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.     

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.     

Assessment of criticality and burn up behavior of candu reactors with nuclear waste trans uranium fuel

Large quantities of nuclear waste plutonium and minor actinides (MAs) have been accumulated in the civilian light water reactors (LWRs) and CANDU reactors. These trans uranium (TRU) elements are all fissionable, and thus can be considered as fissile fuel materials in form of mixed fuel with thorium or nat-uranium in the latter. CANDU fuel compacts made of tristructural-isotropic (TRISO) type pellets would withstand very high burn ups without fuel change.As carbide fuels allow higher fissile material density than oxide fuels, following fuel compositions have been selected for investigations: ① 90% nat-UC + 10% TRUC, ② 70% nat-UC + 30% TRUC and ③ 50% nat-UC + 50% TRUC. Higher TRUC charge leads to longer power plant operation periods without fuel change. The behavior of the criticality k_∞ and the burn up values of the reactor have been pursued by full power operation for > ∼12 years. For these selected fuel compositions, the reactor criticality starts by k_∞ = 1.4443, 1.4872 and 1.5238, where corresponding reactor operation times and burn up values have been calculated as 2.8 years, 8 years and 12.5 years, and 62, 430 MW.D/MT, 176,000 and 280,000 MW.D/MT, with fuel consumption rates of ∼16, 5.68 and 3.57 g/MW.D respectively. These high burn ups would reduce the nuclear waste mass per unit energy output drastically. The study has show clearly that TRU in form of TRISO fuel pellets will provide sufficient criticality as well as reasonable burn up for CANDU reactors in order to justify their consideration as alternative fuel.     

A fuel performance analysis for a 450 MWth deep burn-high temperature reactor

A performance analysis for a 450 MW_th deep burn-high temperature reactor (DB-HTR) fuel was performed using COPA, a fuel performance analysis code of Korea Atomic Energy Research Institute (KAERI). The code computes gas pressure buildup in the void volume of a tri-isotropic coated fuel particle (TRISO), temperature distribution in a DB-HTR fuel, thermo-mechanical stress in a coated fuel particle (CFP), failure fractions of a batch of CFPs, and fission product (FP) releases into the coolant. The 350 μm DB-HTR kernel is composed of 30% UO₂ + 70% (5% NpO₂ + 95% PuO_1.8) mixed with 0.6 moles of silicon carbide (SiC) per mole of heavy metal. The DB-HTR is operated at the constant temperature and power of 858 °C and 39.02 mW per CFP for 1395 effective full power days (EFPD) and is subjected to a core heat-up event for 250 h during which the maximum coolant temperature reaches 1548.70 °C. Within the normal operating temperature, the fuel showed good thermal and mechanical integrity. At elevated temperatures of the accident event, the failure fraction of CFPs resulted from the mechanical failure (MF) and the thermal decomposition (TD) of the SiC barrier is 3.30 × 10⁻³.     

Dynamic Monte-Carlo modeling of thermal desorption of H2 from H irradiated graphite

Thermal desorption of hydrogen molecules from H⁺ irradiated graphite is studied using dynamic Monte Carlo simulation. The purpose of this study is to understand the experimentally observed phenomena that the thermal desorption of H₂ from the graphite exhibits sometimes single desorption peak, sometimes double peaks, and even three desorption peaks under certain circumstances. The study result reveals that the fluence of pre-implanted H⁺, the concentration of trap sites, porosity, and mean crystallite volume are important parameters in determining the number of desorption peaks. It is found that low implantation fluence and high concentration of trap sites easily lead to the occurrence of single desorption peak at around 1000 K, and high implantation fluence and low concentration of trap sites favor the occurrence of double desorption peaks, with a new desorption peak at around 820 K. It is also found that small porosity of graphite and large crystallite volume benefit the occurrence of single desorption peak at around 1000 K while large porosity of graphite and small crystallite volume facilitate the occurrence of double desorption peaks, respectively, at around 820 and 1000 K. In addition, experimentally observed third desorption peak at lower temperature is reproduced by simulation with assuming the graphite containing a small concentration of solute hydrogen atoms.     

