Thorium Fuel Research Articles

Improving the reactor safety aspects by the implementation of (Th-U233-Pu) fuel in a PWR assembly

The aim of this work is not only to demonstrate how to introduce thorium as an alternative nuclear fuel into pressurized water reactors, but also to examine the several advantages that (Th, U233, Pu) has over (Th, U233) and (Th, Pu) fuels when (U233O2) is mixed with (Th-Pu)O2 fuel. To investigate these advantages, three thorium fuels, (Th, U233), (Th-U233-Pu) and (Th-Pu) are simulated and analyzed by the MCNPX 2.7 code. The reactor safety aspects of these fuels are compared with those of the reference UO2 fuel. The concentration of Th-233 and fissile isotopes is adjusted to obtain criticality (Keff = 1.03) at the same burnup value (35 GWd/tonne) for all fuel types. Based on the predicted results, the author revealed that: 1) 56 % of plutonium (238-242) is depleted in the case of (Th-U233-Pu), while 41 % of plutonium is depleted in the (Th-Pu) at 60 GWd/tonne. This result indicates that (Th-U233-Pu) is more proliferation-resistant than (Th-Pu). 2) In terms of improving the quantity of radioactive waste, (Th, U233, Pu) model produces 0.6 and 0.9 times the amount of Am (241-243) and Cm (242-244) in UO2, respectively. On the other hand, (Th, Pu) fuel produces about 7 times the amount of Am and 8 times the amount of Cm in UO2 fuel. 3) The fuel temperature coefficient (FTC) and moderator temperature coefficient (MTC) of (Th-U233-Pu) are more negative than that of UO2 and (Th-U233) at the early stages of fuel burnup (0–5) GWd/tonne.4) the total mass of fissile isotopes in (Th-U233-Pu) is greater than that of UO2 and (Th-U233) and as a result, the fuel life time is longer in the case of (Th-U233-Pu). 4) The conversion ratio provided by (Th-U233-Pu) is greater than that provided by UO2 and (Th-U233).This behavior is expected owing to the hardening of neutron spectrum in (Th-U233-Pu). 5) The peaking factors are lower in the case of (Th-U233-Pu) compared to (Th-Pu). 6) The concentration of xenon and samarium is noticeably lower for (Th-U233-Pu) in comparison with (Th-Pu) and consequently, the time required for restarting the reactor is accelerated.

Neutronic Analysis for The Radial Direction Heterogeneous Core Configuration of GFR with Thorium Fuel

A neutronic analysis has been carried out for the radial direction heterogeneous core configuration in GFR with thorium nitride fuel to obtain the most optimal heterogeneous design has keff > 1 and excess reactivity below one percent. Neutronic analysis was performed using openMC version 0.13.2 based on open-source Monte Carlo simulation with ENDF/B-VII.1 data library. The first calculation is benchmarking openMC and SRAC where the largest error is 2.49% is caused by different data library. SRAC data library is JENDL 4.0. The second calculation is a homogeneous core configuration with 5% to 15% uranium-U233 fuel variation, the stability lies at 8% for the reference of heterogeneous core configuration. The heterogeneous core configuration was performed with three types of fuel percentage and five different geometry variations. Optimum design for geometry F1:F2:F3 = 1 ring:4 rings:1 ring using fuel type F1:F2:F3 = 6.5%:8%:9.5%. Optimum design for geometry F1:F2:F3 = 2 rings:2 rings:2 rings using fuel type F1:F2:F3 = 7.5%:8%:8.5%. Optimum design for geometry F1:F2:F3= 1 ring:2 rings:3 rings using fuel type F1:F2:F3=7%:8%:9%. Optimum design for geometry F1:F2:F3= 1 ring:3 rings:2 rings using fuel type F1:F2:F3 = 7.5%:8%:8.5%. Optimum design for geometry F1:F2:F3 = 1 ring:1 ring: 4 rings using fuel type F1:F2:F3 = 7.5%:8%:8.5%. The most optimum design among the other five designs is the F1:F2:F3 = 2 rings:2 rings:2 rings geometry design because it has a maximum keff of 1.00999 and excess reactivity of 0.99%. This design is carried out for an extended burn-up 15 years and still stable with the characteristics of the neutron flux and fission rate in this design decrease with burn-up time. Fission product in this design also decreased according to the purpose of the generation IV reactor is the prevention of nuclear weapons.

