Utilization of TRISO fuel with reactor grade plutonium in CANDU reactors

Large quantities of plutonium have been accumulated in the nuclear waste of civilian LWRs and CANDU reactors. Reactor grade plutonium and heavy water moderator can give a good combination with respect to neutron economy. On the other hand, TRISO type fuel can withstand very high fuel burn-up levels. The paper investigates the prospects of utilization of TRISO fuel made of reactor grade plutonium in CANDU reactors. TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The fuel compacts conform to the dimensions of CANDU fuel compacts are inserted in rods with zircolay cladding.In the first phase of investigations, five new mixed fuel have been selected for CANDU reactors composed of 4% RG-PuO₂ + 96% ThO₂; 6% RG-PuO₂ + 94% ThO₂; 10% RG-PuO₂ + 90% ThO₂; 20% RG-PuO₂ + 80% ThO₂; 30% RG-PuO₂ + 70% ThO₂. Initial reactor criticality (k_∞,0 values) for the modes , , , and are calculated as 1.4294, 1.5035, 1.5678, 1.6249, and 1.6535, respectively. Corresponding operation lifetimes are ∼0.65, 1.1, 1.9, 3.5, and 4.8 years and with burn ups of ∼30 000, 60 000, 100 000, 200 000 and 290 000 MW d/tonne, respectively. The higher initial plutonium charge is the higher burn ups can be achieved.In the second phase, a graphical-numerical power flattening procedure has been applied with radially variable mixed fuel composition in the fuel bundle. Mixed fuel fractions leading to quasi-constant power production are found in the 1st, 2nd, 3rd and 4th row to be as 100% PuO₂, 80/20% PuO₂/ThO₂, 60/40% PuO₂/ThO₂, and 40/60% PuO₂/ThO₂, respectively. Higher plutonium amount in the flattened case increases reactor operation lifetime to >8 years and the burn up to 580 000 MW d/tonne.Power flattening in the bundle leads to higher power plant factor and quasi-uniform fuel utilization, reduces thermal and material stresses, and avoids local thermal peaks. Extended burn-up grade implies drastic reduction of the nuclear waste material per unit energy output for final waste disposal.     

Thermalhydraulics studies examining the feasibility for introducing slightly enriched uranium fuel into the Embalse CANDU reactor

A joint study on the technical feasibility of using 0.9% slightly enriched uranium (SEU) fuel in the Embalse CANDU reactor was performed by Atomic Energy of Canada Limited (AECL) and Nucleoeléctrica Argentina S.A. (NASA). The feasibility study focused on the following technical areas: reactor physics and fuel management, fuel performance, and safety. Part of the safety assessment involved detailed thermalhydraulics analyses of three accident scenarios for a full core of SEU fuel bundles: (i) slow loss-of-reactivity control (LORC) event, (ii) large-break loss-of-coolant accident (LBLOCA) with emergency core cooling system (ECCS) available, and (iii) end-fitting failure. Other accident scenarios possibly encountered during the demonstration irradiation exercise or transition core have also been examined. It is concluded that introducing SEU fuel into the Embalse CANDU reactor is feasible. Clear advantages (e.g., fuel cost saving, increase in fuel exit burnup, and reduction in spent fuel volume) have been identified. The reduction in maximum bundle powers and the shift of the maximum bundle-power location to the inlet of the channel for the SEU fuel improve operating and safety margins. These margins are higher with the CANFLEX SEU fuel than the 37-element SEU fuel, due to lower linear powers and improved thermalhydraulic design.     

Increased fuel burn up in a CANDU thorium reactor using weapon grade plutonium

Weapon grade plutonium is used as a booster fissile fuel material in the form of mixed ThO₂/PuO₂ fuel in a Canada Deuterium Uranium (CANDU) fuel bundle in order to assure the initial criticality at startup.Two different fuel compositions have been used: (1) 97% thoria (ThO₂) + 3%PuO₂ and (2) 92% ThO₂ + 5% UO₂ + 3% PuO₂. The latter is used to denaturize the new ²³³U fuel with ²³⁸U. The temporal variation of the criticality k_∞ and the burn-up values of the reactor have been calculated by full power operation for a period of 20 years. The criticality starts by k_∞ = 1.48 for both fuel compositions. A sharp decrease of the criticality has been observed in the first year as a consequence of rapid plutonium burnout. The criticality becomes quasi constant after the second year and remains above k_∞ > 1.06 for 20 years. After the second year, the CANDU reactor begins to operate practically as a thorium burner.Very high burn up could be achieved with the same fuel material (up to 500,000 MW·D/T), provided that the fuel rod claddings would be replaced periodically (after every 50,000 or 100,000 MW·D/T). The reactor criticality will be sufficient until a great fraction of the thorium fuel is burnt up. This would reduce fuel fabrication costs and nuclear waste mass for final disposal per unit energy drastically.     

