Burnable Neutron Absorbers Research Articles

A comparative study on the impact of Gd2O3 burnable neutron absorber in UO2 and (U, Th)O2 fuels

The performance of gadolinium burnable absorber (GdBA) for reactivity control in UO2 and (U, Th)O2 fuels and its impact on spent fuel characteristics was performed. Five fuel assemblies: one without GdBA fuel rod and four each containing 16, 24, 34 and 44 GdBA fuel rods in both fuels were investigated. Reactivity swing in all the FAs with GdBA rods in UO2 fuel was higher than their counterparts with similar GdBA fuel rods in (U, Th)O2 fuel. The excess reactivity in all FAs with (U, Th)O2 fuel was higher than UO2 fuel. At the end of single discharge burn-up (~ 49.64 GWd/tHM), the excess reactivity of (U, Th)O2 fuel remained positive (16,000 pcm) while UO2 fuel shows a negative value (−6,000 pcm), which suggest a longer discharge burn-up in (U, Th)O2 fuel. The concentration of plutonium isotopes and minor actinides were significantly higher in UO2 fuel than in (U, Th)O2 fuel except for 236Np. However, the concentration of non-actinides (gadolinium and iodine isotopes) except for 135Xe were respectively smaller in (U, Th)O2 fuel than in UO2 fuel but may be two times higher in (U, Th)O2 fuel due to its potential longer discharge burn-up.

Actinide and non-actinide production from UO2 fuel in W-SMR: Effects of gadolinium burnable absorber

Burnable neutron absorbers are used for efficient reactivity control especially in light and heavy water reactors. Effect of gadolinium burnable absorber (GdBA) on reactivity, actinide and non-actinide content of nuclear waste was investigated. Westinghouse small modular reactor (W-SMR) was studied using MCNPX 2.7 code integrated with CINDER90 for the calculations. Positive reactivity swing was observed in fuel assemblies with GdBA rods. The concentration of actinides was relatively higher in FA without GdBA rod compared to those with GdBA rods. The net activity of 241Pu and 239Np were respectively higher compared to other isotopes of plutonium and neptunium and slightly larger in FAs with GdBA rods. However, a significant increase on the concentration of non-actinide inventory was observed which increased with the number of GdBA rods. Therefore, the use of GdBA for reactivity control should be limited to avoid unrealistic reactivity swing and increased non-actinides content of the spent fuel.

Radiation-Induced Densification Model of Uranium-Gadolinium Fuel

In this paper, we have developed a model of radiation-induced densification of uranium-gadolinium fuel. The fuel is established to be in the brittle state under irradiation in the volume of uranium-gadolinium fuel pellet containing a burnable neutron absorber and in the plastic state out of this volume. The plastic zone volume increases during burnup of the neutron absorber. The relative change in a pellet volume or length of a fuel rod for uranium-gadolinium fuel is shown to be lower than that for uranium oxide nuclear fuel having the same microstructure and irradiated under the same conditions (the neutron flux).

Sintering behavior of UO2[sbnd]Er2O3 mixed fuel

The incorporation of burnable neutron absorbers into nuclear fuel pellets is important regarding reactivity compensation, which enables longer fuel cycles. The dry mechanical blending route is the most attractive process to accomplish absorbers incorporation because of its simplicity. By using this route, the present work has investigated the sintering behavior of UO2Er2O3 mixed fuel. A comparison with UO2Gd2O3 sintering behavior was presented. The behavior of UO2Er2O3 fuel sintering was similar to that reported for UO2Gd2O3 fuel, e.g. two-stage sintering with two peaks in the shrinkage rate curves. The effect showed to be less pronounced for Er2O3. This was attributed to the characteristics of the Er2O3 powder particles used as raw-material, whose agglomerates can be more easily broken and thus better homogenized during the blending with UO2 powder. These results confirmed that sinterability depends directly on the quality of the homogenization of the powders, as seen previously. A second phase was experimentally detected in the UO2Er2O3 system, but its impact on the sintering behavior of this mixed fuel has not yet been clarified.

Optimization of the Canadian SCWR Core Using Coupled Three-Dimensional Reactor Physics and Thermal-Hydraulics Calculations

The Canadian supercritical water-cooled reactor (SCWR) design is part of Canada's Generation IV reactor development program. The reactor uses batch fueling, light water above the thermodynamic critical point as a coolant and a heavy water moderator. The design has evolved considerably and is currently at the conceptual design level. As a result of batch fueling, a certain amount of excess reactivity is loaded at the beginning of each fueling cycle. This excess reactivity must be controlled using a combination of burnable neutron poisons in the fuel, moderator poisons, and control blades interspersed in the heavy water moderator. Recent studies have shown that the combination of power density, high coolant temperatures, and reactivity management can lead to high maximum cladding surface temperatures (MCST) and maximum fuel centerline temperatures (MFCLT) in this design. This study focuses on improving both the MCST and the MFCLT through modifications of the conceptual design including changes from a 3 to 4 batch fueling cycle, a slightly shortened fuel cycle (although exit burnup remains the same), axial graded fuel enrichment, fuel-integrated burnable neutron absorbers, lower reactivity control blades, and lower reactor thermal powers as compared to the original conceptual design. The optimal blade positions throughout the fuel cycle were determined so as to minimize the MCST and MFCLT using a genetic algorithm and the reactor physics code PARCS. The final design was analyzed using a fully coupled PARCS-RELAP5/SCDAPSIM/MOD4.0 model to accurately predict the MCST as a function of time during a fueling cycle.

