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.     

Effect of the heat-treatment temperature of powders on oxide fuel quality

It has been determined at the State Scientific Center of the Russian Federation—Physics and Power Engineering Institute in the course of developing a technology for fabricating various fuel compositions (UO₂+MgO, UO₂+ThO₂, UO₂+Th+ThO₂, PuO₂+MgO, UO₂+Fe+MgO, PuO₂+BaO, and others) for fast-neutron and light-water reactors that structural changes in particle aggolmerates occur at the heat-treatment stage. The optimal properties of the powders are obtained at the temperature of the morphological transformations of the particles. The fuel pellets prepared from these powders possess stable density, porosity, exterior form, mechanical strength, and so on. The total specific surface area of the oxides is an indirect parameter for estimating their quality. Each fuel composition has its own optimal powder heat-temperature temperature. 7 figures, 1 table, 5 references. State Scientific Center of the Russian Federation—A. I. Leipunskii Physics and Power-Engineering Institute. Translated from Atomnaya énergiya, Vol. 86, No. 3, pp. 189–194, March, 1999.     

Neutron Economy of Transmutation of TRU in Thermal and Fast Neutron Fields

Neutron economy of the transmutation of TRU was examined in well thermalized, thermal and fast neutron fields. Burn-up chains of ²³⁷Np, ²⁴¹Am and ²⁴³Am, which are the main TRU nuclides in the high level waste, were calculated in the flux region from 10¹⁴ to 10¹⁷ n/cm².s. Numbers of neutrons absorbed and produced of each chain were calculated using JENDL-3. The net number of neutron produced n _net, which was obtained by the difference of the two numbers, largely varied with the neutron fields, the nuclides and the flux levels. The n _net value in the fast neutron field was positive (0.0–1.0) for ²³⁷Np, ²⁴¹Am, ²⁴³Am and TRU with the nuclide composition in the high-level waste generated by the conventional PWR. The transmutation of TRU by fission can be performed with producing neutrons in the fast neutron field. On the other hand, the n _net value was negative for the well thermalized and thermal neutron fields. For TRU in the high-level waste, the values in those fields were —1.0 at 10¹⁴ n/cm².s and 0.0 at 10¹⁶ n/cm².s. In the high flux region of 10¹⁶ n/cm².s, TRU in the high-level waste can be transmuted by fission without consuming additional neutrons. In the flux region about 10¹⁴ n/cm².s, the transmutation of TRU in the high-level waste by fission requires about one neutron.     

Fundamental Study of the Sulfide Reprocessing Process for Oxide Fuel (I) Study on the Pu,MA and FP Tracer-Doped U3O8

For the recovery of fuel materials from spent nuclear fuel, a novel reprocessing process based on the selective sulfurization of fission products (FP) has been proposed, where FP and minor actinides (MA) are first sulfurized by CS₂ gas, and then, dissolved by a dilute nitric acid solution. Consequently, the fuel elements are recovered as UO₂ and PuO₂. As a basic research of this new concept, the sulfurization and dissolution behaviors of U, Pu, Np, Am, Eu, Cs, and Sr were investigated by γ-ray and α spectrometries in this paper using ²³⁶Pu-, ²³⁷Np-, ²⁴¹Am-, ¹⁵²Eu-, ¹³⁷Cs-, and ⁸⁵Sr-doped U₃O₈ samples. The dependence of the dissolution ratio of each element on the sulfurization temperature was studied and reasonably explained by combining the information of the sulfide phase analysis and the chemical thermodynamics of the dissolution reaction. The sulfurization temperature ranging from 350 to 450°C seems to be promising for the separation of FP and MA from U and Pu, since a clear difference in the dissolution ratio between FP and U was derived by the sulfurization treatment in this temperature range.     

