Fuel Burnup Research Articles

A Comparative Study on Gamma-ray Measurement and MCNP Simulation for Precise Measurement of Spent Nuclear Fuel Burnup

A Comparative Study on Gamma-ray Measurement and MCNP Simulation for Precise Measurement of Spent Nuclear Fuel Burnup

JSI TRIGA fuel rod reactivity worth experiments for validation of Serpent-2 and RAPID fuel burnup calculations

Reactivity worth of fuel rods at the JSI TRIGA research reactor was measured. Differently burned fuel rods were chosen to validate fuel burnup calculations. Two methods of measuring reactivity worth of fuel rods are used, traditional method is compared to newly introduced method using fuel rods swapping. Connection between both methods is described theoretically and the theory is validated experimentally. Fuel rod worth calculated using the newly introduced fuel rod swap method was within 1σ of worth measured using the traditional method. In addition to the recently performed experiments, weekly measurements of reactor core reactivity throughout the operational history are used for validation. The measured data were used to validate the fuel burnup and core criticality calculations. Fuel burnup calculations are performed using three different computer codes: the deterministic TRIGLAV, the Monte Carlo Serpent-2, and the hybrid RAPID. Great agreement was observed for Serpent-2 and RAPID by simulating fuel rod worth and its burnup, indicating that the fuel burnup and criticality calculations are accurate and that reactivity changes due to small burnup differences on the order of 10 pcm can be accurately simulated. In addition it was shown using ex-core detectors and large fission chamber that detector response changes due to fuel swapping are evident for fuel rod burnup differences of 20 MWd/kg. Fuel burnup calculations were further validated on excess reactivity measurements for three mixed TRIGA cores. The calculated burnup reactivity coefficient ΔρBU using Serpent-2 and RAPID was within 1σ of the measurements, showing both codes are capable of calculating burnup for different TRIGA fuel types.

Fuel performance evaluation of two high burnup PWR core designs during normal operation, control rod withdrawal, and control rod ejection scenarios

There is interest among utilities to extend the current, 18-month operating cycle to 24 months. Economically, this extension would require greater than 5 % enrichment and peak rod average discharge burnup levels above 62 GWd/MTU. A notable challenge of increasing enrichment is the resulting additional excess reactivity encountered during the early stages of fuel life. To accommodate, burnable absorbers beyond soluble boron are introduced into the fuel system. In high burnup fuels, the possibilities of cladding lift-off and fuel melting increase due, in part, to increased rod internal pressures and limited fuel thermal conductivity, respectively. This work collaboratively employs PARCS, RELAP5-3D, and BISON to compare the fuel performance of two high burnup fuel candidates with higher than 5 % enrichment. The fuel performance parameters were compared to current NRC guidance. The results demonstrate an annular fuel design with homogenously blended gadolinium as a burnable absorber operates with greater safety margins during normal operation, allowing for additional operational flexibility. During normal operation, the core design utilizing Integral Fuel Burnable Absorber pins contained fuel pins which reached plenum pressures above 15.5 MPa by the end of the first fuel cycle and fuel pins experienced cladding hoop strains above 1 %. In the Gd core design, only two observed pins experienced plenum pressures above 15.5 MPa and no pins exceeded 1 % cladding hoop strain. During the control rod withdrawal scenario, plenum pressures for pins in both designs marginally exceeded system pressure, however neither experienced excessive hoop strain. The Gd core design experienced a maximum fuel temperature of 2418 K, which is significantly higher than the Integral Fuel Burnable Absorber design at 2157 K, but still within regulatory guidance. We predicted that the fuel in both could return to service after the CRW event. We also predicted that cladding would not fail during the Control Rod Ejection in either core design. Generally, the Integral Fuel Burnable Absorber core design performed with greater safety margin with regards to temperature during normal operation and the transient events. However, the Gd core design performed with greater safety margin regarding plenum pressure and hoop strain limits during normal operation and both transient events.