Protactinium-231 – New burnable neutron absorber

Burnable neutron absorbers such as gadolinium and erbium are used for compensating excess reactivity in nuclear reactors. Their daughter nuclides resulting from neutron absorption by erbium and gadolinium do not play important role from the viewpoint of neutronics processes occurring in the reactor core. Selection of such burnable neutron absorber, daughter nuclides of which would favorably impact fission chain reaction, is of significant interest.The aim of the present study is to investigate neutronics properties of 231Pa – the new burnable neutron absorber – and possibilities of its producing in significant quantities. The chain of isotopic transformations starting from 237Np is the analogue of the chain of isotopic transformations started from 231Pa. However, improvement of neutron-multiplication properties in the 237Np-chain can only be achieved in fast neutron spectra while in the case of 231Pa-chain positive neutron balance can be achieved both in fast and thermal neutron spectra. From this viewpoint the chain starting from 231Pa is unique. In addition, 237Np can be produced in nuclear reactors as the result of neutron radiative capture by 235U while significant amounts of 231Pa can only be produced through the threshold (n,2n) and (n,3n)-reactions on 232Th under its irradiation with super high-energy neutrons. Such neutrons with super high energies are practically absent even in fast spectrum reactors, but, however, these neutrons are available in fusion reactors. Breeding of 231Pa in fusion reactors and further use of 231Pa in nuclear power reactors can make it possible to realize potential capabilities of fusion facilities for radical increase of nuclear reactor fuel burn-up. Thus, 231Pa isotope is the new and unique burnable neutron absorber never suggested for the purpose before.Evaluated nuclear data libraries JENDL-4.0 and ENDF/B-V, as well as computer software system SCALE-4.3 were used in the implementation of the present study.The following results were obtained.(1) In contrast to conventional burnable neutron absorbers on the basis of gadolinium and erbium, the protactinium isotope suggested in this paper appears to be more attractive because it allows us not only compensating initial excess reactivity, but, also, ensuring high fuel burn-up due to good multiplication properties of its daughter nuclides.(2) Significant quantities of protactinium could be produced in hybrid fusion–fission reactors acting as sources of neutrons (not sources of energy) with parameters already achieved by the present moment by experimental facilities in USA, Japan, UK.

Fueling Study of a CANDU Reactor Using Fuels Containing Burnable Neutron Absorbers

Trace amounts of burnable neutron absorbers (BNAs) were used to tailor the reactivity of the 37-element, natural uranium (NU) fuel bundle used in CANDU reactors. The BNAs of interest included Gd2O3 and Eu2O3, which were added to the fuel in variable quantities and combinations. The fuel lattice was modeled using the WIMS–AECL 3.1 code, and core simulations were conducted using the Reactor Fuelling Simulation Program (RFSP). The fuel model assumes an equivalent and uniform distribution of BNAs in the CANLUB layer of each fuel element.The incorporation of BNAs is designed to improve CANDU reactor operating margins during on-power refueling by eliminating the fueling transient (FT) and reducing the magnitude of the plutonium peak (PP) that is characteristic of NU fuels. By adding an optimal combination of “fast-burning” and “slow-burning” BNAs, the FT and PP can be selectively reduced, and a significantly flatter trend in the burnup-dependent evolution of fuel reactivity can be achieved.The results of the study indicate that by adding ~150 mg [~8 parts per million (ppm)] of Gd2O3 and ~300 mg (~15 ppm) of Eu2O3 per fuel bundle, the best gain in the operating margins of a 2650-MW(thermal) (480-channel) model CANDU reactor can be achieved. Based on the simulation of refueling events, it was shown that the magnitude of average postrefueling channel power ripples can be reduced by an average of 100 kW and a maximum of 220 kW for powers observed immediately after refueling. This reduction in postrefueling powers was also shown to allow the average liquid zone controller level to decrease from ~48% to 10%. This decrease implies a potential relief on overpower protection (an operating margin).