Core-Optimization of Ritm-200 Reactor Fuel Considering Thorium Fuel Cycle

Core-Optimization of Ritm-200 Reactor Fuel Considering Thorium Fuel Cycle

Study of Uranium and Thorium Fuels in Breed-and-Burn Mode

The formation of fissile nuclei through breeding conversion is a hot topic in academic research, as it provides a continuous source of nuclear fuel for nuclear reactors. Fast neutron reactors, which have been extensively studied, use natural uranium or low-enriched uranium as the nuclear fuel, achieving burning after uranium-plutonium conversion. Thorium, as another potential fissile fuel, can theoretically be converted into nuclear reactor fuel through the thorium-uranium cycle. In this study, the physical evolution process of nuclear fuel in a specific core parameter is simulated using the Monte Carlo program, and the performance differences of the thorium-uranium cycle and uranium-plutonium cycle in achieving in situ breeding-burning (Breed-and-Burn, B&B) mode are analyzed using neutron balance analysis method. The optimal ratio conditions for achieving self-sustained B&B burning in a thorium-uranium fuel mixed core are investigated. The study shows that for traditional solid-state nuclear reactor cores with a lower fuel proportion, thorium-breeding fuel has poor neutron economy compared to uranium-based breeding fuel, making it more difficult to achieve the B&B mode.

Neutronic analysis of ABR-1000 sodium cooled fast reactor with radially and axially heterogeneous core configurations

This study presents two proposed models for an ABR-1000 SFR metallic fuel core utilizing different combinations of Thorium, Uranium, and Plutonium fuels. The axial and radial models were evaluated using the OpenMC, open-source Monte Carlo simulation tool. The results of the depletion analysis revealed that ThUPu and UPu models demonstrated a higher cycle length than the ThU model, with the ThUPu(A) model being the most favorable option regarding safety and environmental sustainability. Furthermore, only ThUPu and UPu exhibit fissile inventory increment as time progresses. The simulation of the axial and radial models also demonstrated a higher neutron flux for all axial models and negative reactivity coefficients for all models. UPu(A) exhibits the highest beta effective value, followed by ThUPu(A) then by ThU(A). It is concluded that the axial model is the most favorable option, with the ThUPu(A) model being the optimal choice among the six proposed models.

Radiation and Nuclear Physics Aspects of the Use of the Thorium Fuel Cycle in a Hybrid Fusion Facility

This paper presents the results of the experimental determination and computational simulation of the ambient dose equivalent rate for a metallic thorium cylindrical miniblock and the (n,2n), (n,f), and (n,γ) reaction rates in a thin 232Th metal foil irradiated with neutrons of the NG-24M generator spectrum. The ambient dose equivalent rate was determined by dosimeters-radiometers. The reaction rates were determined by the activation method using Ge spectrometers without destroying the irradiated samples. Computational simulations of ambient dose equivalent and reaction rates were performed, respectively, using the radiation transport codes PHITS, MCNP5, and KIR2, which use various nuclear data libraries: JEFF-3.2 and -3.3; JENDL4.0; ENDF/B-VII.0, -VII.1, and -VIII.0; ROSFOND; FENDL; and TENDL. The authors give an estimate of the 232U/233U relative accumulation upon natural thorium irradiation in a fusion facility blanket with defined neutron spectrum. The nonirradiated and irradiated thorium nuclide composition change simulation and visualization were performed using analytical solutions of an ordinary system of homogeneous linear differential equations describing nuclide transmutations.

Comparative Assessment of the Range of Spectral Regulation of Reactivity Margin for Fuel Burnup in Pressurized Water Reactors Using Zirconium Displacers for Uranium and Thorium Fuel Cycles

Comparative Assessment of the Range of Spectral Regulation of Reactivity Margin for Fuel Burnup in Pressurized Water Reactors Using Zirconium Displacers for Uranium and Thorium Fuel Cycles

Review of Methods and Results for Reactor Physics Analysis of Thorium-Based Fuels From Irradiation Experiments Conducted in the National Research Universal Reactor