Dynamic analysis of a thorium fuel cycle in CANDU reactors

The thorium fuel recycle scenarios through a Canada deuterium uranium (CANDU) reactor have been analyzed for two types of thorium fuel: homogeneous ThO₂UO₂ and heterogeneous ThO₂UO₂–DUPIC fuels. The recycling was performed with dry process fuel technology which has a proliferation resistance. For the once-through fuel cycle model, the existing nuclear power plant construction plan was considered up to 2016, while the nuclear demand growth rate from the year 2016 was assumed to be 0%. After setting up the once-through fuel cycle model, a thorium fuelled CANDU reactor was modeled to investigate the fuel cycle parameters. In this analysis, the spent fuel inventory as well as the amount of plutonium, minor actinides and fission products for the multiple recycling fuel cycle were estimated and compared to those of a once-through fuel cycle.     

Conversion of high enriched uranium in thorium-232-based oxide fuel for light and heavy water reactors: MOX-T fuel

This paper discusses the possibility of using military high enriched uranium and plutonium in thorium oxide fuel for light and heavy water reactors (LWRs and HWRs). It is shown that such a fuel has several important advantages: (i) ²³⁹Pu and other long-living actinides are generated in quantities which are at least 100 times less than in conventional fuel; (ii) neutron emission is lower by a factor of more than 100; (iii) ²³³U is generated and burnt (the conversion factor for LWRs is 0.64–0.68 and for HWRs about 0.88); (iv) thorium is utilized and the total available amount of nuclear fuel is increased. The problem of non-proliferation of fissile material is also discussed and it is shown that the supervision of such fuel does not differ essentially from the supervision of low enriched uranium fuel with plutonium generation.     

Advanced mixed oxide fuel assemblies with higher plutonium contents for pressurized water reactors

Optimizing fuel cycle costs by increasing the final burnup leads to reduced generation of plutonium. Under properly defined boundary conditions thermal recycling in mixed oxide (MOX) fuel assemblies (FAs) reduces further the amount of plutonium which has to be disposed of in final storage. Increasing the final burnup requires higher initial enrichments of uranium fuel to be matched by an advanced design of MOX FAs with higher plutonium contents. The neutronic design of these MOX FAs has to consider the licensing status of nuclear power plants concerning the use of MOX fuel. The Siemens Nuclear Fuel Cycle Division, with more than 20 years' experience in the production of MOX fuel, has designed several advanced MOX FAs of different types (14 × 14 to 16 × 16) with fissile plutonium contents up to 4.60 w/o.     

Pyrochemical conversion of weapons plutonium into plutonium mononitride,monocarbide and mixed uranium-plutonium fuel for fast reactors

The main directions and results of research on pyrochemical reprocessing of weapons plutonium in fuel for fast reactors are presented. It is shown that this technology is economical and ecologically validated, compact, fire and explosion safe, especially for reprocessing in carbide-nitride as well as oxide fuel for fast reactors. It satisfies the principle of nonproliferation. For reprocessing weapons plutonium in oxide fuel with deep removal of ²⁴¹Am and Ga, a combined process which combines pyrochemical conversion of plutonium into oxide or nitride powder, and dissolution in acids and extraction of impurities. It is shown that the fuel kernels made from nitride, carbide, and oxide powers both from individual PuN, PuC_0.86, and PuO₂ powders as well as mixed plutonium compounds with uranium are fabricated by means of the conventional regime and provide the required density and content of gallium of <0.001 wt. %.     

Plutonium credit in the low enriched uranium fuel of PARR-1

The amount of plutonium (Pu) isotopes and the resultant savings of ²³⁵U due to their production were calculated in the low enriched uranium (LEU) fuel, being utilized in Pakistan Research Reactor-1 (PARR-1). Further the importance map and relative importance map for different isotopes of Pu were also determined. Equilibrium PARR-1 core was achieved for these calculations. MTR-PC26 package was used to generate the microscopic cross-sections data for 45 elements including fissile/structural materials and also the fission products. Finite difference reactor core analysis code CITATION was employed for the fuel management analysis and static depletion calculations.The results indicated that PARR-1 core has attained its equilibrium state after eleven cycles with each cycle of duration about forty full power (10 MW) days. Further, the results showed that at the beginning of equilibrium cycle (BOEC) of the PARR-1 core, net reactivity addition due to all isotopes of Pu was 4.86 × 10⁻³Δk/k. Amount of ²³⁵U equivalent to this value of reactivity was found to be 15.58 ± 0.021 g. Plots of importance and relative importance maps predicted higher isotopic concentrations of Pu in the fuel elements located in the vicinity of central water box.     