New fuel rod design based on LEU/Th fuel and an assessment on in-core fuel cycle length of the MHR core

The current Modular Helium Reactor (MHR) fuel cycle uses fissile LEU (19.8 wt% U-235) and depleted uranium in separate TRISO particles, in a single fuel rod within a graphite matrix. The TRISO particle volume packing fraction (PF) in the fuel rods is 29%, of which the LEU particle PF is 62%. The lifetime between refuelings is about 476 effective full power days (EFPD). In this paper we assess the possibility of replacing the depleted uranium TRISO particles with thorium TRISO particles, and evaluate the impact of such replacement on fuel cycle length. A preliminary scoping study was performed to determine the most promising fuel rod/zoning configurations. The scoping study indicates that there is advantage to separating the thorium TRISO particles from the LEU particles at the fuel rod level instead of mixing them within a single rod. An axial checkerboard distribution of the fuel rods where all uranium and all thorium rods are interchangeable along the axial direction within the graphite block is the most promising configuration that was identified in this study and can be lead to a fuel cycle length extension of 50-80% relative to the current design, with only a modest increase in the fissile material loading (15-20%). To this advantage can be added the benefit of a significant reduction in nuclear waste and in health risk. This study also lays the foundation for improving the fuel rod arrangement within the graphite block and the graphite blocks within the entire reactor core. The analysis is limited to a once - through fuel cycle based on in situ fissioning of the U-233, without further separation and reprocessing. The preliminary heat transfer analysis indicates that the maximum temperature in the fuel will be raised by about 10-15% over that of current MHR design.     

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.     

Immobilization of carbon-14 from reactor graphite waste by use of self-sustaining reaction in the C-Al-TiO2 system

As a result of long-term neutron irradiation, the long-lived ¹⁴C is produced in the reactor graphite. The exothermic self-sustaining reaction 3C(graphite) + 4Al + 3TiO₂ = 3TiC + 2Al₂O₃ was proposed for processing of such waste. In doing so, the carbon, including the ¹⁴C, is chemically bound in the stable TiC. The reaction products in the C-Al-TiO₂ system were investigated both by thermodynamic simulation and experimentally in the course of this work.     

The inhibiting effects of hydrogen on the corrosion of uranium dioxide under nuclear waste disposal conditions

A number of electrochemical experiments were employed to investigate the effects of hydrogen on the corrosion of UO₂ under nuclear waste disposal conditions. A combination of corrosion potential (E_CORR) measurements and cyclic voltammetry have indicated that dissolved hydrogen can polarize the UO₂ surface to reducing potentials; i.e., to E_CORR values more negative then those observed under anoxic (argon-purged) conditions. A comparison of the behaviours of SIMFUEL specimens with and without incorporated noble metal ε-particles indicates that these particles may act as catalytic electrodes for H₂ oxidation, H₂ ↔ 2e⁻ + 2H⁺. It is the galvanic coupling of these particles to the UO₂ matrix which suppresses the fuel corrosion potential.     

Performance of TRISO particles fueled with plutonium and minor actinides in a PBMR-400 core design

The TRISO particle design of high temperature reactors fueled with plutonium (Pu) and/or minor actinides (MAs) is investigated by calculating the failure fraction of TRISO particles during irradiation. For this purpose, a fuel depletion, neutronics and thermal-hydraulics code system, which delivers the fuel temperature, fast neutron flux and power density profiles, is coupled to an analytical stress analysis code. The latter is being further developed for the calculation of a reliable and realistic failure fraction. The code system has been applied to a PBMR-400 design containing TRISO particles fueled with 1st and 2nd generation plutonium and with a target burn-up of 700 and 600 MWd/kgHM, respectively. It is shown that the pebble-bed type high temperature reactor under consideration is a promising option for burning Pu and MAs if very high burn-ups can be achieved. The TRISO particle failure fraction is also calculated for both Pu and MA fuels, and compared to U-based fuel. It is shown by the present stress analysis code that the Pu-based fuel particles need a better design and this has been achieved for the MA-based fuel, in which helium gas atoms have a significant contribution to the buffer pressure.     

Thermal–hydraulics features of the use of LEU U3Si2 fuel in low-power reactors

Calculations for the use of the U₃Si₂ LEU fuel in low-power research reactors were made. The design basis accident was simulated using the feedback coefficients calculated by the BMAC system. Usability of this fuel in low-power research reactors was demonstrated for both normal daily and accidental operation conditions even if the power of the reactor touches 142 kW during the design basis accident simulation. Both HEU and LEU fuels behave similarly in the normal operation, the temperature of the cladding reaching about 60 °C while higher temperature are obtained for the accidental conditions in the case of the LEU fuel (about 113.7 °C against 98.6 °C for the fuel center temperatures).     