Rapid Transmutation of High-Level Nuclear Wastes in a Catalyzed Fusion-Driven System

The aim of this study is to investigate the high-level waste (HLW) transmutation potential of fusion-driven transmuter (FDT) based on catalyzed D–D fusion plasma for various fuel fractions. The Minor actinide (MA) (²³⁷Np, ²⁴¹Am, ²⁴³Am and ²⁴⁴Cm) and long-lived fission product (LLFP) (⁹⁹Tc, ¹²⁹I and ¹³⁵Cs) nuclides discharged from high burn-up pressured water reactor-mixed oxide spent fuel are considered as the HLW. The volume fractions of the MA and LLFP are raised from 10 to 20% stepped by 2% and 10 to 80% stepped by 5%, respectively. The transmutation analyses have been performed for an operation period (OP) of up to 6 years by 75% plant factor (η) under a first-wall neutron load (P) of 5 MW/m² by using two different computer codes, the XSDRNPM/SCALE4.4a neutron transport code and the MCNP4B Monte Carlo code. The numerical results bring out that the considered FDT has a high neutronic performance for an effective and rapid transmutation of MA and LLFP as well as the energy generation along the OP.     

Criticality investigations for the fixed bed nuclear reactor using thorium fuel mixed with plutonium or minor actinides

Prospective fuels for a new reactor type, the so called fixed bed nuclear reactor (FBNR) are investigated with respect to reactor criticality. These are ① low enriched uranium (LEU); ② weapon grade plutonium + ThO₂; ③ reactor grade plutonium + ThO₂; and ④ minor actinides in the spent fuel of light water reactors (LWRs) + ThO₂. Reactor grade plutonium and minor actinides are considered as highly radio-active and radio-toxic nuclear waste products so that one can expect that they will have negative fuel costs.The criticality calculations are conducted with SCALE5.1 using S₈–P₃ approximation in 238 neutron energy groups with 90 groups in thermal energy region. The study has shown that the reactor criticality has lower values with uranium fuel and increases passing to minor actinides, reactor grade plutonium and weapon grade plutonium.Using LEU, an enrichment grade of 9% has resulted with k_eff = 1.2744. Mixed fuel with weapon grade plutonium made of 20% PuO₂ + 80% ThO₂ yields k_eff = 1.2864. Whereas a mixed fuel with reactor grade plutonium made of 35% PuO₂ + 65% ThO₂ brings it to k_eff = 1.267. Even the very hazardous nuclear waste of LWRs, namely minor actinides turn out to be high quality nuclear fuel due to the excellent neutron economy of FBNR. A relatively high reactor criticality of k_eff = 1.2673 is achieved by 50% MAO₂ + 50% ThO₂.The hazardous actinide nuclear waste products can be transmuted and utilized as fuel in situ. A further output of the study is the possibility of using thorium as breeding material in combination with these new alternative fuels.     

First nuclear transmutation of 237Np and 241Am by accelerator-driven system at Kyoto University Critical Assembly

This study demonstrates, for the first time, the principle of nuclear transmutation of minor actinide (MA) by the accelerator-driven system (ADS) through the injection of high-energy neutrons into the subcritical core at the Kyoto University Critical Assembly. The main objective of the experiments is to confirm fission reactions of neptunium-237 (²³⁷Np) and americium-241 (²⁴¹Am), and capture reactions of ²³⁷Np. Subcritical irradiation of ²³⁷Np and ²⁴¹Am foils is conducted in a hard spectrum core with the use of the back-to-back fission chamber that obtains simultaneously two signals from specially installed test (²³⁷Np or ²⁴¹Am) and reference (uranium-235) foils. The first nuclear transmutation of ²³⁷Np and ²⁴¹Am by ADS soundly implemented by combining the subcritical core and the 100 MeV proton accelerator, and the use of a lead-bismuth target, is conclusively demonstrated through the experimental results of fission and capture reaction events.     