Abstract The development of advanced fuels and fuel cycles for conventional Generation III+ reactors, advanced reactors (such as Generation IV), and small modular reactors will help ensure the long-term sustainability and safety of the use of nuclear energy. Thorium-based fuels are an example of an advanced fuel that could augment and extend uranium resources. As part of previous efforts in Canada to investigate the use of thorium-based fuels for potential use in pressure-tube heavy-water reactors and other technologies, irradiation experiments with various thorium-containing fuels had been conducted in the National Research Universal (NRU) reactor at the Chalk River Laboratories. The NRU reactor physics computational tools and methods, developed initially for conventional enriched uranium fuel, had also been extended for analysis of thorium-based fuel experiments, with most satisfactory results. In this study, a number of physics and operation factors have been analyzed to quantify their effects on the production and burnup of 233U in thorium-based fuel, as well as to improve both the analysis and the use of thorium fuels and fuel cycles.

Neutronic Analysis of the New Proposed Configuration of Carem-25 Advanced Small Modular Reactor with Enriched Thorium Fuel

Neutronic Analysis of the New Proposed Configuration of Carem-25 Advanced Small Modular Reactor with Enriched Thorium Fuel

Fissile utilization of uranium, thorium, and plutonium fuels for a 60 MWt block‐type high‐temperature gas‐cooled reactor

High-temperature gas reactor (HTGR) is one of the Generation IV nuclear reactors currently undergoing rapid development in the world. The utilization of fissile material becomes an important parameter in HTGR technology because of the use of layered particle-type fuel that makes it challenging to reprocess. This research aims to compare and optimize the fissile utilization of HTGR 60 MWt using UO2, (Th-U)O2, and (Pu-U)O2. The utilization of fissile material on HTGR is being studied in this research by using the SRAC (Standard Reactor Analysis Code) version 2006 and the Japanese Evaluated Nuclear Data Library (JENDL) version 4.0. Parametric surveys are performed for important neutronic parameters, including analyzing the effects of fuel enrichment and the coated fuel particle (CFP) amount. Fissile materials are analyzed on a 1% to 20% enrichment level, and the CFP number is varied from 4470 to 26 830. The neutronic parameters taken into account are k-inf, conversion ratio (CR), the density of fissile and fertile material, and neutron energy spectrum. The results suggest that 11% of fuel enrichment and CFP number between 8950 and 11 180 are optimal for 60MWt HTGR. The highest fuel utilization level at the end of the reactor period was achieved by Pu-based fuel, with 81.6% of utilized fissile material.

Neutronic Performance of the VVER-1000 Reactor Using Thorium Fuel with ENDF Library

In this paper, neutronic calculations and the core analysis of the VVER-1000 reactor were performed using MCNP6 code together with both ENDF/B-VII.1 and ENDF/B-VIII libraries. The effect of thorium introduction on the neutronic parameters of the VVER-1000 reactor was discussed. The reference core was initially filled with enriched uranium oxide fuel and then fueled with uranium-thorium fuel. The calculations determine the delayed neutron fraction βeff, the temperature reactivity coefficients, the fuel consumption, and the production of the transuranic elements during reactor operation. βeff and the Doppler coefficient (DC) are found to be in agreement with the design values. It is found that the core loaded with uranium and thorium has lower delayed neutron fraction than the uranium oxide core. The moderator temperature coefficients of the uranium-thorium core are found to be higher than those of the uranium core. Results indicated that thorium has lower production of minor actinides (MAs) and transuranic elements (mainly plutonium isotopes) compared with the relatively large amounts produced from the uranium-based fuel UO2.

Comparative Study of Spectral Regulation Range of Excess Reactivity Control in Pressurized Water Reactors Using Zirconium Displacers for Uranium and Thorium Fuel Cycles

Comparative Study of Spectral Regulation Range of Excess Reactivity Control in Pressurized Water Reactors Using Zirconium Displacers for Uranium and Thorium Fuel Cycles

Transitioning to a Sustainable Thorium Fuel Cycle in Pressurized Water Reactors Using Bimetallic Thorium-Zirconium Alloy Cladding