Gas emission from the UO2 samples, containing fission products and burnable absorber

The process gas released from the fuel pellets of uranium fuel during fuel burn-up reduces the thermal conductivity of the rod-shell gap, enhances hydrogen embrittlement of the cladding material, causes it's carbonization, as well as transport processes in the fuel. In this study a technique of investigating the thermal desorption of gases from the UO2 fuel material were perfected in the temperature range 300-2000 K for uniform sample heating rate of 15 K/min in vacuum. The characteristic kinetic dependences are acquired for the gas emission from UO2 samples, containing simulators of fission products (SFP) and the burnable neutron absorber (BNA). Depending on the amount of SFP and BNA contained in the sample thermal desorption gas spectra (TDGS) vary. The composition of emitted gas varies, as well as the number of peaks in the TDGS and the peaks shift to higher temperatures. This indicates that introduction of SFPs and BNA alters the sample material structure and cause the creation of so- called traps which have different bonding energies to the gases. The traps can be a grid of dislocations, voids, and contained in the UO2 matrix SFP and BNA. Similar processes will occur in the fuel pellets in the real conditions of the Nuclear Power Plant as well.

High temperature dilatometry of simulated oxide nuclear fuel

High temperature dilatometry of model systems based on uranium dioxide with additives of burnable neutron absorbers both as Gd2O3 and as AlGdO3, and fission products simulators (FPS) was performed. It shown that in some cases instead of high temperature samples shrinkage there is a sharp transition to the expansion, which is associated with an increase of the samples volume due to the formation of liquid phases. The beginning of a complex composition eutectic melting starts at temperatures from 1950 to 2250 °C in the uranium dioxide samples containing significant amounts of Al, Gd, and FPS. Thus, in the analysis of oxide nuclear fuel behavior at high temperatures should be considered that the formation of liquid phases is possible at a temperature of 1000 °C lower than a melting point of pure stoichiometric uranium dioxide if its initial composition became more complex.

Thermal properties of (Al,Gd)O3 doped uranium dioxide

In this paper were defined thermal characteristics (thermal diffusivity up to 2000 °C, heat capacity up to 1250 ° C and thermal conductivity of up to 1250 °C) of oxide nuclear fuel with additives of burnable neutron absorber in the form of (Al,Gd)O3. Also for comparison samples UO2 with gadolinium Gd2O3 and pure uranium dioxide were studied. It is shown that the thermal diffusivity and thermal conductivity of UO2 with dopant of gadolinium in the form of (Al, Gd) O3 is considerably higher than those of Gd2O3 -doped uranium dioxide, due to the fact that (Al, Gd)O3 does not form a solid solution with UO2.

Cubic Phases in the Gd2O3-ZrO2 and Dy2O3-TiO2 Systems for Nuclear Industry Applications

Neutron absorbers are elements with a high neutron capture cross section that are employed at nuclear reactors to control excess fuel reactivity. If these absorbers are converted into materials of relatively low absorption cross section as the result of neutron absorption, they consume during the reactor core life and so are called burnable. These elements can be distributed inside an oxide ceramic that is stable under irradiation and thus called inert. Cubic zirconium oxide is one of the preferred materials to be used as inert matrix. It is stable under irradiation, experiments very low swelling, and is isomorphic to uranium oxide. The cubic phase is stabilized by adding small amounts of dopants like Dy2O3 and Gd2O3. As both dysprosium and gadolinium have a high neutron cross section, they are good candidates to prepare burnable neutron absorbers. Pyrochlores, like Gd2Zr2O7 and Dy2Ti2O7, allow the solid solution of a large quantity of elements besides being stable under irradiation. These characteristics make them also useful for safe storage of nuclear wastes. We present a preliminary study of the thermal analysis of different compositions in the systems Gd2O3-ZrO2 and Dy2O3-TiO2, investigating the feasibility to obtain oxide ceramics useful for the nuclear industry.

Variation of Burnable Neutron Absorbers in a Heavy Water–Moderated Fuel Lattice: A Potential to Improve CANDU Reactor Operating Margins

A CANDU lattice cell has been modeled using the Los Alamos National Laboratory’s MCNP 6 code and Atomic Energy of Canada Limited’s WIMS-AECL 3.1. Models for the CANDU 37-element fuel bundle have included a CANLUB coating, as a carrier for the neutron absorbers. The objective is to improve CANDU reactor operating margins by adding small amounts (~1 g) of neutron absorbers to each fuel element.For CANDU natural uranium fuel bundle design, the results indicate that (a) the fueling transient (due to the xenon-free effect) could be significantly reduced using gadolinium oxide (Gd2O3), with no significant impact on fuel burnup, and (b) the reactivity peak (due to plutonium production) could be reduced using europium oxide (Eu2O3), with minimal impact on fuel burnup. An appropriate mixture of Gd2O3 and Eu2O3 that will improve operation and safety margins while having a minimal impact on fuel burnup is determined.Reactivity and power calculations for various mixtures of Gd2O3 and Eu2O3 are reported here. It is concluded that ~180 mg Gd2O3 and ~1000 mg Eu2O3 (~4.9 ×10−3 wt% per bundle) are sufficient to suppress the refueling transient and lower the axial plutonium peak, with a 0.27% burnup penalty (which is a small impact).Fuel safety and performance are always important topics for a nuclear utility. This approach of a relatively simple application of burnable poisons to existing CANDU natural uranium fuel design offers the benefits of improving fuel utilization and safety margins.