Criticality Safety Analysis for Cask Design with High Discharged Fuel Burnup Using MCNPX Code

Criticality Safety Analysis for Cask Design with High Discharged Fuel Burnup Using MCNPX Code

Simulating Properties of Canadian Research Reactor Fuels Important to Disposal

Canada has operated 17 research reactors at 11 different locations. The spent fuel from these research reactors differs significantly from CANDU fuel, which makes up the vast majority of spent fuel in Canada, and will eventually require disposal. The focus of this paper is to identify properties specific to Canadian research reactor fuel designs that would impact their suitability for disposal. The radionuclide inventory, hazardous chemical inventory, decay heat, residual enrichment, radiation decay rates, and gas generation of several Canadian research reactor fuel designs were simulated using the SCALE 6.2.4 software suite. The National Research Universal U3Si/Al dispersion rod, the National Research Experimental uranium metal X-rod, the Royal Military College UO2 SLOWPOKE-2 core, and the Whiteshell Reactor 1 uranium carbide bundle were investigated. Fuel burnup is the primary driver for the concentration of most radionuclides, hazardous chemicals, decay heat, and radiation decay rates. Carbon-14, chlorine-36, and mercury are driven by initial impurities in the fuel, which vary by fuel design. Low burnup, enriched fuels constitute a reasonable bounding case for the evaluation of criticality safety and proliferation risks. Canadian research reactor fuels are unlikely to present a greater risk of over pressurization from helium generation than CANDU fuel. Overall, the small volume of Canadian research reactor fuels requiring disposal is an important factor in the evaluation of disposal requirements.

Optimization of Neodymium Separation from Cerium and Samarium for Nuclear Fuel Burn-Up Determination

Sorption behavior of neodymium (Nd) from cerium (Ce) and samarium (Sm) on anion-exchange resins was investigated by column techniques. The anion exchange studies involved the sorption of Nd, Ce and Sm ions onto Dowex 1-X4 (100–200 mesh) in nitric acid – methanol medium. Separation of neodymium from cerium and samarium was carried out by column method to find out the best separation parameter condition with variating HNO3 concentration and weight of resin. The best elution profile was found using 0.3M HNO3 : methanol (1 : 9) as eluent through 1.25 grams of Dowex 1-X4 resin with 97.16 % Sm recovery and 95.59 % Nd recovery, while Ce was retained in the column.

Effects of Evaluated Nuclear Data Libraries on the Calculation Results for Fuel Burnup with Minor Actinides in a VVER Reactor

Effects of Evaluated Nuclear Data Libraries on the Calculation Results for Fuel Burnup with Minor Actinides in a VVER Reactor

A comparison study on the burnup of HTR-10 fuels using radiometric and mass spectrometric methods

Online burnup measurement of a single irradiated spherical fuel element using gamma spectrometry has been applied for years on HTR-10. This reactor has a complex operation history for varies experiments, which raises the uncertainty over the non-destructive measured burnup. In this research, three irradiated spherical fuel elements with different burnup were deconsolidated, and their coated particles were analysed by both radiometric methods based on Cs-137 activity and mass spectrometric methods based on neodymium and actinide mass. The burnup distribution in each spherical fuel element was indicated by fractional U-235 burnup of particles from various radii and relatively uniform. The inventory of Cs-137 in one spherical fuel element was measured by on- and off-line gamma spectrometry and results showed tiny differences. Destructively measured Cs-137 activities from particles indicate negligible portion of Cs-137 in the graphite matrix. Burnup calculated from Cs-137 by radiometric method and Nd-148 by mass spectrometric method were compared. Results showed that the two values were not significantly different for fuels with a clear radiation history, but differences were observed in the fuel with a complex radiation history.

