally studied. In the experiment, the graphite samples were irradiated by high flux hydrogen beam with a maximum flux of 1×10 22 H/m 2 s (H 3 +, 5 keV). Just after the irradiation, hydrogen release from the samples was measured with keeping the sample temperature at the temperature reached by the irradiation. The transient hydrogen release was observed in all cases. The amount of the transient hydrogen release increased with increasing temperature. The amount of maximum transient hydrogen release was 4.4×10 20 H/m 2 at the temperature of 973 K. From this result, the pumping performance of the pebble drop divertor will satisfy the specification of fuel gas feed needed in steady state operation of ITER.

The Reversed Field Pinch: Understanding and Improving Confinement

The reversed field pinch (RFP) is a toroidal, high beta plasma confinement configuration with great potential as an attractive, compact, high power density fusion reactor core. Relatively poor confinement has been a main obstacle in establishing the viability of the RFP. Recent progress in understanding magnetic-fluctuation-induced transport in the RFP has spawned the idea of current profile control to reduce fluctuations and transport. With the addition of inductive poloidal current drive in the Madison Symmetric Torus (MST) device, the energy confinement time is increased five-fold from 1.2 ms to 6 ms. The lowest magnetic fluctuation level and highest electron temperature observed in MST are also obtained with inductive current profile control. These results strongly encourage the development of improved and steady-state current profile control. 17 refs., 4 figs., 1 tab.

High heat removal and structural design for a quickly replaceable element for a high power density fusion reactor

A study of a new technology for the first wall design of a high power density fusion reactor has been performed by applying both a newly devised cooling technology, and a mechanical and structural design using the shape memory alloy (SMA) elements. The first wall is quickly replaceable by using a quickly replaceable-technology (QRT) to remove elements which have suffered severe radiation damage. This QRT will enable the design to employ an alternative cooling technology, liquid metal mist cooling (LMMC), to remove high thermal wall loads of up to 4 MW/m 2 . The LMMC uses a sprayed mist flow containing liquid mist droplets in its carrier gas, which can be operated under low pressure and free from any MHD effect in a strong magnetic field. This new technology is applied here to a cassette compact toroidal reactor (CCTR). Thermal analyses of temperature and thermal stress distributions in the first wall are discussed.

Application of shape memory alloys to compacting and element-quickly replaceable design in high-power density fusion reactors

Abstract Quick replacement has been accomplished by using a shape memory alloy (SMA) coupling in the conceptual design of a cassette compact toroid reactor (CCTR). Further, by using an SMA driving element, a compact large gate valve can be newly devised. This gate valve will enable in-situ handling without breaking vacuum, so that the baking of the vacuum boundary for every replacement becomes unnecessary, except the initial baking. In these applications of SMA, the compacting and quickly replaceable technology that are needed for dealing with the problems associated with very high neutron loading in a compact reactor become available by using presently available or promised materials rather than an assumed material.