Thermal-fluid characteristics of the transuranics fuel in a deep-burn HTR core

The Deep Burn Project is evaluating the feasibility of the DB-HTR (Deep Burn High Temperature Reactor) to achieve a very high utilization of transuranics (TRU) derived from the recycle of LWR spent fuel. This study intends to evaluate the thermal-fluid and safety characteristics of TRU fuel in a DB-HTR core using GAMMA+ code. The key design characteristics of the DB-HTR core are more fuel rings (five fuel-rings), less central reflectors (three rings) and decay power curves due to the TRU fuel compositions that are different from the UO₂ fuel. This study considered three types of TRU kernel compositions such as 100%(PuO₂ + NpO₂ + Am), 99.8%(PuO_1.8, NpO₂) + 0.2%UO₂ + 0.6 mole SiC getter, and 70%(PuO_1.8, NpO₂) + 30%UO₂ + 0.6 mole SiC getter. The first fuel type of TRU kernel produces higher decay power than the UO₂ kernel. For the second and the third fuel types, removing the initial Am isotopes and reducing the volumetric packing fraction of TRISO particles will reduce the decay power. The flow distribution, core temperature and TRISO temperature profiles at the steady state were examined. As a safety performance, this study mainly evaluated the peak fuel temperature during LPCC (low pressure conduction cooling) event with considering the impact of decay power, the annealing effect of the irradiated thermal conductivity of graphite, and the impact of the FB (fuel block) end-flux-peaking. For the 600 MW_th DB-HTR core, the peak fuel temperature of 100%(PuO₂ + NpO₂ + Am) TRU was found to be much higher than the transient fuel design limit of 1600 °C due to the lack of heat absorber volume in the central reflector as well as to the increased decay power of the TRU fuel compositions. For a 0.2%UO₂ mixed or a 30%UO₂ mixed TRU, the peak fuel temperature was decreased due to the reduced decay power, however, it was still higher than 1600 °C due to the lack of heat absorber volume in the central reflector.     

Neutronic Analysis for Transmutation of Minor Actinides and Long-Lived Fission Products in a Fusion-Driven Transmuter (FDT)

This study presents the transmutations of both the minor actinides (MAs: ²³⁷Np, ²⁴¹Am, ²⁴³Am and ²⁴⁴Cm) and the long-lived fission products (LLFPs: ⁹⁹Tc, ¹²⁹I and ¹³⁵Cs), discharged from high burn-up PWR-MOX spent fuel, in a fusion-driven transmuter (FDT) and the effects of the MA and LLFP volume fractions on their transmutations. The blanket configuration of the FDT is improved by analyzing various sample blanket design combinations with different radial thicknesses. Two different transmutation zones (TZ_MA and TZ_FP which contain the MA and LLFP nuclides, respectively) are located separately from each other. The volume fractions of the MA and the LLFP are raised from 10 to 20% stepped by 2% and from 10 to 80% stepped by 5%, respectively. The calculations are performed to estimate neutronic parameters and transmutation characteristics per D–T fusion neutron. The conversion ratios (CRs) for the whole of all MAs are about 65–70%. The transmutation rates of the LLFP nuclides increase linearly with the increase of volume fractions of the MA, and the ⁹⁹Tc nuclide among them has the highest transmutation rate. The variations of their transmutation rate per unit volume in the radial direction are quasi-concave parabolic.     

Main results of the development of dispersion type IMF at A.A. Bochvar Institute

At A.A. Bochvar Institute a novel conception of IMF to burn civil and weapon’s grade Pu is currently accepted. It consists in the fact, that instead of using pelletized IMF, that features low serviceability and dust forming route of fuel element fabrication, the usage is made of dispersion type fuel element with aluminium or zirconium matrices.Dispersion fuels feature a high irradiation resistance and reliability; they can consequently reach high burnups and be serviceable under transient conditions.Three basic fuel element versions are under development in VNIINM for both thermal and fast reactors.The first version is a fuel element with a heterogeneous arrangement of fuel (PuO₂ or YSZ granules) within an Al or Zr matrix. The second version of a fuel element has a heat conducting Al or Zr alloy matrix and an isolated arrangement of PuO₂ in a fuel minielement more fully meets the ‘Rock Fuel’ requirements. According to the third version a porous meat of zirconium metallurgically bonded to a fuel cladding is formed through which a PuO₂ powder is introduced. All the versions are technologically simple to fabricate and require minimal quantities of process operations related to treating MA and Pu. Preliminary in-pile tests of IMF prototypes are presented.     