Utilization of TRISO Fuel with LWR Spent Fuel in Fusion-Fission Hybrid Reactor System

HTRs use a high performance particulate TRISO fuel with ceramic multi-layer coatings due to the high burn up capability and very neutronic performance. TRISO fuel because of capable of high burn up and very neutronic performance is conducted in a D-T fusion driven hybrid reactor. In this study, TRISO fuels particles are imbedded body-centered cubic (BCC) in a graphite matrix with a volume fraction of 68%. The neutronic effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on the fuel performance has been investigated for Flibe, Flinabe and Li₂₀Sn₈₀ coolants. The reactor operation time with the different first neutron wall loads is 24 months. Neutron transport calculations are evaluated by using XSDRNPM/SCALE 5 codes with 238 group cross section library. The effect of TRISO coated LWR spent fuel in the fuel rod used hybrid reactor on tritium breeding (TBR), energy multiplication (M), fissile fuel breeding, average burn up values are comparatively investigated. It is shown that the high burn up can be achieved with TRISO fuel in the hybrid reactor.     

TRISO-fuel element thermo-mechanical performance modeling for the hybrid LIFE engine with Pu fuel blanket

A TRISO-coated fuel thermo-mechanical performance study is performed for the fusion-fission hybrid Laser Inertial Fusion Engine (LIFE) to test the viability of TRISO particles to achieve ultra-high burn-up of Pu or transuranic spent nuclear fuel blankets. Our methodology includes full elastic anisotropy, time and temperature varying material properties, and multilayer capabilities. In order to achieve fast fluences up to 30 × 10²⁵ n m⁻² (E > 0.18 MeV), judicious extrapolations across several orders of magnitude of existing material databases have been carried out. The results of our study indicate that failure of the pyrolytic carbon (PyC) layers occurs within the first 2 years of operation. The particles then behave as a single-SiC-layer particle and the SiC layer maintains reasonably-low tensile stresses until the end-of-life. It is also found that the PyC creep constant, K, has a striking influence on the fuel performance of TRISO-coated particles, whose stresses scale almost inversely proportional to K. Conversely, varying the geometry of the TRISO-coated fuel particles results in little differences in terms of fuel performance.     

AFM characterization of model nuclear fuel oxide multilayer structures modified by heavy ion beam irradiation

This work explored a potential new model dispersion fuel form consisting of an actinide material embedded in a radiation tolerant matrix that captures fission products (FPs) and is easily separated chemically as waste from the fuel material. To understand the stability of this proposed dispersion fuel form design, an idealized model system composed of a multilayer film was studied. This system consisted of a tri-layer structure of an MgO layer sandwiched between two HfO₂ layers. HfO₂ served as a surrogate fissile material for UO₂ while MgO represented a stable, fissile product (FP) getter that is easily separated from the fissile material. This type of multilayer film structure allowed us to control the size of and spacing between each layer. The films were grown at room temperature by e-beam deposition on a Si(1 1 1) substrate and post-annealed annealing at a range of temperatures to crystallize the HfO₂ layers. The 550 °C annealed sample was subsequently irradiated with 10 MeV Au³⁺ ions at a range of fluences from 5 × 10¹³ to 3.74 × 10¹⁶ ions/cm². Separate single layer constituent films and the substrate were also irradiated at 5 × 10¹⁵ and 8 × 10¹⁴ and 2 × 10¹⁶, respectively. After annealing and irradiation, the samples were characterized using atomic force imaging techniques to determine local changes in microstructure and mechanical properties. All samples annealed above 550 °C cracked. From the AFM results we observed both crack healing and significant modification of the surface at higher fluences.     

The evolution of hydrogen from proton irradiated glassy carbon

Glassy carbon was irradiated with 15 keV H⁺ ion beam. It was observed that the implanted hydrogen is unstable in material and evolves as H, H₂ and H₂O. Post-irradiation evolution of H, H₂ and H₂O from proton irradiated glassy carbon was monitored by temperature programmed desorption (TPD) in the period of 30 days. In between irradiation and TPD measurements the irradiated samples were stored in air. The evolution of the molecular hydrogen, although the protons are implanted deeply below the surface of the disordered glassy carbon, proceeds over the same mechanism as in the case of low-energy H-atoms chemisorbed on the very surface of an ideal graphite structure.     

Analysis of fuel element matrix graphite corrosion in HTR-PM for normal operating conditions

The corrosion of spherical fuel element by oxidizing gases will degrade their thermal and mechanical properties in pebble-bed reactors. Oxidizing impurities in primary helium coolant may have significant influence on the gasification of graphite during the long service time even though their concentrations remain relatively low. This paper deals with the corrosion of matrix graphite by steam and oxygen in Chinese high temperature gas-cooled reactor pebble-bed module (HTR-PM). The influence of steam contents was first analyzed, and then the effect of oxygen contents was also taken into account. The results show that the corrosion by steam was weak and it would not threaten the normal service of spherical fuel elements for expected steam contamination levels. On the contrary, the corrosion would be more severe while the oxygen content was as high as currently expected. Finally, the upper limits of steam and oxygen in primary helium coolant were recommended to be 1.0 and 0.05 cm³ m⁻³.     