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

Effect of the Burnable Absorber (Gd) on the Multiplying Coefficient (K<sub>∞</sub>) in the Process of Fuel Burnup for Full-Scale and Poly Cells Models for the WWER Reactor

The paper considers various concentrations of burnable absorber (BA) Gadolinium (Gd) in the system of compensation of excess reactivity in the reactor of WWER type at the extended campaigns. It is analyzed the influence of the method for the Full-scale and Poly-cells model placing the burnable absorber in the fuel with gadolinium rods (tveg). The strong influence of the BAs composition in the fuel with gadolinium rods (tveg) dependence on the multiplication factor of the fuel burnup is shown.

A study on nuclear hydrogen production using a novel approach cobalt-chlorine thermochemical cycle in a laser driver fission fusion blanket for various molten salt fuels

This study presents the hydrogen production potentials of laser inertial fusion engine (LIFE) reactor using two different molten-salt (ThF4 and UF4) fuels. For the hydrogen production method the Co–Cl cycle, which is one of the latest thermochemical cycles, was selected in this study. The neutronic calculations were performed with MCNP nuclear code. Flibe as the main constituent was mixed with 2 vol % ThF4 and 2 vol % UF4 heavy metal molten salt. The initial tritium breeding (TBR) values were computed as 1.124 and 1.133, respectively, for corresponding heavy metal molten salt vol. fractions. Blanket initial energy multiplication values (M) have been calculated as 1.277 and 1.339, respectively, as a result of exoenergetic neutron absorption in 6Li. The highest M values over plant operation was approximately be found as (at the end of the 34th and 35th years) 9.882 and 5.921, respectively. Fuel burnup (BU) in the blanket was computed as 7.9 GWd/tM and 40.7 GWd/tM for the ThF4 and UF4 fuels, respectively. The amount of the hydrogen production (m˙H2) was calculated as 8.987 kg/s and 4.942 kg/s for the same fuels, respectively. According to the numerical results that were acquired from the simulations it can be said that the considered LIFE molten-salt fusion reactor has a high neutronic performance and can produce a considerable amount of the hydrogen production.

Effects of azimuthal temperature distribution and rod internal gas energy on ballooning deformation and rupture opening formation of a 17 × 17 type PWR fuel cladding tube under LOCA-simulated burst conditions

ABSTRACT In order to contribute to better modeling and evaluation of fuel fragmentation, relocation, and dispersal expected under loss of coolant accident (LOCA) conditions, LOCA-simulated cladding burst experiments were performed on as-received nonirradiated 17 × 17 type Zircaloy-4 cladding specimens that were internally pressurized. The experiments were designed to terminate at burst occurrence to focus on ballooning and rupture opening formation and to investigate the effects of various factors. The postburst cladding hoop strain decreased with the increase in azimuthal temperature difference (ATD) of the cladding, as found previously. The rupture opening size increased with the increase in ATD and the increase in energy of the pressurized gas stored inside the pressure boundary of the test sample system. Comparison with the existing database, which included tests on irradiated rods containing fuel pellets, suggested that formation of the rupture opening was influenced by the characteristic behavior of high burnup fuels, such as limited gas migration in the cladding tube due to fuel–cladding bonding and interaction of the ejected fuel fragments with the cladding tube.

Physics Assessment of the Impact of Modified End Pellets on Axial Power Peaking for Advanced/Nonconventional Uranium-Based Fuels in Pressure Tube Heavy Water Reactors

Abstract Axial power peaking is a phenomenon with safety implications for pressure tube heavy water reactors (PT-HWRs). Since PT-HWRs use shorter (∼50 cm) bundles, there are small axial gaps, which expose the ends of the fuel elements to more neutron flux, and therefore results in higher power density levels occurring in the ends of the fuel elements. Power peaking has the potential to cause fuel damage and failure, if the local linear element rating (LER) exceeds 57 kW/m, and may be of greater concern for advanced, higher burnup fuels. Earlier studies have been done using three-dimensional mcnp models of a PT-HWR fuel bundle with slightly enriched uranium; they demonstrated that ThO2 could be used to reduce axial power peaking in fresh fuel. This result was achieved by replacing some of the UO2 with ThO2 in the last 3 cm of fuel pellets at each end of a fuel bundle. This study extends the previous work by performing 3D neutronics and burnup calculations using serpent, to evaluate how power peaking changes with burnup. In addition, alternative dilution materials (such as depleted UO2, ZrO2, and MgO) were also evaluated. It was found that axial power peaking can be significantly reduced by using the ThO2 dilution material for fresh fuel, while ZrO2 or MgO are even more effective at higher burnup levels. Dilution materials have little impact (less than 2%) on the exit burnup of the fuel.