Neutronics analysis of the power flattening and minor actinides burning in a thorium-based fusion–fission hybrid reactor blanket

A neutronics analysis has been performed for a thorium fusion breeder with a special task of burning minor actinide ²³⁷Np, ²⁴¹Am, ²⁴³Am, and ²⁴⁴Cm, and production of ²³³U for the future PWR application. Under a first wall fusion neutron wall loading of 0.1 MW/m² by a plant factor of 100%, preliminary neutronics calculations have been performed using the one-dimensional transport and burnup calculation code BISONC and the Monte Carlo transport code MCNP. To obtain a quasi-constant nuclear heat production density, 11 fuel rods containing the mixture of ThO₂ and minor actinides are placed in a radial direction in the fissile zone where ThO₂ is mixed with variable amounts of minor actinides. Calculation results show that the tritium breeding ratio is greater than 1.05 for both investigated Cases A and B, and the hybrid reactor is self-sufficient in the tritium required for the (DT) fusion driver in those models during the operation period. The blanket energy multiplication factor M, varies between 13.8 and 29.6 depending on the fuel types at the end of the operation period. The peak-to-average fission power density ratio (Γ) is less than 1.66 and 1.68 for both Cases A and B, respectively during the operation time. After 720 days of plant operation, the accumulated ²³³U is 1277 and 1725 kg in the blanket for the Cases A and B, respectively.     

MgO-pyrochlore composite as an inert matrix fuel: Neutronic and thermal characteristics

Inert matrix fuels are an important component of advanced nuclear fuel cycles, as they provide a means of utilizing plutonium and reducing the inventory of ‘minor’ actinides. We examine the neutronic and thermal characteristics of MgO-pyrochlore (A₂B₂O₇: La₂Zr₂O₇, Nd₂Zr₂O₇ and Y₂Sn₂O₇) composites as inert matrix fuels in boiling water reactors. By incorporating plutonium with resonance nuclides, such as Am, Np and Er, in the A-site of pyrochlore, the k_infvs. burn-up curves are shown to be similar to those of UO₂, although the Doppler coefficients are less negative than UO₂. The Pu depletion rates are 88-90% (²³⁹Pu) and 54-58% (total Pu) when the inert matrix fuels experience a burn-up equivalent of 45 GWd/tHM UO₂. Because of the high thermal conductivity of MgO, the center-line temperatures of the MgO-pyrochlore composites at 44.0 kW/m are lower than those of UO₂ pellets. After burn-up, the A-site cation composition is 15-35 at.% lower than that of the B-site cations in pyrochlore (e.g., A_1.84B_2.17O_7.00) due to the fission of Pu in the A-site and the presence of fission product elements in the A- and B-sites of the pyrochlore structure.     

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.     

Time-dependent Neutronic Analysis for High Level Waste Transmutation in a Fusion-driven Transmuter

This study presents time-dependent transmutations of high-level waste (HLW) including minor actinides (MAs) and long-lived fission products (LLFPs) in the fusion-driven transmuter (FDT) that is optimized in terms of the neutronic performance per fusion neutron in our previous study. Its blanket has two different transmutation zones (MA transmutation zone, TZ_MA, and LLFP transmutation zone, TZ_FP), located separately from each other. High burn-up pressured water reactor (PWR)-mixed oxide (MOX) spent fuel is used as HLW. The time-dependent transmutation analyses have been performed for an operation period (OP) of up to 10 years by 75% plant factor (η) under a first-wall neutron load (P) of 5 MW/m². The effective half-lives of the MA and LLFP nuclides can be shortened significantly in the considered FDT while substantial electricity is produced in situ along the OP.     

Fabrication and study of the properties of PuO2−MgO fuel pellets

Conclusions Defect-free PuO₂−MgO pellets with a density of 4.4 g/cm³ (90% of the computed density of the composition, in which the volume fractions of PuO₂ and MgO equal 15 and 85% respectively), were obtained. Work with plutonium-containing material showed that the technology developed for fabricating UO₂−MgO fuel pellets is suitable for fabricating PuO₂−MgO fuel pellets. Main Science Center of the Russian Federation — A. I. Leipunskii Physics and Power-Engineering Institute. Translated from Atomnaya énergiya, Vol. 82, No. 5, pp. 355–358, May, 1997.     