Dopant solubility and lattice contraction in gadolinia and gadolinia–chromia doped UO2 fuels

Gadolinia doped UO2 fuel is widely used as burnable neutron absorber in Light Water Reactors to reduce power peaking and excess reactivity during the first reactor cycle of fresh fuel assemblies. The thermal conductivity of gadolinia doped fuel is substantially lower than that of standard UO2. To maintain safety margins later in life, some design or operating restrictions can be defined, for example to compensate higher fission gas release levels. Development of large grain U/Gd fuel by suitable doping, e.g. Cr2O3, could offer a solution to such restrictions, but solid state information about the double doped (U1−x−yGdxCry)O2 system is very scarce. In the present paper, we present X-ray diffraction and microstructure results of standard U/Gd fuel and chromia doped U/Gd fuel manufactured by powder metallurgy. The dissolution of chromium in (U1−xGdx)O2 as a function of Gd content, the role of free UO2 and the lattice contraction at different Gd and Cr doping levels of (U1−x−yGdxCry)O2 is studied both for the single doped and double doped system. On the basis of lattice contraction and precise measurements of the composition of the solid solution phases, the evolution of theoretical density with dopant concentration is derived.

An investigation of some models and approximations used in calculating fuel assemblies of a high-temperature gas-cooled reactor of the GT-MHR type

The paper describes basic neutron-physics models developed in the Division of Advanced Nuclear Power Systems of the Institute of Nuclear Reactors, Russian Research Center Kurchatov Institute, as design models intended for calculating the characteristics of block fuel assemblies of a high-temperature gas-cooled reactor GT-MHR, namely, models for calculating burnup of fuel and isotopes of burnable neutron absorbers and calculating fuel assemblies at fixed points with respect to burnup with preparation of the neutron constants in a preassigned number of energy groups for full-scale design of a reactor. A model problem for investigation of calculated approximations is proposed. The outcome of this investigation is a developed stage-by-stage procedure of preparing group homogeneous cross sections of a fuel assembly and its parts that has been introduced into the practice of design calculations of a GT-MHR reactor.

Development of Neural Network for Predicting Local Power Distributions in BWR Fuel Bundles Considering Neutron Absorber

A neural network model is under development to predict the local power distribution in a BWR fuel bundle as a high speed simulator of precise nuclear physical analysis model. The relation between 235U enrichment of fuel rods and local peaking factor (LPF) has been learned using a two-layered neural network model ENET. The training signals used were 33 patterns having considered a line symmetry of a 8x8 assembly lattice including 4 water rods. The ENET model is used in the first stage and a new model GNET which learns the change of LPFs caused by burnable neutron absorber Gadolinia, is added to the ENET in the second stage. Using this two-staged model EGNET, total number of training signals can be decreased to 99. These training signals are for zero-burnup cases. The effect of Gadolinia on LPF has a large nonliniality and the GNET should have three layers. This combined model of EGNET can predict the training signals within 0.02 of LPF error, and the LPF of a high power rod is predictable within 0.03 error for Gadolinia rod distributions different from the training signals when the number of Gadolinia rods is less than 10. The computing speed of EGNET is more than 100 times faster than that of a precise nuclear analysis model, and EGNET is suitable for scoping survey analysis.

Development of Neural Network for Predicting Local Power Distributions in BWR Fuel Bundles Considering Burnable Neutron Absorber.

A neural network model is under development to predict the local power distribution in a BWR fuel bundle as a high speed simulator of precise nuclear physical analysis model. The relation between 235 U enrichment of fuel rods and local peaking factor (LPF) has been learned using a two-layered neural network model ENET. The training signals used were 33 patterns having considered a line symmetry of a 8x8 assembly lattice including 4 water rods. The ENET model is used in the first stage and a new model GNET which learns the change of LPFs caused by burnable neutron absorber Gadolinia, is added to the ENET in the second stage. Using this two-staged model EGNET, total number of training signals can be decreased to 99. These training signals are for zero-burnup cases. The effect of Gadolinia on LPF has a large nonliniality and the GNET should have three layers. This combined model of EGNET can predict the training signals within 0.02 of LPF error, and the LPF of a high power rod is predictable within 0.03 error for Gadolinia rod distributions different from the training signals when the number of Gadolinia rods is less than 10. The computing speed of EGNET is more than 100 times faster than that of a precise nuclear analysis model, and EGNET is suitable for scoping survey analysis.