ADS for HEU burning and 233U breeding in HEU-Thorium Oxide system
In this paper, a preliminary neutronic design study of an accelerator-driven subcritical system is presented. The conceptual design of coupling the reactor core “Mixed HEU-thorium oxide” with proton accelerator spallation target is investigated, and its feasibility as a neutron source for HEU burning, and isotopes production is evaluated. A 3D model of the JSA reactor, the accelerator beam, and the Pb target was developed based on actual reactor parameters.
MCNPX calculations were carried out to calculate the multiplication factor Keff and the production of233U and depletion of 235U. Introducing thorium fuel into the core with HEU offers an effective way to produce more 233U isotopes and to burn 235U isotopes more efficiently. In addition, production of less minor actinides (MA) and generation of energy can be achieved from this process.
In various countries, the accelerator-driven subcritical system, popularly known as the accelerator-driven system (ADS), is recently being studied with specific objectives.
In USA, Europe, and Japan, the main goals of ADS are incorporating inherent safety in nuclear energy systems and providing long-term solutions to nuclear waste disposal by burning plutonium and minor actinides and transmutation of long-lived fission products. Themain goal in India is to develop ADS for safe and efficient breeding of 233U from the abundant resources of thorium and provide sustainable nuclear energy security [1].
In addition, the ADS has heat removal and electricity generation equipment.
The claims of enhanced safety characteristics are based on the fact that the ADS operates in a non-self-sustained chain-reaction mode which reduces criticality concerns. The ADS is operated in a sub-critical state and stays sub-critical, regardless of the accelerator being on or off. Moreover, the accelerator may provide a convenient control mechanism for sub-critical systems that would reduce the need of control rods. The sub-criticality itself adds an additional level of operational safety with regard to criticality insertion accidents [5].
Accelerator-driven subcritical assembly system (ADS), with a family of energy amplifiers (EA), and thorium as the breeding fuel offers potentially significant advantages in thorium fuel cycle (‘once-through’ or ‘closed’) in terms of minimizing radiotoxicity of nuclear waste and ensuring proliferation-resistance at the same time.
Denaturing the thorium with the addition of 238U may somewhat mitigate risks of proliferation as the 238U lowers the effective enrichment in 233U of the reprocessed uranium (as the chemical reprocessing cannot separate the 238U from the 233U). Dilution (or denaturing) of fissile materials by their isotopes can complicate the use of fissile materials in nuclear explosive devices (NED)[6].
In the second stage of the India’s three-stage programme, Fast Breeder Reactors (FBRs) are fuelled by fuels based on Plutonium mixed with reprocessed Uranium recovered by reprocessing of the first stage spent fuel and Thorium-232 will be introduced as a blanket material to be converted to Uranium-233 [7].
In this paper, a conceptual design of ADS for burning high enriched uranium (HEU)in addition to power generation utilizing thoriumfor233U breeding will be applied. This can be considered as an advantageous option for countries that do not have spent nuclear fuel but have a huge amount of thorium.
The model of ADS used in this simulation depends on the typical fast neutron spectrum, lead-bismuth accelerator-driven transmutation system in Trellue research [8]. By bombarding 1GeV proton beam into the LBE cylindrical target, neutrons are produced from the spallation reaction.The cylindrical core of 140 cm radius is loaded with 180 fuel assemblies. The fuel rods are introduced in the oxide form and contained in hexagonal assemblies.
However, in this work, instead of loading the whole core with reprocessed fuel as in the original design, 96 seed assemblies of HEU and 84 blanket assemblies of thorium fuel were inserted into the individual regions. This heterogeneous approach will help to simplify the fuel assembly fabrication and in-core fuel management.
By spatially separating the fertile and fissile materials, ideally, there would be no competition for neutron absorption between them; thus, the capture rate in fertile fuel would be optimized, and 233U conversion ratio in the blanket could be enhanced. The breeding233U from thorium fuel will compensate the burnt EU in the HEU fuel, thus reducing the reactivity swing.