Although η, the number of neutrons released per neutron absorbed in a 232Th-233U (thorium) fuel cycle, is greater than 2 in the thermal spectrum and therefore the possibility of breeding in a water-moderated reactor exists, it has been found difficult to achieve in practice. It is useful to relax the constraint for breeding and examine a thorium cycle for pressurized water reactors PWRs, denoted as PWR-Th, with the provision that the shortfall be made up by 233U bred in a PWR operating on a uranium fuel cycle, denoted as PWR-U, both of which utilize bimetallic thorium-zirconium alloy cladding as part of the fuel rod design. The number of complementary PWRs that could be sustainably operated on a thorium cycle was seen to critically depend on the moderator-to-fuel ratio (MF). Detailed cycle-by-cycle analysis shows that at the end of the first cycle, the sustainability ratio, namely, the ratio of sustainable PWR-Th reactors to PWR-U reactors, is 1.07 at an MF of 1.91, 1.4 at an MF of 1.43, and 4.45 at an MF of 0.954. The shortfall in 233U was found to decrease continually in subsequent cycles with the sustainability ratio increasing to 1.45, 2.01, and 28.3 at the respective MF values of 1.91, 1.43, and 0.954 by the 25th cycle. Although the sustainability ratio increases with lower MF, the achievable discharge exposure decreases necessitating larger material throughput in reprocessing. Detailed evaluations for fuel thermal, mechanical performance and nuclear reactivity feedback parameters require a further narrowing of potential design parameters based on holistic considerations arising from reprocessing. The PWR-Th reactors generate only trace amounts of transuranic (TRU) waste, and combined with a PWR-U design with bimetallic thorium cladding that generates only a fourth of the TRU waste compared to the standard all-UO2 fuel cycle, a significant reduction in TRU waste is possible.

The Effects of the Number of Coated Fuel Particles on the Neutronic Aspects of 25 MWt Pebble Bed Reactor with Thorium Fuel

High-Temperature Gas Reactor (HTGR) is a type of reactor that continues to be developed because of its advantages in terms of economic aspects, proliferation resistance, and safety aspects. One of the safety aspect improvements is due to the use of the Coated Fuel Particle (CFP). A coated fuel particle is a fuel with a diameter smaller than 1 mm and is protected by several carbon layers. In the Pebble Bed Reactor (PBR) type of HTGR design, the CFP is placed in a 6 cm fuel ball. How much CFP is put into the fuel ball will determine the neutronic characteristics of the reactor. In this study, the effect of the amount of CFP in the fuel ball on the 25 MWt PBR design using Thorium fuel and its impact on several important neutronic aspects, such as the effective multiplication factor, the amount of fuel enrichment, the utilization of fissile material, and the density of the fissile material formed. The calculation was performed by the Monte Carlo MVP / MVP-BURN code. This study found that the coated fuel particle fraction of 15% was the optimum value for the studied neutronic parameters.

Experimental measurements to evaluate the effects of fuel microstructure on fission gas release in thoria

Canadian Nuclear Laboratories (CNL) has many decades of thoria research and development experience, including fabrication, irradiation testing, and post–irradiation examination. CNL has previously shown fission gas release (FGR) to be dependent on fuel microstructure in thoria fuels. Lower density fuel with large granules surrounded by porous regions results in higher FGR. Higher and more uniform density fuel results in superior FGR results in thoria fuels with a similar irradiation history. In this paper, this effect of fuel microstructure on FGR is shown empirically using the most recent post–irradiation examination (PIE) results of CNL’s DME-176 and DME-221 thoria experiments. Previously reported PIE results of CNL’s BDL-422 and DME-221 thoria experiments are also compared.

Optimal Monazite Concentration Processes for the Extraction of Uranium and Thorium Fuel Material

The optimal conditions for the nitric acid dissolution of precipitates of hydroxides and hydrated oxides of rare-earth elements, uranium, and thorium obtained after autoclave alkaline opening of samples of monazite concentrate have been determined. The distribution of radioactive impurities between the solid phase and the solution in the processes of alkaline opening, dephosphorization, and acid dissolution of the pulp was studied. Two options are proposed for the extraction of uranium and thorium in the presence of rare earth elements, followed by separation of the components using tributyl phosphate of various contents in the carbon diluent.