Analytical calculation of the fuel temperature reactivity coefficient for pebble bed and prismatic high temperature reactors for plutonium and uranium–thorium fuels

We analytically evaluated the fuel coefficient of temperature both for pebble bed and prismatic high temperature reactors when they utilize as fuel plutonium and minor actinides from light water reactors spent fuel or a mixture of 50% uranium, enriched 20% in ²³⁵U, and 50% thorium. In both cores the calculation involves the evaluation of the resonances integrals of the high absorbers fuel nuclides ²⁴⁰Pu, ²³⁸U and ²³²Th and it requires the esteem of the Dancoff–Ginsburg factor for a pebble bed or prismatic core. The Dancoff–Ginsburg factor represents the only discriminating parameter in the results for the two different reactors types; in fact, both the pebble bed and the prismatic reactors share the same the pseudo-cross-section describing an infinite medium made of graphite filled by TRISO particles. We considered only the resolved resonances with a statistical spin factor equal to one and we took into account 267, 72, 212 resonances in the range 1.057–5692, 6.674–14485, 21.78–3472 eV for ²⁴⁰Pu, ²³⁸U and ²³²Th, respectively, for investigating the influence on the fuel temperature reactivity coefficient of the variation of the TRISO kernel radius and TRISO particles packing fraction from 100, 200 to 300 μm and from 10% to 50%, respectively. Finally, in the pebble bed core, we varied the radius of the pebble for setting a fuel temperature reactivity coefficient similar to the one of a prismatic core.     

Overview of high-temperature fuel behaviour with relevance to CANDU fuel

This paper provides an overview of high-temperature phenomena in nuclear fuel elements and bundles, with particular relevance to the CANDU fuel design. The paper describes heat generation, fuel thermal response, and thermophysical properties of the fuel and sheath that can affect the thermal and mechanical response of the fuel element. Sources of chemical heat that can arise during accident conditions in the fuel element are also detailed. Specific phenomena associated with fuel restructuring, fuel sheath deformation, fuel-to-sheath heat transfer, fuel sheath failure criteria, oxidation, hydriding and embrittlement of the Zircaloy sheath, gap transport processes in failed elements, fuel/sheath interaction and fuel dissolution by molten cladding are detailed as important phenomena that can impact reactor safety analysis. Fuel behaviour during a power pulse and fuel bundle behaviour that occurs during a severe reactor accident are further considered. The review also points out areas of further research that are needed for a more complete understanding.     

The irradiation test of inert-matrix fuel in comparison to uranium plutonium mixed oxide fuel at the halden reactor

Conservative modelling for pin layout shows that the relatively low thermal conductivity of Inert-Matrix Fuel (IMF) causes higher temperatures and therefore higher fission gas release than in uranium plutonium mixed oxide (MOX). According to neutronic calculations, performance differences will also arise from different evolutions of the respective radial power and burnup distributions. Modelling of these effects as well as a 10% greater production of Xe in the thermal spectrum of the Halden reactor is well within the capabilities of appropriate codes. Some of the data and models used for the pre-calculations are preliminary and will be revised after the first experimental data have become available.     

Calculation of fuel burn up and radioactive inventory in a CANDU reactor using WIMS-D4 code

A study of fuel burn up and radioactive inventory for the CANDU reactor is carried out to validate the computer code WIMS-D4 for cluster geometry. The infinite and effective multiplication factors are calculated as a function of burn up and compared with the available results. A good agreement is observed among the present calculations and the previous published results. The code is then used to calculate the ²³⁵U and ²³⁸U consumed and the ²³⁹Pu produced in the fuel bundle. The inventories and the corresponding activities of some important fission products are also calculated.     

Pyrochemical method of salvaging weapons plutonium in oxide for fabricating mixed fuel for fast reactors

Experimental results of investigations of pyrochemical conversion of weapons plutonium into plutonium oxide for fabricating fast-reactor fuel are presented. Weapons plutonium was hydrogenized by hydrogen at 220°C, after which the plutonium hydride obtained was acidified at 550–560°C with the formation of PuO₂. To increase fire and explosion safety of the process, a mixture of oxygen with nitrogen, helium, or argon was used or nitriding with nitrogen and oxidation of plutonium mononitride were introduced. The particle size of the plutonium oxide powders obtained was less than 15 μm. The powders showed poor flowability, but after granulation they were suitable for fabricating kernels with mixed fuel. The gallium was removed by reduction of Ga₂O₃ by hydrogen to Ga₂O, which was sublimated. The mixed-fuel kernels sintered at 1600–1700°C in a hydrogen atmosphere contained <0.001 wt.% gallium, and their density was 94–97% of the theoretical value.