ECRIX-H experiment: Synthesis of post-irradiation examinations and simulations

The purpose of the ECRIX-H experiment is to study the behaviour of a composite ceramic target made of AmO_1.62 microdispersed in an MgO matrix irradiated for 318 EFPD in the Phenix sodium-cooled fast reactor (SFR), in a specific carrier sub-assembly equipped with annular blocks of CaH_x acting as a neutron moderator. Results indicate that magnesia-based inert matrix targets display satisfactory behaviour and moderate swelling under irradiation, even for significant quantities of helium produced and a high burn-up. On this basis, the design of transmutation fuel pins for recycling of minor actinides (MA) in accelerator-driven systems (ADS) or in fast neutron reactors (FR) could be optimised so as to increase their performance level (initial MA content, burn-up, etc.).The measured Am fission rate (25 at.%) was found to be lower than that predicted by neutronic simulations probably due to the inaccuracies linked to the complexity of neutron modelling and the uncertainties on nuclear data related to moderated neutron spectrum. In addition, as most of the initial Am transmuted into Pu under irradiation, a PuO_x-type phase was created within the initial AmO_1.62 particles, leading to the incomplete dissolution of the irradiated targets under standard reprocessing conditions. This issue will have to be considered and investigated in greater detail for all transmutation fuels and targets devoted to the multi-recycling of MA.     

Mechanical properties and XRD studies of silicon carbide inert matrix fuel fabricated by a low temperature polymer precursor route

The mechanical properties of silicon carbide (SiC) inert matrix fuel (IMF) pellets fabricated by a low temperature (1050 °C) polymer precursor route were evaluated at room temperature. The Vickers hardness was mainly related to the chemical bonding strength between the amorphous SiC phase and the β-SiC particles. The biaxial fracture strength with pre-notch and fracture toughness were found to be mostly controlled by the pellet density. The maximum Vickers hardness, biaxial fracture strength with pre-notch and fracture toughness achieved were 5.6 GPa, 201 MPa and 2.9 MPa m^1/2 respectively. These values appear to be superior to the reference MOX or UO₂ fuels. Excellent thermal shock resistance for the fabricated SiC IMF was proven and the values were compared to conventional UO₂ pellets. XRD studies showed that ceria (PuO₂ surrogate) chemically reacted with the polymer precursor during sintering, forming cerium oxysilicate. Whether PuO₂ will chemically react in a similar manner remains unclear.     

Prediction of fission mass-yield distributions based on cross section calculations

For the first time, fission mass-yield distributions have been predicted based on an extended statistical model for fission cross section calculations. In this model, the concept of the multi-modality of the fission process has been incorporated. The three most dominant fission modes, the two asymmetric standard I (S1) and standard II (S2) modes and the symmetric superlong (SL) mode are taken into account. De-convoluted fission cross sections for S1, S2 and SL modes for ^235,238U(n, f) and ²³⁷Np(n, f), based on experimental branching ratios, were calculated for the first time in the incident neutron energy range from 0.01 to 5.5 MeV providing good agreement with the experimental fission cross section data. The branching ratios obtained from the modal fission cross section calculations have been used to deduce the corresponding fission yield distributions, including mean values also for incident neutron energies hitherto not accessible to experiment.     

Analysis of the characteristics of <Superscript>238</Superscript>U, <Superscript>232</Superscript>Th, <Superscript>237</Superscript>Np,and <Superscript>231</Superscript>Pa radiators for fission ionization chambers for measuring the neutron flux density in the ITER diverter

Computational results, obtained by analyzing possible schemes of nuclear transformations of each of four threshold fission radiators ²³⁸U, ²³²Th, ²³⁷Np, and ²³¹Pa, for fission ionization chambers are presented. The influence of the nuclear reactions (n, ƒ), (n, γ), and (n, 2n) on the characteristics of fission ionization chambers is taken into account in the nuclear transformation schemes for all four radiators. The results are presented in the form of a dependence of the sensitivity of the fission ionization chambers on the neutron fluence in the range 10²¹–10²⁴ cm⁻². The effect of 0.2 and 1 g/cm² thick boron screens is examined. Translated from Atomnaya énergiya, Vol. 106, No. 1, pp. 42–47, January, 2009.