Three-dimensional (X-Y-Z) Core Design of Long-Life Pressurized Water Reactor Using (Th-U)O2 Fuels with The Addition of Gd2O3 and Pa-231 as Burnable Poisons

Pressurized water reactors (PWRs) are one of the most dominant types of nuclear power plants that have been operated commercially to produce electricity in the world. The purpose of this study was to perceive a three-dimensional (X-Y-Z) core design of long-life PWR using Thorium-Uranium dioxide ((Th-U)O2) fuels with the addition of Gadolinium (Gd2O3) and Protactinium-231 (Pa-231) as the burnable poisons. A combination of Thorium and enriched Uranium fuels have a higher conversion ratio than other fuels, therefore can guarantee the reactor to operate longer. The burnable poison isotopes could be used to reduce excess reactivity due to the very high thermal neutron absorption cross-section. For core geometry analysis, a three-dimensional (X-Y-Z) geometry and a fuel volume fraction of 40% were applied. The computer code of SRAC 2006 from the Japan Atomic Energy Agency (JAEA) and the JENDL 4.0 as a nuclear data library were used for calculation. In this study, different fractions of Uranium dioxide, Uranium-235, Gadolinium, and Protactinium-231 in fuel were carried out. The result of this study was a three-dimensional core design of 800 MWt PWR using 60% Uranium dioxide fuel with enriched Uranium-235 of 12%-11% and the addition of 0,025% Gd2O3 and 1,0% Pa-231 which could operate for ten years without refueling. This research is expected to be a reference for long-life PWR design using the Thorium and Uranium fuel cycles.

Analysis of Mo-99 production as function of CAMOLYP reactor power

CAMOLYP or critical assembly for molly production is a nuclear reactor based on mixed uranium thorium fuels, which is currently being developed in BATAN Yogyakarta. This study aims to know how much the amount of 99Mo isotope production can be achieved on 6 days reactor operation as function of different power levels. The method is a calculation of reactor physics simulation by using MCNPX computer code. The calculation is done for the reactor core configuration of 23 thorium dioxide fuel rods in the outer ring with the total mass of 73.12 kg and 1.706 kg of 235U in the annulus. The analysis result shows that the CAMOLYP’s 99Mo production can be formulated as linear function of the reactor power level in watt. From the analysis result, it can be concluded that the CAMOLYP reactor can produce 39.91 curie 99Mo at 1000 watts and 6 days operations. This analysis result indicates that the CAMOLYP reactor as a low power reactor has succeeded in producing 99Mo isotope which is sufficient for domestic needs.

DIFFERENCES IN EXIT BURNUP, DEPLETION, AND FUEL REACTIVITY CALCULATIONS ARISING FROM THE CHOICE OF MODELLING CODE FOR BUNDLES WITH THORIUM AND URANIUM

A computational benchmark, using the deterministic codes WIMS-AECL and WOBI, and the stochastic code SERPENT, is made for burnup calculations of advanced thorium fuels in heavy water moderated reactors. Exit burnups and the concentration of the longer-lived actinides from the deterministic code set of WIMS-AECL and WOBI, which are 2-D, were compared to those from a full 3-D calculation in SERPENT. Results for reactivity vs. time are in general agreement to within a few mk (<1% in overall neutron multiplication) and appear to be systematic. Results for exit burnup were larger, in the 3%–6% range, because small reactivity effects can be amplified here.

Study on the Characteristics of Effective Delayed Neutron Fraction (βeff) for Pebble-Bed Reactor with Plutonium Fuel

A current effort focused on mechanisms of maximizing plutonium consumption is one of the solutions to control large stockpile of plutonium produced from nuclear power plants and military programs. The pebble-bed reactor with plutonium fuel is a unique case in the high-temperature reactors intended to reduce the concentration of plutonium that has been accumulated in the world. However, the plutonium core with tendency to produce a lower effective delayed neutron fraction (βeff), which is a crucial parameter in reactor safety, should be examined. The purpose of this study is to investigate the characteristics of effective delayed neutron fraction (βeff) for pebble-bed reactor with plutonium fuel. The uranium and thorium contents with a 1:1 and 1:2 ratio of heavy metal mass per pebble were combined into the plutonium core to study their effect on the characteristics of βeff. A series of calculations with various mass of plutonium per pebble were performed by using the MCNP6 code and ENDF/B-VII library. The βeff of pebble-bed reactor with plutonium fuel and its combination with uranium and thorium fuels were analyzed in term of core safety aspect because it could affect the transient behavior into a fatal accident situation.