FLUKA study of the actinides induced irradiation damage in Zircaloy-4 cladding during interim dry storage

Among the factors influencing the degradation of the spent nuclear fuel cladding in interim dry storage, the irradiation history and the average burnup at discharge must be considered. In fact, due to the pellet-cladding contact particularly affecting high burnup fuels, the inner surface of the cladding becomes increasingly exposed to the damage caused by the alpha decaying actinides present at the rim of the pellet. Moreover, due to the low temperature conditions characteristic of the interim dry storage, thermal recombination of the produced defects is not expected to occur. Here, we investigate the irradiation damage accumulated inside an irradiated Zircaloy-4 cladding 32, 55 and 100 years after the end of irradiation and discharge from the reactor core. The considered cladded UO2 pellet belongs to a Pressurised Water Reactor (PWR) fuel rod consisting of five segments and having an average burnup at discharge of 50.4 GWdtHM−1. The calculations performed with FLUKA 2021.2.0 Monte Carlo code show that the volume mostly affected by the irradiation damage corresponds to the ZrO2 layer formed between the pellet and the cladding. The actinides which are responsible for the alpha damage are mainly 242Cm, 244Cm, 241Pu and 238Pu. The recoiling daughter nuclei during the alpha decays produce irradiation damage only within the first µm of oxide layer.

Radiometric Assessment of Nuclear Fuel Burnup in the Romashka Reactor

Radiometric Assessment of Nuclear Fuel Burnup in the Romashka Reactor

Study of the hardness and Young's modulus at the fuel-cladding interface of a high-burnup PWR fuel rod by nanoindentation measurements

During irradiation in a nuclear reactor, the fuel swelling and the cladding creep are at the origin of the contact between the fuel and the cladding, which in turn leads to the initiation of Zr oxidation by the UO2 fuel. In high burnup fuels, the zirconia layer formed at the Fuel Cladding Interface (FCI) presents numerous circumvolutions and a heterogeneous microstructure within its thickness. In this paper, nanoindentation measurements are carried out in the fuel-cladding interface of a PWR fuel rod irradiated up to a high burnup of 61 GWd/tU. The measured hardnesses and Young's moduli of the internal zirconia layer are compared to reference measurements obtained on a non-irradiated sample of commercial Y-TZP. The study of the evolution of mechanical properties across the zirconia layer led to the identification of four distinct zones related to the zirconia microstructure (phase, grain size, porosity and fission product recoil). Close to the cladding, the zirconia hardness HZrO2tends toward the zirconium hardnessHZr. Then, the zirconia hardness increases significantly with the grain size, until reaching a plateau at the middle of the zirconia layer, where the maximum hardness is obtained. Following the same trend, the elastic modulus progressively increases in the zirconia layer and reaches a plateau near the fuel High Burnup Structure.

Analysis of Fuel Burnup and Transmutations at High Burnup of Sodium Fast Breeder Reactor

In this paper, the Monte Carlo N-Particle extended computer code (MCNP) were used to design a model of the European Sodium-cooled Fast Reactor. The multiplication factor, conversion factor, delayed neutrons fraction, doppler constant, control rod worth, sodium void worth, masses for major heavy nuclei, radial and axial power distribution at high burnup are studied. The results show that the reactor breeds fissile isotopes with a conversion ratio of 0.994 at fuel burnup 70 (GWd/T), and minor actinides are buildup inside the reactor core. The study aims to check the efficiency of the model on the calculation of the neutronic parameters of the core at high burnup.