Figure (1) shows the vertical and horizontal sectional views of the seed-blanket ADS. In order to improve the neutron economy, most of the thorium blanket assemblies are placed at the periphery of the core. The rest of the 232Th assemblies are distributed inside the core along with the HEU seed assemblies around the ADS target.
Sodium is employed as a coolant in the system. In Table (1), the main design parameters of the ADS reactor are presented.
Figure (1) Vertical and horizontal sectional views of the seed-blanket ADS design (scale is given in cm) [5]
Table (1) Main design parameters of the ADS reactor
Radial power peaking factor at BOC
Radial power peaking factor at EOC 3.04
1.71
Cycle length (days) 1100
Keff
BOC 0.96044 ± 0.00070
EOC 0.89798 ± 0.00056
The initial content of the core (thorium and HEU fuels) was determined in order to achieve approximately initial Keff of 0.96. This was achieved by attempting different 235U enrichments and calculatingthe effective multiplication factor Keff in each case. The isotopic composition of the core is shown in Table (2).
Table (2) Fuel composition of HEU, reprocessed fuel and thorium
MCNPX is a general-purpose Monte Carlo radiation transport code designed to track particle types over broad ranges of energies. MCNPX program began in 1994 as an extension of MCNP4B and LAHET 2.8 in support of the Accelerator Production of Tritium Project (APT). The work envisioned a formal extension of MCNP to all particles and all energies; improvement of physics simulation models; extension of neutron, proton, and photonuclear libraries to 150 MeV [9].
The growth of computing power caused a sustainable increase of the share of Monte Carlo codes in nuclear reactor and nuclear criticality research and development. These Monte Carlo codes can provide the most accurate locally dependent neutronic characteristics in realistic 3D geometries of any complexity. Among them, only the general-purpose radiation transport code MCNPX [9, 10] is fully capable to treat ADS-related problems, since it tracks almost all particle types of nearly all energies.
In the upper energy region (above 20?MeV) it relies on the model calculations using different intra-nuclear cascade, pre-equilibrium and equilibrium model combinations. However, recent progress in nuclear data library extensions to higher energies (up to 200?MeV) allows implementing evaluated nuclear data in neutronic analysis of ADS systems [10].
The calculation of neutron multiplication factor for ADS was performed by MCNPX code using KCODE option. In an ADS, Keff must be between 0.95 and 0.98 indicating that ADS is a subcritical system [1]. ADS reactors, as a matter of fact, are sub-critical in all conditions and power levels.
The system was simulated with Th-HEU fuel assemblies loaded and the multiplication factor was calculated. The loaded core is shown in Figure (1) and the multiplication factor was adjusted to be 0.9604 at the beginning of cycle (BOC) where the chain reaction could sustain. The variation of Keff with EFPD is shown in figure (2).
Figure (2) Keff vs. EFPD in ADS system
The mass evolution is an important parameter to prove the HEU burning and 233U breeding in the seed and blanket ADS fuel. Figure (3) shows the element inventory in the 840 MWth ADS using thorium – HEU fuel. At the end of the cycle (EOC), a significant amount of EU is burnt. This can be seen from the figure as the amount of 235U was about 1.841E+03 kg at the BOC while it was about 9.638E+02 kg at the EOC, meaningabout 52.4% of the initial content of 235U was burnt.
Mean while a great amount of 233U was bred. From figure (3), it can be noticed that 5.700E+03 kg of thorium existed at the BOC; at the EOC, about 5.281E+03 kg of thorium, 2.638E+02 kg of 233U in addition to small fractions of other actinides such as 233Pa were present.This gives a ratio of about 4.63% of thorium converted into 233U. Important contribution accounted for239Pu (mainly due to the conversion of 238U).
Figure (3) isotopic inventory for ADS fuel
In this case, enriched 235U starter core is surrounded by a blanket of fertile fuel (thorium). Enriched fuel produces neutrons that generate power and convert fertile fuel to fissionable one. Figure (4) shows the depletion of 235U content due to the external neutrons from the ADS system and fission process.