Numerical model for separation of H-D gas mixture in batch-type concentric-tube thermal diffusion columns

The modeling simulation for the separation of H-D gas mixture in batch-type concentric-tube thermal diffusion columns have been analyzed from the transport equation coupled with the application of mass balance. The most important assumption is that the concentrations of H₂, HD and D₂ are locally equilibrium at every points in the column as H₂ + D₂ ↔ 2HD. The concentration distribution equation was derived and the concentration difference between the bottom and top ends of the column could be estimated. The degree of separation and separation factor for recovery of deuterium from H-D gas mixture in the batch-type cryogenic-wall thermal diffusion column were estimated.     

Corrosion of irradiated MOX fuel in presence of dissolved H2

The corrosion behaviour of irradiated MOX fuel (47 GWd/tHM) has been studied in an autoclave experiment simulating repository conditions. Fuel fragments were corroded at room temperature in a 10 mM NaCl/2 mM NaHCO₃ solution in presence of dissolved H₂ for 2100 days. The results show that dissolved H₂ in concentration 1 mM and higher inhibits oxidation and dissolution of the fragments. Stable U and Pu concentrations were measured at 7 × 10⁻¹⁰ and 5 × 10⁻¹¹ M, respectively. Caesium was only released during the first two years of the experiment. The results indicate that the UO₂ matrix of a spent MOX fuel is the main contributor to the measured dissolution, while the corrosion of the high burn-up Pu-rich islands appears negligible.     

Modeling spent TRISO fuel for geological disposal: corrosion and failure under oxidizing conditions in the presence of water

High temperature gas reactors (HTGRs) are being considered for near term deployment in the United States under the GNEP program and farther term deployment under the Gen IV reactor design (U.S. DOE Nuclear Energy Research Advisory Committee, 2002). A common factor among current HTGR (prismatic or pebble) designs is the use of TRISO coated particle fuel. TRISO refers to the three types of coating layers (pyrolytic carbon, porous carbon, and silicon carbide) around the fuel kernel, which is both protected and contained by the layers. While there have been a number of reactors operated with coated particle fuel, and extensive amount of research has gone into designing new HTGRs, little work has been done on modeling and analysing the degradation rates of spent TRISO fuel for permanent geological disposal. An integral part of developing a spent fuel degradation modeling was to analyze the waste form without taking any consideration for engineering barriers. A basic model was developed to simulate the time to failure of spent TRISO fuel in a repository environment. Preliminary verification of the model was performed with comparison to output from a proprietary model called GARGOYLE that was also used to model degradation rates of TRISO fuel. A sensitivity study was performed to determine which fuel and repository parameters had the most significant effect on the predicted time to fuel particle failure. Results of the analysis indicate corrosion rates and thicknesses of the outer pyrolytic carbon and silicon carbide layers, along with the time dependent temperature of the spent fuel in the repository environment, have a significant effect on the time to particle failure. The thicknesses of the kernel, buffer, and IPyC layers along with the strength of the SiC layer and the pressure in the TRISO particle did not significantly alter the results from the model. It can be concluded that a better understanding of the corrosion rates of the OPyC and SiC layers, along with increasing the quality control of the OPyC and SiC layer thicknesses, can significantly reduce uncertainty in estimates of the time to failure of spent TRISO fuel in a repository environment.     

Preparation of composite polymer electrolytes by electron beam-induced grafting: Proton- and lithium ion-conducting membranes

Two classes of composite polymer electrolyte membranes, one conducting lithium ions (Li⁺) and the other conducting protons (H⁺) were prepared using simultaneous electron beam-induced grafting. Porous poly(vinylidene fluoride) (PVDF) films were impregnated with styrene and subjected to electron beam (EB) irradiation to obtain polystyrene (PS) filled PVDF precursor films that were subsequently treated with either chlorosulfonic acid/1,1,2,2-tetrachloroethane mixture to obtain H⁺-conducting composite membranes or LiPH₆/EC/DEC liquid electrolyte to obtain Li⁺-conducting composite membranes. The properties of the obtained membranes were evaluated using Fourier transform infrared spectroscopy (FTIR), scanning electron microscopy (SEM) and AC impedance measurements. The obtained membranes were found to achieve grafting content up to 46% with superior Li⁺-conductivity of 1.91 × 10⁻³ S/cm and H⁺-conductivity of 5.95 × 10⁻² S/cm. The results of this work show that simultaneous radiation-induced grafting with EB is a promising method to prepare high quality ion-conducting membranes for possible use in fuel cells and lithium batteries.