Application of metaheuristics optimization in fuel rod design: A case study for helium charging pressure

In this paper, it is suggested that the design parameters of the fuel rod can be optimized by using a metaheuristics algorithm so that the fuel rod is entirely safe in its normal operation and has a more efficient performance than usual. We applied Artificial Bee Colony (ABC) in fuel rod design. The fuel burnup, irradiation history, and operation data of the heavily burned fuel rod are obtained for typical VVER-1000 NPP by modeling core cycle design data from the first to the fourth cycle in the KASKAD package. This package intended for core calculation of VVER reactors consists of several primary modules such as BIPR-7A for assembly-wise core calculation, PERMAK-A for pin-by-pin core calculation, and TVS-M for macroscopic cross-section generation.The conventional initial gas pressure in the VVER reactor is 2 MPa. In this paper, considering the effects of gap gas pressure on the heat distribution of the fuel rod, the objective function has been selected to reduce the maximum temperature and the average temperature of the fuel pellet while maintaining safety standards. The optimal pressure obtained in this paper for this particular reactor with a specific power history is 0.467 MPa which is entirely safe in terms of safety throughout the reactor operation. It reduces the average fuel temperature during the process. Furthermore, the core calculations show that using this optimal pressure increases the quantity of energy produced. Fuel burnup increases at optimum pressure with an overall decrease in fuel temperature.

Analysis of Fuel Burn-up Calculations of 3600 MWth Sodium cooled Fast Reactor Core

Fast spectrum reactors are essential for the future of nuclear energy. Therefore, there is a need for continuous research and development of the design and safety of current and future nuclear fast reactors. The aim of this work is to analyze the process of fuel burnup in a large scale (3600 MWth) Sodium cooled Fast Reactor (SFR) core. This design is called European Sodium Fast Reactor (ESFR). It is proposed in the 7th Framework Programme within the Euratom Framework. A new version (version 2.7) of Monte Carlo neutron transport code (MCNPX) was used to design a 3D model of the ESFR core to evaluate and analyze a number of burnup-relevant characteristics. These include the flux and power distributions across the ESFR core as well as the reactivity changes and fuel transmutation during burnup to take into account the changes in fuel composition during burnup. Obtained results can serve as up to date evaluation for the design and well allow for a detailed assessments of the fuel performance inside the ESFR core.

The burnup raising feasibility by studying the effects of Th and U contents on the thermophysical and mechanical properties of Th1-xUxO2 solid solutions

A comprehensive molecular dynamics study was carried out to investigate the role of thorium and uranium contents on improving the thermal and mechanical properties of the Th1-xUxO2 nuclear fuels. In this regard, several thermophysical and mechanical properties of the Th1-xUxO2 solid solutions including the lattice parameter, density, thermal expansion coefficient, specific heat capacity, thermal conductivity, thermal diffusivity, linear (heat/power) rating, elastic constants (C11, C12, C44, bulk, shear, Young's moduli, and Poisson's ratio) and the maximum radial thermal stress applied to fuel pellets were numerically investigated in a wide range of temperature and compared with experimental results. The results indicated that the Th1-xUxO2 fuels have a lower density, specific heat capacity, and thermal expansion coefficient than pure UO2 at any temperature. The results also showed that the Th1-xUxO2 fuels could increase or decrease the thermal conductivity, thermal diffusivity, and the linear heat rating of nuclear fuels depending on the uranium content (x). The amount of x in which all of the mentioned parameters can rise calculated in a wide range of temperatures. Maximum radial thermal stress was another important parameter that explicated in this paper. Results demonstrated that the Th1-xUxO2 fuels have lower maximum radial thermal stress, especially at high temperatures. These improved the thermo-mechanical properties provide the possibility of rising the fuel burnup in the thorium-based fuels compared to the uranium-based fuels.