The exposure to neutron flux gradually decreases 235U concentration. Each neutron absorbed in 235U decreases the fuel concentration and thus the macroscopic fission cross section, which causes the reactivity decrease.
Figure (4) Depletion of 235U in ADS core
On the other hand, the neutron capture by232Th forms 233U. The percentage change in 232Th concentration is not significant due to its high concentration in the ADS core. However, production of the small amount of 233U is extremely important for the core performance. Figure (5) shows the breeding rate of 233U from the BOC to the EOC in the ADS core.
Figure (5) Breeding of 233U in ADS core
ADS is capable of burning any type of fuel and the choice of thorium potentially provides low radiotoxicity, fuel diversity, and proliferation resistance.
An important target for using thorium fuel is to breed as much fissile isotope 233U as possible in addition to decrease the amount of 239Pu produced for safeguards concerns. The use of ADS with HEU and Thorium blanket fuel achieves both goals.
The ADS used in this work resulted in the conversion of about 4.63% of initial content of thorium isotopes into 233U. Moreover, the amount of 239Pu produced in the system (about 328kgafter one cycle) is much lower than that produced from the other reactors like the PWR (about 1700kg after one cycle [11]). ADS also shows a high capability for 235U isotope burning (about 52.4% of the 235Uinitial content was burnt during the first cycle).
Therefore, energy generation can be achievable by the use ofADS fueled withHEU and Thorium that also results in a significant fertile-to-fissile conversion as well as burning of HEU.
[1] “Thorium fuel cycle – Potential benefits and challenges,” IAEA-TECDOC-1450, May 2005.
[2] C. Rubbia, J.A. Rubio, S. Buono .et. al,”CERN-group conceptual design of a fast neutron operatedhigh power energy amplifier,” European Organization for Nuclear Research, CERN, Geneva, Switzerland, Sincrotrone Trieste, Trieste, Italy, Laboratoire du Cyclotron, Nice, France.
[3] Charles Bowman .et. al,”Accelerator-Driven Transmutation of High-Level Waste from the Defense and Commerical Sector,” Los Alamos National Laboratory.
[4] Mats Sk?lberg, et al., “PARTITIONING AND TRANSMUTATION (P&T),”technical report, Department of Nuclear Chemistry, Chalmers University of Technology, Department of Neutron and Reactor Physics Royal Institute of Technology Stockholm, 1995.
[5] “Accelerator driven systems: Energy generation and transmutation of nuclear waste,” IAEA-TECDOC-985, November 1997.
[6]Kryuchkov E.F., Tsvetkov P.V.1, et al., “Isotopic Uranium and Plutonium Denaturingas an Effective Method for Nuclear FuelProliferation Protection in Openand Closed Fuel Cycles,” National Research Nuclear University “MEPhI”,Texas A&M University Russia.
[7] S A BHARDWAJ, “Indian nuclear power programme – Past, present and future,” Nuclear Power Corporation of India Limited, Anushaktinagar, Mumbai, India.
[8] H. R. Trellue, “Reduction of the radiotoxicity of spent nuclear fuel using a two-tiered system comprising light water reactors and accelerator-driven systems [Ph.D. thesis],” University of New Mexico, 2003.
[9] D. B. Pelowitz, Ed., “MCNPX user’s manual, version 2.6.0,” Tech. Rep. LA-CP-07-1473, 2008.
[10] D. B. Pelowitz et al., “MCNPX 2.7.C extensions,” Tech. Rep. LA-UR-10-00481, Los Alamos National Laboratory, 2010.
[11] Thanh Mai Vu and Takanori Kitada, “Transmutation Strategy Using Thorium-Reprocessed Fuel ADS for Future Reactors in Vietnam,” Division of Sustainable Energy and Environmental Engineering, Osaka University, Japan, 2013.
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